ML20070U703
| ML20070U703 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 12/31/1990 |
| From: | Broughton T GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| C311-90-2156, NUDOCS 9104090273 | |
| Download: ML20070U703 (55) | |
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GPU Nuclear Corporation n U Nuclear :::,me:reo s Middletown. Pennsylvania 17057-0191 717 944 7621 TELEX 84 2386 Writer's Direct Dia! Nurnber; yggy (717) 948-8005 C311-90-2156 U.S. Nuclear Regulatory Commission Attention Document Control Desk Washington, D.C. 20555 Dear Sira
Subject:
Three Mile Island Nuclear Station, Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 10 CFR 50.59 Report for 1990 In accordance with the requiremente of 1C CFR 50.59, enclosed are summaries of the changes to TMI-l systems and procedur 9, for the period of January to December 1990, ao described in the Safety Anaylein Report (SAR). Attachment 1 of this report addresses those activities which directly affected eyatemo/componento described in the SAR. Attarhment 2 of this report addresses those activities for which a GPU Nuclear safety evaluation was performed, due to the potential for the activity to adversely affect nuclear safety or safe plant operations, but which do not directly impact SAR systems /componenta.
Sincerely, W
T.
/A G. Bro nton Vice President & Director, THI-l TGD/RDW cca R. Hernan - Senior Project Manager T. Martin - Regional Administrator J. Stolz - Director,. Plant Directorate I-IV F. Young - NRC Sr Resident Inspector TMI Enclosure 9.104090273 PDR 901231 i R ADOCK 05000289 'F l r
PDR 4
qqt:*gM v, GPU Nuclear Corporahon is a subsidiary of General Punhc Unkbes Corporation
=
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C311-90-2156 Page 1 Attachment 1
- 1. Tests and Experiments Activity: Special Test Procedure (STP) 1-90-039 Once Through Steam Generator (OTSG) Operating Range (OR) Level Verification (SE 000224-009 Rev. 1)
Description of Activity:
The purpose of this test was to increase the OTSG OR level to demonstrate thats a) safety analysis assumptions are not exceeded even if the downcomer is full of water; and b) loss of feedwater heating will not adversely affect plant safety.
(See Section II, " Procedure Changes", of this report for discussion of the results of this test).
Safety Evaluation Summary:
The OR level indication was originally conceived as a relative measure of OTSG inventory during power operation. Although the OR level is of ten ref erred to as downcomer level, both field data and testa show that there is no steam-water interface in the downcomer, but rather a steam water mixture that extends from just below the main feedwater nozzle down to the downcomer orifice. Raising of the OTSG high level to maximum had the potential to af fect the feedwater heating function if the feedwater nozzles were flooded and could increase the secondary side steam and water mass inventory as related to main steam line break analysis.
As a result of raising the OTSG high level limit, two (2) possible effects on system performance were considered: 1) downcomer flooding as related to maintaining the feedwater heating function; and 2) secondary side stoam and water mass inventory as related to the design basis accident analysic.
This test procedure was developed to slowly increase power (thus the downcomer level) to approach the maximum achievable level. The purpose of this test was to determine the indicated level at which the plant can be operated without losing feedwater preheating. The main steam line break (MSLB) accident from hot shutdown conditions 10 not af fected by a change in the high level limit since the steam generators are operated in a flooded nozzle condition. Calculations determined that at 100.2% OR level at 100% power, the feedwater nozzle will not be flooded and proper feedwater heating is attained. The margin of safety defined in the SAR was not reduced since the total inventory at the proposed higher level is less than the value of 62,600 lbm assumed in the SAR and 55,000 lbo assumed in other analyses. The running of an OTSG high level test did not increase the probability of occurrence of any accident or transient. Neither increased downcomer inventory /nor decreased downcomer temperature initiate accidants or transients.
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4 C311-90-2156 Page 2 Attachment 1 Activitvt Lithiam Tracer Testing for Venturi Nozzle Calibration (SE 412565-002 Rev. 0)
Description of Activity:
The purpose of the CE Chem Trae lithium tracer test was to calibrate the replacement feedwater flow venturin by accurately measuring feodwater flow to each OTSG using a chemical tracer (see related modification BA 412565 under Section III of this report).
Safety Evaluation Summary:
The only aspect of the chemical tracer test which could have impacted an accident probability related to secondary side lithium injection. The OTSG tube rupture is the only accident evaluated in the SAR whose probability could have been impacted by secondary side chemistry alteration. Based on the analysis in this SE, lithium had no degrading effect on the OTSGs; thus, the probability of occurrence of any previously evaluated accident was not increased. The conduct of this test did not introduce any new accident precurs,'rs. Reactor operations were maintained within licensed limits. Administrative controle were in place to preclude any adverse condition resulting from improper control of equipment to conduct the test. Thn nuc30ar instrumentation was calibrated based on conservative estimates of core powt r. Thus, the margin of safety as defined in the bases for the Technical Specifications relative to core power was not reduced.
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3 C311-90-2156 Pago 3 Attachment 1 II. Procedure Changes
- pgocedure: AP 1035 Control of Transient Combustible Materialo (PCR 1-EG-90-0003)
Doncription of ChaA928 This change incorporated a control rod storago box (untreated lumber) to be stored in the fuel pool area approximately three (3) months af ter each refueling outago.
Safety Evaluation Sugnar,yi The TMI-1 Firo }!azards Analynlo Roport (FHAR) identifies the type of transient combustibloo stored in each fire aron / zone. This proceduro chango will be incorporated into the next updato of the FHAR. All changos in the fire loading are within the limite catablished to support the 10 CFR 50 Appendix R ovaluation.
Procedurg: AP 1038 Administrativo Controla - Fire Protection Program (PCRn 1-EG-90-0001, 1-EG-90-0004, and 1-EG-90-0015).
Deceriotion of Changgs
- PCR 1-EG-90-0001 corrected procedural references, and updated Exhibit 2, Table 1 to show reduced number of firo detectors associated with the Mod Comp Halon syntom.
PCRs 1-EG-90-0004 corrected a procedural reference f rom OP 1105-21 to OP 1105-20, Romoto Shutdown System."
PCR 1-EG-90-0015 implomonted fire drill changes dincussed in CPU Nuclear lotter
- C311-90-2073 dated June 1,_1990 and reviewed (prior to irnplomontation) by HRC i
- lotter dated November 21, 1990. An unrelated chango was also mado to Soction 5.8 l - to doloto the requiremont to issue a quarterly fire drill reoultu nummary.
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Safotv EvaluftL{on Sunnary:
The correction and update of procedure referencon is administrativo in nature.
The dolotion of one f tre detector annociated with the Mod Comp Halon System was performod under DA 412281 and evaluated in SE 412281-005, Revision 2, and FPE TI-412281-004, Rovision 1. The minimum number of required operable dotectore romainn at throo (3) for the halon syntem, ono (1) detector was ollminated as the result of a reduction in the-sizo of the monitored area. Sinco the firo detection system la a priority matrix syntom and since the remaining tronch area
. La exposed to the obnorvation room undorfloor, the dotection nyntom will ntill
C311-90-2156 Page 4 Attachment 1 Safet.y Evalu_a_ tion Summary (Cont'd):
operate as required. The deletion of the quarterly drill results summary is administrative in nature since drill experiences are addressed in monthly
! training sessions. Therefore, these changes do not adversely affect nuclear safety or constitute an unreviewed safety question.
Procedulq: Revised OTSG liigh Level Limit Following STP l-90-039 (SE TI-ll5403-005)
OP 1101-1 Plant Limits and Precautions (PCR l-OS-90-359)
OP 1101-2 Plant Setpoints (PCR 1-OS-90-357)
SP 1303-11.37 A-D llSPS-0TSG Level and Pressure Channel Tests (PCRs 1-MT-90-8559, 8560, 8562, and 8563)
Alarm Responso Procedures 11 1 and 11 2 Main Annunciator Panel H (PCR l-OS-90-358)
Description of Chance:
The purpose of the above procedure changes was to establish a maximum steady state OTSG downcomer level and associated control system settings based on the results of STP l-90-039 (see Section I, " Tests and Experiments"). The results of the plant test to raise OTSG level were used in establishing the ICS high level limit and high level alarm setpoints.
Safety Evaluation Summary:
STP l-90-039 was conducted to determine the maximum OR level for operation without flooding the Feedwater (FW) nozzles. The test was performed by slowly raining unit load, thus downcomer level, while monitoring for a significant reduction in FW preheating caused by flooding the nozzles. The results of this test indicated that the downcomer level was beginning to encroach on the mixing region below the FW nozzles above approximately 99% OR level. A significant decrease in downcomer temperature did not occur during the conduct of this STP.
The plant response during this STP indicated that there is about a 14 inch (4.8%
OR level) margin available to accommodate slow and small upsets in plant conditions and not cause FW nozzle flooding. Downcomer temperature was relatively constant until OR level increased above approximately 95%. In summary, there were no indications of unacceptable plant response during the elevated level test. Steam superheat, cold leg temperature, OTSG pressure, core power and downcomer temperature responded within acceptable 1imits.
l C311-90-2156 Page 5 Attachment 1 EAfety Evaluation Summarv (Cont'd):
A significant reduction in lower downcomer temperature during a rapid transient could af f ect tube stresses, but the operator can reduce FW llow bef ore the shell to tube aT could increase above the compressive stress limit. This activity did not decrease the margin of safety as defined in th' bauls of the Technical Specifications. The TMI-l Technical Specifications do not have limits on either oTSG level or shell-to-tube differential temperature. However, a procedural limit was established to assure that thermal stresses were kept within design value. Nuclear saf ety and safe plant operations were not adversely af fected by raising the high level limit to a point close to flooding the FW nozzles during steady state operations. The OR levels experienced during the STP will not be exceeded during steady state plant operations. This STP validated that operating at the new high level limit would not cause nozzle flooding.
Procedures OP 1104-5 Reactor Building (RB) Spray System (TCN 1-90-0134)
Description of Chance:
The procedure was temporarily revised to provide for sodium hydroxide additions using a chemical addition pump instead of a crane, safety Evaluation Summary:
This procedure change created a potential pathway by which sodium hydroxide could drain from Sodium Hydroxide Storage Tank (BS-T-1). The sodium hydroxide tank level was maintained lower than the BWST 1(vel such that the BWST/SHST dif f erential level was maintained within Tech. Spec. limits. This change did not af fect the major components of the RB Spray System. A temporary rig was attached to the recirculation piping which had its own isolation valves. The inventory of the sodium hydroxide tank was maintained throughout the entire chemical addition; therefore, this change did not increase the probability or consequence of an accident previously evaluated in the FSAR.
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Procedure: Abnormal Pro :edure 1203-19 River Water System Failure, Decay Heat &
Secondary Se rvices (DR/SR) (PCR l-0S-90-0260) i Description of Chpnce:
This procedure change provided guidance to cross-connect Nuclear Services River Water (NR) to Secondarf Services River Water (SR) in the event of a total and sustained loss of SR.
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4 C311-90-2156- Page 6 Attachment 1 Safety Fvaluation Summary:
- his p ocedure change implements a design feature which would permit emergency cooling to secondary components while plant shutdown was in progress. While implowentation of this design feature degrades the safety classification of the NR f y utem, long term operation in this mode is prevented by the existing Tech.
Spec. requirements. Use of this option is considered to be safe since suf ficient NR cooling will be maintained during the shutdown process. Since this emergency condition would require a plant nhutdown within the time constraints established by Tech. Specs., this change had no adverso impact on the THI-1 FSAR.
Procedyr,3 SP 1302-1.1 Power r a nge Calibration (PCR 1-OS-90-0576) pesertnti u e< cnanaci Tid c. procedure change provided guidance for not performing power range calibration when the lluclear Instruments (NIs) exceed heat balance power by more than two (2) percent during temporary power reduction.
SafetyJ,yahqalion Summary:
This change did not aficct nuclear safety. It allowed the power range NIs to be greater than heat balance power for plant conditions of temporary power reduction. This provides a protective systam actuation at a lower actual power than if the NIo were adjusted down to agree with the heat balance. The probabiltuy of occurrence or the consequences of an accident were not increased by this change because it affects only calibration and does not change any protective functions.
The margin of safety was improved by this change. FSAR eection 7.3.1.3 states, "the sum (of upper and lower power range detectors) will be recalibrated whenever it.is rf Mrmined that the sum disagrees with the heat balance by 2 percent or more." This change created a larger margin to safety for the condition of a temporary power reduction. This change remnine in compliance with Technical Specification Table 4.1-1.
l Procedurq: Alarm Response Procedure H&V A-1-9 Reactor Compartment Air Tomperature High (PCR 1-OS-90-0343) l l pSecription of Chance I
this procedure chango increased the setpoint of TR-802 points 1,4,7, and 11, f rom 210 F to 235 0F.
