ML20095B405
| ML20095B405 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 12/31/1991 |
| From: | PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | |
| Shared Package | |
| ML20094P025 | List: |
| References | |
| NYN-92041, NUDOCS 9204220185 | |
| Download: ML20095B405 (176) | |
Text
1 e e New ilampt. hire Yankee M arch 31,1992 ENCLOSt1RIT 1 TO NYN 02041 Report of I'acility and Procedure Chung-: Made at Seabrook Station Purt.uant to 10C1350 !,
for the Peri J A;>ril 1,1991 - De ce rn. ; 3 8 1N l
l l
l l
l l
t 9204220185 920331 PDR ADOCK 05000443 w p pp, 1 R
- . . - - - . _ . . -. . .. --.- _ . .- . .~-.. ~ _ .--. .
. e TABLE OF CONTENTS l
l I
1 Drsign hiodificationt. . 1
. Design Coordinatior. Reports (DCRs) hiinor hiodificatior.s (hth10Ds)
- 2. Ternporary hiodifications . .
108 I 3a Temporary Setpoint Changes . . .
1.12
- 4. Procedures. . . IU
- 5. Procedure Revisions .
140 143
- 6. Tes's and Estariments. ,
7 Technic
- dequirements hianual 4 , 144
- 8. FS A R/UFS AR . . . . . . . . . 152
- 9. hiiscellancous 10CFR50.59 Evaluations . . . . 165 I
I ma - -
'-r+r= - -
g
. . . - - . - . - . .. . .. . . . -_ . . - . - - - - . - . - . . - - - . - - . - . ~ . - .-
' 1. Deslan Mocil_fications The following design modifications were ituplement d at Seabrook Station pursuant to the requiretuents of 10CFR50.59.
i 1
l
(
e .
1 Dru,10N COORDINNI10N R EPORT: Number 86 024 TIT 1.te Reactor Caulant Pump (RCP) N10 tor Lif ting Rig SUMMnRY DP.SCRIPTION: This Design Coardination Report (DCR) provided the engineering basis to utilize on a permanent basis the Reactor Coolant Pump (RCP) motor lifting rig which was utilized during initial plant con (truction. The entirr rig was evaluated agair.st the NHY commitments made in response to NUREG-0012, Revi, ion 2 (Control of licavy Loads). As a result if this evaluation, slings, shackles and turnbuckles associated with this rig were upgraded to meet the guidelines of NUREG 0012, Revision 2.
The DCR also corrected incompatibilities with the dimensions of the slings, shackles and turnbuc} les.
PURPOSit This DCR ensured (but the RCP Motor lift ing rig ceuld b. used with irradiated fuel in the reactor vessel without violating Operating License rcquirements or compromising NilY commitments.
S Arim' EVAt.UATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review deteru.ined that this DCR diri not make changes in tha facility as described in the FSAR. Ilowever, as a conservative measure, a safety evaluation was performed. The safety evaluation determined that certain revised dimensions did not reduce the required safety factors of the rig and that no safety related equipment was relocated. The safety evaluation concluded that tbc DCR did not create an unreviewed safety question.
2 I.____--_-_-_--___--_-___--__-_--_______-_____---___-___-____-_ _ - - - _ - _ - _ _ - - _ _ - _ - _ _ - _ - _ _ - _ _ _ _ _ _ _ - _ _ _ - - _ _ _ _ -
.. . . _ . . - _ -. . . . . - . - - . - . - _ _ - - ~ . - - - . - - . . - - . -
I, e
DnSIGN COORDINATION Hl! PORT: N u n.be r 86 036 Tr!LI:: Steam Generator (SG) l'rimary Alanwa) Studs i
SUMMARY
Dl'.SChil'110N: This Design Coordination Report (DCR) replaced the bolts with studs as the SG primary manway cover fasteners. l ne studs will '
be installed and removed using a multi-stud tensiocing tool.
1 l'URroste This DCR was implemented to benefit from the many adsanlages of using studs versus bolts as SG manway cover fasteners. Time to install anti remove the manway is greatly reduced. A 4.8 man rem dose sn6p per manway cover removal or installation is expected. Fastener preload and gasket compression are more uniformly applied, thus minimizing the potential for leaks. Stud tensioning avoids turning and torquing fasteners under high load, reducing the potential for thread seizure. Studs with tighter thread design tolerar.ces cr.n be used, further reducing the potenti.il for thread seizure.
S Al'Ir1T Evat.UN' ion
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review determined that the FSAR was not directly e.ffected in tnat the text, figures and tables in the FSAR did not L reouire changes. The safety evaluatiot determined that the SG manway cover studs are Lafety Class 1 components. The studs meet the original design requirements of the bolts. i The stud threads will exceed the tolerances requirements of the bolts. The use of studs will produce a significant dose savings and reduce the potential for manway cover leakage when installed. The safety : valuation concluded that the DCR did not create an unreviewed safety question.
3 5
Dl! SIGN COORDINAT!nN Rt! PORT: Number 86-163 t
Trn.11: R e p t.sce m e n t of 4" Diaphragm uhes in the Spent Fuel Pool Cooling e.nd Cleanep System
SUMMARY
Df!SCIUPGON: This Design Coordination Report (DCR) replaced two 4 inch diaphragm salves originally installed in the drain line of the reactor cavity cleanup section of the Spent Fuel Pool Cooling and Cleanup (SFP) System.
The replacement valves were plug valves. The DCR also provided the enginecting basis to revise the Normal Condition I maximum temperature value of the Reactor Cavity Cleanup
- System frota 125 degrees F to 140 degrees F and to upgrade the frequency of occurrence of a refueling loss of cooling event from Upset Condition 11 to Emergency Condition Ill.
PURPGsit The two 4 inch diaphragm valves originally installed in the drain line of the reactor cavity cleariup section of the SFP System were replaced based on ALARA considerations. Operating experience at other plants indicated that diaphragm valves
- n this application were likely to create crud traps causing high radiation dose rates. The other changes resolved discrepancies between the system design basis and applicable design standards.
SAPITY EVA1.UATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety. evaluation applicability review determined that the DCR made changes in the facility as described in the FSAR. The safety evaluation -
determined that the Reactor Cavity Cleanup - System is non-nuclear safety class! seismic Category I except for - the Containment penetration piping which is Safety Class 2/ Seismic -
Category 1. The replacement plug valves were eval.iated to be equivalent substitutes for the
- diaphragm ulves. The updated design calculations are conservative and consistent = with Technical Spuifications. The safety evaluation concluded that the DCR did not create an unreviewed sai sty question. .
FCR 90 033 4
l we a --n.-- ..~w - . ~ . - n. -, - , . , ..,A,,. , .. -- - ..wN---,,n-- - . , ,- , , , - - - ,-e>-..--, ve~
e .
l DESIGN COORDINATION RiilonT: Number 86-210 ,
l 1
i TITLE: Spent Fuel Pool Level and Leakoff Sump Level Modification
SUMMARY
DILSCRWrlON: This Design Coordination Report (DCR) replaced the existing ultrasonic level measuriag rystem for the Spent Fuel Pool with a capacitance type system. The replacement system provided narrow range monitoring of a ,
nairow band around the normal water level. The DCR also relocated the spent fuel leakoff sump level switch. ,
i FURPOSin The existing SFP level measuring instrunant failed tc maints.in the required accuracy and physically interfered with SFP bridge crane operation. The spent fuel leak off sump level switch was inaccessible when sump water level was high. The ,
objective of the DCR was to improve the accuracy of the Spent Fuel Pool level measurcineat with equipment that is more accessible and does not interfere with opeiation of the SFP bridge crane, This DCR addressed the recommendations of NRC Information Notice 88-65. -
sal'InY EVALUATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review determined that the DCR made changes in the facility as described in the FSAR, The safety evaluation
' determined that the design changes do not affect the safety functions of the structures, systems or components associater' with the cooling of spent fuel. The changes enhance .
existing design by improving instrument accuracy and removing interference with bridge cranc ,
operation. The reduction in the indicating range of the SFP level instrumen_t was evaluated c to be acceptable with respect to Technical Specification action requirements. The safety i evaluation concluded that the DCR did not create an unreviewed safety question. ?
FCR 90-019 L
i
(
l t
i l
5 i
l'.
l l
Dt'.slON COORJ11 NATION Hl! PORT: Number % 227
, T m .le Relocation of Bulk liydrogen Storage Facihty
SUMMARY
Dl!$CHtP110N: This Design Coordination Report (DCR) installed an upgraded vertion of the originally designed bull hydrogen storage facility.
The facility consists of two mobile tube trailess located northwest of the Unit 2 Turbine Building, a regulating station and distribution piping. The DCR used equipment and previously-installed oiping which were part of the original bull storage facility. (The original bulk storage facility was abandaned in a partially completed state during initial construction.
Hydrogen bottles stored in racks in the Turbine Building and the Waste Precessing Building were the alternetive source of hydropeu gas.) The DCR also converted an unused hydrogen header in the Administration Duilding to an argon gas header, and made miscellaneous other changes.
l PURrosin The bulk hydrogen storage facility was installed primarily to eliminate the handling and movement of hydrogen bottles. The conversion of the unused hydrogen header to an argon header resolved an unsafe condition regarding storage of argon gas bottles in the Chemistry Laboratory Mechanical Room.
SArttry EVALUATION
SUMMARY
. A safety evaluation was performed for this design change.
The ssfety evaluation applicability review determined that -
the DCR made changes in the facility as described in the PSAR. The safety evaluation determined that the affected piping an.! equipment is designated non nuclear safety class and met the requirements of applicable design codes and standards. Major equipment, such as the bulk sto age trailers, were the equipment originally purchased for this application. Other equipment added or upgraded met applicable requirements. Conversion from bottled hydrogen storage to bulk hydrogen storage did not affect the systems which use the hydrogen.
The safety evaluatio*: concluded that the DCR did not create an unreviewed safety question.
FCR 90 85 I
6 e
._ _ __ _ - . . . . _ _ _ _ ___ _ . . _ _ _ . . _ . . _ . _ _ . . . _ _ ~ . . __ _ _ _ _ _ . .
. e l
UtiMON COORDINA110N RI? PORT: N u m bt'r bo-228
)
l TITLie Main Control Board Meter Banding
SUMMARY
D15CRil"IlON: This Design Coordination Report (DCR) added pressure sensitive, graphic tape to meters on the Main Control Board. The graphic .
tape contains
- bands" indicating parameter limits and normal operating points. The DCR also relocsted a fire panel alarm born Pulu'Osn: This DCR resolved human engineering deficiencies identified in the Control Room Design R e vie w. The intent of the bands was to indicale when the parameter was out of its full power value or range.
sal'lfrY EVALUATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review dete.rmined that the FSAR was not directly affected in that the text, figures and tables in the PSAR did not require changes. The safety evaluation determined that the pressure sensitive tape would not affect the proper operation of the meter. The relocated fire panel horn would be more
~
audible and would not affect the proper operation of other Control Room equipment, The safety evaluation coccluded that the DCR did not create an unreviewed safety question.
{
7 l
- ~ _ , . . - . . - . , , -
4 .
i D1510N COORDINATION RILPORT: Number 86-238 ,
Triu: Addition of Check Whc to Refueling Cavity Drain Piping i
SUMMARY
D15CRifTION: This Design Coordination Report (DCR) added a check valve to the refueling cavity drain line to the Floor Drain Tank.
PURPOS11/ Addition of a check valve to this refueling cavity drain line prevents possible contamination of reactor coolant water during refueling by preventing backflow of water containing oil or chemicals from the Floor Drain Tank to the refurling cuity.
J S4 3rrY EVA1.UATION
SUMMARY
- A safety evaluation was performed for this, design change.
The safety evaluation applicability review determined that the DCR made changes in the facility as described in the FSAR. The safety evaluation determined that the refueling cavity drain line is part of the non nuclear safety, Seismic Category 1 Reactor Cavity Cleanup Syrtem. The system is used only during refueling shutdown plant conditions. The added check valve meets or exceeds the design requirements of the original system. The safety evaluation concluded that the DCR did not involve an unreviewed safety question.
FCR 90 049 i
p 1
8
-- ^, . . ,.. ._, . . . , . . . - . _ _ . , . . - - - . . . - - . -
. , . . -. . . _ . . _ . , .--..- . . . . , m -- --- , . .
e
- DESIGN COORDINA110N REPORT: Number 86 239 Tm.n: Residual Heat Removal (RHR) Autoclosure Interlock Removal
SUMMARY
DESCR!! TION: This Design Coornination Report (DCR) removed the design feature of the RHR System motor-operated suction isolation valves which causes automatic salve closure on high pressure. Redundant RC5 high pressure '
alarms were also deleted.
PURPOSE: Removal of the automatic closure feature eliminated the potential for spurious closure of the RHR Sysicm suction isolation valves. Spurious isolation of the RHR suction isolation valves could contribute to Reactor Coolant Sys t etu (R CS) overpresst.rization as a resul. of loss of RHR cooling and isolation of the RHR System relief vahes from the RCS.
SAnn EVAI.UATioN
SUMMARY
- A sdfety evaluation was performed for this design change, The safety evaluation applichbility review determined that the DCR made changes in the facility and in the procedures as described in tbc FSAR. It also deterinined that a change to the Operating License was required. The safety evaluation determined that the DCR did not involve an unreviewed safety question.
FCR 90 090 i-9 f
u -- , ___ _ . . _ .,.__ _
DitSIGN COORDINATION Rt! PORT: Number 56-355 l
TITt.it Reserve Auxiliar3 1ransformer (R AT) Disconntct Switches and Sulphur llexafloueride (SI 6) Bus Replacement
SUMMARY
Dl!SCKWM ON: This Design Coordination Report (DCR) replaced the bus duct segments of the SF 6. insulated bus sptem in Gas System Zone
- 2 (GS
- 2) with Asea Brown B os e ri (ABU) bus duct segments. It also added ABB disconnect and ground switches to Gh#2 near the R ATs.
PURPOSt'.: Replacement of the Gh#2 bus duct segments with segments manufactured by ABB resobed reliability concerns with the replaced bus duct segments. The addition of the disconnect and ground switches to GS#2 permits isolation of one RAT for testing or in the event of its f ailure while the other RAT remains energized.
SartnY EVAL.UATION
SUMMARY
- A safety evaluation was performed for this de ign change.
The safety evaluation applicability review detern'ined that the DCR made changes in the facility as described in the FSAR. The safety e 'aluation determined that the replacement bus duct segments in GS#2 would have a very low failure rate, and thus improve the reliability of offsite power. The replacement equipment conforms to applicable design criteria. The safety evaluation concluded that the DCR did not create an unreviewed safety qu stion.
FCR 90-126 10
DESIGN COORDINATION REPORT: Number 86-447 TrrLn: Primary Component Cooling Water (PCCW) Chemical Addition
SUMMARY
DFSCRIPTION: PCCW System coolant chemistry is maintained by periodic, manual additions of hydrazine to the PCCW bead tanks. This Design Coordination Report (DCR) provided permanently installed tubing and valves to facilitate the addition of hydrazine to the head tanks from the mezzanine platform, elevation 65' 9' of the Primary Auxiliary Building (PAB) using a portable pump.
PURPOSH: By providing permanent tubing and valveb, this DCR facilitated a safer method of adding hydrazine to the PCCW bead tanks. The former method required personnel to climb on equipment at unsafe heights within the PAB to install temporary tubing each time a chemical addition was needed. Therefore, this DCR eliminated personnel safety concerns.
S AFIGY EVALUATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review determined that the DCR made changes in the facility as described in the FSAR. The safety evaluation determined that the DCR would not affect the reserve coolant volume of the PCCW System and would not, therefore, affect the cooling capability of the PCCW System. The permanent tubing conforms to applicable Code requirements and does not affect the previously-plugged connection point to the head tanks. The safety evaluation concluded that the _ DCR did not create an unreviewed safety question, i
FCR 91-070 t-11
Dl!SION COORDINATION RI! PORT: Numbte 86 f 54 I
i Trn.It: Containment Enclosure Ventih tion Area (CEVA) Test Connt etions
SUMMARY
DIISCRil'rtON; This Design Coor din atic,n Report (DCR) provided test connections accessible otaside th: CEVA for rr.casuring negalise pressure at ten different locations inside the CEVA. Instrument tubing screened on the inside and capped on the outside was installed to provide a permanent location to sneasure differential pressure and thus determine the negative pressure of the CEVA at the sampled location.
PU RPOSti: The DCR was implemented to provide a permanent penetration from which to measure CEVA pressure at ten different locations. The DCR eliminated the need for repeated entries into the CEVA during the conduct of the 18 month surveillance test of CEVA negative pressure produced by the Containment Enclosure Ventilation System.
SArtrrY EVALUATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review determined that the DCR made changes in the facility as described in the FSAR. The safety evaluation determined that the test connections do not interact with or affect the function of the Containment Enclosure Ventilation System. The integrity of the CEVA penetrations through which the test instrument tubing is installed is maintained by the use of seismically qualified supports and quality tubing. Leak tightness of the CEVA is maintained by caps on the outside of the instrument tubes. The safety evaluation concluded that the DCR did not create an unreviewed safety question.
FCR 91014 i
FCR 91056 l
l I
l' l
12 l
i
. . - . . , . , , ,- .,, - - . . . . , , - - , - , . - - . - . - , --n- , , . , - . ,
J Dl'.5tGN COORDINNriON RI'. PORT: Number 86-634 i.
Trrtje Primary Drain Tank (PDT) Pressure Transmitter Rescale/ Heat Tracing of S11 and 13RS Instrument Lines
SUMMARY
DPSCRil"flON: This Design Coordination Report (DCR) replaced the pressure transmitters for the Primary Drain Tanks (PDTs), which had an indicating range between 0-35 psia. it also revised the heat tracing for instrument tubing in the Boron Recovery (BRS) System and the Steam Generator Illowdown (SB) System. The DCR also deleted the requirement for heat tracing of certain instrument tubing in which the boron concentration is sufficiently low that precipitation without heat tracing will mat occur.
PURPGste The replacensent PDT pressure transmitters have an indicating range between 5 and + 15 psig. The resised range will provide more useful indication of tank internal pressure relative to atmospheric pressure to help prevent tank damage due to internal vacuum, The revised heat tracing corrected an improper heat tracing installation which covered the instrument diaphragm seal and part of the capillary tubing.
SAPIrrY EVALUNIlON
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review determined that i the DCR made changes in the facility as described in the FSAR. The safety evaluation determined that the PDT pressure transmitter rescaling did not change the BRS System design and was a human factors improvement. The heat tracing changes did not affect the process connections and improved the ability of affected instrumentation to function. The safety evaluation concluded that the DCR c"i not create an unreviewed safety question.
FCR 90-069 h
L
- _ - . - . . ..-. -... ~ . ~ . - - - . .. - - - .--..- - - - . - _ . - - . ~ . _ . - -.- .
Dt! SIGN COORDINA110N Hl! PORT: Number 87 294 Tm.it: Spent Fuel Pool Sampling l
l
SUMMARY
D1tSCRilTION: This Design Coordination Report (DCR) enhanced the Spent Fuel -1 Pool sampling lines by connecting 3/8 inch tubing to the existing three sample lines. The 3/8 inch tubing was routed te a new sample sink.
PURPGst!: The purpose of this DCR was to facilitate a more controlled sample finw and to minimize the potential for contaminated spills during sampling.
SAr1rIY FVALUA110N
SUMMARY
- A safety evaluation was performed for this design change. -
The safety evaluation applicability review determined that the DCR made changes in the facility as described in the FSAR. The safety evaluation determined that the new sample tubing was non nuclear safety-related equipment which did <
not interact with or affect safety related equipment. The safety evaluation concluded that the DCR did not create an unreviewed safety question, i FCR 90-064 l
l a
)
l-i
'P 14 V
- _, , .-,. . - .-g, wmwr.-,c,-_ ---
.%-_. ,,--c,y-w m, .,-n ,,,,.,m ,, __ m-,.7,w,
.~.~.___.__.-.__.__m.___.___ . - - _ -- . - - _ _ . _ m______._
l . . _ _ _ _ ..-.___ . .
DliSIGN COORDINA110N Rt! PORT Number 87 316 l
I i
Tritti: Safety injection Accumulator Tank Pressure lustrumentation l l
l l
SUMMARY
Dl!SCR!l'110N: This Design Coordination Report (DCR) upgrades one of the existing pressure tr.ansmitters installed c,n each Accumulator Tank to Design Category 2 instrumentation as defined by FSAR Section 7.5.4.4.c.
PURPOSil: This DCR was implemented to fulfill NHY's commitment to provide envitc.nmentally qualified instrumcntation to monitor Accumulator Tank level or pressure S AFirrY EVALUA'110N SUMMAkY: A safety evaluation was performed for this design change. .
The safety evaluation applicability review determined that the DCR made changes in the facility as described in the PSAR. The safety evaluation determined that the design requirements of the upgraded transmitter and tubing installation are equal to or greater than those cf the original design. The safety evaluation concluded that the DCR did not create an unreviewed safety question.
FCR 89-037 i
G 15
,,,.e, , g. , , -, _g--,--m,
-_.__-_-._....__....__.__._m._.__ -___.__m_..-._.-_. _ . . . _ _ . _ _ _ _ . . . _ . . ~ . . _ . _ _ - .
9 4 DI! SIGN COORDINA110N RI!!' ORT: Number 87-318 TIT 1 ll: Containment Sump Water Temperature SUMM ARY Dl!SCR!l"IlON: This Design Coordination Report (DCR) upgrades existing thermocouples installed on the inlet side of each Containment fluilding Spray (CBS) beat exchanger to Design Category 2 instrumentation as defined by IEEE 3231974 and referenced standards.
PURPOSil: This DCR was implemented to fulfill N H Y's commitment to provide environmentally qualified instrumentation to monitor Containment Sump Water Temperature.
. S APITY EVAI UAllON '
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review determined that the DCR made changes in the facility as described in' the T'SAR. The safety evaluation '
determined that the environmentally qualified Containment sump water temperature monitoring channels are non nuclear safety related equipment and do not interact with or affect safety related equipment. The design requirements of the upgraded instruments are equal to or greater than:those of the original instruments. The scfety evaluation concluded that the DCR did not create an unreviewed safety question. .
FCR 90-058 F
E L.
e 16
_ . .,.m _. _ . _ . - _ _ _ _ _ - - _ _ _ . - _ _ . . _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ - _ . . _ _ _ _ _ _ _ _ . . _ . . ~ . . _ . =
I- l I
1 DitSIGN COORDINATION Hl!!' ORT: Number 87.122 i TITL11: Replacement of Residual Heat Removal (RHR) Miniflow Valves,1-RH FCV-610
& 1-R H-FCV-t>ll S UMM ARY Dl!SCRl!"nON: This Design Coordination Report (DCR) revises the design of the Residual Heat Removal (R!l) Pump A and B minimum flow recirculation isolation valves. Motor operated gate valver, were replaced by motor-operated globe valves. The minimum flow restricting orifices were re-sized to accomtandate the more restrictive globe valve flow characteristics. The DCR ah,o relocated these valves in and revised the piping cenfiguration of the RilR minimum flow recirculation lines. Miscellaneous other mechanical changes and associated electrical und instrumentation design changes were also made.
PURPOSti: A excessive pressure drop existed across the flow restricting orifice in the f RHR rninimum flow recirculation lines. This pressure drop created high vibration. Leakage at the pressure taps of this flow element was attributed to this high vibration. The purpose of the design change was to redistribute the pressure drop within the minimum flow recirculation 'line and thus reduce the piping vibration.
SAFETY EVALUATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review deterrnined that the DCR made changes in the facility as described in the FSAR. The safety evaluation
- determined that the replacernent globe valve and the re configured recirculation piping meet the applicable design criteria of and perform the same function' as the original gate valve and piping. The safety evaluation concluded that the DCR did not create an unreviewed safety question.
FCR 89-018 17
e e DEstGN COORDINATION Kl!POR't: Number 89-047 l
I TrrtA Startup $ trainers for Condensate c.d ticater Drain Pumps
SUMMARY
DESCRIF110N: This Design Coordination Report (DCR) replaced the original suction strainers for the Condensate and lleater Drain Pumps.
This DCR also provided strainer differential picssure indication and made changes to the Heater Drain Pumps stuffing box leakoff, mechanical seal and seat injection pressure control vahe.
PURPOSE: The use of suction strainers during power ascension testing protects the pumps from damage caused by debris that is expec'ed to be swept into the hotw ell during initial operation. The original Condensate Pump suc. inn straincts had inadequate free surface area to support full condensate flow . The origiaal Heater Drain Putnp suction strainers had a mcsh opening larger than that recommended by the pump manufacturer.
Strainer diff e r ential pressure indication f acilitated monitoring strainer cleanliness and pruided flush and cleanout connections for expedited strainer cleaning.
S ArtrrY EVAL.UATION
SUMMARY
- A safety evaluation wit p?tformed for this design change.
The safety evaluation applicability review determined that the DCR made changes to the facility as described in the FSAR, and ideutified the affected FSAR Figure. The safety evaluation determined that the DCR made changes to portions of the Condensate and Heater Drain Systems that provide no safety function; and that the changes did not affect saf-ty systems. The safety evaluation concluded that the DCR did not create an unreviewed safety question.
FCR 89 068 18
I D11 SIGN COORDIN ATION Rt'I' ORT: Number 89-03)
I T m _In Mid I.oop Ultrasonic Les el Measurement Sptem
SUMMARY
DitSCRif"110N: This Design Coor dinnt;on Report (DCR) provided a narrow range indication of Reactor Coolant Syr. tem (RCS) level. The Ultrasonic Level Measuring Systein (U1.315) utilites piezoc!cctric ultrasonic trant,ducers dry coupled the hot leg piping of RCS Loops 1 and 4. The lesel signal is brought to the Main Plant Computer System (MPCS) for indication in the Main Control Room. The indicating range of the ULMS covers from approximately four inches above the bottom of the hot leg to the top of t Ne hot leg.
PU RPOSl!: The ULMS provides an independent method of measuring RCS inel during reduced RCS inventory conditions, especialiy when RCS water level is at "mid-loop / The ULMS provides the RCS level hardware enhancements committed to in the NilY response to Generic Letter 8817 (NHY letter NYN-89012). The instrumentation provides additional aids to the operator to help prevent loss of decay heat removal capability while shutdown.
SAirrY EVALUATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluetion determined that the DCR changed the facility as described in the FSAR and identified the affected FSAR Figure The safety evaluation applicability review determined that the ULMS consists of non-safety related instrumentation which is not part of any system identified in FSAR Chapter 7 as important to plant safety in the operational or post accident modes. The safety evaluation determined '
that the DCR is a design enhancement, will decrease the probability of certain accidents and will not adversely affect the function or operation of structures, systems and components important to safety. The safety evaluation concluded that the design change did not create an unreviewed safety question.
FCR 90-079 19
l e e (el$tGN COOllDINAT10N Rl! PORT: N u rn ber 89 065 Trn.tr Circulating Water D* *.a - T Upgrade
SUMMARY
Dl3 Crit" DON. This Design Cootdination Report (DCR) replaced the resistance temperature detectors (RTDs) in flic Circulating Water (CW) intake aad discharge structures. It also replaced their associated processing circuitry for display and alarm functians through the Main Plant Computer System (MPCS), The DCR also .clocated one of three temperature detectors used as input to the MPCS for calculation of CW average discharge water temperature.
PURPOSin The purpose of this design change was to provide detectors and circuitry which would more accurately measure the CW temperature rise across the Main Con d e nse r. An accurate measu_rement of tnis parameter is required to ensure compliance with the National Pollution Discharge Elimination System (NPDES) Permit. The relocated temperature element provided a more reliable indication of CW discharge water temperature.
(CW average ditcharge water temperature is calculated by the MPCS using this and two other measurements.)
SAITTY EVALUNMON
SUMMARY
- A safety evaluation was perforrned for this design change, The safety evaluation applicability review determined that the DCR changed the facility as described in the FSAR and identified the affected FSAR Figure. The safety evaluation determined that the CW System temperature measurement channels are not nuclear safety-related; and the changes did not affect safety-rela'ed equipment. The safety evaluation concluded that the DCR did not create an reviewed safety question.
FCR 91-009 I
. . . _ . . _ . . _ _ . . - _ _. . - . _ - _ _ _ _ _ _ . . - _ . . _ . . _ . . _ __.. _ m. ._._.. .- _____.m_..
9 -
DESIGN COORI)! NATION REPORT: N u m ber 59-067 TrrLt!: Waste Gas (WG) System Replacement Valves
SUMMARY
DIISCRW110N: This Design Coordination Report (DCR) replaced ten globe valves installed in the Radioactive Gaseous Waste, Vent Gas and Hydrogen Gas Systems. The original ten globe valves varied in size between 1/2 inch and 1 inch and had Teflon seats. The replacement valves were 1/2 inch ball valves of a design which has been successfully applied to hydrogen gas service in other industries.
Punrosf3: The purpose of the design change was to improve the leak-tightness of the valves and reduce required corrective maintenance.
r a
SAPIrry EVALUATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review siated that tbc DCR changed the facility as described in the FSAR, and identified the affected FSAR Figur e. The safety evaluation stated that the affected systems are non-nuclear safety class.
The safety evaluation determined that a failure of the affected systems would not result in I a failure of safety related equipment. The safety evaluation concluded that the design change l did not create an unreviewed safety question.
FCR 90 091 C
r d
4 l
21 l
e wE'-M - w- **vs 4--cevwv'w1 e e v --
'-etev t '"rw v yee1- ye e-rr--'1- ---3 w +---Tw m
_ ~ _ . . _ _ _ _ __ . . ... . . - _ . . _ . m -.__ _ .
Dus1GN COORDINNI'ON REA' ORT: Number 89 070 Trrt.E. Steam Traps for Auxiliary Steam System Adjacent to Valve AS-V.38 S UMM ARY ' DESCRir' TION: This Design C oor dinatic.i Re ,-t (DCR) provided for the installation of two additional stear, trap stations in the Auxiliary Steam ( AS) Systero, with condensate return to the system via the Auxilian Steam Condensate (ASC) System. Three change authorizations made changea to the piping, pipe supports and t.dded a salve.
pi1RPOSE: Thit DCR provides for the remcval of condensate collected in low points of the AS piping, The valve added by one of tbc change authorizations was to facil; tate on-lioe in:tallation of the DCR.
SAlvrY EVALUATION StJMM AR1i: A safety evaluation applicability review determined that this DCR did not make changes in the facility as described in the FSAR. However, as a conservative measure, a safety evaluation was performed- for the change authorizations associated with this DCR. The saft.ty evaluation determined that the revised piping, pipe supports and added valve did not alter the function of the AS System. The safety evaluation concluded that the change authorizations did not create an unreviewed safety question.
4 k
- M i
i l
1 L
x, .*
Dt1SION COORDfNATION R!iPORT Number 89-0?$
Ttrut: Unit 2 M ain S'eam isointion Valve I MSIV) Actuator Rebuild for Unit ' 1 Installation
SUMMARY
DESCIUrrlON: The actuators for the Unit 2 MSIVs were rebuilt and refurbis t d by the original manufaciarer. This Design Cootdinatiou Report l
(DCR) installed the rebuilt and refurbished actuators on the Unit 1.MSIV actuators. The
- replaced, Unit I actuators were returned to storage, 4
PURPOSP: . The qualified life of the Unit 1 MSIV actuators would expire several months '
af ter restart follow:ing the first refueling outage. The rebuilt and ref urbished Unit 1 actu e>rs were restored to ths new condition by the original manufacturer and installed on the Unit 1 MSIVs per this DCR to permit continued plant operation witn environmentally-quahfied equipment.
t F
SAFinY EVALUATION St,MMARY: A safety evaluation was performed for this design change
" e safety evaluation applicability review determined that
- the FSAR was not directly affected in that the text, figures and tables in the FSAR did not require changes. The MSIVs are Safety Class 2, active components. Safety related electrical l components of the actuator are Clau IE. The safety evaluation determined that the change was essentially an "in-kind" replacement. The ability of the MSIVs to close within the time '
r equirements stated in the FSAR and with the existing level of reliability were unaffected
- by this change. Minor subcomponerit changes were evaluated to verify that they created no adverse affects on 'the function of the MSIV.and that tne actuators were mounted seismically.
The safety evaluation cancluded .that the DCR did not create an unreviewed safety question.
L i
U l I
_ _ _ _ _ _ _ . . . . - _ _ _ _ - _ . _ _ _ .~. .. _ _ . - . - , , - _ . - _ ,
. - - - - . . - . . . - . -. . - ~ . . - _ . . . . _ . . . .
. . - . - . . - . - . _ . - . - . - . ~ . ~ . . . _
y .,-
[)tGlGN ' COORDIN A '.10N RE!' ORT: Number 89 079 TrrLl;. Unit 2 Personuci Tunnel Detection System
SUMMARY
DESCR!!'I1ON: Details of this design modification are not provided in this repcr?
since they might involve safeguards information.
PURPOSE: . The purpose of this design modification is not stated in this report since it might involve .cafeguards information. .,
J SAFirTY EVALUA110N SUMM ARY: A safety evaluation WRS performed for [nis desigt. Change.
Details of the salety evaluation for this design raodification are not provided in this report since they might involve safeguards information. The safety evaluation concluded that the DCR did not create an unreviewed safety question.
F 1
4 e
1 1
t e
l'
~
4 24 1 i
I j
l- I L
V . . - .- - . . - - .l .
- . . ... .- .-, . - . .. . . . _ - . .. ___ _ . _ . . . . - . - - . - . . - . . . ~ . . , .
~
[e .
d.
- DESIGN COORDINATION REPORT: Number 89-080 Tn't.E: Control Building Air (CBA) WL. Air Intake Structure Relocation
SUMMARY
DESCR11*nON: Details of this design modification are not provided in this report since they might involve safeguards information. >
PURrosu: The purpose of this design modification is not stated in this report since it might involve safeguaids information.
R SAFrm' EVALUATION
SUMMARY
- A safety evaluation was performed for Ibis design change.
Details of the safety evaluation for this design modification are not provided in this report eince they might invo;ve safeguards information. The safety evaluation concluded that the DCR did not create an unreviewed safety question.
FCR 90-026 1
0 n
u h
4 9
9
'i.
W
.>w= m -ar =,w4 .wv v -- i--<-,,e ue-- v-
. _ . . . . . _ . . . _ . . em .. _ _ _ _ _ . . .m _. ._ . . . _ _ ._ _ . _ _ . .
4 4 Dl' SIGN COORDINATION REPORT: Number 89-081 Trrun Unit 2 Circulating and Service Water P aing S UMM ARY Di'.SCRil' MON: Details of this design modification are not provided in this report since they might involve safeguards information.
PURPOSE: The purpose of this des;gn modification is not stated in this report since it might involve cafeguards information.
S AFErr EVALUNDON
SUMMARY
- A safety eva'uation was performed for this design change.
Details of the safety evaluation for this design modification are not provided in this report since they might involve safeguards information. The safety evaluation concluded that the DCR did not create an unreviewed safety question.
I l.
Lc l
I 20 l
1
~ . . . . . ,
.4' o.
g .Di$1GN COORDINATION RuroitT: Number 90-007' ,
TrlLE: Reactor Makeup Water (RMW) Nitrogen Blanket
SUMMARY
DI5CRITTION: This Design Coordination Report (DCR) converted two temparery modifications (TMODs89-011 and 89 019) to permanent desiga.
The DCR provide.d nitrogen' cover gas to tbc Reactor Makeup Water Storage Tank from the Nitrogen Gas System and a piped supply of demineralized water to the water seals on the tank vent and overflow pipes. It also prosided an revised tueans of sampling the tank contents for dissolved oxygen concentration.
PURPOSE: The purpose of the deAign chanbes was to make permanent the temporary modifications which were proven effective in reducing the dissolved oxygen content of the tank ' contents.
' S AFIrrY. EVALUATION
SUMMARY
- - A safety evaluation was performed for this design chan;,e.
Tbc safety evaluation determined that the. DCR changed
.the facility as described in the FSAR and identified the affected FSAR Figure. The safety evaluation concluded that the des,gn change did not create an unreviewed safety question,
'FCR 90 068 i
i' I-( '. '
l l
27 l
L
_. -. , ~. . _ , , . , . - . r .
. ._-.- .- . .. -. . - , , . . . . ~ . - . - . . - ,. - ~ ~ . - - . . ..- . . - _ . . .
A a DESIGN COORDINATION Rl! PORT: N u mbs.r 90-009 ,
TrrL11: Condensate System Tie-in for Portable Demineralizer Return
SUMMARY
DILSCRll"rlON: This Design Coordination Report (DCR) added a connection and a gate valve to the Condensate System to permit condensate which has been purified by mobile demineralizers during plant startup to return directly to the Main Condenser rather than indirectly via the Condensate Storage Tank.
PURPOSE: This design change will increase the maximum rate of condensate cleanup by demineralization during startup. .
SAFITTY EVALUATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation determined that the design change
. modified a non-safety-related system described in the FSAR in such a way that .no adverse impact on safety related structures, systems -or components was created. The safety evaluation concluded that the design char:ge did not create an unreviewed safety question.
L FCR 90 032 i
o p.
l t.
L i
, ,..-_v.. -
- ~ _ ..-.. - - . . -. -- ~ . - = . _ _ _ _ . . . ~- .
'o - .* l I
l DustGN COORDtNATION REPORT: Number 90 023 i TmA Additional Demineralized Water Storage
SUMMARY
DESCRII' TION: This Desigr Coordination Report (DCR) provided an additional 500,000 gallon storage tank to the Demineralized Wnter System.
In addition to the storage tank, the DCR added piping interconnections, a tank heating system, instrumentation and controls to monitor tank parameters and prevent freezing of the contents. The DCR also documented a probabilistic evaluation of the consequences to the r Control Building Air (CBA) east air intake from flooding due to the postulated failure of the new tank. As a result of this evaluation, a splash shield was provided to protect the CBA cast air intake from the effects of a potential water jet that could be produced by a puncture of the new tank. Finally, the DCR provided a continuous level transmitter to the new tank with readout locally and in the Main Control Room via the Main Plant Computer System (MPCS).
PURPOSE: The purpose of this DCR was to provide 500,000 additional gallons of demineralized water to be immediately available in storage for use in the event of a major tube . leak in the main condenser resulting in seawater contamination of the secondary system. The additional demineralized water would be used for expeditious secondary plant clean up (rapid reduction of chloride concentration) following repair of tne tube leak in support of plant restart.
l SAlitrTY EVAL.UATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation stated that the DCR changed the facility as described in the FSAR. It identified the affected systems as non-nuclear Safety Class. Safety-related equipment was not directly affected. The indirect, potential effect of flooding of the CBA cast air intake resulting from loss of tant contents was evaluated and the effects were determined to be within the limits of the bounding flooding analysis- A splash shield was determined to be needed to protect the east air intake 2 rom the effects
- of a potential water jet produced by a puncture of the new tank in certain locations. The
!- safety evaluation concluded that the design change did not create an unreviewed safety question.
FCR 90-077 P
3 29
. . . _ . _ _ _ _ . . _ _ _ . ._._. _ _ .. _ _ __ ~ _ _ . ~ __ _ _ _ _ _ _ _ . __. _ _ _ _ . .
I.
e_.
i 4
L
_ DI!stGN' COORDINATION flEPORT:
Number . 90 024 Trrt.E: RCA - Exit Area Modifications
SUMMARY
DFJCRil'rlON: This Design Coordination Report (De'R) reconfigurated _the rit zone of the Radiologically-Controlled Area (RCA) of the Station.
The DCR modified the Administratiou 13uildir.g room arrangement in the RCA exit arca, '
added two new exit contamination monitors and a self-contained heating, ventilating and cir-conditioning (HVAC) system fcr the relocated Count Room.
PURPOSit The purpose of this DCR was to imptove the efficiency of the RCA check point and enhance the ability to expeditioesly process personnel exiting the RCA.
SAFIrrY EVAI,UATION
SUMMARY
- A safety evaluation was performed for this design change. '
The safety evaluation determined that the design change affected several Figures of the FS AR and made minor correction; to the description of the Conat Room Air Conditioning System in the FSAR. . The modificatious did not affect safety-related systems. . The safety evaluation concluded that the design change did not ceate an unreviewed -safety question. ,
i FCR 91-tJ05-i.
l~
3 1
4 g.
I a
- -- - , a a r ~ . ~.,e.,+ .. _ , , w<, a . . . . , ,. r y- r
l_
DPsiGN COORDINATION REPORT: Number 90-025 Tm.E: Main Control Board (MCB) Alarms for Turbine Runback / Setback! Trip
SUMMARY
DESCRIrrlON: This Design Coordination Report (DCR) made numerous changes.
M ain Control Board (MCB) indication and alarm of inputs relating to Main Turbine runback, setback and trQ were added. Time delays were added to Main Turbine setback signals originated by Condensate Pump logic and Isolated Phase Bus Doct Cooling logic. A Main Turbine Serback signal in response to a trip of either Main Feed Pump was added. Miscellaneous corrections to electrical drawings were also made.
PURPOSE: Main Control Board indication and alarm of inputs relating to Main Turbine runback, setback and trip were added to enhance the ability of the operators to assess plant conditicas following Main Turbine automatic action. The time delays added to the Main Turbine etback signals originated by Condensate Pump logic and Isolated Phase Bus Duct Cooling logic block unwant-d Main Turbine setback during startup of standby equipment. The Main Tutbine setback in response to a trip of either Main Feed Pump reduces the potential for low ster.m generator level and consequent reactor trip following trip
. of either Main Feed Pump.
1 S AFETY EVA1.uATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation determined that the design changes are enhancements that do not adversely affect systems or equipment. The DCR affected a Figure of the FSAR. The safety evaluation concluded that the design change did not create an unreviev ed safety question.
FCR 90-106 31
B J DESIGN COORDtNATION REPORT: Nu ruber 90-030 Trn.E; High Energj Drain Line to Condenser
SUMMARY
DESCRil"flON: This Design Coordination Report (DCR) upgraded the design of the hiain Condenser. Design changes included: a) the addition of flow control orifices to the warm up, main steam and steam chest drain lines to limit the energy level of finid entering the. Main Condenser to moderate levels; b) the addition of internal condenser shields and baffles to protect the condenser tubes from impingement and erosion damage; c) size reduction of valve htSD-V52 and line (wbich bypasses steam from the gland steam supply header to the condenser) to limit the energy
-level of fluid entering the blain Condenser throegh this line; d) changes in tube stake design to increase stake effectiveness; e) removal or abanJonment-in-place of unused Main Condenser instrumentation taps and connection fixtures.
PURPOSE: The purpose of this DCR was to address the recommendations of EPRI Report CS-2251 regarding the admission of high energy fluids to the Main Condenser, and to incorporate other Main Condenser design enhancements.
SAFETY EVALUATION
SUMMARY
- A safety evaluation was performed for this design change.
l The safety evaluation states that the design change affects a Figure of the FSAR but does not affect the function of the systems as described in the FS A R. The safety evaluation concluded that the DCR did not create an unreviewed safety
- question.
TCR 91-006 l
l l
32 l
des 1GN COORDINNrlON RUPORT: Number 90-031 TrrLit: Add Fixed Camera for Gate 1 and VHF Receiver at S AS S UMM ARY D ESCR11"rION: Details of this design modification are not provided in this report since they may involve safeguards information.
PURPOSE: The purpose of this de.dgn modification is not stated in this report since it might involve safeguards information.
\
i SAITirY EVALUATION
SUMMARY
- A safety evaluation was performed for this design change.
Details of the safety evaluation for this design modification are not provided in thi.; report since they may involve safeguards information. The safety evaluation concluded that the DCR did not create an unreviewed safety question.
4 4
I h
33
s .
l l
DESIGN COORDINNr10N REPoltT: N u tu ber 90-032 4
TITt.tc Emergency Feedwater (EFW) Pump Turbine 5 team Supply Containment Isolation Valve Change S U MM ARY DP. SCRIPT 10N: This Design Coordination Report (DCR) converted the EFW pump turbine steam supply /GDC 57 Containment isolation valves (M S- V 127 and MS-V128) to manually operated valves, upgraded downstream manually operated valves (MS-V393 and MS-V394) to Safety Class 2, GDC 57 Containment ivlation valves, and upgraded the intervening pipine to Safety Class 2. The DCR extended the GDC 57 Containment isolation barrier for the EFW pump turbine branch headers downstream to pneumatically operated globe valves MS V393, MS V394 and their manual bypass valver The DCR replaced the obsolete pneumatic operators on gate salves MS-V127 and MS-V128 with gear-operated manual at:uators. Tbc DCR re routed drain lines, provided nitrogen gas back-up to the pneumatic actuators for GDC 57 Containment isolation valves MS-V393 and MS-V394, changed the EFW System inoperable alarms and made other miscellaneous changer PURPOSE: The pneumatic operators for valves MS-V127 and MS Y12S were causing the valves to stick in the closed position. Spare parts foi hese obsolete operators were not readily available. The DCR cohanced operational ret. ability at minimum cost by transferring the GDC 57 rontainment isolation function to downstream pneumatically-operated globe valves MS V393, MS-V394. Maintenance flexibility is enhanced by the ability to use MS-V127 and MS V128 as manual uolation valves for MS-V393 and MS-V394.
M SArtrIY Eval.UATION SUMMAkY: A safety evaluation was performed for this design change.
The safety evaluation applicability review deterrnined shat the DCR made changes in the facility as described in the FSAR. The safety evaluation determined that the revised configuration of the EFW steam supply /GDC 57 Containment isolation valves meets regulatory criteria for closed systems. The EFW turbine steam supply system is unaffected. Design modifications met original design requirements. The safety evaluation concluded that the DCR did not create an unreviewed safety question.
FCR 91-011 34
l s -e t
DL1 SIGN COORDINATION REPOR'T: Number 90-041 TITt.E: Steam Generator Level Channels Filter Card Addition and Set Point Program Deletion SUMM ARY DESCRIPTION: This Design Coordination Report (DCR) added eight lead!!ag circuit cards configured for a lag function to the output of the steam generator (SG) narrow range level transmitters which share the same tap as the steam flow transmitters. The DCR also added series relay circuit cards to allow routine testing of S/G level bistable trip _ values without the use of jumpers and circuit modifications.
Finally, it removed the S/G level set point circuitry.
PURPOSH: The purpose of lag function is te prevent pressure waves generated in the impulse lines of the steam flow transmitters from creating talse high high cr low low S/G level signals in transmitters which share the same tap. Such false signals' could unnecessarily isolate Main Feedwater or trip the reactor. Removal of the level setpoint program circuitry eliminated a potential, multiple loop feedwater malfunction.
SAFErrY EVALUNTION SUM:AARY: A safety evaluation was performed for this design change.
The safety evaluation applicability review determined that the _DCR made changes in the facility as described in the FSAR. The safety evaluation determined that the _ replacement circuit cards were Class 1E equipment uned for protective functions. The change was evaluated for its effect on the safety analysis. Pctential failures i of the added equipment were evaluated. The safety evaluation contluded that the DCR did '
not create an vnteviewed safety question.
FCR 90-112 35 r F
s a DESIGN COORDINATION RIIPORT: Number 90-042 Trrt.11: Alternate Spent Fuel Pool Cooling
SUMMARY
Dt!SCRIFFION: N o r mally, heat from spent fuel in the Spent Fuel Pool is transferred by the Spent Fuel Pool Cooling (SF) System to the Primary Co mponent Cooling Water (PCCW) Sys'em. The PCCW System transfers heat to the Sersice Water (SW) System which transfers it to the ultimate heat sink, the Atlantic Ocean. During the first refueling outage, with spent fuel from Core No. I located in the Spent Fuel Pool, both Trains of PCCW were unavailable due to heat exchanger retubing (DCR 90-045). This Design Coordination Report (DCR) provided an alternative means of removing heat from the SF System. A third SF heat exchanger was installed in the Fuel Storage Building (FSB). This heat exchanger was cooled by a temporary, non-safety-related cooling tower located adjacent to the FSB. Piping was installed to make available backup ecoling capability, if needed, from either the Service Water Cooling Tower (preferred back-t:p) or the Atlantic Ocean via the SW System (contingency back-up)
P URPOSit The purpose of this DCR was to provide an acceptable method of cooling irradiated spent fuel in the Spent Fuel Pool while retubing both PCCW heat exchangers simultaneously (DCR 90-0045).
l i
S AFIrrY EVA1.UATION
SUMMARY
- A safety evaluation was performed for thie design change.
The safety evalution applicability review determined that the DCR made changes to the facility and procedures as described in the FS AR. The safety evaluatioc. identified the FSAR text, Tables and Figures affected by the design change. The safety evaluation determined that the Alternate Spent Fuel Pool Cooling System did not impact the limiting fuel handling accident described in the FSAR. Limitations on the system maintained credible accidents and resultant consequences within the bounds of those previously antlyzed in the FSAR. The safety evaluation concluded that the design change did not create an unreviewed safety question.
FCR 90-114 i
l l
36
- . - . . . . ~ . ~ . .- . . - - .- -- - . . . _ . . . ~ . . . - . - . - . _ -
- .s; '*'
DESIGN COORDINATION REPORT: Number 90-045 TITLit: Primary Component Cooling Water Heat Exchanger Retubc/Tubesheet Coating
SUMMARY
DESCR!!YION: This Design Coordination Report (DCR) performed an in kind replacement of the prematurely-degrading tubes of the Primary Component Cooling (PCCW) Heat Exchangers. This DCR also provided technical input for the addition of a protective coating to provide a corrosion barrier between the rolled jois of the tube and the tube sheet of the heat exchangers.
PURPOSE: The purpose of this DCR was to restore the cooling capacity of the PCCW beat exchangers to the originally-designed level, and to provide additional i- protection against damage from crosion/ corrosion.
t
! SAITTY EVALUNDON
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review determined that the design change affected the text of the FSAR. The safety evaluation determined that the design change _ upgraded the design standards of the heat exchanger and enhanced its corrosian resistance by the addition of .he protective coating. The safety evaluation concluded that the design change did not create an unreviewed safety question.
! -- FCR 90127 1:
l l
37 l-l
. _ ..-. . ~ . . . . . . _ - -. . .. .-. - .. . . _ - - _ - . . . . .. .
i .
DESIGN COORDINATION Rt!!' ORT: Number 90-047 Tm.c: Reactor Cavity Cleanup System Filter Installation S UM M ARY DESCRII"I1ON: This Design Coordination Report (DCR) replaced the Reactor Cavity Cleanup System's unshielded strainer with a variable cartridge shielded filter unit. An existing gate valve was also replaced with a plug valve.
_. Additional changes to the system valving and piping configuration were also made.
PURPOSin This design change was implemented to address ALARA concerns associated with the existing system.
SAFETY EVAL.UATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review determined that the design change affected the text and several Figures of the FSAR. The safety evaluation determined that the design change upgraded the design standards of the involved components.
The safety evaluation concluded that the design change did not create an unreviewed safety question.
FCR 91018 t
i e
d 4
. . - . .. . - ~ . . _ _ . . . . - _ - . - - . . _ - .- - - . ..-.-g.~-
s -e
)
DilSIGN COORDINATION Rl! PORT: Number 90-049.
TITI.11: Diesel Generator Jacket Water Temperature Control SUMM.ARY Dt!SCRWrlON: This Design Coordination Report (DCR) modified the response characteristics of the Diesel Generator jacket water pneumatic temperature controi system, and installed several design enhancements to diesel generator support systems. The design enhancements included a) fuel oil filter vent and drain connections; b) fuel oil strainer DP switch test connections; c) revised setpoint for the low fuel oil pressure switch; . d) revised rocker arm lube oil high level alarm Urcuit; e) protective covers for the pneumatic temperature and differential pressure controller .
adjustment knobs. The DCR also made editorial corrections to diesel generator drawings.
PURPOSII: The purpose of the DCR was to improve the performance of the Diesel Generator jacket water temperature control system, to improve maintenancc capabilities and alarm function reliability of diesel generator support systems, and to incorporate various enhancements into the design.
i S AITIY EVALUATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review determined that the design change affected the text and several Figures of the FSAR. The s .fety evaluation determined that the design change improved system performance, and concluded that the -
design change did not create an unreviewed safety question.
FCR 91-024 I
l-39
DEstGN COORDINNI'lON REPORT: Number 90-050 TITLE: Turbine Gcnrator Control Valve Test Bias Addition s
SUMMARY
DL'SCRitTION: This Design Coordination Report (DCR) added a control valve test bias (CVTB) circuit and a speed error filter (SEF) circuit to the Turbine Generator Electro. hydraulic Control (EHC) Cabinet.
PURPOSE: The purpose of this DCR was to permit weekly turbine-generator control valve surveillance testing without the necessity of entering the turbine-generator-EHC Cabinet to make voltage adjustments, which increases the risk of turbine /reactc,r trip. The CVTB circuit allows a bumpless transfer when the stage pressure signal is switched into and out of the. control valve amplifier circuit for testing.
SAlmTY EVALUATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review determined that i the FSAR was not directly affected in that the text, figures and tables in the FSAR did'not l' require changes. The safety evaluation determined tizat the design bases for the new circuit boards are the same as the design bases for the existia boards. Elimination of the need ,
to enter the EllC Cabinet and make voltage adjustments would reduce the probability of L
several Condition II events analyzed in the FSA'R, and thus enhance the margin of safety for '
4 these events. The safety evaluation concluded that the DCR did not create an unreviewed safety question.
e p
l-i l.
40
,y-. y , . - - = .y ,-- - r - - , ~ , .
j; DESIGN ' COORDINATION REPORT: Number 91-009 TIT 1.n; Feedwater Regulating' Valve Trim Change
SUMMARY
Dil5CRIPTION: This Design Coordination Report (DCR)-replaced the trim of the
.feedwater regulating' valves with a trim of balanced, single seat design. The replacement trim consists of the cage, cage spacer, plug and stem assembly, 4
baffle assembly, gaskets, 0-rings and various seals. Replact sent valve actuator components are also included within -the scope of the DCR. The replacement trim is a proven design with excellent performance at other nuclear -plants, This DCR also made permanent the temporary setpoint change which revised the differential pressure range for the M ain Feedwater Pump speed control program from 80-195 psid to 80165 psid. ,
PURPGsu: The purpose of this DCR was to eliminate. feedwater regulating valve stem oscillations experienced dur:ng the first operating cycle and, following trim replacement, to enable the valve to operate -in _ a more fully open position where valve hysteresis has less effect on flow cor. trol.
l- . SAITTY EVALUATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review determined that the UFSAR was not directly affected in that the text, figures and tables in the UFSAR did not require changes. The safety evaluation determined that the feedwater regulating valves are classified as non-nuclear safety equipment but. are important to safety based on their closing function in response to a . feedwater isolation . signal. The :nodified internal components of - the feedwater regulating valve will not change the originally-designed performance objectives of the valve or affect their ability to close in response to a feedwater isolation signal. The replacement components are procured in accordance with the original specification and purchase order for ASME Code Class 3' trim components with one authorized exception. The safety evaluation concluded that the DCR did not create an unreviewed safety _ question.
r-I r
t
! 41 l
l n
l l-
DestoN COORDINATION REPOnT: Number 91 011 Trrtz: Personnel Hatch Area Modifications F
i
SUMMARY
DESCRIPTION: This Design Coordination Report (DCR) reconfigured the ana immediately surrounding the Containment personnel batch. The DCR relocated a support, a lighting panel, a telephone jeck and two area radiation detectors.
i PURPGsn: The purpose of this DCR was to relocate items which significantly restricted access to the Containment personnel batch and thus enhanced the ability to move material and equipment through the Containment personnel batch.
i.- S AFITTY EVALUATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review determined that
' the UFS AR was not directly affected in that the text, figures and tables in the UFSAR did not require changes. The safety evaluation determined that the modifications did not affect any safety-related _ equipment. The relocated radiation detectors remained aligned with the personnel hatch.as described in the UFSAR. The safety evaluation concluded that the DCR did not create an unreviewed safety question.
p i
a 42 i
4
, ,.,-.e i _ . - , - - . - . . , - , . - , _
I t= . , :.
DESIGN COORDINATION RUPORT: Number 91-012 Tm.E: Reheater Drain Tank Level Controls P
SUMMARY
DESCRIPTION: This Design Coordination Report (DCR)' revised the design of the Reheater Drain Tank level control " trees." Design features which were provided by this design change included: a) Double valve isolation for each instrument; b) Welded connections; c) Separate electrical power sources for normal and high level dump controls;' d) Primary and back-up controls for normal and high level dump controls.
PURPOSn: The purpose of the design change was to improve equipment reliability and facilitate on line maintenance.
SAFLTY EVALUATION
SUMMARY
- A safety evaluation was performed for this d. sign change.
The safety evaluatioc st race that the design change affected a Figure of the FSAR but did not affect the function of .the systems as described in the FSARc The safety evaluation concluded that the derign cb .nxe did not create an unreviewed safety question.
1 - FCR 91048 l
1
.i.'
P J
43 i
s
,w- ,-c - y . ,-,. --. .-,-ny,.-- -. - , . , , _ - --, - - - - -
~. . . - . - - - . - . - - - -. . .. . . ~ -. - . -- .-. - - .- - - - - - . _ . .
e =
l DESIGN COORDINATION REPORT: Number 91-035 Trn.U: Feedwater Check Valve Modification
! St>MMARY DilSCRIPTION: This Design Coordination Report (DCR) modified the internals of -Main Feedwater System check valves FW-V330, FW-V331, FW-V332 and FW-V333. The diameter of the sixteen dash platellocking ring attachment bolts was increased from 3/8 inch to 5/8 inch. The DCR also included the necessary changes to the lock ring and dash plate to accommodate the larger diameter bolts. This DCR was implemented during the first refueling outage.
PURPOSit This DCR was a follow-up to Minor Modification (MMOD)91-529, which increased the number of dash plate / locking ring attachment bolts from the original eight to sixteen. The DCR prosided a more conservative redesign giving the dash plate / locking ring joint the. capability to withstand system design pressure and maximum L differential pressure across the dash plate in the opening direction.
SAFLTI*Y EVALUATION
SUMMARY
- ' A safety evaluation was performed for this design change.