C311-90-2156 Page 7 Attachment 1 1
Safety Evaluation Summary:
The temporary setpoint increase provided a means to return TR-802 recorder pointo 1,4,7, and 11, to service such that the alarm input is active. This was a tcmporary setpoint change to one (1) of two ( 2 ) indicators for each penetration dnd both remain functional and operational. Previously, the pointo did not provide alarm input since they cycled on and of f at the 210*F eetpoint which was a distraction and nuisance for operators. No conclusive data has been obtained that would provide indication of what concrete temperate.~e profile in relative to air temperature exiting the penetration. A one time test r9vealed concrete at 10@F and air temperature at 200 to 205 F. This suggested a ., of 15 to 20*
between actual concrete temperature and per. ' ration air exit temperature. Thus, an alarm setpoint of 235' would lead to a I . c oible concrete temperature of 215-220 which exceeda the FSAR limit of 200 F. ' rom a materiale standpoint, if the local concrete temperature would approach 220" there is not a otructural or strength concern. Thus, a temporary setpoint increase does not caulo operator distraction, provides a meaningful alarm indication for unrepresentative recordr points, and does not compromise the structural integrity cf the concrete.
Alternate temperature indication setpoint was not changed a"d continues to provide temperature alarm at original values. Therefore, nuclear safoty in not adversely affected by this change.
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ProceduIg PM M-155 Main Steam (MS) Relief Support System Clearance Readings (PCR l-HT-90-1003)
Description of Chances This procedure change eliminated the_taking of clearance readings during hot plant condittorto due to personnel safety. Gap readings for all the posts except MS-V21D were previously oilminated during hot conditions based on consistent measurementa " hot" and " cold".
Safety Evaluation Summary l
Historical data has shown that gap measurements that are acceptable during cold plant conditions are acceptable also during hot plant conditions. Therefore, gap settings are being monitored during cold plant conditions. The heated post for MS-V-21B was adjusted during the P1 outage no that the gap in within the j acceptance criteria during hot and cold conditions. Meast roments of the gap have shown thin to be the case. Therefore, elimination of '.nio surveillance during hot conditions for personnel safety did not jeopar'ilze the gap clearance or nuclear safety. This change le further addressed in detail in Safety Evaluation
( Sit-000415-001.
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C311-90-2156 .ge B Attachment 1 III. Modifications hgdification: Modification to Waste Level 1ontrollere LC-40A/B (WDL-649A/B)
[BA 123212 (123?16t; 1 Description of Modification This : codification improved the performance of the Waste Evaporator System Level Cm..rollert. by installing an improved model float type regulating valve, float Lugustment mech m.t.sm, eightglass/ access port, and isolation valves.
Safety Evaluation Summary:
This modification serves no safety function other than > r ?ssure boundary for a radwaste system. The modification does not compromiso 1.% sue Disposal (WDL) piping or the operability of who radwaste evaporators or any other system or component. The modification improves the reliability of tho Radwaste Evaporator Systent. The evaporators are r.ot nuclear safety related components and no nuclear safety related components are affected by this change.
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Modification: Installation of Air filters en Ctitical Plant Control Valves (BA 123250)
Description of_ Modification:
To provide a clean, reliable source of instrument air to critical plant control valves, instrument air filters or filters / regulators (with bypass and isolation capability) were installed at the point of use for the followino valves EF-V30 A-D, FW-V16A/B, FW-V17A/B, RS-VS A-F, MS-V4 A/B, MS-V6, HU-VS, MU-V17, MU-V32, AND RR-V6.
Safety Eva lu a t io;_Eumma ry :
Installation of local air filters or addition of a parallel filter / regulator in the supply air lines to the above valves does net af fect the safety function of the Instrument Air (IA), Backup Instrument Air (BUIA), or Two Hour Backup Air Supply L (2HBUI A) Systems. Filters and filter / regulators, with access for pH activities, ensured that a clean, reliable source of air is provided to critical etnerol valves. Filters and filter / regulators are adequately sized to prevent excessive loading of the filter elements within the PM period. This modification servet to improve the quality of instrument air supplied to the specified plant et, trol valves t.nd increases control valve reliability.
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C311-90-2156 Page 9 Atttchment 1 I Eodificatient Interned ate closed Cooling Water (ICCW) Drain Valve l Hodification (BA 128007/WA 51345310) i Descrintion of Modification:
The purpose of this modification was to provido a meane w$th which to drain the ICCW line supplying the control Rod Drive (CRD) motor cooling colle. The line is required to be drained when it is disconnected due to RV head removal. The scope of this modification involved the installation of a 3/8" drain valve and a short piece of tubing and tubing cap on the ICCW line in order to f acilitate l the draining of the line prior to disconnection for RV head removal. ;
Safety Evaluation. Summary:
I The installation of the dre.it itne did not affect the function or operation of the ICCW eystem. The valve is only uood during a plant shutdown prior to removal i of the RV head. The valve is i.ormally closed and capped. Thus, this modification did not adversel effect nuclear safety or safe plant operations The new valve is used only as _n aid to draining the 7CCW system during plant shutdown. The modification did not decrease the margin of safety as described in the SAR or the Technical Specifications since no system cooling functione or containment isolation functions were altered by this modification.
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Modification: Secondary Plant Performance Monitoring Hodification (BA 41227)
Description of Modificard.ga The purpose of this modification was to provide additional instrumentation to enhance the secondary plant performance monitoring. The scope of this modification was to install the necessary pressure and temperature devices to allow for trending ability and the identification of degraded equipment. The additional instrumentation was for the Main Turbine L.P. Section Pressure, I-Condenser Vacuum and L.P. Feedwater Heater Temperature, and various valves in the l seconiary plant.
Safety Evaluation Summarvt This modification did not reduce a margin of safety because there was no adveras affect on the safety features described in the applicable sections of the FSAR for the - af fected safety systems (i.e., Extraction Steam System and the Main condensate System). Additionally, the new instrumentation installed by this task is not governed by the Technical Specificatione; thus, this modification did not result in an unreviewed safety question.
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i f C311-90-2156 Page 10 Attachment 1
- Modifications Demolition of Bailey and Mod Comp Computer (BA 412281)
Description of ModificatiQD8 5 The Mod Comp and Dailey Computer System were replaced with a state-of-the-art 1 could SEL computer system as part of the plant process computer upgrade.
l Following removal, the remaining Dailey cabineto in the Relay Room and Control Room function only as termination cabinete. The "D" SEL computer was connected to the T-Dar cabinet in the Control Tower where the Mod Comp Computer had been previously connected. This modification also removed a fire detector from the section of the raised wirew,.v in the Control Room which was removed.
Safety Evaluation Summary:
This modification did not impa t,t equipment and electrical circuite which are required for 10 CFR 50 Appt adix_ R safe chutdown as _ described in the THI-1 FHAR.
j The fire detection system was not adversely affected since the volume protected by the removed detector did not impact the performance of the halon system since it has a priority matrix detection nyntem, whereby ary two remaining detectors will initiate the halon system. Interfacing systems were reviewed to identify any pertinent nuclear safety related functions or safe plant operatione. To onoure that these nuclear, safety related functions of safe plant operation requirements were not adversely affected, the following provisions were made:
a) Conduit was added in accordance with anta-f alldown support requiremento for Class I Structures.
b) Cable routed for this modification was entirely within solomic trayo within Class I Structures.
The SAR was reviewed to identify any accident scenarios previously evaluated that could be af fected by this modification. The results of this review were that no previously evaluated accident situation could be found that would directly or indirectly be affected by this modification.
}iodifica n lon: Regulatory Guide (R.G.) 1.97 Modifications (BA 412491)
' Description of Modifications
. R.O. 1.97 required that certain plant parametern be connected to the plant computer. The cables for these computer points were routed during 7R outage and terminat ad at the signal conditioning cabineto. Theec cables arc now terminated at ti,a computer I/O cabinet.
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C311-90-2156 Page 11 Attachment 1 i Safety Evalunt,ign Summarvs }
+- This modification did not adversely affect nuclear safety or safe plant operations. The computer is not a nuclear safety related system. The computer input signals were verified by performing a loop calibration. The margli af safety as defined in the basis for any Technical Specification was not reduced.
3 Availability of the new computer points is not assumed in the basis of any Technical Specification.
l Modification Radwaste Solidifiention System /Duilding Upgrade (BA 412521/WA 37521A10)
Descrintion of Modifications This modification involved the upgrading of both the Hittman processing system and the Hittman Buildir.g. Modifications included o change over of electrical power from Unit 2 to Unit 1 for increased reliability and availability.
o Permanent, "hard piped" connection of process vent and dewatering return line to the Auxiliary Building (Aux. Bldg.) HVAC exhaust duct and Aux.
Bldg. floor drains, respectively.
o conversion of the walin of the structure to concrete block for increased strength and durability, increased weather protection, added radiological protection, elimination of combustible materials, and improved decontaminability of building surfaces, l
o Relocation of radwaste supply lines inside the Hittman Building to improve interf ace with the radwaste so) Ldification liners and to provide
(- increased clearance between piping and operations personnel when acceos to the solidification process is required.
Safety Evgluation Summary:
These modifications improve reliabiliti and availability of the system, eliminate reliance on Unit 2 for any support services, reduce the potential for unplanned releases of radioactivity, reduce radiation exposure to operating personnel and to the general public, improve ALARA conditions, and simplify equipment setup and ,
material handling procedures. -connection of the process vent line to the decontamination facility ventilation exhaust duct has negligible impact on the ability of the decontamination facility ventilation system to perform its designed functivn. The Hittman system is non-t' clear safety and is not part of j the engineered safeguards system. All modifications were performed in acco: nee with established quality requirements and seismic classifications. The modified
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$gfety Eval.yA(lon SummaIY lE2Dt.'d) :
system and s t rr ; '.u re comply with applicablo regulations, Technical j specifications, an1 the station's Process control Plan (PCP) for solidification ;
of radioactive wasto. The modifications did not affect the safoty f unctions of I any interfacing system or structures.
1 Mod iLLg,pt i on : Wacto Oil Storage, System Hodificat ion (BA 412521/WA 37521830) 1 Descrint ion e t.Mqdi f icat io.D 8 !
The purpose of this modification was to install a storage system for handling wanto oil produced by plant componente. The new system provides a clean and officient means for storing, transferring, and campling the watto oil prior to 1
, ultimato disposal. The new system replaced the previous method which utilized a portable pump and cask linore for trenofor and temporary storage of wanto 011.
gafety Evahlation Syfangry: ,
The installation of the Wasto 011 Storage System did not af fect the operation of
. the Radioac.ive Wasto (Radwanto) Solidification System, including the liittman j facility. The modification installed a permanent storage system to replace the temporary ogalpment. The modif1 cation did not increans the probability of l occurrence or consequences of an accident since the syntom replaced a temporary, i
makoshif t oil storago system with a dedicated, permanent system. The now system providon better control and handling of wanto oil transfar and storage. This modification did not adversely affect nuclear safety or safo plant operatione cinco the safety functions of Radwasto nyetomo, the Auxiliary Building Structure, and plant electrical systomo were not advernoly af fect "d. .
Modificati2D: Bailey BY Transmitter Replacement (BA 412522)
Deocription of ModificatinD The purpoco of this modification was to replace six (6) Bailey BY transmittero l with now Rosomcunt differential pronouro instruments. Ono BY transmitter, RC1-LT2 was deleted and not replaced. The transmitters af fected by this modification were the proosurizer level instruments Rci-LT1, 2, and 3, and four (4) MPW flow inntruments SP8A/B-DPT1/2.
j Safety Evaluation Summary:
The prosaurizer lovel transmittore provide information to the plant computer system, indication to the control r oom operators, and provide a pressurizer lovel
-eignal to the ICS system. The feedwater-flow transmittera provide an input
C311-90-2156 Page 13 Attachment 1 Safety Evalua11on Summarv (Cont'd):
signal to the ICS system and provide indication cf feedwater flow to the control room. This modification does not decrease any margin of safety since this activity does not impact any systems' saf ety functions, the wrk 10 perf ormed in 69eordance with existing seismic requirements, and electrical / physical separation was ptovided. New tubing and associated fittings for the pressurizer instrument impulse lines were installed and inspected in accordance with existing design criteria no as not to provide a leakage path for reactor coolant.