, The safety evaluation applicability review determined that the UFSAR was not directly affected in that the text, figures and tables in the.UFSAR did not require changes. The safety evaluation determined the safety function of the feccIwater check valves to be a controlled closing function. The design change did not affect this safety function. It also did not affect the opening stroke. The conservative redesign increased bolt size and consequently reduced-individual bolt stress. This redesign was expected to eliminate further instances of bolt failure due to forces that occur during opening. . The safety evaluation concluded that the DCR did not create an unreviewed safety question.
l l-l lt l.
u
. , -~- . . - - -. . . . . . - . .- . .- .. . -... . . . . . ~ . . -
DESIGN COORDINA~ MON REPORT: Number 91-043 Trn.E: Coupling Capacitor Replacement Parts
SUMMARY
DESCRIPTION: This Design Coordination Report (DCR) evaluated the differences between aiench Electric Types TEC345 and TEV345 coupling
. capacitor voltage transformers (CCVTs) and General Electric Type CW30D. coupling l capacitors, It approved the' use of Trench Electric Types TEC345 and TLV345 coupling capacitor voltage transformers (CCVTs) as replacements for the existing General Electric Type CW30D coupling capacitors used in the 345 kv air termination yard as part of the Power 1.:ne Carrier (PLC) portion of the transmission line protective relaying schene,
' PURPOSE: The replacement CCVTs were required because General Electric Type CW30D coupling capacitors are no longer available from the original manufacturer. 4 p
i SAPLTY EVALUATION
SUMMARY
- A safety evaluation was performed for this design change.
The safety evaluation applicability review determined that -
the DCR made changes in the facility as described in the FSAR, although no FSAR revisions were required. The offsite power system and transmission line protection eheme, including the _ power line carrier equipment are classified nonsafety-related. The safety evaluation determined that the replacement CCVT did not change the function or operation of the affected systems and was an acceptable replacement for the G.E. Type CW30D coupling capacitor. The safety evaluation conclue :d that' the DCR did not create an unreviewed safety question. .
I i
45 l
l l
MINOR MODIFICATION: Number- 89-604 Bubbler Tube Matenal I TrrLE:
I i
I s
SUMMARY
DESCR!! TION: - Water level difference on either side of the traveling screens in the Circulating Water and Service Water Pump - bays is .
4 detected by a differential pressure system utilizing bubbler tubes. This Minor Modification (MMOD) replaced the monel bubbler tubes with polyethelene tubing.
PURPOSn: The modification provided bubbler tubes which were not susceptible to sea water corrosion.
SAFETY EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The 2
safety evaluation applicability review determined that the modifications made changes to the facility as described in the FSAR and identified the g affected FSAR Table. . The bubbler tubes are designata non safety class; those in the
, Service Water - Pump House are seismically-supported. The safety evaluation determined
- that the modifications did not affect the design or function of the Circulating Water (CW) or . Service Water (SW) System or affect other safety-related equipment. The safety l' - evaluation concluded that the MMOD did not create an unreviewed safety question.
i I
I' 1-l 46 l
, . . - ._ . . - . - . . . - - _ , . . - - . . .. . .~
=
s
' MINOR MODII'ICATION: ' Number 89-616 TrrII: Condensate Pump Demineralized Water Supply
SUMMARY
DESCRIPTION: This Minor Modification (MMOD) added an isolation valve and a pressure reducing valve to the backup seal water supply from the Demineralized Water (DM) System to the Condensate Pump seals.
PURPOSE: The modification was a design enhancemes provide a more accessible
- isolation valve and a pressure reducing valve to regulate the pressure of this ,
backup source of seal water for the Condensate Pumps.
SAPIrrY EVAL,UATION
SUMMARY
- A safety evaluatian was performed for this MMOD. The safety evaluation applicability review determined that the i modifications made changes to the facility as describ.d in the - FSAR and identified' the p- affected FSAR Figure. The affected portions of the- Condensate and DM Systems are .]
designated non-safety class, non-seismic. The safety evaluation determined that the modifications did not affect the design or function of the Condensate or DM System or '
affect safety-related equipment. The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
FCR 90-070 4
G P
l 47 m _-
s .
l MINOR MODIFICATION: Number 90-515 TrrLE Chain Hoist for North Residual Heat Removal (RH) \'ault
SUMMARY
DESCRII"rlON: This Minor. Modification (MMOD) provided a permanently installed electric chain hoist mounted on an existing monorail i located in the North RH Vault. Electrical power, restraints and supports associated with the hoist were also provided. The former design provided for the use of a hand-operated chain hoist during maintenance periods. The manual hoist was removed from the monorail when not in use.
PURPOSu: The purpose of the chain hoist was to assist maintenance personnel with lifting heavy objects in and out of the sault areas.
SAFETY EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modifications did not directly affect FSAR in that the text, figures and tables in the FSAR did not require changes The chain hoist is designated non-safety-related, but is seismically supported. The safety evaluation determined that the permanently-mounted, electric chain hoist is vithin the capacity of the monorail, meets seismic design requirements and satisfies bravy load criteria based on NUREG-0612. The safety evaluation concluded that the Af MOD did not create an unreviewed safety question.
t.
r h
I 48
a .
e- . ,
MINOR MODIFICARON: Number 90-519 TrrLE: Replacement of Current to Pneumatic Pressure (I/P) transducers for the Atmospheric Steam Dump Valves sSDVs)
SUMMARY
Dl!SCRIPTION: This Minor Modification (MMOD) replaced the I/P transdurers originally provided for the stcarn generator ASDVs with 1/P transducers manufactured by a different vendor. Lubsequently, as part of this same MMOD, the original type I!P transducers were re-installed on the ASDVs.
PURPOSE: The original type 1/P transducers were replaced with the expectation o!
eliminating excessive drift. The original type I/P transducers were subsequentil restored to the ASDVs because the replacement I/P transducers introduced unacceptable instability. The _ drift problem with the original type I/P transducers was minimized by utilizing calibration tolerances i
SAFiflY EVAL JADON
SUMMARY
- A safety evaluation was perfortned for this MMOD. The L safety evaluation applicability review determined that the l modification did not directly affect the FSAR in that the text, figures and tables in the
(- .FSAR did not require changes. The ASDVs are designated Safety Class 2 components. The I/P transducers are designated non-safety related, but are seismically supported. The safety evaluation determined that the replacement transducers conformed to the original design criteria and did not affect the function, operation or failure modes of-the ASD":. However, the final phase of the modification restored the original design; so, in the end, there was no change. The safety evaluation concluded that the MMOD did not create an unreviewed
- l. safety question.
l' j FCR 91-027 49 P
s .
K MINOR MoulPICA110N: Nomber 90 523 TrrLl!: Ser6 ice Wi.ter Yahes Strol.c Time itevition.
SUMM ARY Db5CR!riluN: This Minor Modification (MMOD) provided the basis to revise the stroke ttrue trquirements for r,ix motor opunted valves in the Service \ Dater (SW) System. The NilY Data Sherts for Motor and Air Operated Valves
(
and Dampers were resised to reflect this change.
L Punrosin Tne stroke time requirements for the affected valves were revised to provide margin for the Primary Component Cooling Water (PCCW) high terr.perature q trin delay such that spurious PCCW pump trips do not occur. The revised stroke time rep irements were consistent with analysis assumptions.
[
Sarta Y EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The '
safety evaluation applicability review determined that the modifications made changer, to the facility as described in the FS AR and identified the affected FSAR Table. Tba SW System is designated Safety Class 3, beistric Category 1.
The safety evalcation de. :ned that the revised stroke time requirements did not affect the design of function o affected valves or the SW Systr.m. The rafety evaluation concluded that the MMOL d not create an unreviewed safety question. ,
FCR 90 072 _
b I
r E
50 E"
l E
\ _ - - _ _ _ _ - _ - _ - _ _ _ _ - _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _
. - ~ - . ~ . - - -.- -. .~- - _ - --._.~.- .- - ,-- _ . .-_
a MINOR Moull'ICAT10N: Number 90 529 Tm.it Charging System instrumentation Modifications
SUMMARY
DIL4.RilTioN: This M!nor Modification (MMOD) provided several instrumentatior, and control changes affe ctin,; the Charging Subsystem of the Chemical and Volume Control (CS) S); tem. The cht.uges included a new capillary for a replacement level transmitter far the Volume Control Tank ( V CT),
rearrangement of the bellows and tubing / piping for the VCT level transmitiers, and addition of instrument suubtsers on process connections tiear the pressure indicators associated with the boric a:id transfer ;iumps and the boric acid filter.
l PURPOSu: The new capillary for the replacement VCT level transmitter was needed because the capillary for the replacement transmitter was aborter than that of the original transmitter. The purpose of the rearrangement of the bellows and tubing / piping for the VCT level transmitters was to eliminate low points which accumulate moisture which affected instrument accuracy. The purpose of the instrument sunbtus was to c!iminate pressure pulsations to the gages.
i-SAITTY EVALUA110N
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the ,
tuodifications did not directly affect FSAR in that (be text, figures and tablesin the FSAR did not require changes. The affected instruments, car 'laries, tubing and piping are 8 designated non safety related but are seismically mounted. The safety evaluation d'termined -
that - the modifications conformed to the original design - criteria - and did not affect the ,
function,' operation or failure modes of the affected instruments. The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
FCR 91027 ,
l I-h 51 .
tx i
i 9
, ,,,y..~ . . -, y ..-,w., ,, , w.~,,,.,<.,w.,,, .,.,..,.c y.. ., # ,, .,,n-1, , + . , , - _ , _ . . - , . . - ,-..-6 ,%.... . -~.
+
e l MINOR MOD 111CN110N: Number 90 $44 i
TITI.I': lissential and Non Ener:tial Swi . becur Room Access-SUM M ARY D l'.S C R1 t*I'lO N . This Minor MLA !ication (MMOD) tdded a doar between the Turbine liuilding and the Non essential Switchgear Room, it also added conc.ete ramps at this door and at an existing door to the Essential 5witchgear Room.
PUIJ Oste The purpese of the modifications was to f acilit ate the safe mmement of electrical test equipment between the affected rooms.
S Al'IrfY EVA1.UATION
SUMMARY
- A safet) evaluation was performed for this MMOD. Tbc safety esaluation applicability review determined that '.be modifications made changes to the facility as described in the FSAR and identified the affected FS AR Figures. The changes affect non safety-related, non seisn ic structures. The safety evaluation concluded that the MMOD did not create an unreview,d safety question.
FCR 90115 52
I MINOR MODil'jCATION: N u nibt r 90 515
- 1111.1 : litoudown Ilash Tank Subcooling injection Line S U M M AR'i DILNCRJITION This M..ior Modification (MMOD) added a water injection line from the Demin ralized Water ( D h') bystem to the Steem Generator filuwdown System Flash Tank. The MMOD also provided the basis f or increaung the normal opeinting level of the Blowdown Flash lant.. Finally, it :nadt othei miscellaneous changer. to Ilosh Tank hardware, instrumentation and docntnentetiot PU KI*osti: 5 team generator blowdown flow rate was restricted to 60 percent ut d. sien blowdown flow rate because, at blowdown flow rates above this rate, flasling was occurring at the flar.h tank outlet. The water injection line was added to the 1110wdowa Flash Tank to enhance subcooling margin.
S Arrrt EVAt.UATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modifications inade changes to the facility as described in the FSAR and identified the affected FSAR Figure. The DM System piping and injection nozzle are designated non-safety class. The safety evaluation determiner. that the MMOD enhanced the ability to a chies e the design blowdown flow rate and therefore maintain proper sicam ge n:: tat or
-hemistry control. The changes to non safety class piping did not affect safety-related systems. The safety evaluation concluded that the MMOD did not create an unreviewed r,alety question.
FCR 90 024 1
53
m m._.. . _ _ _ . , _ _ - _ _ _ _ . - . _ ._, _ _ _ _ _ _ _ . - . _ . - --
, 4, i
MINOR Mot >II1 cation: Number 90538 TITI.tu A ldition of Hypass/ Test Connection for the Emergency Feedwater (EFW) bystem Cross Tic Check Valves ,
4
SUMMARY
DilSCRil'IlcN: This Minor Modification (MMOD) added a bypass / test conner. tion a to the EFW Systern to permit backflow testing of two EFW ,
System check valves.
Pulu'ose The two EFW System check valves are required by the ASME Code, Section Xi to be backflow tested. This MMOD provided the means to conduct the testing.
SArtnY FVA1,UATION
SUMMARY
- A safety evatuation was performed for this MMOD. The ,
l safety evaluation applicability review determined that the modifications made changer to the facility as described in the FSAR and identified the affected FSAR Figure. A portion of the piping and valving of this MMon is designated #
Safety Class 3, Seismic Category 1; and a portion is desigaated non safety class. The safety evaluation determined that the piping and valves which were added resulted in a non-
' func.ional change to the EFW Systemi The safety evaluation concluded that the MMOD did not create an unreviewed r,afety question.
j FCR 90-101 i ,
I l'
l l
t t
p -- - .
1 i
e e MINon Motsil'ICAT10N: f,' u m b e r 90 561 1111.!!- Renm al of Signa! Memory Functwn to the Main Fecdwain Pumps
$UMMARY 1:11 SCRIP 110ts: This Minor Modification (MMOD) disabled the signal mer..ory furction (SMF) associated with the Main Fredwater Pumps speed control system This control feature is de.,igned to lock in a previous control signal when the actual control signnt is lost.
PURPosit The SMF feature was disabled because it would not functwn as intended in c the Seaorook Station application. If activated, the SMF feature weald probably reduce ratter than maintain the speed of the Main Feedwater Pump. Furthermore, if actisuted, thia feature would not provide an alarm or other direct indication to Ibc Control R oont that is was actuated
.u S AFETY F.VAUIATioN
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modifications did not directly affect FSAR in that the text, figures ind tables in the FSAR did not require changes. The affected cotitrol circuitry is designated non safety related equipment. The safety evaluation determined that this MMOD improved the reliability of the Main Feedwater Pump speed control circuit, and did not adversely affect the it.nction or operation of the Feedwater Systen. or safety related equipment. The safety evaluation concluded that the MMOD did not crcate an unreviewed safety question. _
_ _. _ _ . _ _ _ _ _ _. _ . _ . -.~ -.._ .____. _ _. _ ._. - _ _ _ _ _ ...__ _ __._ .. _ _ . _ . _
i e s^
\
l MINOR Morme.:ATION: N u rnber 90 567 e
i Tr ra_ti: Steam Generator Blowdewn (Sil) Sample Tubing Personnel Protection I i
SUMM Aav DLECR!l"110N: This Minot Modificativa (MMOD) provided insulation and .
enclosures around segments of SB System sample tubing in the Pritrwry Auxiliary Building (PAB).
1 Pukrust!: The perpose of the insulatiot. and enclosures was to protect personnel from injury resulting from cenract with the bot tubing.
Sart rY EVAL.tJATION $L'MMARY: A safety evaluation was performed for this MMOD.. The safety evaluation applicabiluy review determined that the modifications did not directly affect FSAR in that tL. text, figures and tables in the FSAR did not require changes. The safety evaluation determined that this MMOD provided personnel injury protection and not affect the function or operation of the SB Sptem and-did not adverscay ificct safet) related equ pment. The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
J q
56 p
.-,.e ,w< - , -
r-,,, -
,wr -.-m,-,-,,, -w-,
e a MINon MODIPICATION: dumber 90 582 Tritn: Service Air Float Trap Substitution 4
SUMMARY
DESCRil' TION: This Minor Modifiervion (MMOD) replaced the float traps originally installed on the service air compressors and ibc air receiver tanks. Minor piping and vahe modifications were reade to facilitate installation of the substitute traps.
Puntosu: The original float traps are no longer manufactured. Thus, replacement traps and spate parta, are not available. The purpose of the MMOD was to provide the engineering basis for an equivalent replacement.
SArtrrY EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review cetermined that the modifications mad: changes to the facility as described in the FSAR and identified the affected FSAR Figure. The replacement float trap is designated non safety related u.n nan. seismic. The safety evaluation determined that the replacernent float trap was identical in function as the original and differed only in form, The replacement float trap did not change the performance or operation of the Instrument Air System iu any way that would affect the analysis or conclusions of the FSAR. Therefore, the substitute float trap was determined to be an accepte.ble replacement for this application. The safety eva;untion concluded that tbc MMOD did not create an unreviewed safety question.
FCR 90 050 I
l o
I:
57 L
I l
l.
. . . -, , , _ , , , _ - . , , , , , , ~ . . , _ , _ .
_,.,_,,,,,..,,._.,__m. , _ . , _ , , , _ . . , , , , , , . , . . . _ . . , _ . , , _ . _ , , , , , , , _ ,
e
- Mm9n MoDW3CA110N: Number 90586 TITt 1:: Update of Drawings for the Primary Sample Panel i
SUMMARY
DP.SCR11"I1ON: This Minor Modificction (MMOD) updated drawings associated with the Primary Sample Panel to refleet as-built conditions. It also provided new and/or replacement tag nameplates and replaced pressure indicators associated with the vacuum pump.
Punrosit The draring apdates were performed in onder to document and lobcl the as-built condition of the Primary Sample Panel, inte rfacing salves, piping and fittings. Replacement of the pressure indicators associated with the vacuum pump f acilitated the use of commonly stocked rather than unique components.
SArrrY EVAL.UA'!10N
SUMMARY
- A sately evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modificatians made changes to the facility as described in the FSAR and identified tb-affected FS AR Figure. The Primary Sample Panel is designated non safety class, non-seismic. The safety evaluation detc.rmined that the drawing updates and pressure indicator replacements were enhancements and safety related equipment would not be adversely affected. The safety evaluation concluded that the MMOD did not cicate an unreviewed safety question.
FCR 90-047 r
b 38
-.-. . -.- - ..---- . - - . .- . - _ . . - - - - . . . - ~ ._,- - -. - - . - .. - . .
l i
l l
MINOR MODIFICATION: Number 90-$S8 i TITLin Condensate Storage Tank (CST) Temperature Contrel
SUMMARY
DiiSCR11410N: This Minor Modification (MMOD) revised the main steam line break analysir, to permit a lower CST temperature operating band and implemented various design enhancements to tbc CST teruperaturt contsof sy, tem.
The enhancements included local CST temperature indicatore, CST temperature indication in- the Control Room based on direct CST temperature measurement, revised CST temperature alarm setpoints, and CJT temperature controller adjustments.
I Punrosit The purpose of the main steam line break ana!ysis revision was to lower the j ast,umed minimum enthalpy of condensNe supplied to the Emergency Feedwater j (EFW) System to the steam generators. This revision enabled the operating range of CST i water temperature to be more realistic and within the design range of the temperature control system. The purpose of the various hardware and setpoint enhancements was to l improvc automatic CST temperature control and increasc the operator's ability to monitor ;
CST temperature, ,
1 SAPIGY EVALUAT10N
SUMMARY
- A safety evaluation was performed for this MMOD. The i safuty evaluation applicability review determined that the -
modifications made changes to the facility as described in the- FSAR and identified tbc !
affected FSAR Sections. Affected hardware is designated non safoty class, non-seismic.
The safety evaluation determined that the reduction in the operating temperaturc rangt of >
the CST would r.ot adverse!y affect the ability of the EFW- Systern to perform its sa.fety functions. The safety evaluation also determined that the hardware and setpoint changes of this MMOD enhanced CST temperature monitoring capability and automatic control, and did not adversely affect safety related equipment. The safety evaluation concluded that ti.e '!
MMOD 'did oc,t create an unreviewed safety question.
FCR 90 097
'I l
59 4
e-- e--c-- -we. v -s-= . .m4- , -
, -,..h-+- -..m4- ,.w- w w -w ---+-w. er<vv--,ar-,++ e-e--e m . .-~we m-r-----vn w e - n wwe,=e-ao---v rwe m -,e *vwv 7-vew -+m+-
MINOR MOLLIFICATION: Number 90 597 ITitt:: Post Accident Sampling System (PASS) Panel Modification, SUMMAny DESCRII' TION: This Minor Modification (MMOD) added an inlet filteritrap and check valve to the PASS Panc! vacuum pump suction line.
~~
PURPost!: "Ibe purpose of the modifications was to remove moisture from the sample to ensure the c apability of the PASS to analyze Sample boten concentrations within specified accuracy. The check valve also protects against the possible back tiow of oil from the vacuum pump to the expansion cylinders. ,
S AIT!T EVALUATION
SUMMARY
- A safety evaluation was perfortned for this MMOD. The safety evaluation applicability review determined that the modifications made changes to the facility as described in the FSAR ani identified tbc aff ected FSAR Figure. The Pass Panel is designated non safety class and ron seismic. The safety evaluation determined that this MMOD enhanced system design and did not adsersely affect safety-related equipment. The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
FCR 91007 l
h
-_ . . . . - . _ . . . - - . _ . . ~ . . . . _ . - _ _ . - . . - _ _ - . . _ - _ _ . . - _ _ . . . ~ - _ . .
I MsNOR Monti'IcrrlON: Number - 90 599 Ti rl.U: Auxiliary Boiler Feed Pump Mechanical Seals i
SUMM ARY Di%CRll'rlON: "t his Minor Modification (MMOD) replaced the gland seals and '
packing of the Auxiliary Boiler feedwater pumps with single mechanical seals. ,
PURPOSI': The original gland seals and packing for the Auxiliary Doller feedwater pumps have had a history of packing failures, and have required a high amount of maintenance. The replacement-mechanical seals are expected to reduce the failures and the amount of required maintenance.
SA!'UfY EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the n.odifications did not directly affect FSAR in that the text, figures and tables in the FSAR did not require changes. The Auxiliary Doiler feedwater pumps mechanical seals are designated non-safvtv-related, non seismic. The safety evaluation determined that this MMOD enhanced system design and did not adversely affect safety related equipment. The safety evaluation concluded that the MMOD did not create an unreviewed safety question, l
l l
1
?
61 l.
1.... . - . . . ~ . - . - - - _ . - - . - . . - . -. - . ~ - . - . . . . . . - ., - .- - - . - .
e MINOR f,10t)ll*1 CATION: Number 90 600 Til't.it Resise Magnetic Switch for Door P415
SUMMARY
DilS(lRl! TION: Details of this minor niodification are not provided in this report since they might involve r,afeguards inf ormation.
PURrOsti: 'T he purpose of this minor modification is not stated in this report since it mit .ht involve safeguards inf ormation.
S al*trfY l'v A1.UATION
SUMMARY
. A safety evaluation w as performed for this MMOD.
Details of the safety esaluation for this design niodification are not provided in this report since they might involve safeguards information. The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
62
- _ - . ._.- -.- -. . . - . .._ - . .. .~. -.. . . ~ - - . - -
e e MINOR MoniricATioN: Number 90 018 Trrt.!!: Miscellaneous Tagging and Label Changes SUM uY DFACRil' TION: This Minor Modification (MMOD) made miscellt aus docurnent and label updates. Tagging and/or label information was added so drawings, documents and panels associated with the Loose Parts Monitoring System. The MMOD also corrected a vendor manual and FSAR error relating to this system. The changes affected documentatior; there were no hardware changes. other than adding new and/or replaccruent tag nameplates.
PURPOS!!: The drawing, document und FSAR updates were performed in order to document and label the as-built condition of the Loose Parts Monitoring (LPM)
System. ,
SAPIITY I3 ALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the MMOD mado document changes affecting the FSAR. The LPM System is designated non-safety-related, but is designed to withstand an Operational Basis Earthquake. The safety evaluation determMed that the MMOD made document and tagging / label changes. It did not substantially alter punt equipment, and therefore did not affect the function or operation of the LPM System or . safety related equipment. The safety evaluation concluded that the MMOD did not create an unreviewed safety question, FCR 90 061
- 63 d
- , -.. , , - , , , ,. - n , -- , oe -m
e-
- MINOR MODIFICATION: Nuenber 90-622 j l
l l
TITLin Modification to Turbint-Generator Setback Signal Logic
SUMMARY
DESCRIPTION: This Minor Modification (MMOD) added time delays to two parameters which produce a turbine-generator load setback signal.
A .bree second time delay was added to the Main Generator breaker coolleg banks load setb.ick signal. A thirty minute time delay was added to the Generator Step-up (GSU)
Transformer Cooling System load setback signal. The MMOD also added a three second 6 time delay to the Main Control Room and local alarms initiated by the generator breaker cooling banks overload circuit. !
l l
l
-PURroste The purpose of the time delays added to the turbine. generator load setback
-signal and the alarm circuitt was to pievent unnecessary load setbacks and alarms resulting from the automt. tic startup or switchover of cooling fans in the GSU Transfor.ner and Main Generator breaker cooling systems.
t i
S APINY EVALUATION
SUMMARY
- A safety evaluation was perfortned for this MMOD. The safety evaluation applicability review determined tbst the modifications did not directly affect FSAR in that the text. figures and tables in the FSAR did not require changes. The GSU Transformer and Main Generator Breaker cooling systems are designated non-safety related, no* seismic. Tbc safety evaluation determined that the EHC modifications were design enhancements and did not adversely affect safety-related equipment. The safety evale don ccecluded that the MMOD did not create an uriteviewed safety que:: tion. ,
4 P
i
.-Is - , . , ---,,,.,,-, ,, ,,-,.....,,,_.,y.+,, , , ,v-, ...,w-., , 4.
. -. - - - . - - . . . _ _ - . . - - - . - . - _ _ - ~ - . . .,.
e
- P MINOR MODil'1 CATION: Number 90 624 I
l l
Tin.It Update of Drawings for Steam Generator Sample Chiller Unit Skid ,
1 i
SUMMARY
DI!SCR11'110N: This Minor Modification (MMOD) updated drawings associrted with the Steam Generator Sample Chille Unit Skid and four, associated Radiation Monitoring Skids to reflect as built conditions and provided new and/o; replacement tag nameplates. The changes affected documentation; there were no hardware changes, other than adding new and/or replacement tag nameplates PURPOSil: The drawing updates were performed in order to docurnent and label the as built condition of the Steam Generator Sample Chiller Unit Skid and four Radiation Monitoring Skids, interfacing valves, piping and fittings.
SMrTY EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modifications made changes to the facility as described in the 2c AR and identified the aficcted FSAR Figeres. The Steam Generator Sample Chiller Us - Skid and four Radiation Monitoring Skids are designated non-Safety Class. The safety evaluation determined that l the drawing updates were enhancetnents and safety related equipment would not adversely ,
affected, since no hardware changes, other than the addition or replacement of a nameplate were made. The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
FCR 90 073 t
i 65
- - - - _ - - . - - .. - _ , , . . - ~ . _
. ~ . . - _ .. - ...- - _ _ ..-._- - . -.-_.- - ..-.__ _ .-.
e
- MINOR MODIFICATION: Number 90 635 Tirt.it Electro Hydraulic Cont +ol (EHC) System Trip Solenoid Valve
,$ U M M A RY Dl!SCRIF110N: This Minor Modification (MMOD) replaced the original, two coil type electric trip solenoid valve (ETSV) with an improved, finned single coil type ETSV. The MMOD also added a manifold strainer.
Puneositi The replacement ETSV is an improved design developed by the Turbine-Generator manuf acturer. It has been demonstrated to be more reliable than the two coil type. The manifold strainer is also expected to improve system reliability.
S Artrry EVAI.UA*I1ON
SUMMARY
- A safety evaluation was performed for this MMOD- The safety evaluation applicability review determined that the modifications did not directly affect FSAR in that the text, figures and 1 ables in the FSAR did not require changes. The EHC System is designated non-safety-related, non seismic.
The safety evaluation determined that the EHC modifications were design enhancements and did not adversely affect safety-nlated equipment. The safety evaluation concluded that the MMOD did not create an unreviewed safety question, f
i
. , - - - - . -, ~. . _ , . . - . . - . . . --,..,,.c.- . - . . _ . , , . _ _ . . . , . - . . _ , . . . .,-,,,,.-,--,,.,_,.~..,.cr.,
-- . . _ . . . - _ ~ - - . . . . . - _ - . . - . - . - - - .- -- ..-
I' s
t MINOR MODII'ICA'110N: Number 90 638 Triu:- Alarm System Enhancements and Lens Engraving SUMM ARY DILsCR11'flON. This Minor Modification (MMOD) made miscellaneous changes to the Video Alarm System (VAS) alarm circuits and Main Control Board (MCB) status monitoring light lenses. The modifications to the VAS were to correct drawing errors, enhance alarm logic, and ievise alarm setpoints. The modification, to the status monitoring light lenses were to provide more appropriate terminology, Changes affected alarms and status monitoring light lenses for parameters in the Reactor Coolant (RC), Residual Heat Removal (Ril), Service Water (SW), Containment Enclosure Air Handling (EAH) and the Fuel Storage Building Air liandling (FAH) Systems.
Punroste The purpose of the modifications was to enhance various alarms and status lights and to correct miscellaneous discrepancies in drawings.
SArLT; EVAL.UATION
SUMMARY
- A safety evaluation was performed for this MMOD. The l safety evaluation applicability review determined that the modifications made no changes to the facility as described in the FSAR except for a minor changes affecting several FSAR Figures. The VAS is designated non safety related. The safety evaluation determined that the modifications we e enhancements which would not adversely affect safety related equipment. The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
FCR 90118 l 67
-- . - . . - _ - - .~. . . - . . . - . _ . . - _ - - - ~ . _ _ . . - - - ~ . - . - _ . _ . - . - .
- l l
l MINOR MODIl1CA110N: Number 90 644 Tiltte Feedwater Heater Relief Valves 1
- StJMMARY DESCRW110N: This hiinor Modification (MMOD) replaced the Feedwater Heater tube side relief valves with relief valves of a balanced bellows design, revised the setpoint for the new relief valves and directed their discharge to the fer' vater heater drain piping.
PURPOSic The original design directed the discharge of the Ferdwater Heater tube. side relief valves to an open drain in the Turbine Building Heater Bay. In the event of relief valve actuation, hot water and steam would be released to the Turbine Building, creating a personnel hazard. This MMOD directs the discharge of the relief valves to a closed piping system to climinate the personnel safety hazard. The use of relief valves of balanced bellows design is appropriate for the variable back pressure application.
SAPIrrY EVALUA110N StJMMARY: A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modifications did not directly affect FSAR in that the text, figures and tables in the FSAR did not require changes. The affected portions of the Feedwater, Condensate and Heater Drain Systems are designated non-safety related, non seismic. The safety evaluation determined that this MMOD enhanced system design and did not adversely affect sa'ety-related equipment. The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
l l
l t .
~
1:
t
e .
MINOR MODIFICA*I1ON: Nu rn ber 90-647 Tritu: Feedwater Pump Instrumentation 1
l
, StJMMARY DESCR11"r10N: This Minor Modification (MMOD) made several instrumentation changes affecting the Main Feedwater Pumps. The low suction pressure trip setpoint was reset from 230 p.ig to 210 psig. The feedwater header high pressure alarm was increased from 1185 psig to 1285 psig. Main feedwater pump discharge pressure switches were replaced with switches having a greater range.
I I
l PURPOSE: The purpose of the setpoint changes was to provide mo : margin between the normal operating point of the parameter and the alarm or trip setpoint so that unnecessary trips arid alarms will not result from operational transients. The replacement pressure switches will produce more reliable operation.
SAIYTY EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the
'.iodifications did not make changes to the facility as described in the FSAR except for a i minor change affecting one FSAR Figure. The affected instrumentation is non-safety-related, The safety evaluation determined that the modifications were enhancements which would not adversely affect safety related equipment. The safety evaluation concluded that the MMOD did not create an unreviewed safety question. i FCR 91025 l
l l
69 L
. _ . . . , . . ~ . - . , . . - . .-. ,,.-,,. ,--,-... - , _ ---. -.,. - . - - , - . . , - - .
. = = _ . . . . _ - _ _ - . - _ - - - - - _ . - _ ~ . _- .. -. _
MINOR MODil'ICATION: Number 90 648 Triu!: Update of Drawings for Radiation Monitor Slids
SUMMARY
DESCR'f'rlON: This Minor Modification (MMOD) updated drawings associated with ten Radiation Monitoring Skids to reflect as built conditions.
It also prosided new and/or replacement tag nameplates. The changes af fected documentation only; there were no hardware changes, other than adding new and/or replacement tag nameplates.
PURPGse: The drawing updates were pe r f orr.ie d in order to document the as built condition of the Radiation Monitoring Slids and interfacing valves, piping and ,
fit t ings.
SAtrirrY F. vat,UATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modifications made changes to the facility as described in the FSAR and identified the affected FSAR Figures. The Radiation Monitoring Skids are designated non Safety Class, non seismic except for 1-RM SKD 60 which is Seismic Category 1. The safety evaluation i determined that the drawing updates were enhancernents and safety related equipment would not be adversely affected, since no hardware changes, other than the addition or replacement of a nameplate were made. The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
FCR 90 089 l
l l
l 70
. . . . ~ __ .~ . . , .
o -O MINOR MODIFICATION: Number 90 657 Trn.it Miscellaneous 14uman Factor Changes on the Main Control Board
SUMMARY
DESCRIITloll: This Minor Modification (MMOD) provided miscellaneous modifications to the Main Control Board (MCB), Changes included annunciator lenses with revised engraving, a modified plexiglass cover for Site Area Emergency Alarta Controls, and a revision to the FSAR regarding the color of labels for '
Category 1 variables. The MMOD also included corrections to design documents.
PURrosu: The misce!!aneous changes to the MCB were made based on human factors considerations.
S AITiry EVA1,UATION
SUMMARY
- A safety evaluation was performed for this MMOD. The 4 safety evaluation applicability review determined that the modifications tnade changes to the facility as described in the FSAR and identified the affected FSAR Section. The MCB is Seismic Category 1 and contains Class IE and non Class IE electrical equipment. The safety evaluatica determined that the changes to the MCB were enhancements which would not adversely affect the function of the MCB. The I safety evaluation concluded that the MMOD did not create an unreviewed safety question.
FCR 90-123 1.
71 l.
m.,-s-- m. .. - . , - , , - - - - - - .+,y -, m -. ii-v-w..- ,# -- w %- y e-.- i-
+ - w a n. g.- - p ,. ....,,-..wwv--_vw----
e e j $11NOR h10DI/ICATION: Number 90-664 Tril.it: Fire Panel ! P-CP-4h9 Annunciator biodifications i
SUMMARY
D13CRIPTION: This hiinor hiodification (hth10D) rewired the internal circuitry of Fire Panel FP CP-409 such that a trouble condition within the panel itself (i.e. loss of primary power) would be annunciated in the hiain Contiol Room.
PURPGst2: The original design provided local indication at the panel of trouble conditions within Fire Panel FP CP-409 The h1h10D was implemented to preside Control Room annunciation of trouble conditions within Fire Panel FP-CP-409.
Sn!'ETY EVALUATION SUMMA.RY: A safety evaluation was performed for this hih10D. The safety evaluation applicability review determined that the t-odifications did not directly affect tbc FSAR in that the text, figures and tables in the FSAR did not require changes. Electrical equipment in the Fire Protection System is not designated Class 1E. The safety evaluation determined that this change to panel trouble detection circuitry would not alter the function or operation of the Fire Protection System and would not adversely affect safety related equipment. The safety evaluation concluded that the hih10D did not create an unreviewed safety question.
72
_ . . ~_ . _ . _ _ _ _ _ . . _ . . - . __.._._..-. - _.__.._._._ _ . -
. . _ ._ _ _ . _ . _ . m_. __ __ ._
e
- h!INOR h!ODIFICATION: Number 90 669 Trnx: Fuseblock Insulator Installation I
SUMMARY
Dl!SCR!!"flON: This Minor Modification (MMOD) installed glass epoxy insulators beneath two typc3 of fuseblocks utilized in the Isolation Relay '
Cabinet and the Main Steam Isolation Valve (\1SIV) Logic Cabinets.
PURPOS!!: The MMOD was implemented in response to the recommendations of the Cabinet manufacturer to correct a potentia: problem. The manufacturer's lettes described the possibility of arcing or current leakage from the fuse clip on the assembly to the structure on which these type of fuseblocks are mounted.
SAMTTY EVALUATION
SUMMARY
- A safety evaluation was pctformed for this MMOD. The 4
safety evaluation applicability review deterati..ed that the modifications did not directly affeu FSAR in that the text, figures and tables in the FSAR did not require changes The affected Cabinets are designated Class 1E equipment. The
' safety evalcation determined that the added losulators conformed to applicable design criteria and would not affect the operation or function of any equipment or system. The safety evaluation concluded that the MMOD did not create an unreviewed safety question, n
l l
73 l ..
l
- y w,. ..,r - wc-m-, ,--d-, -
y, e--%- y m p ww y-wwv. m rwww. y--+,---- ---y-s. e-me-e w e i wp --tw ,,v.- rw.- *e -' w4- sm- -'y-7 Y
a e MINOR MODI!7tcA110N: Number 90 670 Tritt!; Waste Processing Building (WPB) Waste Solidification Area Supply Fun Motor Replacement
SUMMARY
Dl3CRIPTION: This Minor Modification (MMOD) replaced the electrical motor for the WPB Solidification Area Supply Fan with a motor manufactured by a different vendor and having slightly different operating parameters than those of the original motor. The MMOD evaluated the application of the new motor,
, provided new electrical protection setpoints and updated the affected drawings.
P U RI OSE. The original electrical motor had failed; but an identical replacement was not available. The purpose of the MMOD was to provide the engineering basis for an equivalent replacement. ,
SAirtrrY EV!.LUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modifications did not directly affect the FSAR in that the text, figures and tables in the FSAR did not require cht.nges. The replacement electric motor is designated non safety related and non-seismic. The WPB Solidification Area Supply Fan does not perform a safety related function and does not affect safety-related equipment. The safety evaluation determined that the replacement motor was the same as the original in terms of fit, form
, and function. The replacement motor did not change fan perfortnance or operation of the WPB Ventilation System in any way that would affect the analysis or conclusions of the FS A R. Therefore, the replacement rootor was d:termined to be an acceptable replacement for this application. The safety evaluation concluded that the MMOD did not create an unt-eviewed safety question.
74 i
, <- , , - - ~ , -. e.-- ,,..
. . . . - . . . _ . _ . - _ _ _ - ~ . - . _ _ . _ . _ _ _ _ _ . . - . . . _ . . - - _ _ _ _ _ _ _ . . . _ _ . . _ _ _ .
4 0 4
h11NOR h10DIPICATION: Number 90-672 TITLii: Appendix R Report Revision l
l l
SUMM ARY DI!SCRIPTION: This Minor Modification (MMOD) updated the Seabrook Station Fire Protection of Safe Shutdow. ^apability ( Appendix R) Report. The Appendix R Report was also updated to include manual operation of the i Fire Protection water supply to the Service Air Compressors. This was a document i change only; no physical changes to equipment were made.
l PURPGst!: A postulated fire in the B Train Switchgear Room could theoretically disable the motor-driven and the turbine driven Emergency Feedwater Pumpt in the event of such a fire, the Start-up Feedwater Pump and its support equipment is available to provide emergency feedwater The Appendix R Report was revised to account for the consequences of a postulated fire in 0 Train Switchgear Room, S APInY EVAt.UATION
SUMMARY
- A safety eva.!uation was performed for this MMOD.
The safety evaluation applicability review determined that the MMOD made document changes affecting the Appendix R Report, which is part of the FSAR. The safety evaluation 9.etermined that the document changes did not affect the function or operatior of the affected systems or other safety related equipment. The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
FCR 90110 L
l f'
L 75 a, a._
. . - . - . - - _ _ -. . - - . ~ _-._ ._- -.- .---. . . . --- _
MINOR MODIFICATION: Number 90 674 TIT 1.!!: Demineralized Water (DM) and Instrument Air (I A) for the Temporary 1&C Hot Shop
SUMMARY
Dl!SCRIPTION: A modified construction office trailer located just cast of the o Fuel Storage Building (FSB) is being utilized as a Temporary
" Hot 11&C shop. This MMOD modified the DM and IA S) stems in the FSil and provided two penetrations t!' rough the wall of the FM to provide for the passage of two 1 inch pipes, one for demineralized water and (bc other for instrument air service to the Temporary Hot l&C Shop. This MMOD also documented a safety evaluation for the Temporary Hot !&C Shop itself.
P URPOSit: The purpose of the MMOD wa6 to provide dem.neralized water and instrument air seruces to the Temporary Hot l&C Shop, and to document '
a safety evaluation for the Temporary Hot I&C Shop itself.
S AI't!TY EVAI,UAT10N
SUMMARY
- Two safety evaluations were performed for this MMOD.
The first safety evaluation pertained to the modifications to the DM and IA Systems and the FSB wall. The safety evaluation applicability re' :w determined that the modifications did not directly affect the FSAR in that tE- text, fig: ires and tables in the FS AR did not require changes. The_ DM and IA Systems in the FSB are designated as non-safety class, Seismic Category 1. The FSB wall is a Seismic Category 1 structure and functions as a safety related ventilation boundary. The safety evaluation determined that the two pipe penetration. in the FSB cast wall did not impact fuel .
handling equipment, the fuel handling process or the ability of the FSB Air Cleaning System to maintain the required negative pressure. This modification is an enhancement to facility design. The safety evaluation also determined that the piping modifications to tne DM and IA Sysams did not adversely affect these systems or impact safety-related equipment. The s, ,ty evaluation concluded that the MMOD did not create an unreviewed safety- question.
The seca,d safety evaluation pertained to the Temporary Hot I&C Shop itself, The safety evaluation applicability review determined that the addition of a Terrporary Hot I&C Shop did not directly affect the FSAR in that the text, figures and tables in the FSAR did not-require changes. The Temporary Hot 1&C Shop is a refurbished and modified -
construction trailer located wathin the plant protected area. . Access to and from the l Temporary Hot I&C Shop is through the east FSB door and via at enclosed walkway.
The Temporary Hot I&C Shop is a radiologically-controlled area, The safety evaluation applicability review determined that all features of the Temporary Hot I&C Shop weie l- consistent with applicable descriptions and commitments of the FSAR. The safety
[
evaluation determined that the Temporary Hot I&C Shop did not adversely impact safety.
L related structures, systems' or components. The safety evaluation concluded that the addition of a Temporary Hot I&C Shop did not create an unreviewed safety _ question.
l 76 l
L - . - - . . __ _ ___._, __ _ , _
- . . _ .- - _ - .. _ . - - _ _ _ . - _ - -. ..__~.- - . . .. - - - - . -_-
MINOR MODII'ICA'!10N: Number 91-505 TITt.it: Fuel Handling Tool Structural Modifications SUMMnRY Di'.SCRl!" TION: This Minor Modifiustion (MMOD) made three changes to the Fuel Transfer System Dei n. It re-mounted the Pod Cluster Control (RCC) changing tot. support bracket at a i._,,ne r elevation. It lengthened the emergency pull cable for the fuel transfer drive system. Finally, it removed the comb lock assemblies from the burnable poison rod assembly (BPRA) handling tool.
PURI'OSl!: The higher mounting elevation of the RCC change fixture support bracket ensured that fuel movement and HVAC operation, and refueling pool water surface ripple would not cause moisture to enter tbt- tool motor housing. The extended length of the emergency pull cable improved its accessibility from the Fuel Storage Building (FSB), elevation 25 ft. Remosal of the comb lock assemblies from the BPRA handling tool improved the performance of that tool.
SArtrY EVA1.UATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the '
modifications made changes to the facility as described in the FSAR and identified the affected FSAR Figures. The RCC change fixture support bracket is designated non Safety Class, Seismic Category 1. The fuel transfer drive system emergency pull cable is designated non-Safety Class. The safety evaluation determined that the modifications did not alter the fucion or operation of the affected equipment and did not affect safety related equipment.
The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
FCR 91-019
-.n. , , , _ .. - , , . . . .-
e e MINOR MOutFICAT10N: Number 91 510 TrT11; loverter 1-2A DC Feed Undervoltage Relay Allowable Orrrating Value
SUMMARY
DESCRIPHON: Technical Specifications require that the operation of the iuerter supplying the Main Plant Computer System (MPCS) be limited to a maximum of fifteen m;nutes when fed from its DC source on loss of its battery charger.
In orde; to f ulfill this requirement, au , .iervoltage relay senses battery discharge, and a t i ming relay trips the cirecit breaken feeJi...g computer inverter. This Ms Modification L "tOD) provided the engineering basu. f u, the undersoltage relay setpoint of 125 VDC and -
ti.e iiming relay setpoint range of between 12.55 cad 14.45 minutes. The relay setpoints were
~
unchanced.
purpose: The purpose of the MMOD was to document the engiceccing basis for the undervoltage and tinng relay setpobts.
/
a S AmrY EVA1.UATION
SUMMARY
- A ::afety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the setpoints did not directly affect FSAR in that me text, figures and tables in the FSAR did not require changes. The affected relays are designated Class 1E. The safety evaluation determined that the MMOD did not involve physical, functional or performance changes to e q uip me nt. Tbc setpcints remain cachanged; the MMOD simply documents the sound, ,
engit.eering basis for the setpointi. The safety evaluation concluded that the MMOD did not ,
create an unreviewed safety question.
78
._ _ _ _ _ . .- _ _ _. ._ _ _ . _ . . . . . - . . . - - _ _ _ . . _ _ . _m._ _
9 4-MINOR MODil1CA110N: Number 91-512 Trrt.E: Fire Protection Pump Instrumentation Changes
SUMMARY
DESCR!! TION: This M nor Modification "iMOD) provided several instrumentatier, and control chan e,es . affecting Fire Protection Systern equipment and documentation. Hardware changes include revised alarm and control switch serpoints affecting the Fire Pumps and the Fire Storage Tank freeze p r ot e ci.on.
Document changes included the corre ct!rm of information discrepancies between design documents.
t 4
Pt! arose: The purpose of the MMOD was to provide resolution to several Fire Protection System instrumentstion und control concerns and to provide the basis for selpoint and documentation changes.
SAPIrry EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modifications made changes to the facility as described in the FSAR and identified the affected FSAR Section. The Fire Protection System is designated es non-Safety Class. The revised setpoints met the requirements of applicable codes and/or rt.gulatory references and ,
did not affect safety-related equipment. The safety evaluation determined that the modifications did not compromise the perfortnance of Fire Protection System equipment.
The safety evaluation concluded that the MMOD did not create an unreviewed safety-question.' ;
FCR 91-027 e
t
MINOR MODIFICATION: Number 91-519 TIT 1.F: Steam Pipir>g Upgradt S UM AI A RY DESCRIPTION: This Minor Modification (MMOD) replaced a carbon stee' elbow ;n the 16 inch extraction steam line with an elbow fabricated from chrome-molybd:num steel. Tbc MMOD also replaced piping downstream of Main Stc .m condensate drip leg flow restriction orifices with piping fabricated f rom erosion-resistant, staiuless steel, Pur. -'e ifted piping. and elbow were fabricated from me.erials which are stunt to crosion caused by wet steam f!ow. The materials conform to int is for erosion / corrosion contained in EPRI Report NP-3944.
S AFIHY EV ALU ATION ' SUMM ARY: A safety evalitation was performed for this MMOD. The safety evaluation applicabill:y review determined that the modifications did not directly affect the FSAR in that the text, figures and tables in the FSAR did not require changes. The affected piping and elbow are designated as non-
- f. Safety Class. The safety evaluation determined that tlie nodifications enhanced the l Extraction Steam System integrity by providing a :nore crosion/ corrosion-resistant mate ial.
The modifications did not affect safety-related equipment and did not change the form or function of the Extraction Steam System. Tbc safety evaluation concluded that the MMOD did not create an unreviewed safety question.
l I
L 80 l
l.
~
O
e
- MINon Moott'icATioN: Number 91-526 TrrL!t Larg, 2cre Hydraulic Snubber Modifica, ion Sit'.tM ARY DLISCRIPTIOr This Minor Modification (MMOD) provided two manuf acturer-recomtuended enhancernents to 1000 kip hydraulic snubbers used on the stwu gericrators. The first cubancement was the addition of a carbon steel sprcer to the packing assembly. The t.econd was the removal of unnecded tubing between the snubber body and fluid reservoir.
P U RPOSil: The carbon ste el spacer provided additional support for the packing.
Elimination of thr- unueeded tubing significantly reduced the qua'..tity of snubber ft id that would be spilled during chevron packing replacements.
S AITTY EVALUA*10N SUMMAPY: A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modifications did not directly affect the PS AR ia that the text, figures and tables in the FSAR did not require changes. The affected snubbers are designated as Safety Class 1 Corapenent Supports. The modifications did not afftet the function, load capacity or design basis of the affected snubbers. The safety evaluation coacluded that the MMOD did not create an enreviewed safety questiou.
rt 1
81
- c. e MINOR MonmCATION: Number 9 ' ~N 4
Tm.rt: Teedwater Check Valves Internals Modifications
+
SUMMARY
DESCIUmON: This Minor Modification (MMOD) modified the internals of Main reedwater System chcek valves FW V330, FW- V331, FW-V332 and FW- V333. The number of 3/8 inch diameter bolts holding the dash plate / retainer ring assembly was doubled from eight to sixteen. The dash plate assembly was alto modified to ensure that the disc did not impose impact loads on the dash plate.
PURPosn: Follcwing an outage in April 1991, plat >' full power could not be restored due to limited feedwater flow. Investig, Jon resealed that one of the four feedwater check vaices was not opening properly. Disassembly of this check valve revealed that the disc was jamtr.ed in the nearly-closed position and seven of the eight dash plate / locking ring attachment bolts were broken. Inspcetion or the other three check valves revealed additiona! broken bolts. This MMOD increased the number of dash plate / locking ring attachment bolts such that the total number of bolts would be capable of withstanding postulated differential pressure loads across the disc nd dash plate. This MMOD was ,
implemented prior to return to power following the April 1991 outage. This MMOD was h.: wed up by DCR 91035 which is summarized oc page 44 i
l SAFIrrY EVALUATION SWNLARY: A safety evaluation was performed for this MMOD. The safety evaluation applicability review deterrnined that the modification did not directly affect the FSAR in that the text, figures and tables in the FSAR did not require changes. The safety evaluation determined that the safety function of the feedwater chect valves is a controlled cicsing function. The MMOD did not affect e this safety function. It also did not uffect the opening stroke. The conservative redesign increased the number of bolts end consequently reduced individual bolt stress. This l modificat in was expected to eliminate further instances of bolt failure due to forces that I occur during opening. The safety evaluation concluded that the MMOD did not create un ,
! unreviewed safety question.
l
{
l f
I
- . -- . . - - - ~ . .. - -. . .~ . . - .- .-
A 4-l l
MINOR MODIFICATION: Number 91-531 i 4
TITLE: Door W400 Modifications l
l TUMMARY DILSCRll"flON: This Minor Modification (MMOD) modified the design of the l connections for Door W400, a twin leaf, dutch stpe, tornado- !
res;stant, bulles resistant, alarmed door between the Primary Auxiliary Building (PAB) and ;
a forty-five foot walkway leading to the Waste Processing Building (WPB). l I
PuRrosu: /6 menorail passes through this door. The monorailis used during maintenance periods. The coor modifications repaired damaged bolted fasterers and enhanced future maintenance activities by faciliting a more simplified procedure fnr door breakdown and re-assembly.
n i
SAFinY EVALUNTION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review deterrnined t. hat the modifications did not directly affect FS AR in that the text,- figures and tables in the FSAR
- did not require cFanges. The affected door is designated safety related, Seismic Cstegory 1, The safety evaluation determined that the modified connecrians. did not alter the originel design requirements and function of the door, The safety evaluation concluded that the MMOD -did not create an unreviewed safety question e c'
d i- _z N
1-i- ._ -- _ , ,
- , , , . . - . . . - - - . ~ . .. ~ . - . . . - - ,
a e, l
l
. MINOR hiODIMCATION: Number 91-$37 Triu!: Replacement of Positioner on Fecdwate: Control Valves a
SUMMARY
DESCRflTION: This Minor M odification (MMOD) replaced the obsolste
! positioners on the four Main Feedwater Control Valves. The replacement positioners are updated modela. Th e- MMOD also changed details of air supplies to the positioners.
PURPosn: This MMOD upgraded the Main Feedwater Control Valves' positioners to improve ' positioner reliability acd spare parts support.
SAMHY EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. Ti'e safety evaluation applicability review determined that the mcdifications did not directly affect FSAR in that the text, figures and tables in the FSAR did not require changes. The affected positioners are: designated non safety related and non-seismic. The safety evaluatica determined that the modification did not affect the fit, 1; form or function of the positioner. Three of the main feedwater control valves are assumed in the-FSAR to close in response to a feedwater isolation signal, and one valve is assumed to fail in the open position. This modification was d:termined to not affect that assumption.
The safety evaluation concluded that the MMOD did not create an unreviewed safety l question.
l
, , . -. - . - = . . - . . . . . . .-- . . . = . . .
4
- i MINOR MODIFl(_ATION: Number 91 538 Triu!: Door and Access Barrier for Detector Storage \rea in the Fuel Storage Building.
SUMMARY
DIISCIUI"I1ON: This Minor Modification (MMOD) provided a wire mesh, lockable door and barrier in the Fuel Storage Building.
PURPosII: The lockable storage area was provided to store a shielded pig containing irradiated moveable neutron detectors and spare neutron detectors, SAFETY EVAll'ATION
SUMMARY
- A safety evaluation was performed for this MMOD. The
. safety evaluation applicability review determined that the modific.ations made minor ebanges to the facility as described in the FSAR aud ic ntified the affected FSAR Figure. The door and barrier are designated non-Safety Class, Seismic Category Ic The safety evaluaticu determined that the modification did not impact equipment important to safety. The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
FCR 91-039 4
1
4
- MINOR MODUMCATION: Nutnber 91-545 TITir Technical Support Center (TSC) Emergency Lighting
SUMMARY
DESCRII"TjoN: This Minor Modification (MMOD) reclassified the TSC essential lighting circuits to emergency lighting circuits and electrically powered them from an uninterruptable power supply (UPS) inve: .er. The increased load a non-sufeiy battery 2B, which provides backup power to the inverter, was evaluated to be v.ithin the design margin. The MMOD also made other necessary and related electrical changes.
PURPOSE: This MMOD upgraded the power suppiy to the TSC lighting circuits to ensure lighting of the TSC in the event of e Station Blackout (loss of all AC power).
SAFETY EVALUATION
SUMMARY
- A safety evaluation was performed for thi MMOD. The safety evaluation applicability review dei 2rmined that the modifications did not directly affect FSAR in that the text, iigures and tables in the FS AR did not require changes. The affected lighting circuits are designated on-nuclear safety.
The safety evaluation determined that the modification did not increase the loading of non-safety battery 2B beyond its design capacity and did not impact equipment important to safety. The safety evaluation conchaded that the MMOD did not create an unreviewed safety question,
~
Ef>
. .- - . .- . .. . - - - .. ~ . - - - _ _ - . . ~ . . .-- . . . . - - .~.
+ n MINOR MODIIENI1ON: Nuaber 91 553 Trlin: Snubber Elimination
SUMMARY
D1?SCRIPTION: This Minor Modification (MMOD) eliminated two snubbers:
one originali) installed on an '8 inch Residual Heat Removal (RHR) line and the other originally installed on a 24 inch Service Water (SW) line.
, PURPOSE: Piping stress relanalysis has shown that stresses in the piping and remaining supports with the affected snubbers removed ren.ain within allowable lirnits established by applicable desiga Codes. Removal of the snubbers wculd result in reduced maintenance requirements, = reduced personnel erposure to radiation while performing maintenance, and increased plant reliabi!ity by elimination of a component which could fail, ,
SAFLTY EviwATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the L modifications did rot directly affect FSAR in that the text; figures and tables in the FSAR did net require changes. The affected snubbers were designated safety related, Seismic Category 1.- The safety evaluation determined that the modification did not alter the function or design basis of the affected s stems. The modification was an overall design enhancement since the snubbers were not needed; and their removal eliminated the consequences to the facility of their potential failure. The safety evaluation concluded that the MMOD did not
. create an unreviewed safety question.
4 e
j 87
...a.. 5 , -. . _ _ _
-9 6 l
1 l
N ualber 91-554 MINOR MODIFICATION:
T11'Lt" ' Fuel Pool Material Storage Locking Device l
SUMM ARY DESCRIPTION: This Minor Modificatior. (MMOD) provided a design for a lockable mounting device from which radioactive material could .
be suspended on an " arm" for underwater storage in either the Spent Fuel Pool or the Reactor Cavity or both. Several of these mountirig devices trave been mounted at the perimeter edge of the Spent Fuel Pool but none have been mounted in the Reactor Cavity, h At present, no material is belug stored using these devices. The locks, when used, will be under the control of the llcalth Physics Department.
PURrosE: The lockable mounting device was designed to provide a means, if needed in the future, of safely storing highly radioactive objects such as fasteners or filters. Underwater storage would ensure the appropriate shielding. Storage locations
- anticipated by the design are the Spent Fuel Pool and/or the Reactor Cavity. The loching device is to ecsure that the radioactive objects requiring underwater storage are not inadvertently raised or removed from water depths necessary to provide the required amount of radiation shielding.
S AFIrrY EVALUATION ' S UMMARY: . A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modifications did not directly affect FSAR in that the text, figures and tables in the FSAR
- did not require changes. The lockable mounting device is designated a non Safety Class,
- active - Seismic Category ' 1 compenent. The safety evaluation determined that the modification is an enhancement to facility design. The safety evaluation concluded that -
the MMOD did not create an unreviewed safety question.
i i
f-4 d
4 i
. - . . ~ __.~ . ~ . . _ . . . .
, 4 1
MINOR MODIFICATION: Number ' 91538 TITLE: Maximum Torque Swi;ch Setting for 1-SW-V$4 l
SUMMARY
DESCR11 TION: This' Minor Modification (MMOD) increased the torque switch setting for Service Water Sp:em valve 1 SW-V54 to the maximum :
setting allowed by its limiter plate. The NHY Data Sheets for Motor and Air Operated Valves and. Dampers were revised to reflect this change. 1 i
l
. PURPGsu: The revised torque switch setting was determined by engineering calculations performed as part of the program responsive to NRC Generic' Letter 8910.
SAlm EVALUNFION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modifications did not directly affect the UFSAR in that the text, figures and tables in the l .UFS AR did not require changes. Service Water valve 1-SW-V54 is a Safety Class 3, active
~
~
component. The safety evaluation determined that increasing the torque switch setting provided additionul assurance that the valve _ would ,be capable of- performit.e, its safety function. The safety evaluation concluded that the MMOD. did not create an unreviewed safety question.
1 i s
5 1
89
,,2- y ,- e
. ,,. . . ~ , ... - - . . . . - ~-- - ~.. .. -- - - . - - -
. -( i I
' MtNOR MODIFICATION: Number 91-566 !