Rodification: MW Digital Meter / Transducer Upgrado (BA 412528/WA 30128310) procription of_ Mod _lileatient In order to provide the operator with more accurate indication of main gener . tor MW the analog meter was replaced with a digital meter. The previously existing MW transducers used f or ICS input are no longer available. These transducero (2) were replaced by more state-of-the-art type transducers, ggiqLyDhp11on Summary:
The margin of safety as defined in the basis for any Technical Specification was not reduced. The MW signal is not defined in any Technical Specification er'ety limit basis. Safe plant operations were not adversely affected since the operators are still provided with an indication of MW when loading and unloading th3 main generator. The availability of generated MW signals le not taken credit for in any safety analysis of Chapter 14 of the SAR. There are no saf ety related interlocks associated with MW signal loops.
Mod i f ic tLiip,n Installation of Main Steam to Heated Poet Isolation Valveo (BA 412528/WA 30129310) peserietion of Modification:
A means shall be provided to maintain the heated post system while the pla.it is at power. Previously, all four (4) lines were required to be isolated to permit maintenance. Isolation of all four trains ir nct permitted by che Technical Specifications. This modification inL2alled four inolation valves in the steam return line downstream of the drip leg. This permite short term isolation of individual trains for maintenance of related piping.
Hafety Evaluation Summary:
The main steam to heated poets supports system is designed to maintain the supports for the Main Steam Safety Valve discharge line such that the main steam
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C311-90-2156 Page 14 Attachment 1 1
i
-$11d v Evaluation Summarv (Cont'd) piping is not overstressed when the valves open. This modification did not increase the probability of occurrence or consequences of an accident, since the normally open valves will not af fect system performance and the safety f unction of the heated post supports system. This modification did not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR, since the I integrity of the pressure boundary and the seismic boundary of the main steam to I heated post supports system are still maintained. This modification permits on l line maintenance of individual trains of the main steam to heated post supports and related piping.
Modification: Deletion of Water Detector All-LS-2 4 9 A/B/C for the RD Ventilation Fan Motors (BA 412528/WA 30149310)
[Lescription of Modification:
This modification deletes the water detector, All-LS-2 49 A/6/C for the RB Ventilation f an motors, All-E-1A/B/C, and spares the associated Control Room alarm and computer points. The water seals on the f an motor shaf ts were replaced with a dry seal.
Safety Evaluation Summarvt The margin of safety defined in the FSAR Chapter 6 was not reduced, and nuclear safety and safe plant operations were not adversely affected. The All-E-1 motor design with dry shaf t seals and continuous draining with backflow prevention from check valves protects the motor against motor water induction. Therefore, AH-L-249A/B/C are no longer required to assess the operability of the motors. Removal of the water detectors eliminates nuisance alarms due to repeated f ailures of the moisture detectors. Deletion of the moisture detectors did not introduce any new f ailure mechanisms for the RB f an motors. A pipe plug was installed in place of the moisture detector to prevent a possible new pathway for moisture leakage to the motor cavity.
Modification Heat Sink Protection System (HSPS) Additions (BA 412538)
Descriotion of Modification:
GPU Nuclear committed to install, during the SR refueling outage, different control room indication for the Hain Steam Rupture Detection (HSRD) portion of the HSPS. THI-) Operations, Maintenance, and Plant Engineering had identified other concerns with the HSPS configuration, particularly as it related to the man-machine interf ace. A summary of each of the individual tasks associated with
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C311-90-2156 Page 15 Attachment 1 paggiotion of HodificatLgfL f Cont'dt:
this project in described below 1
j 1. Provided new design to resolve the NRC's SER comment on MSRD control room indice. tion which also includes new provinicno for the CRO to isolate the OTSG independently.
- 2. Provided new channel-check provisione via the new computer system that would allow the CRO to read and compare all analog input parameters to l HSPS. l
- 3. Installed additional instrument tubing isolation valves that would allow ,
transmitter testing without degrading H'SPS logic from 2 of 4 to 1 of 2.
- 4. Installed new and reconfigared existing control room alarme concerning Emergency Feedwater (EFW) Actuated, EFW Defeated, end Main Feodwater (HFW) Isolated to provide clearer and more consistent annunciation of actuation / defeat status.
- 5. Installed new centrol room valve position indication for MFW control valves FW-V16/17; no direct position indication had previously existed.
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- 6. Installed new capability to determine status of back-up powei supply to HSPS cabinete.
- 7. Installed new capability to determine status of isolation fuses for MFW isolation valve solenoids.
- 8. Installed additional provisions to allow V/I module testing and Resistance Temperature Detector (RTD) test input without lif ting leads in HSPS cabinets.
- 9. Installed new capability to test HSPS status ligh e in the control room and on the HSPS cabinets.
- 10. Provided a rescaled startup OTSG level range to control room recorder for enhanced readability at low levele,
- 11. Changed the default signal to che EFW control valves such that the controller will revert to the OTSG startup level signal rather than the OT50 operate range level signal.
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C311-90-2156 Page 16 Attachment 1 pfseriotion of Modifleation fCont'd):
) 12. During testing of this rnodi f icat ion, an as-found condition was discovered which required a change to be engineered and installed prior
. to restart. It was found that partial losses of power within the HSPS l cabinets could cause MrW isolation. Blocking diodes were innta11ed in l the HSPS cabineto, and additional confirmatory testing was performed.
Safety Evaluntion Swamam The installation of this modification improved the capability of maintaining the OTsce as the primary heat sink and provides additional information to the
! operator in other related areas associated with the HSPS. -his modification did not introduce any accident or malfunction not previously evaluated, nor did it l inerence the likelihood of occurrence or consequences of any accident as analyzed in the SAR. No safety-margine were reduced as a result of this modification.
Therefore, this modification did not adversely impact nuclear patoty and does not represent an unreviewed safety question.
Modification: MU-P-1B Lubs 011 Pump Power Supply Changeover
- (BA 412543) pfagfiplion of Modifleation
Tt'I - 1 Plant Operations and Probabiliutic Risk Assecament (PRA) identified a condition which could cause a total lone of High Pressure Injection (HPI) due to a single f ailure. This condition arose when either MU-P-1 A or 1C is unavailable, ;
MU-P-1B is selected for ES operation, and the IC ES Valve MCC supplies power to the auxiliary and main lube oli pump (MU-P-2B/38) either of which is required for operation of MU-P-la. The modification changed the power supplien for HU-P-2B/3B from the previoun source, 1C ES Valve MCC to the followings o MU-P-2B -- 1A ES Valve MCC o MU-P-3B -- 1B ES Valve MCC This new power supply configuration enouros that MU-P-1B always has an energized lube oil pump to support operation.
Safety Evaluation Summ4 m This modification does not adversely affect nuclear safety or safe plant
. operations and does not reduce a margin of safety because the purpose of this modification was to ensure that MU-P-la always has an energized lube oil pump to 9upport ite operation. Reconfiguring the power supply of the lube oil pumpo ensures that MU-P-2B in available to support MU-P-1B when powered from the ID
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t C311-90-2156 Page 17 Attachment 1 Modificat1GD BR Underprotected Cable Fix Modification (BA 412561) 5 Demeription of Modification: '
This task implemented during the BR outage the modifications recommended in the Underprotected Cable Study Interim Report, Revision 1, dated June 5, 1989. This report identified the underprotected cables which represented the greatest risk to safe plant operations and/or premature cable degradation. This modification encompasses the following cable, breaker, and overload heater replacements.
-Replacement Change Change Powered Item Circuit ELQal._ Toi From Lp, gA Cable CB4 #6 AWG #2 AWG 1B Turbine B Generator Plant - Unit Bus Duct Yan ID Breaker CG54 70 AMP 50 AMP 1A ES MCC-Unit Heating coil SDL AH-E-29A i Breaker CS63 30 AMP 20 AMP 1C ESV MCC- Instr. Air -
Unit 98R Dryer IA-Q-1 I Breaker EA371 30 AMP 20 AMP 120 VAC-PNL HSPS Cabinet VDD-#20 B Breaker EA771 30 AMP 20 AMP 120 VAC-PNL HSPS Cabinet VBD-#20 D 0ver1oad CL41 H51 FH50 1A Reac. Plant Reaetor l Heater H&V MCC-Unit Operating SA Floor Supply Fan AH-E-3A Eafety Evaluation Summary:
The margin of safety defined in the FSAR was not reduced because this modification was designed and installed in accordance with established procedures and criteria such that 'the safety function of associated systems were not adversely compromised. This modification did not decrease the margin of safety as described in the bases of the Technical Specifications because this modification was a replacement-in-kind. Class 1E circuit breakers powering HSPS sbineta were replaced with Class 1E breakers with lower trip ratings to properly protect the power cable in accordance with its design capacity.
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I C311-90-2156 Page 18 Attachment 1 i
Eglely__ Evaluation Summary The margin of safety defined in the FSAR was not. reduced because this )
modif,' cation was designed and installed in accordance with established procedures and eritoria such that the safety function of associated systems were not adversely compromised. This modification did not decrease the margin of safety as described in the ba set, of the Technical Specifications because this modification was a replacement-in-kind. Class 1E circuit breakers powering HSPS cabinets were replaced with Class 1E breakers with lower trip ratings to properly protect the power cable in accordance with its design capacity.
Modifit.pt ien t Main Generator Relay Protection (DA 412563)
Eg C stion of Modification:
This modification replaced the underf requency relay scheme with a new two-out-of-thtee relay matrix. The new matrix provideo the same trip and alarm functions as the previous relays, and also includes the breakers for the incoming line from the 500KV substation. The primary function of these relays is to protect the turcine generator under lov frequency conditions.
Eafety Evaluation Summary:
The underf requency relays perform no safety related functions or affect the operation of any safety related equipment or systems. The two-out-of-thre, matrix enhances reliability, and eliminates the possibility of a malfunction of the single underfrequency relay causing a turbine and reactor trip. By tripping breakers to separate the turbine generator from the grid under low frequency conditions, which is consistent with original design philosophy, the underfrequency relays prevent damage to the low pressure turbine bladec. Thus, i t.here is no increase f.n the probability of occurrence or the consequences of an j accident previously evaluated in the SAR.
i Modification: Relocation of Decay Heat closed Cooling Water (DC) Valves DCV-2A/D & 65A/B Controls (BA 412569)
Description of Modification:
The purpose of this modification was to alleviate operator burden for a task that wan necessaty during all normal plant shutdowns. Previously, valves DC-V-2A/B and DC-V-65A/B are controlled during cooldown using local manual pneumatic controllers located above the decay heat vaults. This required statinning an-auxiliary operator at the controllers and calling down from the control room the necessary adjustments. The pneumatic controllers were replaced by current to pneumatic converters driven f rom manual loading stations located on main control l
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C311-90-2156 Page 19 Attachment I l 1 Description of Modification (cont' die I
consolce CC and CR. Nuclear safety related solenoid valves and key locked control switches were added to disable the control loops during normal operation.
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! Safety Evalugtion Summary:
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There are no accidente described in Chapter 14 of the FSAR which are applicable to this modification. The probability of occurrence of an accident previously eyeluated or malfunction of a different type than any previously identified in the SAR was not J r.troduced. Failure of a control component in any one train (A or B) of the new system configuration creates the same condition as it would be created by a failure of a component in any one channel of the existing control system. The consequences af a control component failure would be the inability to remotely adjust the DC water flow and therefore the cooling rate. However, for the LPI operation (i.e. , an_aecident condition), maximum cooling rate would be achieved by de-energized solenoid valves that vont the valve operatore.
Required cooling rates can also be achieved during control component failure by positioning the vrilve operatore using the handwheel at the valves.
I Modification: Modification of Penetratione 414, 415, and 416 (BA 418736)
Description of Modification:
The purpose of this modification was to install blank flanges on penetration 414 both inside and outside containment and to replace the existing elbowe with flanged end elbows in penetratione 415 and 416. The flanged end elbown can be removed ts provide accessible openinge for the temporary routing of electrical
- _ wires, cables, and -hoses required to support outage related work. The l modification of penetration 414 also eliminated Leak Rate System and Penetration
! pressurization System piping and valves which are no longer used for Integrated Leak Rate Testing (ILRT).
Safety Evaluation Summary:
The blank flanges outside and inside containment are the new isolation barriere for penetration 414 and eliminate valves LR-V-4, $,.and 6 ao containment loolation valves. The new flanged spool pieces at penetratione 415 and 416 do not degrade their containment inolation barriere. Containment isolation requiremente per the Technical Specifications were met prior to operation following this modification. This modification did not increase the probability of occurrence. or the consequence of an accident or malfunction of equipment important to safety previously evaluated in the SAR since the integrity of pressure boundary and solemic boundary of the RB ILRT system are still maintained, n.erefore, this modification did not adversely af fect nuclear safety
( or safe plant operation.
l' C311-90-2156 Page 20 Attachment 1 Modificationt Control Room (CR) Watch Station Upgrade (BA 418*146) ,
Description of ModificatioD8 This modification upgraded the TMI-1 CR Wateb Station area located in the north end of the CR. This area was redesigned so that the Shift Supervisor, Shift Foreman, and Shift Technical Advisor can ef ficiently and of f atively perform their functions during normal and accident conditions. The Watch Station was redesigned to incorporate new technical equipment such as personal computers, printers, storage devices, and keyboards, in addition to the normal desks, cabinets, and telephones. The Watch Station was also redesigned so that the CR Supervisory team has greater visibility of the operations in the CR by providing the desk area on an 8 inch raised platform. A study area consisting of three (3) cubicles was installed in the CR behind the main panels which is utilized by of f-shift CRos.