Tm.n: Modifications to the Fuel Transfer Drive System Components ,
SUMMARY
DESCRll'110N: This Minor Modificatiou (MMOD) modified equipment in the Fuel Transfer System. An extension was added to the driveshaft and a new bracket was installed to support the drive motor, gear redurer and Torq Gard unit. A new driveshaft key, fasteners, wiring changes and other n.iscellaneous items were included in the MMOD.
PURPOSE: As originally designed, the aluminum housi 4 of the Torg Gard unit, a carbon steel key and carbon steel bu u'up . would have been partly submerged when the refuelitig cavity was fully flooded for refueling. These materials should not be in contact with borated water. The MMOD revised the design of the fuel transfer drive unit to ensure that the aluminum and carbon steel componcnts remained above the fully flooded refueling
- cavity water level.
SAFETY EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The l- safety evaluation applicability. review determined that the modifications did not directly affect UFSAR in that the text, figeres and tables in the
- UFSAR did not require changes. The-fuel transfer drive components are designated non-Safety Class, Seismic Category L The safety evaluation determined that the replacement equipment would meet applicable design criteria and would not affect the ope. ration of safety-
- related equipment. The- safety evaluation cone'.uded that the MMOD did ~ not create an unreviewed safety question.
b l
l 3 y w y..- y-- -i7-s7- - - - , , _w-
3 .--4 MINOR MOD!l'ICATION: Number 91568 T m.u: Modifications to the Residual Heat Removal (RHR) Vault Elevator a-St/MMARY DESCRII"I1ON: This Minor Modification .(M M OD) added minor - structural nhancements to the enclosure and car of the RHR V.iult clevator.
PURPOSM:. The modifications were to ensure the structural integrity of the c!cvator enclosure .snd car. The enhancements were made in response to an elevator
. inspection performed by a representative of the State of New Hampshire.
SAITTY EVA1.UATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modifications did. not directly affect UFSAR in that the text, figures and tables in the UFSAR did not require changes. The safety evaluation determined that the modifications
! would meet applicable design criteria and would not affect the operation of safety-related equipment. The safety evaluation concluded that the MMOD did not create an unreviewed safety question, I.
4 l-91 l
.q. . - , , . '
=vpy wr_.y---,a drc +e
_. . . . . . _ _ _. . ._ . . _ . . - ~ . ._ ..__. _ _ _ ,, .
c c ,
i MINOR MODIFICATION: Number 91-569 5
Trrttii First Refueling Outage Motor-Operated Valves ,
4
SUMMARY
DESCRIMlON: This Minor Modification (MMOD) replaced tbc limiter plate set in the motor operator for valves CBS-VS and CBS-V14, which connect the Containment Sump to the Residual Heat Removal (RH) and Containment Building Spray (CBS) Systems. It also provided the engineer'ing basis to revise the NHY Data Sheers for Motor and Air Operated Valves and Dampers for these valves. Finally, it provided the engineering basis .for miscellaneous other hardware and document changes
~ affecting motor-cperated valves.
PURPOSE: The hardware modifications aad document changes were Ue result of implementing the NHY program responsive to NRC Gtceric Letter 89-10.
SAPITTY EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modific.ations did not directly affect the UFSAR in .that the text, figures and tables in the
~UFSAR did-not require changes. The Containment sump isolation valves are Safety Class
- 2 gate valves; The replaced limiter plate set and revised settings did not change the ability
, of the valves to operate as designed. The higher limiter- plate setting allows the motor
- . operators to produce a higher thrust output. Changing the limiter plates in the Containment sump isolation valves does not affect any other safety-related equipment. Updating NHY documentation . defining motor-operated valve parameters based on calculations and other sources of information further ensures that the motor-operated valves will be capable of '
performing their design function. The safety evaluation concluded that the MMOD did not create en unreviewed safety question.
2 f
- 92 1
f r -
e,,, --n.,- , , , - - , ,m. w- -.-,--~nygg,,n,- , ,,--,.ev-yg- 1g y
ja , :s 4
MINOR MODIFICATION: Number 91-577 Trn.E: Battery Rooms A & B Thermocouple Shield Termination Change -
SUMMARY
Dt!SCRIPTION: This Minor Modification (MMOD) grounded the cable shields c for the Control Building battery room temperature measurements to the Main Plant Cotuputer System (MPCS).
t PURPOSE: The computer manufacturer recommended that cable shields be left floating
-(ungrounded) at the field end. The MPCS Intelligent Remote Terminal Unit ,
(IRTU) cad has guard circuits designed to cancel interference. However, false computer alarms occurred due to electrical' noir.- on these circuits. Therefore, the cable shield at the field end was grounded to eliminate the noise, i
r SAFl!!Y EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The-safety evaluation applicability review determined that the modifications did not directly affect the UFSAR in that the text, figures and tables in the
- UFSAR did not require changes. The temperature elements are not safety-related but are reismically mounted. The ' safety evaluation determined that the modification affects ,
temperature indication only and not control or protective functions. The modification improved 'the reliability of the battery room temperature indication and alarm function.
The. safety evaluation concluded that the MMOD did not create an unreviewed safety qt.cstion.
E f
93
~. .,%.. . . . , , , . . .
. .,. . , - . . . + ,. . . . _ . - . .
..~ ..-
MINOR MODIFICATION: Number 91-582
.TrrLE: Pressurizer Gas Sample Line Modification
SUMMARY
DESCRIPTION: This Minor Modification (MMOD) replaced or isolated all Raychem "CryoFit" couplings installed ou selected sampling and instrumentadon tubing. CryoFit couplirigs installed in pressurizer and reactor coolant loop sampic ! ices inside Containment, in pressurizer _ instrument Lubit:g and in other, selected applications were replaced or isolated. Welded or compression type fittings were used as replacement couplings.
PURPOSE: During the first refueling outage, it was discovered that tht. CryoFit couplings installed in applications in which they could be exposed te b;gh temperature and high hydrogen concentrations could fail. For more specific mfor m ation on the background,_deteils and resolution of Cryofit coupling issues at Seabrook Staden, please refer to Docket No. 50-443, Licensee Event Report (LER) No. 91010, Revisiot /l, dated October 2, =1991 (forwarded to the NRC via letter NYN 91160 dated October 2, MtX i.
SAFLTY EVALUATION
SUMMARY
- A safety evaluation was performed for tels MMOD. The safety evaluation applicability review Cuermir,ed that the modifications made minor changes to the fr,cility as described in the UF3 AR and identified the affected UFSAR Table and Figures. The affected sample and instrument tubing is Safety Class 2. The safety evaluation determined that the affected instrument 1.ut sample lines are not new; the modification only revised the method of joining the tubing segments. The replacement couplings (either welded or compression type) were acceptable, recognized couplings and were installed in accordance with applicable design requirernentr.. The safety evaluation determined that.the affected tubing could perform its designed functions with the replacement couplings. The safety evaluatioe concluded that the MMOD did not create an unreviewed safety question.
- UFCR 91063 94 1
,..w - v -e ---
m ..
MINOR MOOll'ICATION: Number ' 91599 Trrtm: Seismic Restraint of Diesel Generator Jacket Water Heat Exchanger Tube Storage / Shipping Containers
SUMMARY
DESCRIPTION: This Minor Modification (MMOD) seismically anchored the Diesel Generator Jacket Water Heat Exchanger tube
< storage / shipping containers on the floor of the Primary Audliary Building (PAB), Elevation 53*. It also revised the design of a ladder and handrail serving the platform at Elevation 63*-4* (located above the storage / shipping containers).
s PURPOSE: In anticipation of future replacement of the Diesel Generator Jack'et Water Heat Exchanger tubes, the decision was made to store a replacement tube bundle for_ cach heat exchanger in a storage / shipping container anchored to the floor of the PAB immediately east of each heat exchanger. Because of their length, the only way to get-the tube bundles into the PAB was to pass them through the roof plugs for the PCCW beat exchangers. These roof plugs were open during the first refueling outage. This opportunity
- contributed to the' decision to store the replacement tube bundles in the PAB. Modification of the ladder to the platform at Elevation 63'-4" was needed because of interference with a storage! shipping container. Modification of a handrail was done to facilitate movement of the storage / shipping containers.
l l
SAFI?TY EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the L modifications did not directly affect the UFSAR in that the text, figures and tables in the l UFSAR did not require changes. The storage / shipping containers are not safety related but-are scismically anchored to ensure that they would be restrained and would not damage safety related equipment during a seismic event. The safety evaluation determined that the replacement tubes in their storage / shipping containers were seismically restrained in accordance with the plant design basis; and that_the modification would not adversely affect safety related equipment. The safety evaluation concluded that the MMOD did not create l an unreviewed safety question.
95
- _ _ _ - . . _ - _ _ _ _ _._ _ -. . _ _ _ _ _ .. -~.- . _
.- a MINOR MODIFICATION: N u m be f- 91-664 TrrtE: Mc,dification of Main Steam Line Support MS-2 in Turbine Building
SUMMARY
DESCRJI"110N: This Minor Modification (MMOD) revised the design of the attachment for Main Steam Line Support MS-2. The MMOD removed the flange cover plate attached to the bottom flange of a structural beam in the Turbine Building, to which Support MS 2 was originally attached. A stiffener plate was added; and support MS-2 was re-attached to the beam.
PURPOSE: The flange cover plate, to which support MS-2 was originally attached, had yielded, rendering support MS-2 con-functional. The purpose of this modification was to remove the unneeded flange cover plate and re attach support I1S-2 to the structural beam in accordance with design requirements.
n SAFITTY EVAL.UATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modifications did not directly affect the UFSAR in that the . text, figures and tables in the UFSAR did not require changes. The Turbine Building structural steel is non-seismic and not safety . elated. The safety evaluation determined that the modification replaced the
- existing design with an equivalent or better design, and would not adversely affect safety-related systems. The safety evaluation concluded that the MMOD did not create -an unreviewed safety question, l
i-
,.m._ -. ._
MINOR MODtt'ICATION: Number .91-609 l
Trrtu: _ Primary Component Cooling Water (PCCW) Pump Discharge Check Valve Disc Anti-Rotation Lugs 4
, SUMM ARY DESCRIPTloN: This Minor Modification (MMOD) added anti-rotation lugs to the internals of each PCCW Pump Discharge Check Valve.
The purpose of this modification was to eliminate rotation of the valve discs PURPOSB:
and c. equent wear of the disc mounting stud and hanger.
S AMfrY EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modifications did not directly affect the UFSAR in that the text, figures and tables in the UFSAR did not require changes. The PCCW Pump discharge valves are Safety Class '3 components. The safety evaluation determined that the anti-rotation lugs were a design enhancement which would help ensure that the PCCW Pumps function properly and would
, not adversely impact the proper operation of the PCCW System. .The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
5
- N
.~~, ,_ m ..
.. .. -. . - - - -.~ . - ~ _- - . _ - . .. - .
MINOR MODIFICATION: Number 91-610 Trn.n: Reactor Coolant Drain Tank (RCDT) Pump Suction Line Vent
SUMMARY
. DESCR!! TON: This Minor Modification (MMOD) added a 3/4 inch vent. line and an isolation valve to the - RCDT Pumps' suction line downstream of check valve WLD-V54 in the Containment.
P URPCSt!: The purpose of this modification was to permit venting of a high point in abe RCDT Pumps' suction piping downstream of check valve WLD-V54. This piping high point could accumulate nitrogen gas used as cover gas for the RCDT. Nitrogen gas accumulating _ at this high po:nt cenld be swept into _ the RCDT Pumps, causing gas binding.
S AFirry EVAI,UATION
SUMMARY
- . A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modification made _ changes to the facility as described in the UFSAR and identified the fiected UFSAR Figure. The affected Equipment and Floor Drainage System piping is non-nuclear Safety Class - Seismic Category I, The. safety evaluation determined that the modifications complied with applicable codes and did not affect safety related equipment
' The modifications would improvt: the reliability of the RCDT Pittnps.- The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
UFCR 91-062 t
i L
. 98 i
i i
,- , - - , , , - - n - - . ,- we -- . . - - - - - - -
- - =- - - - - - _ -. .
l-h!!NOR 510DIFICATION: Number 91 611 TITLE: Revised Pipe Support
SUMMARY
DF,SCRtPTION: This hiinor hiodification (AthiOD) revised a pipe support for a 2 in h segment of Oil Collection (OC) piping in the Containment. The modification made the pipe support removable using a bolted connection.
PURPOSn: The purpose of this modification was to eliminate interference between the pipe support and the davit arm swing used to replace the seal cartridge of Reactor Coolant Pump 1B.
SAFETY EVALUATION
SUMMARY
- A safety evaluation was performed for this hih10D. The safety evaluation appliability review determined that the modifications did not directly affect UFSAR in that the ;-xt, figures and tables in the UFSAR did not require changes. The safety evaluation determic.-ri that the modified pipe support complied with original design requirements. The safety evaluation concluded that 1- the hihiOD did not create an unreviewed safety question.
- . ..~ -
, e MINOR MODIFICATION: Number 91-612 Trrt.e: Meteorolof i cal Tower Base Modification i
SUMMAKY DESCRII" TION: This Minor ModFication (MMOD) added an angle / plate assembly-to the base (i the Meteorological Tower.
PURPOSE: An inspection of the Meteorological Tower revealed significant erosion of the center pin which provides horizontal shear resistance. Replacement of the pin
- would require an extend:d tower outage. The purpose of the angle / plate assembly was .o '
provide the required degree of horizontal restraint at the tower base without replackg the center pin, thus maintaining the tower in service.
SAITn' EVA1,UATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modifications did not directly affect UFSAR in that the text, figures and tables in the UFS A'R 'did not require changes. The safety evaluation determined that the modification was a structural enhancement which met applicable design criteria for this structure. The modification would not result in a change to the function or perforrnance of the
- Meteorological Tower. The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
i l
l 100 1
i
. - .. . _ -. .- .__. _ _ __. m_. ._ _ _. ._ . _ . _ . .
- ez ._ s 63 MINOR MODIFICATIC,N: Nu riber 91-613
. . Frn.n: Diesel Generator Reverse Power Relay Wiring Change
SUMMARY
DESCR.ll"flON: This Minor Modification (MMOD) corrected the sc h e ma tic -
diagrams for the Diesel Generator protective circuits which incorrectly depicted the wiring connections for reverse power relays.
PURPOSE: The purpose of the MMOD was to correct the schematic drawings for the Diesel Generator protective circuits as committed to in the Diesel Generator Special Report (NYN-91156) dated September 25, 1991. The connections for the reverse
. power relays were not in agreement with the vendor instruction manual. Proper functioning of the protective circuit was achieved by reversing two connections in the relay. Relay modification was authorized by an Engineering Change Authorization (ECA) during initial l- construction. However, when the modified reverse power relay for Diesel Generator I A was replaced by an unmodified relay from. inventory, a trip of the Diesel Generator occi. ed during post-maintenance testing. This trip was the subject of the above-referenced Special Report, t i
SAFETY EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the I modifications did not directly affect UFSAR in that the text, figures and tables in the l UFSAR did not require changes. The safety evaluation datermined that the modification
- 1would not result in a ' change to the function,- performance, protection or control of the Diesel Generator The modification was to restore the wiring to that specified by original
- design. The safety evaluation concluded that the MMOD did not create an unreviewed '
safety question.
f I.'
l
- l. 101 i.
1
--.e... ._
- s
- MINOR MODIFICATION: Number 91-615 TrrLE: Cooling Tower Portable Pump Relocation i
J
SUMMARY
DESCRIFIl0N: This Minor Modification (MMOD) redesignated the storage location for the diesel engine-driven Cooling Tower basin portable make-up pump and associated equipment (hose and strainer) from the Cooling Tower Unit 2, Train A switchgear room to the Service Water Pumphouse PURPOSEi The purpose of re-designating the storage location for the Cooling Tower basin portable make-up pump and associated equipment was to clear the Cooling Tower Unit 2, Train A switchgear room for use as a temporary storage location
- _ for dry, activated, low level radioactive waste. (A safety evaluation for the temporary storage of dry, activated waste in the Unit 2 Cooling Tower is summarized on page 170 of this report.)
SAFL'rY EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The
- safety evaluation applicability review determined that the modifiestions did not directly affect UFSAR in that the text, figures and tables in the UFSAR did not require changes. The new storage location for the Cooling Tower basin portable make-up pump met the requirement that it be designated Seismic Category I. The safety evaluation determined that the relocated storage location for the Cooling Tower basin portable make-up pump would -not reduce the pump's ability to perform its function as described in the UFSAR. The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
l 1.
l
't 102
_ ~ . . ...-_ - - . . . . --. .
I blINOR h10DIFICATION: Number 91-617
. TITLE: h1ain Generator Current Transformer (CT) Support Hardware Enhancement
~
SUMMARY
DESCRIPTION: This Minor Modification (MMOD) added a washer and nut set to each Main Generator CT mounting connection; and specified the torque range.
PUR"OSB: The purpose of the revised CT support hardware was to reduce the potential for damage to the cts from vibration.
f SAFETY EVALUATION
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined that the modifications did not directly affect UFSAR- in that the text, . figures and tables in the ,
UFSAR did - not eequire changes. The Main Generator is non-safety-related; and the modifications did not affect safety-related equipment. The safety evaluation determined that enhanced : support hardware exceeded the original design requirements and reduced the i potential for failures. The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
0 i
9 i
G 103
, . . _ - - . _ ~ - _ -- -. . . -.
MINOR MODIFICNDON: Number 91 618'
, Trrui: Support Modification for Weld Radiography Accessibility
SUMMARY
DI!SCRII"flON: This Minor Modification (MMOD) replaced a wall-mounted pipe support for a 3 inch segment of Chemical and Volume Control System (CS) piping with a. modified. pipe support. The support and pipe segment were located in a _ concrete pipe enclosure in the Primary Auxiliary Building (PAB). The MMOD also provided a wall opening in the concrete ' pipe enclosure. A lead shield retainer was provided to plug the wall opening after completion of radiography.
PURPOSII:' The pipe support was removed and the wall opening was made to facilitate weld radiography of a field weld in the 3 inch, CS pipe. The radiography was part of the Weld' Record Re-verification Program, The pipe support was modified to simplify re-installation.
S AFirrY EVA1.UATION
SUMMARY
- A safety evaluation was performed for this MMOD, The safety evaluation applicability review determined that the modifications did not directly affect UFSAR in that the text, figures and tables in the UFSAR did not require changes. The safety evaluation determined that the pipe support revised design met original design requirements. The safety evaluation concluded that the MMOD did not create an unreviewed safety question.
a k
e d -
104 4
4
~ - -- - "
-n -
. 4 -- ,
MINOR MODIFICATION: Number 91-620 Coolant Pump (RCP) U nde rvolt age and Tnu: Replacement of -- Reactor I Underfrequency Relays-b
SUMMARY
DESCIUlmON: This Minor Modification (MMOD) provided the engineering basis to replace the underirey2ency time delay relay for RCP IB with an identical relay obtained from tue RCP IB undervoltage detection circuit _ ind to replace the RCP 1B undervoltage time delay relay with a relay of the same tyue, but differe.nt catalog number and having an adjustable range.
PURPOSn: The underfrequency time delay relay for RCP IB (Agastat Type 7022PJ) was
' defective and required replacement, An identical relay- was installed in the undervoltage c.tection circuit, and became the replacement. An Agustat Type E7022PA had been evaluated by MMOD 90-508 as acceptable for use as the undervoltage time delay relay.
The parts substituticns were made to utilize existing stock.
..SAFIn't EVALUAnow
SUMMARY
- A safety evaluation-was performed for this MMOD. The safety evaluation applicability review determined that the modifications did not directly affect the UFSAR in that the text, figures and tables'in the UFSAR did not require changes. The replaceme relays were environmentally .and 4- seismii. ally qualified and ' met the functional an> 2alification requirements of the applications. The setpoints of the time delay rela),
2 the. maximum allowable response times of the circuits were unchanged by the MMOD. he safety evaluation determined that the replacement relays were acceptable for the applic. ons. The safety evaluation concluded
- that the MMOD did not create an unreviewed safety question.
4 d:
kb 1
105
- 4 d '
s
- - -m n ,-- n ~, m n- , y vim ,
_ . . _ _ . _ _ _ . _ . _ _ - . _ . _ _ _ _ . - _ _____...-.__,__.____,..__..-____s 1
MtNOR MODIFICNrlON Number 91 625 l 1
l l
Trrtin Demineralized Water / Radiation Monitoring Skid Isolation i l
SUMM ARY DP.SCRII#rION: This Minor Modification (MMOD) added spectacle Hanges in the demin?ralized water purge supply ' lines for three radiation monitor skids: the reactor coolant gross activity monitor, the Boron Waste Storage Tank lulet Activity monitor and the Auxiliary Steam Condensate monitor.
PilRPO511: The purpose of the apectacle flanges was to preclude backflow from a '
radioactive or potentiall) radioactive process stream to the Demineralized Water System durin; normal radiation monitor operation.
SAJT:TY EVALUNt10N
SUMMARY
- A safety evaluation was performed for this MMOD. The safety evaluation applicability review determined . hat the i
modifications made changes to the facility as described in the UFSAR and identified the
. affected UFSAR Figures, The affected Demineralized Wales System piping is non nuclear Safety Clau, Piph' to the scactor coolant gross activity monitor is scismic Category L The safety evaluation determined that the modifications did not affect safety related equipment and imptsv ed the margin of protection against inadvertent, radioactive contamination of the Demineralized Water System. The modifications slightly increased the potential for minor leakage o' radioactive fluid from process piping to the buildings in
' which the radiation monitors aie located. This potential, minor rete.se could occur as a result of rotating the spectacle flanges prior to purging operations. However, this potential radioactive leakage would be contained and would be well within the limits analyzed in the UFSAR. The sa":y evaluation concluded that the MMOD did not create-an unreviewed r.afety question. ,
V VFCR 91069 l
l' l~
106 l
l l-.-_-.._--- _ _ _ . . . , _ _ _ _ . _ _ . , _ , , , , . , _ _ _ . . , , _ _ _, ,; ,
. _ . - . ; ~ . . . _ - .. _ . _ . . _ - . _ _ - - _ . ___ ~ _ . _ . ._ _ __
_ . _ .._ _ _ . . ,l
. o i MINOR MODIMCATION: Number 91 626 TITLE: Turbine lluilding Gump liigh R9diation Trip 1
I SUM:i:ARY DESCRIPTION: This Minor Modification (MMOD) changed the control logic of j the Turbine Building Sump Pumps such that a high rndiation ,
level detected in the common discharge line to the oil / water separator vault would provide '
pump trip
- scal-in.'
PURSOSE: The pump trip
- seal in* feature prevents automath pump restart following clearance of a high radiation trip until the operator resets the seal in. This modification prevented sump pump cycling in response to radiation monitor
- spike" trip signals and possible release of radioactive fluid to the oillwater separator vault following receipt of a high radiat an trip.
SAFETY EVALUATION
SUMMARY
- A safety evaluation was pelformed for this MMOD, The '
safety evaluation applicability review determined that the modifications did not directly affect the UFSAR in that the text, figures and tables in the UFSAR did not require changes. The Turbine Building Sumps and Sump Radiation Monitor are not nuclear safetj related; and the modification did not affect safety related equipment.
The safety evaluation determined that the modifications did not increase the potential for release of radioactivity to the environment. The safety evaluation concluded that the MMOD -
did not create an unreviewed safety question.
W p
107 e
4 --%-, -.<,-m,v-< .w.-ww-, ww , e' M w n' ne-- ~ T ~+t- e-~'-'w a~ # 5-e's- w-- e'4 e+m-t" 'e-- 'e'-v--t v '= **=-a--** *ea ='" = "* = -
- e ** * * * - " * "
- 3 o
- 2. Temporarv [1q_d_.litcations to The following t-rnporary modifications were iinplemented at Seabrook Station pursuant the requirements of 10C1'R50.59.
108
. o 1
TnMronARY MOlbt'lCATION: Number 91-006 TITt.te Cron connection between Containment Ser$ ice an . Instrutnent Air ]
1 i
SUPMARY Di$CRi!" TION: This Temporary hiedification (TA10D) instal'ed a jumper to l cress-connect the Contaiernent A and 11 instrument air headers; cud provided u temprary supply lir : tc the Containment A and 11 instrument air headers i frem ibe Containment Service Air licader through a temporary filter matifold and air dryer.
PURrosth The Contahment Instrument Air System consists of two independent ait headers, each supplied by its own air compressor and air dryer. Cooling water to the Containment Air Compressors is provided by the Prirnary Component Cooling Water (PCCW) System and was unavailable during the first refueling outage due to the re..ubing of the PCCW heat exchangers. The purpose of th: temporary cro>s connect r,d air supply
'or the Containaient instrument air headers was to maintain instrument air pressure to control the Containment Air Purge (CAP) isolation valves, Containment Air llandling (Call) dampers arid other :omponents requiring instrument air for proper operation during the period of unavailability of the Containment Air Compressers. The TMOD provided backup to one Containment Air Compressor and header during Modes 5 and 6 and provided the sole source of Containment iirstrument air when the reactor core was oil-loaded to the spent luel pool.
SAITTY EVALUATION
SUMMARY
- A salety evaluation was performed for this TMOD. The safety evaluation applicability review determined that the modifications temporarily changed the inellity as described in the PSAR and identified the affected FSAR Sections. The Containment Inst +ument Air System is a non-safety related system. The safety evaluation deterrained that the TMCD could provide cican, dry air to maintain proper operation of coinponents served, and that during the period of dependence on this TMOD, Containment isolation capability could be maintained as required. The safety j evaluation concluded that the TMOD did not create an unreviewed safety question, i
I L
109
T!!M"ORARY M o nll'icN. ,i,; N u m be r 91-008 Ti rl.1:: Steam Generator Blowdown (SB) Demineralizer Drain Valves SUMMany IHLsCRIPf1ON: This Temporary Modification (TMOD) installed a drain .alve between the two isolation valves in each SB mixed bed c.emineralizer outlet line. The drain lines we re open drains to the Waste Holdup Sump.
PURPost!: The purpose of the TMOD was to provide a means of directing leakage of regenerant chemicals past the fit outlet isolation valve of cacb demineralizer to the Waste Holdup Sump. Yhis TMOD will reduce the possib lity that regenerant chemicals will enter the SB process stream. The intent is to maintain this TMOD in place until it is rnade permanent by a design modification.
S Altry EVALUATION
SUMMARY
- A safety evaluation was performed for this TMOD. The safety evaluation applicability review determined '. hat the modifications temporarily changed the facility as described in the FSAR and identified the affected FS AR Figure. The SB demineralizers are non safety related equipment. The safety evaluation determined that the TMOD was a design enhancement which would not adversely affect pr >per operation of SB components. The safety evaluation concluded that the TMOD did not create an unreviewed safety question.
3 110 j l
l
- c. e r-l TEMPORARY MODil'ICATION: Number 9b000 Triue Temporary 480 Volt AC Power to Containment 1.ower pacel ED PP-7A and Lighting Panel L18 SUMMAny DESCRIITION: This Temporary Modification (TMOD) installed a temporry electrical power cable from Unit Substation US 23 to Unit Substation US 11 and a temporary cable t etween Contain: rent Building Power Panel ED-PP.7/s and Liguting Transformer ED X-16E.
PURI'OSn: The purpose .>f the two temporary power cables was to provide a temporary source of 480 VAC electrical power to Containment Building Power Panel ED.
PP-7A and Lighting Panel L18 during a preventive maintenance outage period for US 11, the normal power source fe* these panels.
S Al'LTY EVALUATION
SUMMARY
- A safety evaluation was performed for this TMOD. The safety evaluation applicability review determined that the modifications did not directly affect the FSAR in that the text, figures and tables in the FSAR did not require changes. The electrical equipment affected by this TMOD is naa.
safety related.- The safety evaluation determined that the affect of the added electrical load on US-23 and ED.PP.7A was acceptable. The temporary elenrical configuration met original design requirements regarding capacity and _ circuit breaker coordination, External souting of temporary cable precluded potential circuit independence concerns. The snfety evaluation concluded that the'TMOD did not create an unreviewed safety question.
I I
s f
l 111 i l
_ _ _ _ . _ _ _ _ _ , _ _ . _ _ . ~ . . _ _ _ _ _ _ . . _ . - . , _ . - . , , - . - . . _ . . . . . , , . - - . , . .- _ . - . . . . . , , _ .
O *
+
7tiMPOlunY Mottif1 CATION: Number 91 010 TITI.in Temporary Power to Non sital Ilattery Chargers ED !!C-I4 and ED-IIC-213
SUMMARY
DIL%CfdP110N: This Temporary Modification (TMOD) provided tetiiporary 480 VAC power to non vital battery chargers ED UC 2A and ED BC-2fL 4
PURPGsti: The purpose of the TMOD was to maint:in non-vital battery chargers ED liC 2A nd ED IlC 2D in operation during a maintenance outage of 4160 volt Dus 5 which powers MCC E523, the normal power supply for these chargers. The TMOD would be in place only during Mode 6.
J sal'LTY EVALUN!10N StfMMARY: A safety evaluation was perfolmed for this TMOD. The ufety evaluation applicability review determined that the modifications temporarily changd the facility as described in the FSAR ar.d identified the affected FSAR Figure. The affected equipment is non safety related. The safety evaluation determined that the TMOD would not create a safety concern because the reactor core would be offloaded to the spent fuel pool, the alternative source of power was very reliable and the TMOD would be in effect for a short time. The safety evaluation concluded that '
the TMOD did not create an unreviewed safety question.