Safety Evaluation Summary:
The CR Watch Station raised floor config nation was evaluated to ensure that thw design maintains it integrity during i seismic event. The raised floor was supplied with a seismic brace assembly for approximately one-third of the internal pedestal supports for additional lateral stability. This modification did not impact the accident scer.arios in Chapter 14 o. SAR nor did it adversely nuclear safety or safe plant operations.
Modifications. Fire Service Water System Connection for the North Of fice Building (NOB) (BA 419648)
Description of Modification:
This modification provided the fire service water connection to the NOD. It provided a water supply for the fire suppression systems in this building and surrounding yard area. The scope of this modification included the tie-in to the existing 12" underground fire main and extending a branch line form the new Post-Indicator Valve (PIV) at the fire main to the second new PIV located outside the NOB. Two (1, new hydrants were added to the yard north of the building.
Safety Evaluation Summary:
This modification did not impact the safety function of any affected systems.
This modification did not increase the prcbability of occurrence or the consequences of a malfunction of equipment since the installation of the PIV allowed the isolation of the-branch line in the event of a line break without
' loss of water to the remainder of the fire main, u -
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C311-90-2156 Page 21 Attachment 1 i
) Modification Fire Service (FS), Service Air (SA), and Domestic Water Tie-Ins for the Outage Support Fabrication Shop (OSFS) (BA 414675)
Deecr.iption of Modifications i
This modification provided the FS water, SA and Domestic Water System connections for the OSFS located on the north side of Fuel Oil Storage Tank FO-T-1 within the protected area fence. The FS water connection provided a supply line for the sprinkler system installed in the building. The SA supply line feeds a header within the OSFS for operation of air tools and other compressed air services.
The Domestic Water connection provided a supply line for a safety shower /cye wash station and cold water supply source for fabrication processes. ,
Safety Evaluation Sunparv The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in Chapter 14 of the SAR is not increased. The tie 'n to the fire service yard main to supply the OSFS did not impact the design capability of the yard fire main. This modification did not decrease the margin of safety of the affected systems since the eyetem performance requirements of the FS water system are unaf fected.
Modifications CA-V-2 Fluid Block Piping (CMR-88-074 Revision 1)
Description of Modification:
This uodification cut and capped the previously abandoned Fluid Block (FB) connection on CA-V-2. The FB System was placed out of commincion and the two (2)
FB head tanks were removed and discarded per a previous modification. This modification also provided a potential fluid out leakage path and, in the case of CA-V2, a suspected air in-leakage path. Thus, it was desirable to disconnect the abandoned piping from operational systems and remove any bandoned piping and equipment which could cause Access problems.
Safety Evaluation Summary:
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The elimination of the FB function from the containment isolation valves was evaluated by SE-412464-001. The safety evaluation for this modification had the limited scope of considering the ef fect of cutting and capping the abandoned fluid blocking connection on any of the previously fluid blocked containment isolation valves and the removal of sections of the FB piping. The abandoned FB system piping in NITSt thus, the installation of the new cap and the elimination of the piping sections has no safety significance. There ama no adverse ef fects on safety related systems or equipment-as a result of this modification.
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C311-90-2156 Page 22 Attachment 1 III. Temporary Jumpers / Modifications Modification: MS-V1065 Temporary Donnet Installation (TMM 1)
Descriction of Modification This temporary modification installed a bonnet assembly and new spiral wound gasket in place of the existing bonnet. The purpose of this modification was to provide a leak tight joint between the body and bonnet of MS-V-1065 prior to the Integrated Leak Rate Test (ILRT).
Safety Ev a lu a t ion Sunp_gy:
The temporary installation of the spare non-QC valve bonnet had no affect on nuclear or safe plant operations. The spare valve bonnet assembly was equ 8. valent to the assembiv being removed. Following the Ocupletion of the ILRT, i.av1065 was replaced with a new valve under CMR 88-138.
Modification: Temporary Dypass of Powdex Sample Control Valves co-V-80C&F (TM'i 2 & 5)
Description of Modification:
This modification allowed on-line cation conductivity monitoring of the ef fluent f rom the "C&F" powdex units to continue while replacement valves were procured.
Safety Evaluation Summary:
The system and other components af fected by this temporary modification are classified a3 "Other". Nuclear safety or safe plant operations were not adversely impacted by this temporary modification, The system was returned to its normal configuration.
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Modification: Temporary Modification to FW-V-95A (TMM 4) 01gnietion of Modi fIcatlon:
The purpose of this temporary modification was to cap-off FW-V-95A (OTSG Drain Cooler Drain) to isolate steam leakage. This valve could not be completely closed due to the valve stem being shearea.
Safety Evaluation Summary:
1 This modification had no ef fect on nuclear safety or any adrerse af fect on plant operations. The steam leak was a personnel safety hazard since it was located i
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I C311-90-2156 Page 23 Attachment 1 Saf gt v Evalunt ion Cummat . ICont'di on the 281' olevation of the RD near the elevator. Thus, this modifleation enhanced personnel saf uty. The valvo was restored to its original configuration.
02dificatioD.L Temporary Hodification to Provido Alternate Make-up Flow to Industrial Coolere from the Fire Service Syntom. (THH 6)
Desetintion.ql ModifIcatio.D8 This temporary modification provided continued water make-up to the induutrial coolero (clocod and open loops) while the normal source was isolated.
1 EAlgtv Evaluation Summary:
The equipment / components af fected by thin temporary modification are not cafety related equipment. The potential for lose of spray side of the coolore would not have climinated all RD cooling. The air sido can still maintain RD temperaturo low onough for sovoral hours to permit repairs of normal make-up path. The system was restored to its orig'.nal configuration.
LigAlflpAligni. Temporary Hodification to Fire Sprinkler Pipe (THH 7)
DooerinA Lof ModiflEDALEnl The purpose of this temporary modification was to permit removal of the ono inch firo sprinkler pipo from above TG-RV-5 to support overhaul of this component during the BR outage.
.cfety Evaluation Summarva The af fected system was maintained in service af ter the removal of the oprinkler head. The surrounding heads provided adequato protection for the area. The oprinkler piping in the Turbino Building is classified as "Other." The system was restored to its original configuration.
Modificatient Temporary Installation of a Larger Suction Lino on WT-P 'l to Toet Pump (THH 19) pjtgrlptjon of Hodification:
The purpose of this temporary modification was to install a larger euction lino on pump WT-P-13 for troubleshooting purpocoo.
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C311-90-2156 Page 24 Att actunent 1 g,3fetv Evaluation Summary:
This modification was temporary to increase the auction pipe to WT-P-13 to troubleshoot the reason for the pump losing suction and becoming air bound.
There were no adverse af fects on nuclear plant safety or safe plant operations.
The system was restored l'te original configuration.
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Modification: Temporary Modification to Repair vacuum Leak (TMM 23 t, 30)
Descriotion of Modifie nigf}:
A leak was found in the vacuum breaker / auto vent piping for the Intermediate Waterbox. This modification installed a welded plate in the waterbox and a flange outside the condenser to isolate the leak.
Safety Evaluation Summary:
Installation of this TMM has no af fect on plant safety. If excessive leakage in identified during plant operation either by loss of vacuum or indication of Circulation Water (CW) inlockage, the loop of the CW system will be removed f rom service, if required, and the problem corrected. If CW s y s t er. shutdown is required, vacuum breakers / auto vents are installed on the inlet and outlet waterboxes. The Intermediate Wat6rbox will be vented through these vents and the tubes in the tube bundle. Proper operation or shutdown of the CW system le not offected. This modification is still in effect.
o............................................................................. s tigd i f icat ion : Provide Temporary Connections to Iron Sampling Rig (THM 25)
@seriotion or Modifications the purpose of this temporary modification was to provide conaections to iron sampling rig for CE-2, 4, 6,12, CE-13A-P, CE-15/16, in support of iron traneport study.
Safetv Evaluation Summarva The installation of those sample rige was not impor . to safety. This installation did not interfere with the ability to sample for secondary coolant activity as required by the Technical Specifications and did not impact any SAR evaluation. The system was restored to its original configuration.
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I C311-90-2156 Page 25 Attachment 1
, Hodification: Temporary Modification to Sluice Duratek Media into the Spent Rosin (WDL-T-4) (TMM 26)
Description of Modification The purpose of this temporary modification was to provide connections in order to eluice opent duratek media into the opent resin (WDL-T-4) or used precoat tank. The temporary spool picco replaced the spool piece on the suction of WDL-P-10 located on the south wall in the decant /elurry pump room.
Safety Evaluation Summary:
This change had no ef fect on any nuclear safety related (NSR) component or reduced the margin of nuclear safety on any NSR component. This change did not adversely affect plant operations. The system was restored to its original configuration.
Modification: Temporary Modification to the containment Vessel Integrated Leak Rate Test (ILRT) Air Supply Piping (TMM 40)
Denerietion of Modification:
The purpose of this modification was to temporarily chsnge the containment Vescol ILRT Air Supply piping service pressure. A low pressure relief valve was blanked and a high pressure relief valve was installed on LR-Q-1 to allow the supply tool air at 100-110 poig.
Safety Evaluation Summary:
This temporary modification did not impact safety related equipment nor did it l adversely affect containment integrity. Operating the system during irradiated fuel movements wac acceptable since the piping was maintained greater than 100 psig which provided an adequate ventilation barrier from the possibility of RD l atmosphere leakage through the leakrate oyotem piping. This system wan made a
( permanent installation.
l godification Temporary Modification to Permit Continuous Instrument Air (IA) Supply Downstream of Air Dryer (IA-Q-1) (TMM 41)
Description of Modification:
The purpose of this temporary modification was to permit continuous IA supply downstream of IA-Q-1 during installation of isolation valves IA-V2104 A&B.
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C311-90-2156 Page 26 Attachment 1 Safety Evaluation Summary:
This temporary modification provided continuous IA supply dow- ' roam of IA-Q-1 and did not change the intended f unction of the I A System. Ti. .ttings and home were designed for pressures greater than the operating pressure of the IA System.
The system was restored to its original configuration.
Modifheationt Temporary Modification to Install Gagging Devices on Selected Precoat Devices (TMM 58 through 61)
Description of Modlfications r The purpose of thin temporary modification was to maintain the precoat vessele (FDL-P I Ar.D) operable during a scheduled 1H 480V bue outage. Cagging Javicen were installed on eclected precoat valves to prevent the valven from closing which could have caused loss of precoat.
il DALety Eval.ygtion Summarve This temporary modification did not reduce a margin of safety. The precoat filters are not nuclear safety related and are not required for reactor shutdown.
This modification allowed the precoat filters to remain operable or in standby during a brief bus outage which would have otherwiec chutdown the cleanup of the BWST and/or the RCBT. The system was restored to its original configuration.
1 Modification: Temporary Modification for Hypochlorite Addition to River Water Systems (TMM 67)
Description of Modification The purpose of this temporary modification was to permit blocide application of codium hypochlorite, in place of gaseous chlorine, to treat the river water systems for microbiological control.
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Safety Evaluation Summary:
The river water chlorination system was intended to aid in controlling microbiological growth and subsequent fouling of servlee water heat exchangers.
Hypochlorite -is an equivalent method of treatment compared to chlorine gas; therefore, material compatibility with the new chemical was acceptable and its use did not adversely affect safety related equipment. 'The use of gaseous I
chlorine.has been discontinued
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1 C311-90-2156 Page 27 Attachment i l
Modi f iCA119.D Temporary Jumper for MDCT Control Panel (EJ 3,8,9,12,16,20,65)
Egigrlotion of ModificatioD This modification installed a temporary RTD converter in the Mechanical Draft Cooling Tower (MDCT) control panel nest 4/elot 9. This change allowed data collection for juotification to permanently relocate the River Water (RW) inlet resistance temperature detector (RTD) f rom its previous location on wing wall to the NR piping.
Hafety Evaluat\9D Summarls The RW inlet temperature instrumentation le classified ao NITS. Additionally, test arrangemento did not af f ect the control room station and unit T indication.
The system was restored to its original configteration.
Modification: Temporary Installation of a Manual Transfer Pushbutton ac d Lif ted lead for Inverter 1E (EJ 13 and 50; LL 27 and 28)
Egge d ilon of Modificat19D A momentary pushbutton (normally open) was installed betwoun wire 301 on terminal L1 of relay RLB and terminal 7 of relay RL10 wire 302. Additionally, this modification lif ted the 301 wire on terminal L1 of relay RL8. The purpose of this modification in to provide e, safe meano to retransfer inverter IE back to ATB from TRB, and to prevent the overheating of resistor R2 on static switch seneing board V082D4.
Safety Evaluation Summary:
Installation of the manual transfer pushbutton on Inverter IE will not prevent it f rom transferring to the alternate source (i.e., TRB) in the event of inverter f ailure or overcurrent/undervoltage condition. The Control Room will be notified that the inverter has transferred to the alternate cource by alarm H-1-4 (ICS Power Trannfor). The margin of plant safety is not compromleed. Inverter IE does not support a vital bus. This modification in still in effect and in to be completed during a planned outage.
tiod i f ic at ioD Temporary Jumper for Testing of a Spare MW Traneducer (EJ 15)
Eeocription of Modification:
1 This temporary modification permitted testing of a spare MW transducer to j determine the cource of spiking of the MW signal.
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- C311-90-21$6 Pogs 28 Attachment 1 Safety Evaluntion Summarva This jumper posed no safety concerns for plant operation. ICS-auto power was I l protected via the 1/4 amp fuse and CA breaker. If CA-auto power was lost, it I
would not have disturbed normal plant operations. The system was restored to its original configuratLon.
- tig.d i f icat ion
- Provide Temporary Power to Distribution Panel D-22 (EJ 19)
Description of Modifications s.
The purpose of this modification was to provide temporary power to distribution panel D-2? during the outage of ATB to repair 1E inverter static switch.
Safety Evaluation summarve o
The plant was in a cold shutdown condition during this modification. Temporary
- loss of the EmergMncy Notification System (ENS) during installation and removal of this modification did not af fect plant safety. The NRC was notified prior to the loss of the ENS phone system. - The system was restored to its original configuration.
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Modification: Temporary. Jumper for HSPS Testing (EJ 32)
Description of Modification The purpose of this temporary jumper was to allow train A.& D modification and
- subsequent loss of power testing of the HSPS. The jumper maintal' ed control of the MS V13A/B valves f rom the control Room and prevented automatic actuation f rom the HSPS.
gAfety Evaluation Summary:
The HSPS is required to be opt.rable prior to criticality. The temporary jumper was removed and tested prior to criticality. The jumper maintained control Room operability of MS-V13A/B; if the junper has f ailed .the automatic actuation would be restored. The system was restored to its original configuration.
.............................................................................t Modification: Temporary Jumper for Testing of CRD Transfer Switches (EJ 46 and.47)
Description of Modifications i
The purpose of this temporary modification was' to permit-CPD Transfer'Swl.tch y=mer w rr$*vry vd=mmp.g.--e,i-7,,,..vmy,.,4,,,-rw,,-9-eg,,,9 -y,>.e g#9gr. wy ..,,9-,g-y r-.g,w. .,,g,e,,y_ 9% .e.,yg w p,, ar.g., y, 9 9,y-o wwy9 y-y.--3,.-pr-,w e; ,yp,-,5p-4y9w
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I C311-90-2156 Page 29 Attachment 1 i l
pecerlotion of _tinglifiention (Cont 'd,11 tooting without motor power. CRD Trancfor Switches can be operated with the normal and at.xiliary power supplice of f.
l S a f etv... Eyg.lgt ion Summarv e i
The plant was in a cold shutdown condition during this modification. The normal and auxiliary power supplies were doenergized, only control power was used for thin tost. CRD stators were not energitod by the testing. Failuro of the jumper would have only provented the relay from being energized which would have provented any .further testing. The uyatum was rostored to Ato original configuratlon.
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ligs11LIEa11QDe Temporary Jumpor to Maintain Power to RMA-14 (EJ 73) j QcagriptLon o[Jiodification The purposo-of thin temporary jumpor was te ntain power to RMA-14 when the ICESVHCC was removod from nervice for prover' .vo maintenanco. This modification allowed continued compliance with Technical Specification tablo 3.21-2, item 6.
higly_Evalua_tton Summaty:
RMA 14 was poworod from a temporary cource when the ICESVHnC was removed irom service for proventive maintenance. The foodor breaker was tagged to provent inadvertent doonorgization. The above actions allowed the plant to be maintained in a normal plant statun and not degrado safe operation. The ayatom was reatored l
to its original configuration.
l Modification Lif ted Lead to provide Electrical Isolation from Excitor Field L Breaker (LL 6 & 7) p.g.scrintion of Mqdificat19D8
'. The purpcae of this temporary modification was to provide electrical Apolation from excitor field breaker during the BR outago.and still permit testing of the field breaker.
.hinty Evaluation Summary:
The gonorator excitation system, which includeo the excitor'and fiold breaker, is not a safety related system. This tempor ary modification was unod only during plant shutdown conditione. The syntom was restored to its original configuration.
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C311-90-2156 Page 30 Attachment 1 Modifications Lifted Leads for Feedwater (FW) Valves (LL 112-116, 118-127)
Descriotion of Modification:
The purpose of this modification was to prevent the cloning of IV v31ves 5A, SD, 92A, and 92D on an actuation from the HSPS during construction and testing.
Safety Evaluation Summary:
The HSPS is required to be operable prior to criticality. The lif ted leads were roterminated and tested prior to criticality. Failure of the lif ted leads would have prevented automatic closays of the above FW valves during hot shutdown conditions. The system was restored to it.e original configuration.
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j C311-90-2156 Page 31 Attachment 2 Activities For Which A Safety Evaluation Was Perf ormed but Did Not Impact SAR Systems / Components Modification: Installation of Elapsed Time Meters (BA 123220 (123240))
Description of Modification:
This modification involves the installation of elapsed time meters on various Balance of Plant (SOP) 4160V switchgear motors.
Safety Evaluation Summary:
The components affected by this modification do not perform a nuclear safety related function. The installation of the elapsed time meters will allow the Maintenance Department to accurately track equipment run time. Thus, the scheduling of Preventive Maintenance (PM) and balancing of equipment operating time is enhanced by this modification. Since the elapsed time meters only track time equipment run time, their failure to operate will not adversely affect equipment operation.
lioillfigation Use of B&W Rolled an4 Ribbed Plugs in TMI-l OTSGs (BA 123253)
Description of Modifications The purpose of this tube plugging was to remove from service certain THI-l OTSG tubes using B&W rolled and ribbed plugs. The plugs are utilized in defective tubes to establish an adequately leak tight pressure boundary between the OTSG l . primary and secondary side.
j Safety Evaluation Summary:
The plug qualification test results demonstrated acceptable primary to secondary leak rates and plug retention capability under normal as well as postulated accident conditions. A pressure test of greater than 10,000 psi performed on both ~ ribbed and rolled plugs indicated no significant plug movement. This demonstrated that a plug is unlikely to be ejected under any adverse conditions since 10,000 poi pressure is greater than the calculated tube rupture pressure
- l. for a nominal thickness tube.
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l Since the plugs meet the same criteria as the remainder of the RCS pressure boundary, the probability of any analyzed accident occurring has not been increased. The design maximum allowable primary to secondary leakage, if assumed to occur for all current and installed plugs, is a small fraction of the Technical Specification limit on total leakage through the OTSG tubes during
C311-90-2156 Paga 32 Attachment 2 Safety Evaluation Summ_Arv (Cont'd):
plant operation. Additionally, since the plugs are designed to maintain their integrity under normal, transient, and accident conditions to the same criteria as the rest of the RCS, the margin of safety associated with the premeure boundary le unchanged.
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Modification: Mini-Mod for Support of Pneumanic Tubing (BA 128086/WA 55?3-51130)
Descriptien of Modification:
The purpose of this modification was to upgrade the support configuration of the instrument and control tubing in the control Building HVAC System located in the A and B Ventilation Equipment rooms.
Safety Evalyation Summarvi This modification resulted in improved rullability of the pneumatic copper tubing and provided a verification that the support configuration conforms with the functional and solemic requiremente of the associated instrumente and controle.
The tubing support attachment upgrado does not directly af feet system operations or performance except for the corresponding increano in the reliability of the associated instrumente and controle. No changes to the existing instrumente or control devices were required.
Mg_dt Lf i cat ion : Makeup (MU) Valves MU-V-11A/B Operator Supporte (BA 128086/WA 51137310)
Eescription of Modification:
The purpose of this modification was to install supporte for the valve operatore for pneumatic-operated valves MU-V-11A/B. The supporte eliminate the extra loading placed on the valve stem by the weight of the operator, enabling the stem to travel without added resistance. This modification allows more reliable and 1recise operation of the valves.
Sal 9tv Evaluation Summary:
The ' unction of valves MU-V-11A/B is to control the flow path of the MU System flow through the filtere MU-F-1A/B to tank MU-Y-1 in order to control system chemittry and inventory. This modification improves the reliability and control of vals es MU-V-11A/B by reducing the resistance to stem movement. The margin of safety or the af fected system was not reduced since the innetion of valves MU-V-11A/B was t ot adversely af fected. Nuclear safe &y and safe plant operations were
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C311-90-2156 Page 33 Attachment 2 Safety Evaluation Summarv (Cont'dit not adversely affected by this modification since the function of the system as described in the FSAR was not impacted.
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Modification HU-V-36 & 37 Controle Relocation (BA 128087/WA 51142310)
Description of Meditlention:
Following an ES actuation, Makeup pump (HU) Recirculation valves MU-V-36 & 37 go closed. To reopen these valves, the operator had to go behind the console to the panel switches. Failure to open these valves could result in damaging the running HU pumps. In order to improve operator response and reduce the possibility of HU pump failure, the control switches for Hu-V-36 & 37 were replaced and relocated to the control Room console center.
Safety Evaluation Summarvs Nuclear safety was not reduced by this modification since the control logic of the valves was not changed. Safe plant operation is assured since the operator is able to more quickly respond to opening the affected valves as required af ter ESAS actuation. Relocation of the control switches does not change the existing ESAS logic for these valves. These valves will still close on ES actuation and are still operable from the Remote Shutdown panel.
'Modificationi. Inverter Radio Frequency Interference (RFI) Choke Bypass Installation (BA 128087/WA 51143310)
Descriotion of Modification
. This - modification added wire jumpers to inverters 1A-E to bypass a RFI inductors (RFI chokes) to improve the voltage at the output of the inverters.
The chokes were determined to not be required and caused an undesirable voltage drop.
Safety Evaluation Summary:
Since the RFI choke bypass -installation did not reduce the ability of the inverters to supply reliable power to safety related loads, no margin of safety defined in the SAR or Technical Specifications was reduced. The RFI choke bypass installation did not reduce the ability of the inverters to supply reliable power to safety related loads; therefore, there was no potential for nuclear safety or safe plant operations to be adversely affected.
C311-90-2156 Page 34 Attacheent 2 Modification: Emergency Diesel Generator (EDG) Up-to-Frequency Relay Replacement and Air Start Valve Test Switches (BA 128087/WA 51150310)
Descript_ ion of ModifiC Sil2D:
To increase EDG reliability, the obsolete Westinghoaco type CF 1 electromechanical up-to-f requency relays were replaced with Brown-Boveri type 81 solid-state relays. Additionally, a test switch was 1.netalled to allow eeparate tenting of the air start valves.
Sef etv.Inluation Sucma_n:
The replacement of the up-to-frequency relays did not alter the deigned operation of the EEGs as described in the GAR. The air start valve test switches are administrative 1y controlled and are used cally periodically to test the opening of the air start vslves. This modification doon not adverooly nue! ar safety or safe plant operations since it elimina"e the contact burning problem and provides a means for testing air start va.<en.
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tiodi f igAtlp.M Test Lights for ES Testing (BA 128087/WA 51152310) pagpq; jot ton of _Modificatign IE Notice 88-03 addressed the potential for inadequate testing of relay concacto during sa t'e ty related logic system functional testing. A review of ESAS surveillance procedures revealed that these were sore relay contacts not specifict.lly addressed by Technical Specification surveillance requirements which were either partially tested or not tested. Only one 2 cat of 3 ES contact logic combination for the 480V bus IP and 1S Auto-Start lockout schema was tested in a refueling interval by the Engineered Safeguards (ES) System Emergency Sequence and Power Transfer tost. This modification connected an amber tout light in eerles with the normally open ES auto start lockout contact at bue IP and 1S.
For ES testing, this light is illuminated when the ES contact is closed.
Previously, the ES blocking contact to the stop pushbuttons on the skid-mounted control panel was not included in a surveillance procedure. The contact 10 currently toeted by a surveillance procedure and, in order to verify contact operation, an amber test light was connected in series with the normally closed contact.