?
I 4
i
\
l i
112 ,
f
(- $
.._...u _ . , . . _ ,_ ._, _ _ _ . . . . , _ . . _ . _ . _ . . , . . , . - . _ _ . . . _ _ . - . . _ , _ . _ , ~_
Tl!MPORARY MODil'ICNilON: N u rnbe r 91 011 T t I I.tt: Temporary Power to Motor Coru <pl L.:nters (h!CCs) E513 and E514 S UMM AR Y Ditscal.*rION: This Temporary Modification (TMOD) provided lemporari 4S0 VAC power to MCCs E513 and E514 from MCCs 111 and 271 PURI Osti. The purpose of the TMOD was to maintain MCCs E513 and E514 in operation .
during the maintenance outage of 4160 volt Bus 5, the normal power supply for these MCCs Maintaining MCCs E513 an.) E514 in operation maintained security ar.d fire protection loads fed from these MCCs energized. The TMOD would be in place only during Mode 6.
~
S Arttry EVALUAT1 ors
SUMMARY
- A safety evalur.: ion was perfortred for this TMOD. The
<afety evaluation applicability review determin:d that the r:odifications temporarily changed the facility as described in the FSAR and identified the affected FSAR Figure. The affected equipment is non safety related. The safety evaluation determined that the TMOD would not create a saiety concern because the reactor core would be offloaded to the spent fuel pool, and the TMOD would bc in effect for a short time. Fire proteuion panels are provided with a bi.ttery backup feature. The safety evaluation concluded that the TMOD did not create an unreviewed safety quest.on. i 113 b___________________________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ __________ _ _ _ _ _ _
_ .-- -- - ~ . - - - . - - . . . . . . - . _ _ - . - ~ - - - . _ . - _ . . . . . - _ . . .
a TEMPORAF.Y MODIFICA110N: Number 91-012 e
1rrt.u: Tcmporary Power to Fire Protection Control Panels <
g 't
SUMMARY
1)(!SCRIFr10N: Thit Temporary Modification (TMOD) provided temporary 120 VAC power to fire protection control pancis normally powered '
- from distribution panels supplied by 4160 Volt llus E5.
PURrose: The purpo e of the TMOD was to maintain electrical power to fire protection control panels during a maintenance outage of 4160 volt Dus ES which-is the normal power supply for these panels. The TMOD would be in place only during Mode 6.
SAFirrY EVALUATION
SUMMARY
- A safety evaluation was performed for this TMOD. The safety evaluation applicability review determined that the ,
modifications temporarily changed the facility as described in the FSAR and identified the affected FSAR Figure. The affected equipment is non safety-related. Tne safety evaluation determined that the TMOD would not create a safety concern because the reactor core would be offloaded to the spent fuel pool,-and the TMOD would be in effect for a short time, in the event of loss of the temporary power source, the battery backup feature would allou the f!:e protection control panel to perform its function. The safety evaluation concluded that the TMOD did not create an unreviewed safety question.
.y t
l_
114
._ .. - . . _- _ , _ . ~ . - - - - . -.- - - ..-. .. - ..- ~
o <
TtIMi>ORAR Y MOnir1cxtloN: Number 91 013 TrrLit: Temporar) Electrical Power for Seismic Monitoring Panet SM-CP-58
SUMMARY
DESCRll'110N: This Temporary Modification (TMOD) provided a iemporary non vital souret of 120 VAC electrical power to . Scistnic Monitoring Panel SM CP 58. This TMOD was in place during Mode 6 with the reac os core off loaded to the spent inel pool.
PURPOSn: This TMOD provided a temporary, non vital source of 120 VAC .clectrical power to Seismic Monitoring Panel SM CP 58 during the outage of power panel 1-1;D.PP 1E, its normal source of 120 VAC power. Seismic Monitoring Panel SM CP 58 provides power to seismic monitoring instrumentation that is required by Technical Specifications to be operable at all times. The Seismic Monitoring System was declared inoperable during the time this TMOD was in place because it was being supplied by a non-vital source of 120 VAC po ier. However, the Seismic Monitoring System war, operating; and the intent of the applicable Technical Specifications was met during the maintenance outage of power panel 1.ED-PP 1E.
SAITTY 1:VALUA110N
SUMMARY
- A safety evaluation was performed for this TMOD. The safety evaluation applicability review determined that the modifications temporarily changed the facility as described i9 the FSAR and identified the affected FS A R Sections. The affected circuits are dutgnated Class IE. The safety evaluation determined that use of the temporarv un vital source of 120 VAC power to Seismic Monitoring Panel SM.CP-58 was accep:able considering the short dutation of time
! it would be in place and the fact that the plant would be in Mode 6 with the reactor core defueled. The safety evaluation conclud.a that the TMOD did not create an unreviewed safety question.
e
- t i
TnuroRAAy MonirtcATioN: Number 91 015 Trn.E: Temporary Electrical Power for the Emergency Diesel Generator 1A Harring l Devic:
SUMMARY
DitSCRIPTION: This Temporary Modification (TMOD) provided a temporary non safety related mutee of 480 VAC electrical power to the barring device for the Emergency Diesel Generator IA.
. PUkrosin This TMOD nermitted operation of the barring device to rotate the shaft of
- Emergency Dier,el Generator 1A while the normal source of oower to the barring device (MCC E511) was de-energi
- ed as the result of the outage of 4160 volt Dus
- 5. Rotation of the shaft of Emergency Diesel Generator IA was required to set tolerances as part of the process of overhauling this machine.
SAPIITY EVALUATION
SUMMARY
- A safety evaluation was performed for this TMOD. Th '
safety evaluation applicability review determined that the modifications temporarily changed the facility as described in tbc FSAR and identified the affected FSAR Sections. The barring device and its normal power supply are designated non safety related. The safety : valuation determined that there would not be a safety concern because the plant would be in Mode 6 with the reactor core defueled, and Emergency Diesel Generator 1A would be under overhaul. The' safety evaluation concluded that the TMOD did not create an unreviewed safety question.
t 116
_ . _ _ . _ . ._ . _ . . _ , . _ _- - , , . , ~ _ . , _- - . ~ , , _ . - .
, I ,
i l
TEMPORARY MODIFICATION: Number 91-024 t 1Trt.le Temporary Pawer to Unit Substation ED-US 16 i
SUMM ARY DP.SCRII"FlON: This Temporary Modification (TMOD) provided temporary 480 VAC power to Unit Substation ED US l6 from an off site source.
PURPOSE: The purpose of the TMOD was to maintain electrical power to Administration '
I: _; 'dir g loads during the maintenance outage cf Bus 1. The TMOD wculd be in place only during Mode 6.
p S ArirrY EVALUATION
SUMMARY
- A safety evaluation was performed for this TMOD, The safety evaluation applicability review determined that the rnodifications temporarily changed the facility as dessribed in the FSAR and identified the affected FSAR Figutt The affected electrical equipment is non safety-related. The safety evaluation deterrained that t.he TMOD would not crea:e a safety concern because the loads were lighting and co:'enience loads not related to safe equipment operation. The safety evaluation concluded that the TMOD did not create an unreviewed safety question.
t d
('
4 117 i
L- .--..U., , . - - , . - _ _ , . _ - . - -~ , .. ,-- ,m -4, - , - . - - - - - v o-,. + -
- e. A l
TEMPORAkY MODIFICNDON: Nurnber 91-025 Tnu Ternparary Power to Fire Protection Control Panels l l
1 l
l
SUMMARY
DESCRII" DON: This Temporary Modification (TMOD) provided temporary 120 VAC power to fire protection control panels normally powered from distribution panels supplied by 13.8 KV Bus 1.
Punrost!: The purpose of the TMOD was to maintain electrical power to fire protection control panels during the maintenroce outage of Bus 1. Tbc TMOD would be in place only during Mode 6. ,
SAFirrY EVA!1% TION
SUMMARY
- A safety evaluation was performed for this TMOD. The safety evaluation applicability review determined that the snodifications temporarily changed the facility as described in the FSAR and identified the affected FSAR Figure. The affected equipment is non safety.related. The safety evaluation determined that the TMOD would not create a safety concern because the reactor core would be offloaded to the spent fuel pool, and the TMOD would be in effect for a short time. In the event of loss of the temporary power source, the battery backup feature would allow the - fire protection control panel to perform its function. The safety evaluation concluded that the TMOD did not create an unreviewed safety question.
l l -
118
/
. . r E,-r
-_ _ - _ - . . - . .. - . _ . _ _ - . . _ . - _ - = . - _ ~ .
d l
l TEMPORARY MODIFICATION: Number 91-033 TITLE: Temporary Cooling Water for Safety injection and Charging Pumps i
SUMMARY
DESCRIPDON: This Temporary Modification (TMOD) insf lied temporary valves, l pipe fittings and hoses to supply demineralized water to the oil '
coolers of the Charging and Safety injection l' amps. It also temporarily re-routed the outlets from these coolers to floor drain systems for discherge to the liquid waste disposal system.
Pt;RPOSB: The purpose of the temporary supply of demineralized water to Ibe oil coolers of the CS and SI pumps was to provide the uceded cooling water to permit operation of the pumps for testing. During the first refueling outage, modifications to the Primary Component Cooling Water (PCCW) System (the normal source of cooling water for these oil coolers) rendered.it unavailable as a source of cooling water.
SAFETY EVALUATION
SUMMARY
- A safety evaluation was performed for this TMOD. The safety evaluation applicability review determined that the modifications temporarily changed the facility as described in the FSAR and idectified the affected FSAR Sections. The Safety injection and Charging Pumps affected by this TMOD are Safety Class 2. The safety evaluation determined that the temporary modifications would be in effect only when the reactor core was fully offloaded to the Spent Fuel Pool; thus the affected pumps would not be providing a safety function while operating under temporary cooling. The safety evaluation concluded that the TMOD did not create an umeviewe6 safety question.
119
1 4- d TEMPORARY MODIFICATION: Number 91-035 Trn.n: Elimination of Funnels from the Drains of Chemical and Volume Contial (CS) and Spent Fuel (SF) Filter Drain Lines ,
SUMMar v. DESCRIFFION: This Temporary Modification (TMOD) removed open entrant funnels from the drain manifold for several filters in the CS and SF Systems. The drain connections of these filters were connected either to the common drain header or directly to floor drains of the Waste Liquid (WL) Syst:m with transpatent flexible tubing secured with hose clamps. The discharge of the drain manifold was also connected to floor drains of the Waste Liquid (WL) System with transparent flexible tubing secured with hose clamps. It is intended to make this temporary modification permanent by future implementation of Minor Modification (MMOD)91-507.
PURPOSE: The purpose of this temporary modification was to prevent splashing of radioactive liquid onto adjacent surfaces and thus help eliminate radioactive contamination during draining and venting while retaining the capabHity to visually observe the liquid discharge.
SAFlfrY EVALUATION
SUMMARY
- A safety evaluation was performed for this TMOD. The safety evaluation applicability review determined that the
-modifications temporarily changed the facility ns described in the FSAR and identified the affected FSAR Sections. The piping affected by this TMOD is designate.d t'on-nuclear safety class, Seismic Category 1. The safety evaluation determined that the temporary modifications '
would not affect the design function of the filter drain lines to route drainage to the WL System. The safety evaluation concluded that the TMOD did not create an unreviewed safety question.
,r i-
=w * ~- rn
TEMPORARY MODII'ICATION: Number 91 036 Triu: Lifted Leads to Primary Component Cooling Water (PCCW) Flow Switches
SUMMARY
DI%CRii .vN: This Temporary Modification (TMOD) lifted leads in the control circuit for the Containment Cooling Fans. The lifted leads are associated with FCCW flow switches which provide a permissive function for Containment Cooling Fan operation, i
PURPosu: The purpose of this temporary modilication was to permit operation of the Containment Cooling Fans in the absence of PCCW flow to recirculate- air through the Containment during the first refueling outage to improve habitabil. v ant' help reduce ambient temperatures.
S AI'ETY EVAL UATION SUMMAkt A safety evaluritieri was performed for this TMOD. The safety evaluatio J applicability review determined that the 1-modifications temporarily changed the facility as desc ibed in the FSAR and identified the affected FSAR Sections. The Containment Cooling Fans are designated non safety related.
The safety evaluation determined that operation tte Containment Cooling Fans without l PCCW flow would not adversely affect performance. The safety evaluation concluded that the TMOD did not create an unreviewed safety question.
ll t
L 121 b
p
. .- __ . =_ ____ ._ __. . . _ _ . _ _ _ _ . ..
. _ - . _ - . . . _ - . . - - _=~ ..
I TEMPORARY MODll'ICATION: Number 91-037 TrrLIE Cross-connection Between Circulating Water (CW) and hervice Water (SW)
SUMMARY
DESCRII"!10N: This Temporary Modification (TMOD) provided a cross connection between the CW System and the SW System to supply cooling water to the Secondary Component Cooling (SCC) beat exchangers during the per;n of outage of the Primary Component Cooling Water (PCCW) System. The cross connection was made with fire hoses.
PURPOSE: The purpose of this temporary modification was to supply cooling water to the Secondary Component Cooling (SCC) heat exchangers during the period of outage of the Primary Compor.ent Cooling Water (PCCW) System, the normal supply of cooling water.
SAFIrl'l EVALUATION
SUMMARY
- A safety evaluation was performed for this TMOD. The safety evaluation applicability review determined that the modifications temporarily changed the facility as described in the FSAR and identified the affected FSAR Sections. The SCC System is designated non. nuclear safety class. The safety evaluation determined that the TMOD did not affect safety related equipment, was not located in a safety-related area, and the affected equipment would be restored to the original configuration prior to attaining Mode 4 following the first refueling outage. The safety evaluation concluded that the TMOD did not create an unreviewed safety question.
122
. . . . . . . - - --.- . _ . - . . . . - . _ - - . . . . . ~ . - -- _ ~ - - - .... _. _ - - - ..- -
TEMPORARY MODil'ICATION: Number 91 040 Tnu: Removal of Carrier Blocking Signal from the System 1 Protection Scheme for ,
the Scobie Line I
SUMMARY
D11 SCRIP 110N: This Temporary Modification (TMOD) removed the coupling capacitor voltage transformer (CCYT) from the *B* phase of the 345 KV transmission line from Seabrook Station to Scobie Pond (the Scobie line). It also turned off the carrier transmitter for the Scobic line. These modifications effectively removed the carrier blocking signal from one of the two independent protective relay systems (System 1).
PURPOsH; This temporary modification was performed because the CCVT failed; and no suitable replacement CCVT was immediately available.
SAFIrrY EVA1.UATION
SUMMARY
- A safety evaluation was performed for this TMOD. The safety evaluation applicability review determined that the modifications temporarily changed the facility as described in the FSAR and identified the affected FSAR Sections. The affected transmission line and equipment designated non-safety related. The safety evaluation evaluated the consequences of operating without the blocking- signal and determined that, in all cases, offsite power would still be available to Seabrook Station thiough the Reserve Auxiliary Transformers (RATS). The safety evaluation concluded that the TMOD did not create an unreviewed safety ques. ion.
1 i
l j-
'm, -
- _ . - - - - - . . . . - . .- - . - . . - . . . . - . . - . - . . . _ . - . - . - - . ~ . . - _ . -- - _ . - . - . - + . - - -
e
- Tl!MPORARY MODIFICATION: Number 91-041 TITL1!: Temporary Electrical Power for isolation System Cabinet FS9.
SUMMARY
DItSCRII"rION: This Temporary Modification (TMOD) provided a temporary non-safety-related source of 120 VAC electrical power to isolation System Cabinet FS9.
PURPOSE: By maintaining power to Isolation System Cabinet FS9, this temporary modification maintained the ability to reset, silence and acknowledge.a group of Control Room hard-wired annunciator alarms during the maintenance outage of Uninterruptable Power Supply (UPS) 1E, the normal source of power to Isolation System Cabinet FS9.
SAITGY EVALUATION
SUMMARY
- A safety evalurition was perfortned for this' TMOD, The safety evaluation applicability review determined that the moolfications temporarily changed the facility as described in the FSAR and identified the affected FSAR Sections. The ~ affected circuits are designated safety-related. The safety evalu'ation determined that there would not be a safety concern because the plant would be in Mode 6 with the reactor core defueled. The safety evaluation concluded that the TMOD did not create an unreviewed safety question. ,
I i .
L i
124 lw , - , - - - , swre e. - - -en-,,- --e <- - .,n ..e -e. , - w+ - - -r-,,-m-.e ,e,r- , - , , , > w ?y-e ~ . - e w
o
- TEMPORARY MODil'ICATION: Number 91 042 T rn.it Temporary Chemical Addition Line for Feedwater System SUMM ARY DILSCRIITION: This Temporary Modification (TMOD) installed a t e rnpo r a r y chemical addition line to the suction flange of Steam Generator Wet Layup Pump 1 FW-P-293.
PURPGsit This temporary ruodification provided the ability to maintain Feedwater System water chernistry during draining and maintenance of the Feedwater System.
S Af71N'i EVAll;NDON
SUMMARY
- A safety evaluation was performed for this TMOD. The safety evaluation applicability review determined that the modifications temporarily changed the facility as described in the FSAR and identified the affected FS AR Figure. The affected pump is designated non-safety-class and is installed in the non-safet) class portion of the Feedwater System. The safety evaluation determined that the chemical addition line did not interact with any equipment im por t t.n t to safety.
The safety evaluation concluded thr.t the TMOD did not create an unreviewed safety question.
125
e N uro ber 91-044 Tl!MPORARY MODtt'1CN110N:
l 11T1.It Jumpers in the Control Circuits for Steam Generator Blowdown (SB) Isola Yahes This Temporary Modification (TMOD) installed two jumpers h e s. in SUMM ARY DI'.SCRll"I1ON:
the control circuits for the four outboard SU isolation lting va The jumpers bypass relay contacts which provide the automatic isolation feature re Irom Emergency Feedwater (EFW) System actuation.
This temporary modification was implemented to preclude unwanted he SB Systern P U RPGsti:
isolations triggered by valve movement while performing modifications on t steam supply line to the turbine driven EFW Pump.
The A safety evaluation was performed for this TMOD.
SartrlY EVALUATION
SUMMARY
- safety evaluation applicability review determined that the FSAR and identified the modifications temporarily changed the f acility as described in the The safety evaluation The affected circuits are safety related.
aficeted FSAR Sections.
determined that the plant would be in Mode 5 while the temporary The safety modification place and, in this Mode, the bypassed isolation feature is not required. i evaluation concluded that the TMOD did not create an unreviewed safety quest on.
126 l
\
._ . . m _. .. . .. _ _ _ , - _ _ - _ _ _ _ _ . . . ~ _ _ . . . . . _ _ _ - _ . _ . _ _ ....m.. . _ _ . _ . . ._ >
TEMPORARY MODil'ICATION: Numb:r 91 047 TrrLn: Jumper in the Control Circuits f or Steam Generator Blowdown (SB) Isolation Valves l
S UMMARY DESCRII' TION: This Temporary Modification (TMOD) installed a jumper in the control circuits for the four inboard SB isolation valves. The jumper bypasses a relay contact which controls the position of the four inboard SB isolation valves besed on level and pressure in the SB Blowdown Flash Tank. A maintenance outage
, of 4160 VAC Bus 5 would de energize the relay and prevent apening of the valves.
PURPosn: This temporary modification was implernented to permit operation of the wet -
layup pump to maintain Steam Generator water chemistry during the maintenance outage of 4160 VAC Bus 5.
l SAFLTY EVALUATION
SUMMARY
- A safety evaluation was performed for this TMOD. The i safety evaluation applicability review determined that the modifications temporarily changed the facility as described in the FSAR and identified the affected FS AR - Sections. The affected circuits are safety-r:la:ed. The safety evaluation
- determined that the plant would be lu Mode 6 while the ternporary rnodification was in place and, in this Mode, the bypassed Containment isolation feature is not required. The safety evaluation concluded that the.TMOD did not create an unreviewed safety question.
l l
l 127
. . - . ~
TEMPORARY MODir1 CATION: Number 91-049 TrrLn: Temporary Main Generator Overcurrent Protection l
SUMMARY
DESCRIPTION This Temporary Modification (TMOD) disabled a portion of the Main Generator _ overcurrent relay protection syst e m s. The disablements were accomplished by removing relay test jacks, which allow the relay to be disconnected internally. A temporary overcurrent relay was substituted for the disabled protective devices. The trip output from the temporary overcurrent relay was connected to be able to trip the Main Generator.
PURPOSE: This temporary modification was implemented to permit post-installation testing of a replaced current transformer associated with the permanent M ain Generator overcurrent relay protection system.
SAITrrY EVALUATION
SUMMARY
- A safety evaluation was performed for this TMOD. The safety evaluation applicability review determined that the modifications temporarily changed the facility as described in the FSAR and identified the affected FSAR Sections. The affected circuits are non-safety-related. The safety evaluation determined that safety-related - equipment would not be affected by this temporary modification. The safety evaluation concluded that the TMOD did not create an unreviewed safety question.
1 i-12S i
1
. . -- . . . - - .- . . . . _ . . - - . - - - --._ -.. ~.-- - -
I TEMPORARY MODIP1CN!10N: Number 91-051 TrrLn: Rinse Flow Path for Steam Gens ir Blowdown (SB) System Dem ineralizers l
SUMMARY
DESCIUPTION: This Temporary Modification (TMOD) prended a combined flow path for the discharge of the Steam Generator Blowdown (SB) demineralizers and tLe discharge of the Waste Holdup Sump Pump to the Circulating Water
,- (CW) System discharge via the Liquid Waste (LW) discharge header. This flow path is continuously monitored for-radiation and flow. A duplex attainer was provided es a backup to the demineralizer resin retention elements to prevent resin from entering the WL or CW Syst e ms.
PURPGsu: The purpose of this temporary modification was to provide the capability to rinse the SB demineralizer after each regeneration cycle to reduce effluent conductivity using liquid from the SB Flash Tank as the rinsing fluid. The intent is to maintain this flow path by a future replacement of this temporary modification by a design change.
SAFlfrY EVALUATION
SUMMARY
- A safety evaluation was performed for this TMOD. The safety evaluation applicability review deterrnined that the modifications temporarily changed the facility as described in the F5AR and identified the affected FSAR S ecticas. 'The affected components are non-safety class. The safety evaluation determined that regenerant chemicals would be discharged to the CW System after recirculation and neutralization, in accordance with the original design. Protection from an unmonitored radiological release is provided by. existing radiation monitoring equipment on the liquid waste discharge header. This TMOD did not affect safety related equipment. The safety e"aluation concluded that the TMOD did not create either an unreviewed safety or an unreviewed environmental question.
1 1
129
---r,
TEMPORARY MODil'! CATION: Number 91-054 TITI.E: Re routing of Protected Area Storm Drains
SUMMARY
DESCRII' TION: This Temporary Modification (TMOD) plugged the storm drain downstream of Manhole No. 9, and installed a temporary pump to direct the unem drains to the Circulating Water (CW) discharge.
PURPOSE: The site drainage system discharges to the settling basin. Effluent from the settling basin is discharged to the Browns River. Radioactive materials cannot be discharged to the Browns River. As a result of the inadvertent radioactive contamination of the Demineralized Water (DM) System, storm drains inside the protected area could potentially become radioactively contaminated. Therefore, this temporary modification re-directed tbc storm drains originating inside the protected area to the CW System, which is recogni7ed by the UFSAR as the discharge path for radioactive effluents. The intent is to replace this TMOD with a design modification (DCR 90-52).
SAFITTY EVALUATION
SUMMARY
- A safety evaluation was performed for this TMOD. The safety evalnation applicability review determined that the modifications temporarily changed the facility as described in the FSAR and identified the affected FSAR Sections. The site drainage system is non safety-related. The safety evaluation determined ti.at safety-related equipment would not be affected by this temporary modification, and that the potential release of radioactive drainage would be directed to a more conservative discharge point. The safety evaluation concluded that the TMOD did not create an unreviewed safety question.
130
- o.
- TEMPORARY h10DII'ICATION: Number 91-057 Trrt.E: Addition of a Temporary Demineralizer Unit to the Secondary Component Cooling Water (SCC) System ,
SUMMARY
DESCRWTION: This Temporary Modification (TMOD) added a temporary demineralizer unit to the SCC System downstream of filter 1-SCC F-18.
PURPG.J: The SCC Systern became inadvertently contaminated with radioactive water as a result of make-up drawn from the inadvertently contaminated Demineralized Water (DM) System. A_ portion of SCC System flow would be routed through the L demineralizer for cleanup of the radioactive contamination.
SAFETY EVALUATION
SUMMARY
- A safety evaluation was performed for this TMOD. The safety evaluation applicability review determined that the modifications temporarily changed the facility as described in the FSAR and identified the affec..cd FSAR Sections. The SCC System is a non-safety related system. The safety evaluation determined that safety-related equipment would not be affected by this temporary modification, and that the radioactive source terrn of the SCC System fluid inventory would be reduced. The safety evaluation concluded that the TMOD did not create an unreviewed safety question, ,
131
O *'
- 3. Temocrary Setooint Chanaes The following temporary setpoint changes v ere implemented pursuant to the requirements of 10CFR50.59.
l l
i l
l l
132 l
L - - - -- -~ . . . _ . _ _ , _ _ _ _ , _ _ . _ . _ _ _ _ , , _ _ _ _ _ , _
o l l
l TEMPORARY SLTPOINr CitANGE: Number 91-003 TrrLE: Main Feedwater Pump Specd Control Differential Pressure Setpoint Change l
SUMMARY
DESCRIPTION: This Temporary Setpoint Change revised the differential pressure range for the Main Feedwater Pump speed control program from 80-195 psid to 80 150 psid. This Temporary Setpoint Change was restored upon implementatioti of DCR 91009 which replaced the trim of the Main Feedwater regulating v7 with a trim of balanced, single seat design.
PURPOSE: Tbc purpose of this temporary setpoint change was to allow the main feedwater ,
regulating valves to operate in a slightly more open position by reducing the pressure drop across them, and by so-doing to reduce the -high frequency oscillations that were being experienced prior to the change.
SAFETY EVALUATION
SUMMARY
- A safety evaluation was performed for this TMOD. The safety evaluation applicability review determined that the modifications temporarily changed the facility as described in the FSAR and identified the affected FSAR Sections. The Main Feedwater Pump and its speed controls are designated non-safety-related equipment. The safety evaluation determined that neither system reliability nor safety-related equipment would be affected by this temporary setpoint change. The safety evaluation concluded that tS TMOD did not ceate an unreviewed safety question.
133 l
. . . , - , . ~ . , , . , , ,
TEMPORARY S!!TPOINT CilANGE: Number 91-004 TinI: Mair' Feedwater Pump Speed Control Differential Pressure Sc. point Change S UMM ARY U t!SCRIFTION: This Temporary Setpoint Change revised the diiferential pressure rauge for the Main Feedwater Pump speed corarol prograra from i fs0-195 p:M to 80-165 psid. This temporary setpoint change was rnade permanent by DCR 91 009.
PURPOSE: The purpose of this temporary setpoint change was to allow the main feedwater regulating valves to operate in a slightly more open p.mtion by reducing the pressure drop across them. Although the plug and stem were no lorger susceptible to flow-induced oscillations, stem packing frictiun was producing an unacceptable degree of valve hysteresis, adversely affecting steam generator level control. This r: vision enabled the valve to operate in a more fully open position there val.e hysteresis has less effect on flow control.
sal'IrTY EVALUATION
SUMMARY
- A safety evaluation was performed for this tetuporary setpoint change. The safety evaluation applicability review determined that the modifications temporarily changed the facility as described in the FSAR and identified the affected FSAR Sections. The Main Feedwater Pump and its speed controls are designated non-safety-related equipment. The safety evaluation determined that this temporary setpoint change would not wersely affect the operation of the main _
feedwater regulating valves or affect their ability to close in response to a feedwater isolation signat No safety-related equipment would b . affected by this temporary setpoint change. ;
The safety evaluation concluded that the temporary setpo. int change did not create an unreviewed safety question.
)
134
4 procedures
'l i c 5.. sowing pr ocedures were approved or implemented - as indicated pursuant to the requirements of 10CFR50.59 I
I!
i l
i l
i l
i t
{
D5 l i
i
.. ,_. . . ,. - - . - , ~-.-- . -- .. -- - .... ~ ..- --- .~.
l PROCPDURn: Number CN91-1-4, Revision 00 (
1 Trrt.n: Elevated Ammonia Program I
SUMMARY
DESCRIITON: During the first operating cycle, feedwater pH was maintained in the - range of 8.8 to 9.2. This range was based on Westinghouse guidelines for secondary water chemistry for plants with copper alloys in the feedtrain. Seabrook Station .has an all ferrous feedwater heater train; however there are copper alloys in other secondary system components (i.e. MSR tubus and condenser ,
- tubesheets). The results to date of Westinghouse studies, has indicated that feedwater pH l can be iicreased to 9.6 in plants with 90/10 CuNi MSR tubes without significant increase in copper corrosion. This procedure was implemented to increase feedwater pH from 9.2 to l 9.6 and monitor copper transport at the elevated pH. -During Cycle 2, feedwater pH will l be gradually increased in increments of 0.1 to a maximum of 9.6. i PURPOSn: The purpose of this procedure was to increase feedwater pH to a higher level with the expectation of reducing the fee .cr iron concentrations and lowering the sludg- burden to the steam generators.