Safety Evaluation Summary:
The margin of safety as defined in the bacio for any Technical Specification was not reduced since these test lights did not alter the designod normal or emergency operation of their respective circuits. Nuclear safety was not adversely affected because these test lights do not perform a safety related
- s .e a_.__m w_ . - - 'l
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C311-90-2156 page 35 Attachment 2 Safety Evaluation Summarv (Cont'd):
function and are passive elements in their ascociated circuits. Safe plant i operations were not adversely affected because these test lights provide verification of ES contact operation during ES testing.
ese.................................e**e .....................................
Modificatient Vertical Bus Bars (B/B) Support Upgrade (BA 128087/WA
$1157310)
Egserietion of Modificat12DI The Appendix R CAI Electrical System Studies Report Vol.1, G/C 2734 had revealed that the short circuit currents on the ES motor control centers (HCC) 1A-ES, IB-ES, IA-SHES, and IB-SHES could exceed the short circuit rating of their vertical B/B. The fault duty rating of these B/B is 22,000A (symmetrical), while the fault current in a worst case scenario could be 24,318A (symmetrical). As a result pSC 87-013 was issued to evaluate the impact on an interim plant operation. The pSC was closed out per JCO-5350-88-310, and LAI-89-9038 was issued to correct this problem. This modification was designed to correct this problem. This modification installed additional B/B insulating support 1 on the vertical buses of these MCC and rearranged the spacing of the insulators, such that these buses are able to withstand up to 42,000A (symmetrical) short circuit currents.
L Fafety Evaluation Summnry:
The safety function of the MCCs is to supply the ES loads duritig a Design Basis Accident (DB A) . This f unction was unaf fected by this modification since the size l of the MCC buses and their normal current carrying capacity were not alt 6ted.
The additional insulators were seismically mounted and improved the electracal
(
- capability of the Mcce to handle electrical fault currents of higher magnitude
- than before. Thus, this modification enhanced the ability of the MCCo to maintain the electrical power to the ES loads.
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l Modificatient OTSG Main Feedwater (MFW) Nottle Replacement (BA 120135) j' Description _of Modification:
Main Feedwater (MFW) nozzle plates, made of carbon steel, at other B&W plants i
have experienced severe corrosion which affects the OTSG performance and could
( impose risk of damaging the OTSG internal components. Visual inspection of the l THI-l MFW nozzles also found indication of erosion. The workscope of this modification was to replace the MFW nozzle plates with B&W new design nozzle head which improves the reliability of the MFW nozzles.
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C311-90-2156 page 36 Attachment 2 Safety Evaluation Summar,y I
Replacerrunt of the MrW nozzle spray plate with Inconel 600 material and the new crafig' ration which increases the size and reduces the number of holes dr.es not impair the integrity of the nozzle. The new design nozzle improves the cort sion and erosion life of this component. The margin of safety as defined in the basis of the Technical Specifications is not reduced since this modification is considered a replacement in kind which improves the life of the MrW nozzle and dc:,s not af fect the performance and safety function of the OTSC MrW System This activity improves the feedwater injection momentum which, in turn, allovs a higher maxituum level and increases the operational margin of safety.
lipdif ication: Containment Hydrogen Monitoring System - Valve Relocatic" (BA 128140)
Description of Modification:
Test connections (valvec HM-VOSA and HM-V08A) were not readil; accessible for Inservice testing (IST) on inner containment isolation valves HM VO3A and HH -
V04A. Thir modification rerouted the test cone.ections to a location accessible f rom platforms that were installed for the int jection cf containment ventilation system purge air valve AH-Vic.
j Safety Evaluation Summuyt l
This modification consisted sf re-routing inaet containment IST test connections for the Cont ai n...e nt Hydrogen Monitoring System. The work was performed in accordance with original design codes a 's classifications and with ident i.ai material specifications. This modification did not decrease the margin of safety as described in the applicable Technical Specifications since system functions were unchanged and containment leak tightness was maintained. The modification did not increase the probability of occurrence or consequence of .:
malfunction of equipment important to safety because the re-routing of the test connections did not affect the safety W 'I any systems.
li2sl[U cat ion: Pressurizer Heater Connector Repair (BA 128141) p uerietion of Modificatign:
Due to pressurizer heater connector failures, it was necessary to replace the failed connectors. The new connectors are a_ permanent connection as opposed to the previous removable type connector.
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t C311-90-2156 Page 37 Attachment 2 HA(ety Evaluation Summary:
The pressurizer heaters serve ro efety function; therefore, this modificatior.
did not increase the probabitity an accident or malfunction of safety related equipment. The new connectors did not altar the designed operation of the pressurizer heaters. Sinct the pressurizer heaters perform no safety related function noi have any ef fect on the operation of safety related equipment, this modification does not adversely impact nuclear safety nor reduce any margin of safety.
Modific- 01: Insulation of Industrial Cooler Systems Piping in the RB (Job Crder 16581)
Descriotion of Modification:
The purpone of this modification was to install insulation on a porcion of the Industrial Cooler system piping in the RB in order to eliminate problems caused
- y pipe swer. ting. The insulation reducea the sipe sweating and the resultant pipe coating degradation and ex t e rJ
- >r corrori u. Thus, this modi. cation minimize., pipe repainting requirements and extenas component servico Itfe.
Safetv Ryaluation Sumparvt The petformance of the Industrial cooler system was unaffr ted by this change since no system configt. ration or operational changes occarren Pio modification did not increase the probability of occurrence or conseque.ces of.an accident since the seismic classification of the Industrial cooler system was ur _f fected, and since the insulation was designed and installed to remain attached to piping i during solamic events.
Modification: Plant Process Computer Reactimeter (BA 412281) i p;,scriotion of Modification l
As cart of the overall plant process computer upgrade and replacement i modifications, the Mod Comp computer with reactimeter was removed. This i'
l modificatica provided installed test connection points and a mobile consolo for rountlig test equipment. The test equipment is used temporarily follcwing
.ref ueling and is connected to the plant via temporary plug-in test cables.
Safety , Evaluation Sumcarva I T; Plant Computer System is not safe shutdown related/ therefore, it does not havn a safety function. Th; Plant Computer System is used by plant operators for alarm monitoring, performance monitoring CRT display, and data logging l
C311-90-2156 Page 38 Attachment 2 Safety Evaluation Summalv icont'd):
activities. This modification did not alter the safety features of any af fected systems. The interface of this modification with the Nuclear Instrumentation System was evaluated in FMEA-TI-602-001 and found to have no impact on safety related functions. Therefore, this modification did not increase the probability of occurrence or consequence of a malfunction of equipment important to ,,.ifety.
This modification did not decrease the margin of safety as defined in the bases of the Technical Specifications because the work did not impact any systems' safety functions.
LigdMication Intermediate Closed Cooling (ICC) Pump Control Circuit (IC FS) (BA 412511)
Descriqtil on of Modification:
-This modification installed time dela Jiays to allow sufficient time for the standby ICC purp to re-establish ICC flow before rending a loss of ICC flow signal in the RC pump trip circuit. Time delay relays were also added to the IC-P-1A/B pump control circuits to prevent tripping of the pump when the indicating flags were matched with the control switch following a pump auto start.
Safety Evaluation Summary:
Tre function of the ICC system is to provide cooling water for various plant components (e.g., letdown coolers, RC drain tank cooler, control rod drive c;ooling coils) . The installation of this modification improva the ability of tha ICC system to provide ICC flow to . system components and eliminate an erraneous trip signal for the RC pumps. The ICC system is not an Engineered L Se futy Actuation System (ESAS) . The only safety consideration of the ICC is that its deactor Building isolation valvos close as tequired on an ELAS signal. This modittcation does not affect the ICC containment isolation valves or the remote shutdoen function of the ICC pumps (i.e., IC-P-1A and IC-P-1B). Safo plant operation is enhar.ced by assuring aute start of the standby ICC pump and eliminating the possibility of a false ICC loss of flow-signal to the RC pump trip.logie Modification: Computer Alarms for Electro H,,raulic Control (
(BA 412512/WA 34512010) l Description of Modific'*lon:
l This modification installed a new digital computer point to provide the control room with an audible computer alarm when an EFC electrical malfur..: tion signal is generated.
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C311-90-2156 Page 39 Attachment 2 Safety Evaluition Summarva The safety function of the plant computer is to display and log data as required by R.G. 1.97. This modification affects the plant computer system by adding a digital alarm input to log. The EHC has no safety related function. The implementation of this modification does not adversely af fect nuclear safety or safe plant operations. The current and new alarms do not interlock with any safety related equipment.
Modification: Main Feedwater Pump (MFWP) Bearing Oil Pressure Trip to Plant computer (BA 412512/WA 34512011)
Dragpf_intion of Modification:
Safety Performance and Improvement Program (SPIP) Recommendation TR-014-MFW addressed the installation of a monitoring system for the MFWP trips. Per the referenced recommendation, this modification connected the MFWP bearir.g oil pressure trip to the plant computer.
pafety Evaluation Summary:
The margin of safety as defined in the basis for any Technical Specification was not reduced since the plant computer and the MFWPs are not in the basis of any Technical Specification. Safe plant operation was not affected since a set of spare contacts was used. The setpoint of the pressure switch was not changed by this modification. The pressure switches connected to the computer do not interlock with any safety related equioment that is addressed in Chapter 14 of the SAR.
Modificationt Class lE Diesel Genera. Synchronous Check Helay (BA 412512/WA 34512012)
Description of-Modificati.on:
.The B&W Owners Group Safety and Pertormance Improvement Program determined that in order to reduce the risk of damage to the Claus 1E diesel generators due to operator error, synchronism check relays e%uld be installed. These relays would prevent an interlock in the EG-Y-1A and IB Class IE diesel. generator breaker l
close circuits, which would automatically prevent breaker closing in.an out-of-
! syn.: condition. This modification installed two (2) Brown Boveri Type 25S sync-check relays in control room panel CR.
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C311-90-2156 Page 40 Attachment 2 Sstfety Evaluatiga_Sunmary:
The margin of safety as defined in the basis for any Technical Specification was not reduced because those relays oliminated the potential risk of damaga to EDCs due to operator error when synchronizing and closing generator breakers. The probability of occurrence or the conscquences of an accident or malfunction of equipment important to safety previously evaluated in the SAR was not increased because the relays are seismically qualified and mounted, and existit.g wiring was used to connect the permissive contacts to the manual control circuit of the breakers.
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Modification: Vent Valve Installation for Testing Air Accumulators (BA 412512/WA 34512013)
Descriotion of Modification:
The scope of this modificativ., was to install a vont valvo on the air supply lino between the check asc_ isolation valves for those air operated valves that aro required to be tested. The Instrument Air (IA) linos for the follt 7 valves were modified to allow accumulator testing: IC-V3,4; IC-V6; and MU-V20. The accumulators for HD-V4, FW-V7A/B, and MS-V4A/B were also tested; however, no modifications were required for theco valves.
Safety Evaluation Sum (pan:
The losa of IA would not adversely af fect any safety related cystem, because all tho safety related air operated valves would-fail in a e.:o position or have accumulators that would maintain the valve in a safe position. The accumulator and associated tubing, required to . provide containment isolation, is safety related. This modification-installed vent valves which allow testing of tnese components. This modification was outside the safety related boundary. Fai ure of the modified tubing, thereforo, would not have increased the probability of occurrence or the consequences of an accident previously evaluated in the SAR.
Modification: Reactor Coolant Pump (RCP) Lube Oil System Upgrado (BA 412512/WA 34512014)
Descriptton of_Hggifleat1o3:
The RCP motor - lube oil system experienced oil 104q9 and dif ficulties when filling the oil accervoire (backup and spillage of on, skow fill rato, and lack of oaoy-to-read normal level reference marks). As a result , the following modificationc were performed:
i I
C311-90-2156 Pago 41 Attachment 2 p_e s_gIipt i on o f &<df.ip_,a t ion (Cont'd):
o Upgraded the piping / tubing to welded joints as much as posolble to minimize leakago, o Redesigned the upper and lower reservoir local fill stations for easier and more rapid filling capability during refueling outages for oil addition capability during forced shutdowns, o Installed rumoto fill statione for on-line oil addition capability during plant operation.
o Replaced existing oil levol night gaugno with ones that are eaolor to read.
o Replaesd level owitches with now upgraded design.
[ o Added capped drain on RCP oil shield drain tank.
o Installed new high pronoure hosos for hhe lube oil lift piping.
o Enlarged / relocated oil drip pans where required as a result of piping / tubing modificationn.
Safety Evaluation Summary:
The purpose of this modification was to improve the RCP .notor lube oil system reliability and to avoid, to the extent possible, plant trJpo and transient conditions that could be initiated by excessive oil leakage. This modification did not af feet the safety function of any associated systemo; thereforo, nuclear safetv cr safe plant operations were not adversely af fected. The solomic design of the lubo oil oystem and the ability to collect c.nd contain any leakage via coismically designed drip pano and leakage collection piping were maintained.
P.odi f icat ion: Int,tallation of Letdown Piping Shielding (BA 412528/WA 30121310)
Doncrintion of Modification-The purpo90 of thin modification was to provide permanent ahleiding around the Reactor Coolant Letdown Piping above the Makeup (MU) Filter 2A/B cubicle at elevation 281' of the TMI-1 Fuel Handling Building (FHT) . This chielding reduces personnel exposure during filter changeout operations which require maintenance personnel to work close to the piping.
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C311-90-2156 Page 42 Attachment 2 Safety Evaluation Summarvt The added support structure and lead shielding added a relatively small mass to the reinforced concrete wall. Attachment of the support structure to the wall
= utilizing concrete expansion anchors did not degrade the structural integrity of the wall. The-support structure and lead shielding have no safety function and are not required to function during or af ter an SSE, but are required to maintain passive integrity. Therefore, they are classified as Seismic class II.
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Modification Installation of Fire Damper FD-66 Access Ladder
--(BA 412528/WA 30162310)
Descriotion of Modifica(1 par-This modification provided a permanent means of access to PD-66 located in the MU-P-1C cubicle. .This access improves accese to FD-66 for purposes of surveillance /mai.1tenance activities and reduces exposure to potential safety hazards by-providing permanent provisions for climbing up to the damper.
Safety Evaluation' Summary:
Th'e. ladder.and . supports have no nuclear safety function' and are not : required to
' function during.or after a safe-Shutdown' Earthquake,.but;they are required to c maintain passive integrity. Therefore, they are classified as Seismic class II,~
anti-falldown. This modification does not impact nuclear safety or safe plant operations..
g Modifications. Installation.of Protective Covers for Cable Trays 4',1 iJ 496;-
E (BAL412528/WA-_31093310) _
Description of Modifications i.
-This modification provided physical' protection for the pressurizor' heater power b cables'~ located : near . the heater- bundles, elevation 312'-7".
' configuration exposed the cables to potential damage and. general wear-and-tear The . previous during maintenance.and inspection-traffic around the pressurizer.
Safety Evaluation Summary --
IThis : modification did--~not _ directly af fect . any plint . systems, subsystems,. or
- components.- It -does indirectly af fec t the RCS- in that the> power cables- to' the-preeluriter heaters receive 1the benefit of_ph;sical protection and-therefore an-increase in reliability. The tray cover 0 and aupports have no safety- function land are not' required . to function during or af ter t'.e SSE, but are required to
C311-90-2156 Page 43 Attachment 2 Safety Evaluation Summary (Cont'd1 maintain pasuive integrity. Therefore, they are classified as Seismic clasu II Anti-Palldown.
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Modification Pump Pecirculation Flow Meters for Inservice Testing (BA 412531)
Description of Modification:
The purpose of this modification was to provide diesel fuel local flow measurement for pumps, DF-P-1A-D in the Emergency Diesel Generat^r (EDG) fuel system (phase 1) and to install in-line flow meters for tc.e Loric Acid Recycle Pumps, WDL-P-13A/B, and the Boric Acj d Pumps CA-P-1A/B (phase 2). This mc31fication resolved a commitment to the NRC regarding in-service testing requirements for the diesel fuel transfer pumps.
Safety Evaluation Summary 1 The installation of the local flow elemente and flow indicators did not affect the performance of the EDG fuel system as it only provided indication and had no control function. The flow indicators and elemente provide local flow indications to the IST program requirements in accordance with the ASME Boller and Pressure Vessel Section XI code. This modification did not adversely af fect nuclear safety or safe plant operations, or decrease a margin of safety since operation and function of the diesel generators was not affected.
hpdificationt Reactor Building Maintenance Platforms (BA 412540)
Description of Modification:
This modification installed platforms at the Reactor Coolant Pumps RCP-1B&C and at the Reactor Coolant Drain Tank (RCDT) to provide safe, permanent vark areas of sufficient size to support maintenance activities for these components.
Safety Evaluation Summarvt This modification affected the secondary shield wal's and their attached compartment walls. No other systems, sub-systems, structures, or ec.7oncato were directly af fected by this u.adification. The additional load of tne platforma plus equipment and personnel Inada on the secondary shield wall and the RB structure were analyzed and found to be acceptable. The platforms and their supports are passive during operations of the plant and did not inhibit the ability to achieve containment integrity. The additional structure loading due I
C311-90-2156 Page 44 Attachment 2 Einfety Evalugtign Sumn1gry (Cont 'd):
to these platforma did not reduce the margin of safety of these supporting structures.
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]igdificat[pa OTSG A&B Upper Manway and AH-Vic Platforms (BA 412546)
Qg.pgrjItton of Modifications l
This modification installed platforme at the AH-V'.0 'rge valvo and extensions to the existing platforme for the OTSG uppnr mar. ways to provide cafo, permanent work areas of sufficient size to support maintet.ance activities for these compononte.
Safety Evciluat ion Summary _L This modification affected the secondar" shield wailt: snd their attached compartment walle. No other r ,e , sub-M + 0, Art: tut,ed, or components woro directly affected by this mo..fatawion. The additional load of the platforms plus equipment and personnel 1 ,ae. on the secondary chield wall and the RB st ru.:tt.re were analyzed and found to be acceptable. 'fhe plat forme and their supporto are pansivo during operations of the plant and did not inhibit the ability to achieve containment ' -tegrity. The additional structuro loading due to these platforme did not red.co the margin of safety of those supporting structe e0.
- eese......eeeeeeeeeee.e............e.eeeee********ee .e.e..ee.eee.eeen.eene Modification: OTSG Skirt Manway Access Enlargement (BA 412562) pfocrinkion of Modification 1_
This modification enlarged a portion of the circular access opening to the OTSG lower manway to improve personnel and material access to the lower head for maintenance an'! Mir operatione. This also added four (4) sets of four (4) small holen in hr r tirt near the access opening to allow the temporary mounting of radiation chie .e for use during non-power operations.
SiAffLY.J,yyJJyat ion Supmgry:
The proposed modification has no ef fect on the safety function of the OTSG eince there is no impact on the ability of the component to provido a finnion product barr:or. Since the modification was wholly external to the OTSO, the nyetem prformanre was unaffected. This modification did not introduce any new I operating mode or subject existing componento to new operatiig challengen outoldo l
C311-90-2156 Page 45 Attachment 2 i l
S,gfetv Evaluation Summarv iCont'd):
of the original operating, safety, and design baeos. No now unanalyzed accidents or malfunctions were introduced, nor made either more probable or of greater consequence.
.o..........................a..........................*......................
Mod i f icat12D: Auxiliary Fuel Handling Bridge (M28) Fiuipment Upgrades (BA 412564)
Descriotion of Modification:
The purpose of those modifications were tot o Install an Inching Motor with the control contor mounted on the auxiliary bridge tower.
o Install programmable geared limit switches with the control console mounted on the AFHB tower.
o Install-fixed lighting on the main and auxiliary bridgen to assist in reading the ZZ position tapon and position scribe marks on the bridge operating f.loor next to the ralle.
o Install a reel type manually operated camera cable otorage assembly permanently mounted on the auxiliary bridge trolley.
o Rewire the interlock statun lights on the main bridge control concolo.
Pact bridge operations had indicated that the lights were working ir: prope rly.
Safety Evaluation Su,tnmary:
The fuel handling system in designed to provido a safo, offectivo means of storing and bandling fuel f rom the timo it reaches the station in an unirradiated condition until it. leaves the station af ter post-irradiation cooling. The syntem in dealgned to minimize the possibility of mishandling or improper operations that could cause fuel' assembly damage, patential finnion product rolseac, o r.
both. These modificatione did not increase the probability of occurrence or conocquence.of a malfunction of " Regulatory Required" equipment. The improved =
upgradou increase fuel handling equipment reliability while decreasing equipment maintenanco. These modificatione did not creato a possibility for an accident or malfunction of a different type than any previously evaluated identiflod in the SAR because operating modos and system control woro unaffected.
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C311-90-2156 Page 46 Attachment 2 Modification: Main Feedwater (MFW) Flow Element Replacement
-(BA 412565)
Descriction of Modification:
This modification replaced the MFW flow elements (SP-3A/B-FE) with new flow elements to increase the accuracy of flow mecouremer.t and to decrease the engradation of measurement accuracy. This modification prceludes the necessity for unnecessary reductions in plant performance caused by artificially elevated estimations of reactor power output.
p_gfety Evaluation Summary:
Following replacement of the ventur18, during power escalation, flow testing was performed to cor firm the accuracy of the signale developed by the newly installed venturis (see oithium Tracer Testing Activity under Section I, " Tests and Experiments"). Nuclear safety or safe plant operations were not advertely af fected by this modification since the safety functions of the af fected systems were not impacted. This modification did not increase the probability of occurrence or consequence of an accident previously evaluated in the SAR. This determination was based on a review of the SAR with particular emphasis on Sections 5.1, 5.4, 10. 5, and 14. Replacement of the flow element did not af fect the function of the FW oystem.
Modification: Control Room Noise Reduction (bA 412566)
Description of Modification:
This modification installed a control room global silence button in order to reduce control room noise during a major plant transient and improve the ability of the Control Room Operator (CRO) to communicate and respond to a transient.
A chime system was also installed to annunciate when any main control room alarm is reset along with a reflash logic system for the RMS interlock' switches. The refiash logic system resulted in the alarm window remaining energized when any
( RMS interlock switch is in defeat and will alarm again if any other interlock switched are placed in defeat.
Safety Evaluation Summarvt.
The safety function of the annunciator system is to alert the CRO to the status of safety related equipment and/or the potential of exceeding pre-determined limits; A global s'.unce pushbutton was installed that silences all annunciator alarms for a pre-determined time when the pushbutton is activated. Eliminating l distracting annunciator noise during the first-minutes of a major plant transient l significantly improves the CRO's ability to communicate and respond to the-l . transient. The global silence button could introduce the risk of missing an l
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-C311-90-2156 Page 47 Attachment 2 Safety Evaluation Summary (Cont'd):
alarm. However, the button will silence the alarms only for a pre-determined time period and the administrative controls governing the use of the button during post-trip conditions will ef fectively limit this risk. Using the button during a major plant transient outweighs the potential of missi.ng an alarm. The operator can determine what alarms are in during and after global silence time frame by visual inspection of flashing alarm w!.rdows. The chimo system will allow the operator to audibly distinguish between when an alarm is received and when.it is reset.
The accidents analyzed in Chapter 14 of the SAR do not take credit for the
.cnnunciator or computer systems. Credit is taken for RMS actuated isolation of release paths to mitigste the consequences of an accident. The changes to the RMS inter' ck defeat alarm did not af fect the ability of the RMS to shutdown any equipment used for isolating a release path since work was only performed on the alarm circuite. The global silence pushbutton and associated relays do not interlock with any safety related equipment that would be used to mitigate or monitor _the consequences of an accident as described in Chapter 14 of the SAR.
.....**........****e....e**....******e.**e....***......... **ee...............
}igdi f icat ion: Replacement of Emergency Feedwater (EFW) Flow Elements (BA 412571)
Descrintion of Modification:
Measurement of EFW flow was previously accomplished with type ANR-76-C21-CSS annubars . in both the A and B loop. Flow indication from the annubare was-considered unreliable due to suspected air entrainment in the sensing lines.
This modification replaced the annubars with flow venturin of proven design whi '
provide a 500 inch aP at 600 GPM and 680 inch aP at 700 GPM which is maximum venturi design flow.
l Eatety Evaluation Summarve l
The margin o* eafety defined in Licensing Baals Documente was not reduced as a result of this modification. System operating parameters and design were not changed e : cept for replacement of the flow element which improves accuracy of flow sensing and indication. Increased system aP does not af fect tha capability of the system to_ deliver required flow under all conditions. Nuclear ea.fety or l- safe plant operations were not adversely affected by this modification because the safety functions of the affected systems were not impacted.