S AI'mY EMLUATION
SUMMARY
- A safety evaluation was performed for this procedure.
The safety evaluation applicability review determined that this pro'cedure would change the administrative control of secondary system chemistry as described in the FSAR. The safety evaluation determined - that the maximum pH to be attained ~ by implementing this procedure was within the established limits for ferrous feedwater trains. Based on studies done by Westinghouse, it was expected that copper transport would not be increased by the elevated pH. The exp:cted end result of this process would be- a reduction in the probability of Steam Generator tube rupture due to corrosion. _ The safety evaluation concluded that implementation of the procedure would not involve an unreviewed safety question.
l i.-
---r- - -- - - - - - - - - - _ . - - - . . . . - - . _ _ _ - - - _ - - - _---_--
. . . - . .- - . . - . - . . .. . ~. . . . - _ - ~ _
o 4-
- PROCEDURE: ~ Number ES1801.00P. .evision 00 TITte; Etnergency Feedwater (EPW) Pump Turbine Overspeed Test Using Auxiliary ;
Steam
SUMMARY
DESCRil'I1ON: This procedure utilized low-pressure steam from the Auxiliary '
Steam System cross connected to the main steam supply header to the EFW Pump turbine to test the EFW Pump turbine overspeed trip setpoint. The turbine and pump are uncoupled for the test. The temporary auxiliary steam cross-connect to the main steam supply header to the EFW Pump turbinc is a steam hose connected between capped connections in the pipe tunnel. Tbc test is performed in either Mode 5 or 6 during each refueling outage.
' PURPOSE: The purpose of this procedure is to employ auxiliary steam rather than Main Steam generated. by the heat from Reactc.t Coolant Pump operation for the test. Operating the EFW turbine with auxiliary steam rather than main steam generated by the heat fror:. Reactor Coolant Pump operation for the test permits the test to be scheduled at a more optimal time during each refueling outage, j- S AF1m' EVA1.UATION
SUMMARY
.: A safety evaluat!un was performed for this procedare.
The safety evahtation applicability review determined _that this procedure would make a temporary change in the facility as described in the FSAR.
The safety evaluation determined that the use of a temporary cross-connect between the Auxiliary Steam System and the Main Steam supply header to the EFW Pump turbine -to conduct t.be test in the manner described by th_e procedure during the refueling outage would
.not introduce any new safety concerns. The safety evaluation concluded that would not involve an unseviewed safety question.
2 9
Q 137 .
^5 .
PROCEDURE: Number ES91-1-31, Revision 00 Trtut BWST/RWST Cross-Tie
SUMMARY
DESCRIrrION: This procedure was implemented to transfer approx.mately 100,000-125,000 gallons of borated water from the Refueling -
Water Storage Tank (RWST) to the 'A" Boron Waste Storage Tank (BWST) and later transfer this water back to the RWST using a temporary pump and hoses.
PURPOSE: This procedure was implemented to lower the Refueling Water Storage Ta'ik (RWST) to a level which would permit the performance of radiographic examination of a weld on a line from the RWST. Reexamination of this weld was determined to be necessary during the Weld Records Re-verification Program. Upon completion of the radiography, the original contents of the RWST were restored SAFETY EVA1..UATION
SUMMARY
- A safety evaluation was performed for this procedure, The safety evaluation applicability review determirmed that this procedure would make a temporary change in the facility as described in the FSAR.
~
The safety evaluation determined that transferring the contents of the RWST to the BWST, temporarily storing this water in the BWST, and transferring it back to the RWST following completion of radiogsaphy could be done without af'ecting safety-related structures, systems or components ann without introducing new safety concerns. The safety evaluation concluded
that would not involve an unreviewed safety question.
138
_ e _.
PROCEDURE:. Number - LS0564.32, Revision 00 TITLE:. . Spent Fuel Pool Cooling Pump-Energizing Backup Power j u
SUMMARY
DESCRIPTION: - This procedure was prepared to provide a standby source of electrical power to operate a spent fuel pool cooFng pump in Mode 6 with the reactor core off-loaded to the Spent Fuel Fool.
PURPOSE: During the refueling outage, maintenance outages of Electrical Buses E5 and E6 were planned. With one of these Buses de-energized for maintenance, a loss of ofl-site power coincident with a failure to start of the operable Emergency Diesel Generator would result in loss of electrical power to the remaining Spent Fuel Cooling Pump, The purpose of .this piocede e was to provide a contingency plan, which would be activated from Station Abncrmal Procedure OS1246,01 to power a Spent Fuel Cooling Pump by the Portable Diesel Generator by means of a pre-staged, temporary feeder cable.
. SAFETY EVAwATION'
SUMMARY
- . A safety evaluation was performed for this procedure.
The safety evaluation applicability review determined that this procedure, if used, would make changes in the facility as described in the FSAR.
Performance of the procedure to demonstrate its adequacy wculd, by strict definition, be a test not- described in the UFSAR. The safety evaluation - determined that testing the procedure in advance of reactor core off load would not degrade safe operation. The safety evalt.atioe also determined that use .of the procedure under conditions in which it was needed would alleviate rather than exacerbate the emergency condition. The safety evaluation concluded that : sting and implementation of the procedure would not invo'ive un unreviewed safety question.
p 4
k t- 139
- 5. Procedure Revisions The following procedure revisions were implemented pursuant to the requirements of 10CFR 50.59 -
i-140
o
SUMMARY
llIGCRil'FION: Procedure EX180 L OO2 implements the requirements of two separate and distinct programs, one of which is the Boric Acid Leakage Monitoring and Prem ' ion Program de scrib e d i .i letter N Y N .85076, the NilY response to NRC Generic Letter 8005. Although Generic Letter 88-05 did not specify qualifications for visual inspectors, NHY committed that personnel qualified as VT-2 examiners would perform certain portions of the boric acid leakage inspections.
In letter NYN-58076. NilY stated it,t personnei qualified as VT-2 etaminers would identify the le aking component, record its tag number and the number of the associated corrective action work request in test docurner,tation, and determine whether or not the test acceptance criteria had been met. Revision 02 to EXIS01.002 went beyond the committnent of NYN-
%076 and required VT-2 qualified examiners to perform the entire inspection, including leak
- rate measurement and t.n initial assessment of the impact of discovered leakage on affected and surrounding components.
Revision 04 restructured the inspection process and removed the requirement that qualified VT-7 examiners perform the inspections. In summary, Revision 04 .estructured the boric acid leakage inspection process into an initial screening inspection, an engineering evaluation and a post-corrective-actic n, VTC inspection. Personnel not qualified as VT-2 examiners, who may perform the screening inspection, will be sufficiently experienced, trained and ,
briefed to ensure their ability to satisfactorily perform the screening inspection. The results of the screening inspection will be reviewed by engineering personnel responsible for the program. Each in:tance o' discovered leakage or boric acid residue will be evaluated by the system engineer. A VT-2 inspection for material degradation of affected areas will bc
)
conducted following completion of corrective action. The above steps ;3rovide a system of checks and balances to et sure that the revised boric acid leakage inspection process is as effective as that required by Revision 03.
PURPosn: The primary purpose of Revision 04 to EX1801.002 was to revise the method .
of utilization of VT-2 qualified examiners in the boric acid leakage inspection process as described above.
S AITFY EVALUATION
SUMMARY
- A safety evaluation was performed for this procedure revision. The safety evaluation applicability review determined that the Boric Acid Leakage Monitocing and Prevention Program for designated systems outside Containment is not described in the FSAR, but is described in letter NY.N- ,
88076, the NHY response to NRC Generic Letter 88-05. Since Revision 04 changed the commitment made in letter NYN-88076 regarding the manner in which personnel qualified as VT-2 examiners are utilized in the conduct of boric acid leakage inspections, a safet) evaluation was performed as a conservative measure. The safety evaluation determined that Revision 04 did not reduce the effectiveness of the overall process of detecting leakage and boric acid residue. Therefore, the safety evaluation concluded that the Revision did not involve an unreviewed safety question.
141
-_ _ _ _ _ _ _ . - _ _ . _ ~ .. - ._. . . ._ _ _ .
, a PROCEDURE: - Nu mber ES 1801.006, Revision 02 Tntin Containment LcIakage Reduction Program
SUMMARY
. DESCRIPTION: Procedure EX1801.006 implements the requiremects of the Boric Acid Leakage Mon,toring and Prevention Program for the Reactor Coolant System (RCS) and portions of systems connectirg to the RCS inside Containment.
This program was described in letter NYN 88076, the NHY response to NRC Generic Letter 88-05. Although Generic Letter 88-05 did not sp:cify qualifications for visual inspectors, NHY committed that personnel qualified as V'l-2 examiners would perform certain portions of.the boric acid leakage inspections.
In'. letter NYN-88076, NHY stated that personnel qualified as VT-2 examiners would idemify the leaking component, record its teg number and the number of the associated corrective action work request in test documentation; and determine whether or not the test acceptance criteria had .bcen met. Revision 00 to EX180L006 went beyond the commitment of NYN-88076 and required VT-2 qualified examiners to perform the entire inspection, inciuding leak rate measurement and an initial assessment of the impact of discovered leakage on affected and surrounding ccmponents.
Revision 02 rotructured the inspection process and removed the requirement that qualified VT-2 examiners perform the inspections. In summary, Revision 02 restructured the boric acid leakage inspection process into an initial screening inspection, an engineering evaluation and a post corrective-action, . VT 2 inspection. Personnel not qualified as VT-2 examiners, who may perform the screeting inspection, will be sufficiently experienced, trained and briefed to ensure their ability t satisfactority perforc the screening inspection. The results of the screening inspection will be reviewed by engineering personnel responsible for the program. Each instance of discovered leakage or boric acid residue will be evaluated by the system engineer. A VT-2 inspection for material degradation of affected areas will be conducted following completion of corrective action. The above steps provide a cystem of checks and balances to ensure that the revised boric acid leakage inspection process is as "
effective as that required by Revision 00.
PURPOSE: The primary purpose of Revision 02 to EX1801.006 was to revise the method of utilization of VT-2 qualified examiners in the boric acid leakage inspection process as described above.
SAFETY EVALUATION
SUMMARY
- A safety evaluation was performed for this procedure revision. The safety evaluation applicability review
' determined that the Boric Acid Leakage Monitoring and Prevention Program for desiguated systems inside the Containment is not described in the FSAR, but is described in letter NYN-88076, the NHY response to NRC Generic Letter 88-05. Since Rei,ision 02 changed the commitment made in letter NYN 88076 regarding the manner in which personnel qualified as VT-2 examiners are utilized in the conduct of boric acid leakage inspections, a safety evaluation was performed as a conservative measure. The safety evaluation determined that Revision 02 did not reduce the effectiveness of the overall process of detecting leakage and boric acid residue. Therefore, the safety evaluation concluded that the Revision did not involve an unreviewed safety question.
142
, , y ._,. .,_-.w ., - . , - . - ,
- 6. Tests and Experiments No tests or - experitnents were conducted pursuant to 10CFR50.59 during this period.
However, as indicated in the summary for Procedure 1.S0564.32, Revision 00, discussed on page 139, by strict definition, testing the procedure could be considered to constitute a test not described in the UFS AR.
143 4
a
- 7. Technical Reauirements Manug The following "Iechnical Requirements Manua! changes have been approved pursuant to the requirements of 10CFR50.59 l'
l
{
\
144
'4-
.. o TUCIINICA1. REQUIREMEfRS MANUAL CitANGE REQUErr: Nuneber 39-04 TirLE: Instantaneous Trip Testing.
SUMMARY
DESCR!! TION and ' PURPOSE:' This _ Technical Requirements Change Request-removed Note 1 in Technical Requirements 13 and 15, This note provided instructions for instantaneous trip testing of molded case circuit breakers. During a test following a trip, the innructions called for attempting to reset the trippe.d. breaker as a means of determining whether the instantaneous trip element or the thermai element causcd the trip. During surveillance testing of molded case circuit breakers, confusion arose as to whether breaker resetting, as described in the note, constituted a r.equirement or was a recommendation. Technical Clarification TS.073 clarified that Note
- 1 did not ecnstitute a requirement. -It was therefore proposed to remove Note 1 from these two TechnicaF Requirements in order to climinate any future cociusion. The surveillance
- testing procedure contains the necessary guidance and defines testing requirements.
SAFETY EVALUATION
SUMMARY
- A safety evaluation was . performeu for -this Technical Requirements Change Request. The safety evaluation
- applicability review determined that the remova! of the note from the affected Technical Requirements constituted a change to the Technical. Specification Improvement Program described in FSAR Section 16.3. The safety evaluation determined that the change cir.rilied
- testing requirements for the affected devices and did not alter hardware or testing methods.
The safety evaluatior, concluded that removal of Note 1 would not involve an unreviewed 'i sufety question.
4 r
t +
145
, - , - - - - , + , , - . , , - - , .
,, , ~ , ,
e
- TECHNICAL REot!!REMENTS MANUAL CIIANGE REQUEST: Number 90 06 TITLE: Service Water Valves Stroke Time Revision
SUMMARY
DESCRIPTION und PURPOSE: This Technical Requirements Change Request updated information in Technical Requirement 14 (Table 16.3-9) regarding the thermal overload protection device for Service Water valve SW-V54 Table 16.3-9 lists thermal overload protection information for motor-operated salves. The stroke time requirements of Cooling Tower Pump discharge valves (SW-V25 and SW-V54) were changed by MMOD 90 523. The stroke time requirements change resulted in a change of the overload heater size for SW-V54 only. Therefore, in Table 16.3-9, the overload heater catalog number and heater current range for SW-V54 were changed to their revised values.
SAFETY EVALUA110N
SUMMARY
- A safety evaluation was performed for MMOD 90-523.
the MMOD associated with this Technical Requirements Manual Change Request. This safety evaluation is summarized in the section of this report covering MMODs.
146 1
TECl!NICAt. REQUIREMUhTS MANUAL CllANGn Recunm Number 90-08 Trnn: Changes in Overload Data for Main Steam Isolation Bypass Valves This Technical Requirements Change Request
SUMMARY
DESCRIITION and PURPosn:
upeated information in Technical Requirement'14 (Table 16.3-9) r$garding the. Main Stear isolation bypass valves' thermal overload devices.
Gear ratios for these valves were changed by NCR 82/937 and ECA 99/117114. As a result of 'his change, motor operator protection was recalculated for these valves by DCR-S6-594 The caleclation led to the replacement of the -overload heaters in the motor operators for these vahes. Therefore, the overload heater catalog. number and-heater current range were changed to their revised values in Table 16.3-9.
SAlcrY EVALUnTION
SUMMARY
- A summary of the safety evaluation for. DCR 86-594 was included with the quarterly 10CFR50.59 report forwarded -
by -letter NYN-90051. A safety - cvaluation was also ' performed for this Technical Requirements Change Request. The safety evaluation applicability review determined that the . changes made did not affect the criteria in the FSAR. However, as a conservative measure, a safety evaluation was performed. Tbc safety evaluation determined that the changes were based on approved criteria for circuit protection; and that proper selection of
. circuit protection- and testing assures proper fun :tioning of the circuits and their motor' operated valves. The safety eva! a tion concluded that the changes would not involve an unreviewed safety question.
i l .
i 147 J
- .,-.r- . , , , . -
,---m- -*r .-~~m, -
. e.
- TEcitNicAL REQUIREMENTS M ANUAL CilANGE REcunsT: Number 9103 Trra.n: Fire Protection Pump instrumentation Changes S UMM ARY DESCRIFrlON and PURPOSn: Technical Requirement No. TR7-4.7.9.1.lf.4 defined the surveillance testing requirement to verify the sequential start of the fire suppression water system pumps in order to maintain proper system pressure. Whi;. attempting to perform this surveillance test, it was discovered that system pressure w i not decrease to the 90 psig start setpoint cf the second pump before the first pump reacu,d a runout condi tion. A review of the fire suppression water system design was conducted. This design review defined the worst-case, safety-related fire g
conditiori and determined that Technical Requirement 7 sbould be changed such that surveillance testing would demonstrate the ability of the fire suppression water pumps to adequately respond to the worst-case, safety-related fire condition. MMOD 91-512 was prepared to resolve several fire protection system concerns and authorize needed setpoint changes. -This MMOD documented the changes needed to ensure correct fire uppression water pump sequential start for maintenance of proper system pressure and flow under the worst case, safety related fire conditions.
NHY calculation C-S-1-69013 determined that- 1791 gpm at 295 feet of TDH was required for the vrorst case, safety-related fire condition. Accordingly, TR7-4.7.9.1.1.f.2) was revised to verify that each pump develops at least 900 gpm at a total developed head of -295 feet.
. Additional changes were made to TR7-3.7.9.1 and TR7-4.7.9.1.1 for mutual consistency and compatibility with the results of the study.
S AFETY EVALUATION
SUMMARY
- A safety evaluation was performed for MMOD 91-512, the MMOD associated with this Technical Requirements Mar.aa! Change Request. This saferv evaluation is summarized in the section of this report covering MMODs.
Y a
I o
149 ,
E
, A; TECilNICAL REQUIREMENTL MANUAL CIIANGE REQUEST: Number 91-06 TITLE: Containment Bypass Leakage Paths i
SUMMARY
DESCRIPTION and PURPOSE: Generic Letter 91-08 provides guidance for preparing a license amendment request to remove certain component lists from Technical Specifications. The guidance of GL 91-08 generally requires that component lists removed from Technical Specifications be relocated to a document which is subject to the administrative requirements of Technical Specifications, Section 6.0.
In a future License Amendment Request (LAR 91-06) NHY plans to implement the guidance of Generic Letter 91-08 for Seabrook Station regarding secondary Containment bypass leakage paths. Upon approval of this LAR, a portion of TRCR No, 91-06 will be implemented relocating the list of secondary Containment bypass leakage paths from Technical Specifications, Table 3.6-1 to the Technical Requirements Manual, under a new Technical Requirement 16, 4
To conform to the guidance of GL 91-08, the portion of TRCR No. 91-06 not requiring prior NRC approval was implemented. This portion added the GL 91-08 definition for administrative control requirements in Technical Requirement No. 6 (Table 16.3-4) for intermittent opening of locked or scaled closed Containment Isolation valves.
SAFETY EVALUATION
SUMMARY
- A safety evaluation was performed for this Technical Requirements Change Request. The safety evaluation applicability review determined that the addition of the administrative control gaidance to the affected Technical Requirements constituted a change to the Technical Specification Improvement Program described in FSAR Section 16.3. The safety tvaluation determined tha' the administrative control guidance provided enhancement and clarification to existing requirements applicable to locked or i,ca.c3 clo<cd valves. The safety evaluation conc!uded that addition of the administrative control guidance would not involve an unreviewed safety question.
e 150
- - _ . , _~ . - - , - - . .. . . - . - . - - - . .-
jf, .4 TECllNICAL REQUIREMENis M ANUAL CllANGE REQUEFr: Number 91-07 TrrLE: Westinghouse Type KD Breaker Replacements for Obsolete ITE Type JL
SUMMARY
' DESCRIPflON and PURPOSn: This Technical Requirements Change revised test setpoints and verification response times for circuit breakers in Technical Requirement 13 (Table 16;3-8). Type JL thermal magnetic circuit breakers are listed in Table 16.3-8, as Containment ' penetration conductor overcurrent piotection_ devices. However, Type' JL circuit breakers are no longer available. Minor Modification (MMOD) No.90-663, determined that Westinghouse Type KD thermal magnetic circuit breakers are qualified replacements for Type JL thermal magnetic circuit breakers in MCC applications. "Iherefore Technical Requirements Change Request #91-07, revised Table 16.3-8, to provide the setpoints for testing Type KD circuit breakers used as replacements for~ Type JL circuit breakers in Containment penetration conductor overcurrent protection applications.
SAFETY EVALUATION
SUMMARY
- A safety evaluation was performed for the MMOD associated with this Technical Requirements Manual Change Request. The safety evaluation applicability review determined that the UFSAR did not specify' the type 'of circuit breaker (i.e. JL or KD) in its discussion of electrical equipment which use this type of circuit breaker. However, Table 163-8 did specifically
! list Type JL circuit breakers. Therefore, the MMOD associated with this Technical Requirements Manual Change Request affected only Table 16.3 8 of the UFSAR. The safety evaluation determined that Westinghouse Type KD thermal magnetic circuit breakers are qualified replacements for Type JL thermal magnctic circuit breakers in MCC applications.
The use of a Type KD circuit breaker as a replacemee.t for a Type JL circuit breaker would not change the function ci operating capabilities of the circuits in which they are installed.
o The Type KD circuit breaker provides the same electrical _ protection for the circuits as the l Type JL circuit breaker. The safety evaluation concluded that the MMOD would not involve an unreviewed safety question.
l-151 l
l - . - ..
.- 4
- 8. Final Safety Analysis Report / Updated Final Safety Analysis Report Change requests to the Final Safety Analysis Report (FSAR) or the Updated FSAR (UFSAR) associated with Design Coordination Reports (DCRs) or Minor Modifications (MMODs) were referenced in Section 1 and 9 of this report. The below listed additional FS/.R or (JFSAR change requests were issued pursuant to the requirements of 10CFR50.59.
4 1
152
- ,. -e FINAL SAHrlT ANALYSIS RITORT CIIANGE RE0tmsT: Number 90-116, Revision 1 TITI.It Updated FSAR Chapter 13
SUMMARY
DESCRIITION: This Final Safety Analysis Report (FS AR) Change Request (FCR) updated the organizational oescriptions in the UFSAR to reflect the approved organizational changes affecting the Maintenance, and the Chemistry /1-lealth Physics Departments.
A TY EVALUATION
SUMMARY
- A safety evaluation was performed for this FCR. The safety evaluation applicability review determined that the Chapter 13 revisions made changes to the facility as described in the FSAR. The safety evaluation determined that the changes were administrative in nature and did not affect plant e quipme nt. The safety evaluation concluded that the changes did not c.eate an unreviewed
- safety question. These changes will be incorporated into the Updated FSAR.
1 I
l l
t i
l 153 1
1 l
l l
l l
. , . . . - - . - - , - - . - . . . . . - . . - , . ~ , . . . . - , - . - - , . . . , ,
FINAL S AITTY ANA1.YSIS REPORT CitANGE REounsT: Number 91-032 i
TITLE: Diesel Generator Lube Oil Pressure Switch Setpoints
SUMMARY
DESCRII' TION: This i nnal Safety Analysis Report (FSAR) Change Request (FCR) revised UFSAR Section 9.5.7.5, " Diesel Generator Lubrication System, Instrumentation." The revision updated the information regarding several pressure switch setpoints and the logic for the low rocker arm tube oil pressure alarm. The revisions bring the UFSAR descriptions into agreement with as built conditions.
S AI'trrY EVAI.UNnON SUMM ARY: A safety evaluation was performed for this FCR. The safety evaluation applicability review determined that this revision constituted a change to the facility as described in the U FS A R. The safety evaluation determined that the revisions to UFSAR Section 9.5.7.5 were editorial and did not represent changes to plant equipment. The revised setpoints were to bring the UFSAR into agreement with the Standard Instrument Schedule (SIS), the official Seabrook Station document specifying instrument setpoints. The safety evaluation concluded that the revisic,n would not create an unreviewed safety question. The revision will be incorporated into the Updated FSAR.
154
e
- l FINAL SAPIrrY ANA1.YSts REPORT CllANGE REOUILvr: Number 91-033 Trrl.E: Startup Test Abstracts
SUMMARY
Di'ZCR11"110N: This Final Safety Analysis Report (FSAR) Change Request (FCR) revised FS A R T able 14.2-5, "Startup Test Abstracts " The revision changed the description of the test abstracts for the " Calibration of Steam and Fe e d"'a t e r Flow Instrumentation" and " Water Chemistry Control' tests. The revised description included additional information requested by the NRC during their review of Amendment 63 to the FSAR.
S Arm EVA1. UNI 10N
SUMMARY
- A safety evaluation was performed for this FCR. The safety evaluation applicability review determined that this revision constituted a change to procedures as described in the U FS A R. The safety evaluation determined that the revised precedural descriptions incorporated procedural requirements that were already addressed in the Startup Program and therefore did not introduce a safety concern. The safety evaluation concluded that the revision would not create an unreviewed safety question.
155
-( ..
FINAL SAITrY ANAL.YSIS REPORT CitANGE REQUEST: Number 91-036 TrrLE: Equipment Required for Safe Shutdown S UM MARY DESCRII'flON: This Final Safety Analysis Report (FSAR) Change Request (FCR) revised FSAR Table 7.4 - 1, " Equipment Required for Safe Shutdown." The revision identified.125 VDC distribution panels PP-113A and PP-113B as Remote-Safe Shutdown =(RSS) locations for the steam generator atmospheric relief valves in
-addition to CP-108A and CP-10SB, already listed. The appropriate circuit breakers at these paucis must be opened to fully isolate the Control Room controls from these relief valves when taking control at the ~ Remote Safe Shutdown Facilitics (CP-108A and CP-108B).
Abnormal Operating procedures have also been revised to ensure that the appropriate circuit breakers are opened.
sal'IrrY EVALUATION
SUMMARY
- A safety evaluation was performed for this FCR. The safety evaluation applicability review determined that this revision constituted . a change to the facility as described in the U FS AR. The safety evaluation determined that the revisions to UFSAR Table 7.4-1 did not introduce a safety concern,- since the revision did no't represent an equipment change. It merely corrects an oversight and makes the information more complete. The safety evaluation concluded that the revision would 'not create an unreviewed safety question. The revision will be incorporated into.the Updated FSAR.
t l-l l
FINAL SAnrrY ANALYSIS REPORT CHANGE REcutum Number 91-043 TrrLE: Accident Monitoring instrumentation
SUMMARY
DESCRIrrION: This Final Safety Analysis Report (FSAR) Change Request (FCR) 4 revised FSAR Table 7.5-1, " Accident Monitoring Instrumentation
. List." "t he; revisions clarified the use of the inadequate Core Cooling (ICC) monitor plasma displays' and d 'eted the requirement for trending certain parameters from the Table. -The AMI List specifies the instruments required to support the Emergency Operating Procedures
! (EOPs). The. lat .st revisions of the EOPs do not require trending of the affected
- parameters. Therefe- UFSAR Table 7.5-1 has been revised to reflect the latest revisions of the EOPs.
SAFlfrY EVALUATION
SUMMARY
- A safety evaluation was performed for this FCR. The safety evaluation applicability review determined that this revision constituted a change to the facility as described in the U FS A R. The safety evaluation determined that the revisions to UFSAR Table 7.51 did not introduce a safety
- - concern, since the operator does__ not depend on trending of the affected parameters in the execution of the EOPs. The safety evaluation concluded that the revision would not create an unreviewed safety question. The revision will be incorporated into the Updated FSAR, l
l I
i l
157 l
I
o .- =
FINAL SAPIrrY ANALYSIS REPORT CHANGE REcursr: Number 91046 Trrt.B: .F.efueling Procedures
SUMMARY
DESCRII' TION: This Final Safety Analysis Report (FSAR) Change Request (FCR) revised certain descriptions of refueling procedure in the UFSAR.
The revised descriptions reflect detailed Seabrook Station design and methodology. The revision includes procedural steps designed to minimize the time that the reactor vessel head is suspended by the polar crane and to sisually verify that all RCCA drive shafts have disengaged from the reactor vessel head prior to commencement of reactor cavity flooding.
t
- SAFETY EVALUATION
SUMMARY
- . A safety evaluation was performed for this FCR. The safety evaluation applicability review determined that this revision constituted a change to procedures as described in the UFS AR. The safety
' evaluation determined that the revised procedural descriptions do not introduce safety.
r concerns. The safety evaluation concluded that the revision would nat create an unreviewed safety question. The revision will be incorporated into the Updated FSAR.
158 l
l o< ew,- m- , _ . - g - + <-
FINAL SAFETY ANALYSIS REPORT CHANGE REQUE5T. Number 91-047 TtTLE: Accident Monitoring lustrumentation
SUMMARY
DESCRIrrioN: This Final Safety Analysis Report (FS AR) Change Request (FCR) revised FS AR Table 7.5-1, " Accident Monitoring Instrumentation List.* The revision deleted the requirement for trending of Containment Building Water Level and Containment Area Radiation from the Accident Monitoring Instrumentation (AMI)
List, UFS AR Table 7.5-1. The AMI List specifies the instruments required to support the Emergency Operating Procedures (EOPs). The latest revisions of the EOPs do not require trending of these parameters. Therefore, UFSAR Table 7.5-1 has been revised to reflect the latest revisions of the EOPs.
3ATETY EVALUATION
SUMMARY
- A safety evaluation was performed for this FCR. The safety evaluation applicability review determined that this revision constituted a change to the facility as described in the UFSAR. The safety evaluation determined that the revisions to UFSAR Table 7.5-1 did not introduce a safety concern, since the operator does not depend on trending of the affected parameters in the execution of the EOPs. The safety evaluation concluded that the revision would not create an unreviewed safety question. The revision will be incorporated into the Updated FSAR.