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C311-90-2155 Page 48 Attachment 2 Modification: Kidney Filter System Modification (BA 412575)
Doocription >f ModificatioD:
The purpose of thin modification was to provide a more ef fective RD atmospheric cleanup capability (1.a. , iodine removal) at THI-l by modifying the configuration of the RB Purge and Kidney Filter System. The scope of the modification involves
, the separation of the kidnoy filter inlet duct (to AH-E-101) from the RB purge inlet duct (supplied through AH-V-lC), snd the rearrangement of the kidney filter duct work to enhance iodine removal during purging operation. This modification allows concurrent operation of the kidney tilter and the purge supply without the systems impeding each other.
fafety Eva_hation Summary:
The function of the Kidney Filter System (Atmospheric Cleanup System) in to reduce the airborne radioactivity levels in the RB during normal plant operations, includi c.g outages. This modification did not increase the probability of occurrence or consequence of an accident since the Kidney "!.lter System performe no safety related function and the operation and function f the Purge system following an accident (i.e., containment isolation) in unaffected by this modification. This modification improves the ability of the Purge and Kidney Filter System to reduce airborne radioactriity levels in the RB during an outage.
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Modification: Provide Separation Between i norgency Diesel Generator (EDG)
Engineered Safety (ES) Relays and Non-ES Componenta (BA 412583)
-Descrictl.3D of Modiflentio :
The voltage and f requency relayn for both EDGs were previously connected to non-ES components.' This modification provided proper separation between ES and non-ES components by:
- 1) -Inctalling fcur (4) 6 amp fuoco in EDG A control cabinet and three (3) 6 amp fuson'in EDG B control cabinet;
- 2) Circuite RYS6 and RYS7 were spared and the new circuits RYS6A and RYS7A were routed in the ES cable trays; and
- 3) The follewing circuito were rectasalf f ed to ES circuite from non-ES circulta RYS3, RYS5, RZ53, R"53D, RZ55, and RZ 56. These circultn '
include the differential protection and relay 46G circuits.
.C311-90-2156 Page 49 Attachment 2 Safety Evaluation Summary:
P The margin. of safety as defined '. the SAR was not degraded by this modification beem.se the fuses and rerou' Ang circuits RY56A and RY57A provide separation f rom non-ES circuits, and adt {uate separation between channel A and channel B is maintained. The f ailv.e of the new components in an open circuit condition does not adversely affect :he Class lE function oecause it affects only the metering function. However, f ailure of the fuse block in dead short position during lose of offsite power could disable the EDG breaker closing circuitry by blowing the upstream fuse. Therefore, the fuse blocks have been dedicated for NSR use by test / inspection. Thus, this modification improves the system by separating the non-ES components from the ES components.
Modification: Installation of a Stop Function or. Control Room Panel PL for Fire Pump FS-P-2 (BA 418695)
Reperintion of Modification:
This modification installed a stop function on Control Room panel PL for FS-P-2.
Thie will allow operators to stop the pump remotely f rom the Control Room af ter spus.ous star fire service malfu.,ctions, and surveillance testing.
Safety Evaluation Summary -
The implementation of this activity did not ad'rersely affect nuclear safety or safe plant operations.bocause FS-P-2 is not classified as Nuclear Safety Related and, although it is fed from a safety related power supply, the elustrical power configuration was not changed. The new control cable is routed through the
- safety related tray and in pr' ely protacted against electrical faults.
l
. Remotely stopping the fire puop doce not increase the possibility for a dif ferent type accident or malfunction.
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Modificatign: Replacement of Regulating Transformers for 120VPT Distribution-l L Panel TAA and TRD'(BA 418695/WA 39695010) l Deteriotion of Modification:
This modification replaced the four (4) 10KVA regulating transfargiers feeding panels TRA and TRB - with two (2) 20KVA regulating transformers. The previous regulating transformers couldn't maintain the required voltage output (120VAC) to the Regulated Voltage Panels TRA and TRB and to the Station Inverters lA and lE..
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C311-90-2156 Page 50 Attachment 2 gafety Evaluation Summarvi
- This modification- did at af fect the normal function of the 120VAC Regulated Distribution ' Panels TRA and TRB since:the loads supplied by these panels are improved by the increased voltage output of the new '9KVA transformera. The
-increased voltage output also improves the performance of the inverter static switch. The margin of safety as defined in the SAR was not reduced because the new 20KVA transformers will continue to support instrumentation, control, and is power loads from panel TRA and TRB. This modification does not increase the probability if occurrence or consequence of a malfunction of equipment important to safety previously evaluated in the SAR because chere was no change in trans#L;:cr size and-no new loads were added to panels TRA and TRB.
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-Medification: Fluid Block (FB) Valves / Penetration Pressurization (PP) vslves for Purge Interspace Pressurization Electrical Removal
& RR-V4C/4D PCR Panel Indication Relocation -(BA 418695/WA
--30695010) p_epn iotion of Modification:
The FB Sy' stem was functionally removed from service by Technical- Funct!r 8 project BA'412465i however, associated electrical controls and hardware were 'ot disconnected or removed. -This modification' disconnected the controls and c : indication-for the;FB valves and spare / remove associated electrical circuits.
Interspace pressurization devices PP-V167; 169,: and 170, and pressure switch PS-o77A was- disconnected v.ith associated . circu it s spared / removed ~ by -this
- modification.1 The PCR Panel indication for MU-V3 was modified:such that on loss e
of-DC control power the indication will fail to show the valve as cpen instead of closed to correspond to the f ailure ' position for MU-V3 on : loss of DC controls power.
i Safety Evaluation Summarvi-
.The FB System was previ ousl y removed from service and performs no' safety function l
' as describe 6 in the SAR.- The electrical removal'was accomplished in a manner l Deuchithat the associated RB isolation-valve position indication-relays remain
- functional when, the pilot solenoid valve is disconnected.- The - automatic -
- funct ne for-the; purge interspace pressurization were previously removed and--
--evaluated -- as acceptable "by 'SE 128124-001. This modification electrically 3 disconnected and removed components that were mechanically. d! cabled. The safety .
function' other'PP cor.ponents was altered by this modification.
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C311-90-2156 Page 51 Attachment 2 Modification: Control Rod Drive (CF.D ) Test Jack Panel Installation (BA 418695/WA 34695110)
Description of Modificatio3:
This modification installed a jack panel in system logic cabinet SL-6. This panel is used to connect recordecs for both Rod Drop Time testing and position
~ indication tube reed stack tacting. This modification permits testing of the position indication and 24% zone reference switches to be performed ir .a reduced period of time.
f,gfety Evaluation Summary This modification did not af fect the safety functions of the CRD System since the
. jack panels are used -only for testing and cal *bration purposes System perfe;mance is enhanced by this modificat!.on since these panels adt ce the
~
- possibility.of terminal board and screw failuro. The installation of ihe test
~
. jack panel did not reduce a margin of safety as def2,ed in the SAR stnce the .
panels are: only used' for the _ connection -of reem ders dur.ng testing of the
- position indicators ' and125%- zone reference switche% Additic-alty, the new 1 scomponents 1were installed in' cabinets -which de- not contain safety related components; thus,_there is no potential for the new components to affect the safety related components of the CRD System.
-***************************e *************e*e **ee .....*e***esee ******e*****
JModification: IInstallation ' of Monorail Beam ar.d Supports for . MU-F-2A/B k Letdown Prefilters-(BA 418695/WA 38695110)
Description of Modification: -
This.madification installed a monorail beam and supports to replace the previous -
! jib beam used for changeout of MU-F-2A/B Letdown Prefilters. The monorail will j facilitate.a motorized-hoist and allow handling'of the shielding pig.from the
} '
' general access hallway adjacent:to the MU-F-2A/B cubicle. -This modif1 cation'-
improves the' efficiency of filter changes and subsequently reauces person-rem expo sure. - ,
Safety " valuation Summary:
.Thisi modification- did- not advermely; affect any safety related systems.
Indirectly, the monorail' af fects the Make-Up and Purification System for handling
-MU-t-2A/E Profilters and the Auxiliary Building (AB)'by virtue of strectural l ~ attachments. Structural ' attachment to the AB concrete walle at apprcximate t elevaElon 296'-8" is designed'in accordance with AISC Structural Code. ' The beam supports-were also designed ' for seismic :II ., anti-f alldown. For personnel safety considerations-a factor-of.eafety of,approximately 5.0 against failure was
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C311-90-2156 Page 52 Attachment 2 Egfety Evaluation Summarv (Cont'dit attained for the passive lif ting cocponente. The overall rated load of 4000 pounds exceeds the heaviest potential lift.
Modification Reactor Vessel (RV) Head Fixed Lifting ?ondant (BA 418704)
Qgnerintion <l Modification:
Th!.o modificacion installed permanent lif ting pendante anc. supports in the TMI RV head and service structure. These pendanto replaced the previous lifting cables installed each time the head was lifted.
Safety Evaluglion Summary:
An analysis was performed by B&W that demonstreted that the pendants have no ef fect on the seismic performance of the RV head and service structure.
Nuclear saf6ty or safe plant operation were not adversely affected by this modification. No additional stresses were induced in the reactar coolant pressure boundary in steady state, transient, or seismic conditions. The RD integrity le ensured since there was no potential increase in RB temperature, increase in hydrogen generation due to an accident, or increase in pressure in the RB during a LOCA as a result of this modification. There was no adverse ef fect on the RB cooling functions during and following a LOCA oue to debria generation.
Modification: Reactor Coolant Pump (RCP) Articulal3c' t.rms Installation (DA 418747)
Egneription of Modification:
This modification installed an articulated arm assembly on each RCP motor stand.
This modification was made as a maintenance enhancement to the RCP motors. The arms provide a more ef ficient means of horizontally transferring RCP parts in and out of the motor stand during RCP seal work.
Safety Evaluation Summary:
. The articulated arms serve to enhance maintenance and perform no safety f unction.
However, the arms are designed to remain secure during noinmic events equal to or greater than the original equipment specification loads. The arms are also mounted to the exterior of the RCP motor stand and do not affect the operation of the RCP or other plant componento. Thus, this modification d.=s not adversely affect nuclear safety or safe plant operationo.
C311-90-2156 Page 53 Attachment 2 Modification: Relocate Transzorbs (CMR-88-075)
Description of Modification:
Makeup valves MU-V-3 and 18, and Intermediate cooling valve IC-V-3 had Transzorbs mounted locally; therefore, the Transzorbs we L:0 required to be on the Environmental Quelification (EQ) list. The Transzorbs suppress EMF produced by the induction of de-energizing relay and solenoid colle. Also, Transzorbs connected to these valves had to be tested for low leakage rate. In order to remove this requirement and help preclude the possibility of non EQ being instelled, the Transzorbs located at MU-V-3 and 18 were removed from the FQ environment and the Transtorb for IC-V-3 was relocated to the Remote Shutdewn Tranufer Switch.
Safety Evaluation Summary:
The safety function of the affected MU valves is isolation containment. The valves are automatically closed by an ES signal. The safety function of the affected IC valve is isolation containment and in automatically closed on any ESAS signal. The function of the Transzorb is protection of contacts and colla located within the circuit.
The above safety functions of the affected systems were not altered. The isolation containment valves still perform their required safety function.
Moving the Transzorb at IC-V-3 did not electrically alter the circuit. Removing the Transzorb on MU-V-3 and 18 did not reduce the protection of the contacts.
The remaining configuration is such that the remaining Transzorb protects all circuits.
- e**e**e****ee******e************ **e**e*eeee*****e****************
.Eqdificationt Control Rod Drive Motor (CRDM) Stator Cooling Lines (CMR 90-019)
Description of Modification:
This modification replaced existing Intermediate closed cooling Water (ICCW) hard tubing lines to the CRDM stators with flexible hoses and quick connect coup 1Lege.
The previous cooling water lines betueen the CRDM stator cooling coils and the
.CRD: service structure concluted of rigid tubing and threeded couplings. .This made reconnection of the cooling water lines dif ficult and tedious. This change was accomplished on only 14 CRDM statore during 8R. The remainder of the cooling
' lines will be converted during future outages.
Safety Evaluation Summary:
. The f unction of the CRDM stator cooling coils is to cool the CRDM motor stators during CRDM operation. Each stator is individually cooled by a cooling coil
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9.
C311-90-2156 Page 54 Attachment 2 Safety Evaluation Su,mm E , g.; ,,g which has cooling water supplied f rom the ICCW. The installation of the flexible hose did not alter the cooling water supply, return and isolation functions of the ICCW system. This modification did not adversely affect nuclear safety or safe plant operation since the function of the CRDM motor and the ICCW system was not affected. The new hose has automatic shutoff on both the stem and body assen.blies minimizing water spillage during and af ter uncoupling.
I Modification RV Head Access Hole (CMR 90-030) peecription of Modifleat.;.ous This modification installed an access hole in the Service Structure Support Assembly (SSSA) for improved access to the top surface of the Reactor Closure Head, kafety EvaluNtion Summary:
The installation of the cecces hole did not adversely af fect the structural adequacy nor alter the functien of the SSSA. The hole was covered with a removable bolted cover to re-establish the physical boundary of the support skirt.
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