159
- , - .a FINAt. SAFirrY' ANALYSIS REPORT CllANGn REQUEST: Numbet 91-049 TrILE: Reactor Cavity Filling for Refueling S UMM ARY . DitSCRII"110N: This Final Safety Analysis Report (FSAR) Change Request (FCR) revised UFSAR Section 5.4.7.2.c.4 regarding the method of filling the refueling ' cavity in preparation for refueling. The section had specified that borated water would be pumped from the Refueling Water Storage Tank (RWST) to the refueling cavity using the Residual Heat Removal (RH) Pumps. The revision is less prest riptive, allowing the transfer to occur .either by pumping or by the gravity feed metbod, at the discretion of operaticas personnel.
S AFETY EVALUATION
SUMMARY
- A safety evaluation was performed for this FCR, The safety evaluation applicability review determi .i that this i revision constituted a change to procedures as described in the UFSA ihe safety evaluation determined that the gravity feed method, may be the preferred ..,nsfer method, has been evaluated by Engineering Evaluation No. 89-10 and does not introduce safety concerns. Tbc safety evaluation concluded that the revision would not create an vareviewed safety question. The revision will be incorporated into the Updated FSAR.
0 L
i- 160
.. . _ _ , . . = _ _ _ . _ . . _ _ . .
f,. 4 FINAL SAFLTY ANALYSIS REPORT CilANoe Rnounst: Number 91-050
- TITLE: Fuel Handling
SUMMARY
DESCRIITION: This Final Safety Analysis Report (FSAR) Change Request (FCR) revised UFSAR Section 9.1.4.2.b.3, ' Phase III - Fuel Handling.'
The revision changed the description of the general fuel handling sequence in the UFSAR.
The revised description encompasses both the full core offload and the core shuffle method of refueling.
SartrrY EVALUATION
SUMMARY
- A safety evaluation was performed for this FCR. The safety enluation applicability review determined that this revision - constitaed a change to procedures as described in the U FS A R. The safety evaluation . determined that the revised procedural descriptions did not change the method of using the-fuel handling equipment at therefore did not introduce a safety concern. The
!~ safety evaluation concluded that the revision would not create an unreviewed safety question.
The revision will be incorporated into the Updated FSAR.-
l l
l l
l l
l-l i-161
.w a FINAL SAFETY ANALYSIS REPORT CitANOB REQUEST: Number 91-057 TrrLE: Seabrook Station Tritium Control Plan
SUMMARY
DESCRIPTION: This Final Safety Analysis Report (FSAR) Change Request (TOR) revised Section 11.1.1.3 of the UFSAR which describes the plan for maintaining tritium in the Reactor Coolant System (RCS) at levels that allow reasonable access to the Containment. The revision replaced the current plan which specifies a 200,000 gallon feed and bleed of the RCS prior to each refueling outage with a revised plan which entails a periodic discharge or feed and bleed based on " tritium control points." The revision was intended to provide more operating flexibility.
SAFETY EVALUADON
SUMMARY
- A safety evaluation was performed for this FCR. The safety evaluation app;icability review determined that this revision constituted a change to procedures as described in the UFSAP The safety evaluation determined that the revision would eliminate unnecessary processing of primary coolant and actually reduce the probability of equipment malfunction. The safety evaluation did not identify any associated safety concerns. The safety evaluation concluded that the revision would not create an unreviewed safety question. The revision will be incorporated into the Updated FSAR.
I.
4 t
162 l
4 4 m
o o FINAL SAPLTY ANALYSIS REPR4T CllANGl! RitOUILm Nurnber 91039 TITI.H: Steam Generato: (SG) Blowdown Demineralizer Usage with Low Level Secondary Side Contamination 1
SUMMARY
Dl!SCRif" MON: This Final Safety Analysis Report (FSAR) Chang Request (FCR) revised various UFSAR sections. The revisir ns reflected the Ol'1 ion of proce:, sing steam generator blowdown using either the blowdown evaporators or the blowdown demineralizer system with low level secondary side contamination. The revisioni specify restrictions applicable to use of the blowdown demineralizer system with low level secondary side contamination.
FArtnT EVALUNnoN
SUMMARY
- A safety evaluation was performed for t'.is FCR. The safety evaluation applicability reviev det-: meed tht this
! revision constituted a change to procedures as described in the ~'NL The safety evaluation determined that operating the blowdown demineralizer system when low levels of secondary side contamination exist would not result in the blowdown demineralizer system operating under conditions exceeding its design capability. The safety evaluation concluded that the revision would not create an unreviewed safety question. The revision will be incorporated into the Updated FSAR.
l 163 l-4
___ . ._ ... _ ___ m-..._ - _ - _ _ _ .-s.. _ . _ _ _ . ._ _ _ _ _ . . _ . _ . . _ . _ - _ _ _ . _ _ . . _
, .=
FINAL S AlvrY ANALYSIS REI* ear CilANOH Hitotnwr Nuraber 91-075 Trrt.tu Ocneral Update of ' Fire Protection Prograra Evaluation' and Cornparison to LITP APCSil 9.5-1, Appendix A' i
SUMMARY
D ESCIUI"rIO*' This Firul Safety Analysis P,eport (FSAR) Change Request (FCR) c'.raisted of a general updtte of the subject document based on a techtncal review. The majority of the chanp> were editorial or administrative. in several lire areas /rones, minor revisions were made to the combustible fire kading to reflect actual plant conditions. These revisions r!!d not affect existing fire protection systems or manual firefighting capability and did not result in recommendations for new fire protection modifications.
SAIVrY I:VAll!ATION
SUMMARY
- A safety evaluation wari performed for this FCR. The safety evaluation applicability review determined that the Fire Protection Program Evaluation and Comparison to BTP APCSil 9.51. Appendix A is incorporated by reference into the FSAR The safety evaluatiot determined that the changes
' did not reduce the effectiveness of the Fire Protection Piogram or affect the conclusions of the 10CFR$0, Appendix R analysis. The safety enkiation concluded that the changes did not create an unreviewed safety question. These changes will be incorporated into the Updated FSAR.
'u k
164
, - - - -* -... , , ,, ,..,a---.--,.w.e.,,,m..+-,.-,,-%= * - - - e+a r. w -
. e
- 4
- 9. Miscellaneous 10CFR50.59 Evaluationg, The following additional safety evaluations were performed pursuant to 10CFR50 59.
165
m e l
TITt.n: Cycle 2 Reload Core Design Safety Evaluations j
SUMMARY
DtiSCR1l'rlON: A safety evaluation was performed for Seabrook Station by i Yankee Atomic entitird the ' Safety Evaluation for Cycle 2 Reload Core Design." This safety evaluation encompassed the Westinghouse safety evaluation und the boron dilution analysis identified bt 'ow. The Westinghouse afety evaluation encompassed all events normally considered by Westinghouse as part of reload esa!uation methodology, except the boron dilution analysis, which is within the Yankee Atomic . ope of analysis for Seabrook Station. A boron dilution analysis was performed for St abrook Station by Yankee ,
Atomic.
The boron dilution event was re analyred by Yankee Atomic using the beginning of Cycle '
2 core physics data. An acceptance criteria specified by NUREG 0800 (the Standard Review Plan) for this analysis is that, if operator action is required to terminate the transient, a minimum of fifteen minutes must be available b-tween the time when an alarm announces an unplanned rnoderator dilution and the time of lo s of shutdcwn margin in Operational Modes 1 through 5. For Seabrook Station, the alarm is provided by the Gammametrics Shutdown Monitor System. The boron dilution re analysis determined that, in order to satisfy ,
this acceptance criteria, the maximum dilution flow rates with filled laops in the early part l of Cycle 2 is 118 gpm during hot shutdown and 107 gpm during cold shutdown. The Seabrook Updated FSAR currently assumes a 150 gpm dilution flow rate for this event. The 150 gpm flow rate corresponds to the capacity of one reactor makeup water (RMW) pump.
In Operational Modes 4, 5 and 6, Technical Specifications require that the fioron Thermal Regeneration System (BTRS) be isolated from the RCS and that reactor makeup systems be inoperable except for the delivery capacity of one RMW pump.
As a result of the boron dilution re analysis findings and condusion summarized above, the Station Staff decided to administratively control the position of valve RMW.V34 (RMW pumps discharge to the boric acid blender) in Modes 4 and 5 to limit reactor makeup flow to the blender to a maximum of 107 gpm until such time in Cycle 2 that the flow rate limitation no longer applies, Also, normally closed valve RMW V36, which provides a flow path from the RMW pumps discharge to the charging pump suction will tw administratively maintained closed.
PURPar.U: The purpose of these safety evaluations was to determine whether or not the Cycle 2 Reload Core design involved an unreviewed safety question.
S AFUTY EVALUATION CONCLUSIONS: The conclusion of the Yankee Atomic Cycle 2 Reload safety evaluation was that introduction of Region 4 fuel and reconstitution of the core in confor lance with Cycle 2 design will neither involve an unreviewed safety question nor require an Operating License amendment. The conclusion was conditioned upon limiting RMW flow to the boric acid blender to a maximum of 107 gpm in Modes 4 and 5 during the early portion of Cycle 2. The conclusion of the boron dilution re e.nalysis was that reconfiguration of the core for Cycle 2 will not result in the creation of any unreviewed safety question provided that flow from a single RMW pump to the boric acid blending ' tee', CS MM 1, in the Chemical and Volume Control System (CVCS) is limited to a maximum of 107 gpm during the early portion of the cycle. The
- Westinghouse reload safety evaluation determined that the Cycle 2 core reload does not result in the safety limits for any accident being exceeded and will not adversely affect the safety of the plant, FCR 91-053 166
. - ~ - .. -- -_- - . - - - . . . . . , - - , - . . . .
s ENGINT!!! RING EVALUATION: Number 91-42 Tr1Lt!: Contamination of Nori radioactive Sptems
SUMMARY
DESCRll' TION: This Engineering Evaluation reviewed the consequences of the inadvertent radioactive contamination of the Demineralized Water (DM) System and certain other fluid systems at Seabrook Station which interface with the DM System. Several of these contaminated or potentially contaminated systems are described as non radioactive in the Updated FSAR (UFSAR). The DW System became radioactively contaminated as a result of reverse flow of reactor coolant frorr. the letdown line to the DM System throogh the Letdown Gross Activity Monitor purge connection. This Engineering Esaluation was performed in. accordance with the guidance of NRC IE Bulletin No. 8010.
PUnrosit: The purpose of this cvaluation was to determine if operation of certain non.
radioactive syst e ms, which had become contaminated as a result of the inadvertent contamination of the DW System, was acceptaole (i.e., did not involve an unreviewed safety question or require a change to the Technical Specifications).
SAFinY EVALUATION
SUMMARY
- A safety evalua*. ion was performed as part of this engineering change. The safety evaluation included a radiological assessment of the consequences of the inadvertent contamination of the DW System and other interfacing systems. The radiological analysis of expected and potential releases to the environment from this upset cotdition concluded that all Technical
- Specification dose limit objectives for routine ope +ation, as well as effluent release j concentration liraits per 10 CFR 20, were complied with.
! The safety evaluation concluded that operation of the DW System and certain other non-t radioactive syste.ns at Seabrook Station with low levels of radioactive contamination did not
! require a change to the Technical Specifications and did not involve an unreviewed safety l quesilon.
(
167
RI!OULWI FOR ENGINt!!! RING St!Rvicus: Number 91 16.5 Tri t ti: Bartlett Nuclear Dry Activated Waste (DAW) Trailer
SUMMARY
DESCRifflON: This Request for Engineering Services (RES) requested a "50.59' evaluation for locating a Bartlett Nuclear Dry Activated Waste (D AW) Trailer in a position adjacent to the south wall of the Fuel Storage Building (FSB),
which is outside the Radiologically Controlled Area (RCA). The DAW Trailer is a 12 foot by 48 foot trailer designated non safety-related, non-seismic. It houses equipment to sort, bag, seal, and monitor dry activated waste material, pORPOS11: The purpose of this evaluation was to determine whether or not locating and opera1g the equipment in the DAW Trailer in its proposed location would require a license amendment or involve an unreviewed safety question.
sal'INY EVALUATION
SUMMARY
- A safety evaluation was performeo as the disposition of this RES to determine whether or not the proposed location and use of the DAW Trailer would require a license amendment or involve an unreviewed safety question. The safety evaluation applicability review determined thet the DAW Trailer, being a temporary facility, would not be specifically described in the FSAR.
The evaluation addressed two radiological concerns: 1) ensuring that dose rates outside the trailer did not exceed restricted area limits, and 2) ensuring that effluents did not exceed limits for an unmonitored pathway. The safety evaluation deterniined that location and operation of the DAW Trailer adjacent to the sooth wall of the FSB would not result in dose rates outside the trailer exceeding restricted area limits. The safety evaluation further determined that radioactive, gaseous effluents from the ' DAW Trailer would he insignificant. Periodic sampling would be performed to confirm this determination. The safety evaluation concluded that the location and use of the DAW Trailer at proposed would not require a license amendment or involve an unreviewed safety question.
l l
168
n t i
RecutisT l'OR ENGINt!!' RING S!!RVIC11s: Numbr 91 255 Tr1 Lit: Temporary Enclosure for Access / Egress to the Radiologically-Controlled Area (RCA)
SUMMARY
Dl!SCRII'llON; This Request for Engineering Services quested en Engineering review of a proposed temporery enclosure (vestibule) to house two personnel rnonitors and a small articles monitor. The vestibule was to be built as an independent unit of wood construction located outside the Containment RCA access door (Door # E M 414). The vestibule war, to be used only during the refueling outage.
Punt'Ost!: The purpose of this evaluation was to provide recommendations regarding construction and use of the vestibule including safety, radiological and ALARA considcrations.
1 SArirri EVAI.UA'I10N
SUMMARY
- A safety evaluation was performed as part of the disposition of this request for engineering services to verify that the vestibule, which would be a temporary extension of the RCA, did not involve an unreviewed safety question. The UFSAR anticipates the use of alternate facilities to support RCA ingress /cgress during maintenance and refueling outages. Adherence to the engineering recommendations and application of Seabrook Station Radiation Protection Prograrn procedures would ensure that use of the temporary vestibule would not violate site radiological safety and other commitments. Tbc safety evaluatio:, determined that the vestibule would not adversely affect safety related structures, systems or coroponents, create a fire hazard, or impact the integrity of barriers designed to contain, process and control the release of airborne radioactivity.
The safety evaluation concluded that the installation and use of the temporary vestibule i
would not involve an unreviewed safety question.
l l
l l
l 1
169 i
e, r TrrLE: Temporary Storage of Dry, Activated Waste in the Unit 2 Cooling Tower PURPOSE: N}3Y is not allowed to ship low-le vel dry, activated waste (DAW) to a repository for disposal. Therefore, it is necessary to temporarily store DAW on site. The purpose of the safety esaluation was to determine whether or not ternporary storage of DAW in the Unit 2 Section of the Service Water Cooling Tower at Seabrook Station involved au unreviewed safety question.
SUMMARY
DESCRIPTION: This safety evaluation addressed the NilY program to store low level dry activated waste (DAW) on an interim basis within the Unit 2 Electrical Smtchgear Room of the Se vice Water Cooling Tower at Seabrook Station.
The safety evaluation applicability review dete..nined that storage of low-level radioactive waste or DAW in the Unit 2 Cooling Tower was not within the scope of the original Cooling Tower Design. The safety evaluation determined that storage of low level radioactive weste or DAW in the Unit 2 Cooling Tower neither affected operation of the Station not revised any system parameters or operating instructions. The storage of DAW in the Unit 2 section of the Cooling Tower revises neither estimates of radioactive waste generation nor other details related to the getieration, handling, processing or cororolling of radioactive, waste as described in the FSAR. The PSAR will be revised to reflect the storage of DAW in the Unit 2 section of the 3ervice Water Cooling Tower. The Unit 2 Section of the Service Wcter Cooling Tower is a Seismic Category 1 structure desi;ned to withstand the derign basis tornado, hurricane and flood. Existing Station programs for radiation and fire protection will be utilized to ensure that exist;ng limits for on site and off-site exposure will not be exceeded. All site beundary doses to the general public will be maintained less than the generally accepted guidance for interim storage of DAW as specified in NRC Generic Letter 81-38, ' Storage of Low Level Radioactive Wastes at Power Reactor Sites."
SAFETY EVALUATION CONCLUSIONS: The safety evaluation concluded that the interim storage of low level radioactive waste or DAW in the Unit 2 Cooling Tower does not involve. an unreviewe 3 safety question.
170 o
r e Trrt.tr Justification for Continued Operation, Field Weld 1-F1188 01-F0150 )
I I
Pt Rrost!: In 1991, NH Y conducted a ' Weld Records Reverification Program? This program serified the accomplisbroent of required radiography and completeness of radiography records for Pullman.Higgins (Pil) field welds for which the applicable Code specified radiography. The program required that "indete minate items" (anomalies) which 1 indicated either an unacceptable wcld or unacceptable weld documentation would be reported i to the NRC. A safety evaluation was also prepared for each weld record anomaly. This I safety evaluation was prepared for the weld record anomaly associated with Field Weld 1-F1-185 01-F0150.
SUMMARY
Dl!SCRlrrION: pullman.Higgins field weld 1-M 188-01 F0150 is a circumferential groove weld on a twelve inch diameter sparc electrical penetration outside the Containment Building. This electrical penetration is ash 1E Ill, h1C '
Class, and the weld is a Category B joint. Thi; weld connects the penetration sleese to a '
prefabricated end cap. This weld joint employs a machined-in t,acking ring, which fits insid-of the end cap assembly. The Pullman-H;ggins Field Wcld Process Sheet demonstrates thai this field weld was radiographed in 1983 in accordance with the Non Destructive Examination (NDE) requiicments contained in the 1977 Edition of ash!E Section III up through and including the Winter 1977 Addenda (the Code applicable to Seabrook Stationb The weld records package for weld 1-FI 183-01-F0150 could not be located. Based on the i lack of weld record documentation, this weld did not meet the record retention requirements of ash 1E N CA-4134.17 and ANSI N45.2.9. However, other available documentation dernonstrated that the radiographic records for this weld were reviewed as part of the Seabrook Station as built records verificatier> nrocess.
Corrective actiona completed in response to this records deficiency were to: 1) complete a radiographic exarainaticn of field weld 1-FI-188 01-F0150; 2) review the film in accordance with the present programmatic requir.:ments; arsd 3) include the radiograph and the required radiographic review forms in the NHY records managema: t system. These actions ensured t.ompliance with the Code.
S AFInY EVALUA110N CONCLUSIONS: The safety evaluation determined that the identified records deficiency neither compromised the bi tegrity of the Containment Building nor affected the operation of the Station. The mere presence of a records deficiency did not introduce a new failure mechat ism nor did it modify the plant in any manner so as to create the possibility of a new accident or malfunction occurring. This record deficiency did not provide any means for an increase in the dose froru any previously analyzed accident and did not make any changes to the plant or its design basis. Therefore, the safety evaluation concluded that the identified records deficiency did not involve an unreviewed safety question.
171 l
I
- O
, l l
J Tm.ti: Justification for Continued Operation, Ficid Weld 1-CS-318 02 F0202 P URPOSti: In 1991, NHY conducted a " Weld Records Reverification Program.' 'I bis program verified the accomplishment of required radiagraphy and completeness of radiography records for Pullman-Higgins (Pl!, sield welds for which the applicable Code specified radiography. 'Ibe program required that ' indeterminate items" (anomalies) which indicated either an unacceptable weld or unacceptable weld documentation would be reported to the NRC. A safety evaluation was prepared for each weld record anomaly. This safety evaluation was also prepared for the weld record anomaly associated with Field Weld 1-CS-31502-F0202.
SUMM ARY Dt!SCRl!"110N: Puliman Higgins ficli Aeld 1-CS 318-02 F0202 is a circumferential butt weld on a three inch diameter section of piping in the Chemical and Volume Control System (CVCS). This section of the CVCS is ASME 111, Class 2, and Safety Class 2 This weld connects a valve to the piping and is alsc adjacent to a reducer. This weld is located in the letdown line of the CVCS downstream of the Letdown lleat Exchanger (Tag number CS-E-4).
The weld records package for weld 1.CS 318 02-F0202 contained a Radiograph Inspection Report (RIR) and the radiographic film. The RIR indicated that the radiograph views for all stations of this weld were of acceptable quality. The RIR also contained the required approval signatures.
A review of the radiogrr;aic film for this weld was conducted to evaluate issues raised by the NRC during a previous inspection. This evaluation confirmed that the required sensitivity was achieved in the necessary penetrameters in the films for all stations of this weld. Additionally, the density through the body of the penetrameters met the requirements l of the ASME Ccde. The weld area of interest in each film also met the density l requirements of the same Code provisions. However, this revitw also revealed that the l comparative densities of the ;enetrameters to those ir. the weld area of interest exceeded the minimucJmaximum density limitation raeges specified in the ASME Code.
Corrective actions corupleted in response to this weld record anomaly were to: 1) enmplete a radiographic examination of field weld 1 CS 318-28 F0202, 2) review the film in accordance with the present programmatic requirements and 3) include the radiograph and the required radiographic review forms in the NHY records management system. These
! actions ensured compliance with the Code.
l r
SAITry EVALl'A110N CONCLUSIONS: The safety evaluation determined that the identified records deficiency neither compromised the integrity of the CVCS nor affected the operation of the Statiou. The mere presence of a records deficiency did not introduce a new failure mechanism nor did it modify the plant in any manner so as to create the possibility of a new sccident or malfunc: on occurring. This record deficiency did not provide any means for an increase in the dose from any previously '
analyzed accident and did not make any changes to the plant or its design basis. Therefore, the safety evalui. tion concluded that the identified records deficiency did not involve an unreviewed safety question.
l2 l
1
- g - *-- - , y --f
= a o Trn.n: Justification for Continued Operation, Field Weld 1 RC-13-02-F0203 poluosin in 1991, NHY conducted a ' Weld Records Reverification Program / 'I his program verified the accomplishment of required radiography and completeness of radiography records f ar Pullman Higgins (PH) field welds for which the c.policable Code gecified radiography. The program required that " indeterminate items" (anomalies) which indicated either an unacceptable weld or unacceptable weld documentation would be reported to the NRC. A safety ev:luation was also prepared for each weld record anomaly. This safety evaluation was prepared for the weld record anomaly associated with Field Weld 1-RC 13-02.F0203.
SUMMARY
D ESCRIPTION: Pullman-Higgins field weld 1-RC-13 02-F0203 it a circumferential butt weld on a twelve inch diameter section of piping in the Residual Heat Removal (RH) Sys'.:m, This section of the RH System is ASME 111, Class 2, and Safety Class 2. This weld is located adjacent to check valve CBS V 55 in line 1209 02, which is the RHR Pump 8A supply from the Refueling Water Storage Tank. This field weld was radiographed in 1961 in accc'rdance with the Non Destructive Examination (NDE) requirements contained in the 1977 Editior. of ASME Section til up through and including the Winter 1977 Addendu (the code applicable to Seabrook Station).
The weld records package for weld 1-RC-13-02-FO203 contaius a Radiograph inspection Report (RIR) and the radiographic film. The RIR and the radiographic film for only one of this weld's four stations (station 3-Oi, contain the information and approvali required by l
the Code. As identified in NHY Corrective Action Request (CAR)91-010, the radiographic l film for stations 01, 12, and 2-3 lacked the identificatian of the exposure date, system /line/ isometric number, weld number, and manufacturer's identification. The only information contained on the film for these three stations was the station number.
i Therefore, the film for these three stations did not rneet Code requirements.
NHY Nuclear Quality Group personnel verified that the radiographic film for stations 0-1, 1-2, and 2-3, were b fact that of weld F0203. F0203 is the only film available for this weld.
Review of the weld process sheets indicates that no repairs were made to this weld before or after the weld wris radiographed.
Corrective actions completed in response to this weld record anomaly were to: 1) permanently identify the Code required information on the radiographic film for field weld 1-RC-13-02-F0203, and 2) Refercca the Corrective Action Request (CAR) on the film package for this weld. These actions ensured compliance with the Code.
SAFirTY EVALUNnON CONCLUSIONS: The safety evaluation determined that the identified records deficiency neither compromised the integrity of the RH System nor affected the operation of the Station, It was verified that l radiographic film existed for all stations for field weld 1-RC-13-02-F0203. 't he weld record deficiency did not int oduce a new failure mechanism nor did it modify the _ plant in any manner so as to create the poscibility of a new accident or malfunction occurring. This record deficiency did not provide any rucans for an increase in the dose from aur previously analyzed accident and did not make any changes to the plant or its design basis i he re for e, the safety evaluation concluded that the identified records deficiency did not involve an unreviewed safety question.
173
e es 8 l f
Tm.in Justification tor Continued Operation, Field Weld 1 CBS 1201-07 F0701 Pulu'osin 'n 1991, NilY conducted a " Weld Records R es e rification Program." This program serified the accompli <hment of required radiocraphy and completeness of radiography records for Pullman liiggins (Ph) field welds for which the applicable Code specified radiography. The program required that indeterminate items" (anomalies) which indicated either an unacceptable wcid or unacceptable weld documentation would be reported to the NRC. A safety evaluation was also prepared for each weld record anomaly. This safety esaluation was prepared for the weld record unomaly anociated with Field Weld 1-CBS 1201 d7-F0701.
SUMMARY
DESCRIITION. Pullman liiggins field weld 1-CBS 1201-07-F0701 is a circumferential groove weld on a fourteen inch diameter section of piping in the Containment Building Spray (CBS) System. This section of the CBS Sprem is ASME Ill, Class 2, and Safety Class 2. This weld is located in a section of piping between the Refueling Water Storage Tanh (RWST) and the suction of the T Train Containment Buildirg Spray Pump. This line also Prmides a source of water for the "B" Train Safety injection Pump z.nd the B' Train Residual llent Removal Purup. The Pullman.
Higgins Field Weld Process Sheet demonstrates that this field weld was radiographed in 19S3 in accordance with the Non Destructive Ex4'niaation (NDE) requirements contained in the 1977 Edition of ASME Section 111 up through and including the Winter 1977 Addenda (the Code applicable to Seabrook Station).
The weld records package for weld 1 CBS-1201-07-F0701 could not be located. Based on the lack of weld record documentation, this weld lid not meet the record retention requirements of ASME NCA-4134.17 anu ANSI N45.2 0 However, other available documentation demonstrated that the radiographic records for this weld were reviewed as part of the Seabrook Station as built records verification process.
Cerrective actions completed in response ta this records deficiency were to: 1) complete a radiographic examination of field weld 1.CBS 1201-07 F0701; 2) review the film in accordance with the present programmatic requirements; and 3) include the radiograph and the required radiographic review forms in the NHY records management system. These actions ensured compliance with the Code.
SMT1T EVA1 UAUON CONCLUSIONS: The safety evaluation determined that the identified records deficiency neither compromised the integrity of the CBS System nor affected the operation of the Station. The mere presence of a records deficiency did not introduce a new failure mechanism nor did it modify the plant in any macner so as to create the possibility of a new accident or malfunction occurring. This record deficiency did not provide any tneans for an increase in the dose from any previously analyzed accident and did not make any changes to the plant or its design basis. Therefore, the safety evaluation concluded t'aat the identified r. cords deficiency did not involve au j unreviewed safety question.
174 l
+b Tnu: Justification for Continued Operation, Field Weld 1-CS-360~08 FUS01 1
PURI'OSit in 1991. NHY conducted a ' Weld Records Reverification Program.' This program verified the accomplishnwnt of required radiogtaphy and completeness of radiography records for Pullman Higgins (PH) field welds for which the applicable Code specified radiography. The program required that " indeterminate items" (anomalies) which indicated either an unacceptable weld or unacceptable weld documentation would be reported to the NRC. A safety evaluation was also prepared for each weld record anomaly. This safety evaluation was prepared for the weld record anomaly associated with Field Weld 1-CS 360 OS FOS 01.
SUMM ARY Dh5 Crit"nON: Pullman Higgins field wcld 1-CS 360 05-F0B01 is a circumferential butt weld on a four inch diameter section of piping in the 1 I
Chemical and Volume Control System (CVCS). This section of the CVCS is ASME 111, Class 2, and Safety Class 2 that connects a 90* long radius elbow to a section oi piping. This weld is lented in the letdown line of the CVCS downstream of the Regenerative Hest Exchanger (Tag number CS-E-2) and upstream of the Letdown Heat Exchanger (Tag nurnber j CS-E-4). 't he weld records package for weld 1-CS 3e.0 08-F0801 could not be located. Based i
on the lack of weld record documentation, this weld did not meet the record retention I requirements of ASME NCA 4134.17 and ANSI N 4 5.2.9. However, other available de tumentation demonstrated that the radiographic records for this weld were reviewed as part of the Seabrook Station as huilt verification process.
Corrective actions completed in response to this wew ucord anomal) were to: 1) comp!cte a radiographic examination of field weld 1 CS 360 08-FOS 01, 2) review the filra in l accordance with the present programmatic requirements and 3) include the radiograph and the required radiographic review forms in the NHY records manacement system. These actions ensured compliance with the Code.
i S AfTTY EVALUATION CONCLUSIONS: The safety evaluation determined that the identified records deficiency neither cornpromised the integrity of the CVCS nor affected the operation of the Station. The mere presence of a records deficiency did not introduce a new failure mechanism nor did it modify the plant in any manner so as to create the possibility of a new accident or malfunction occurring. This l record deficiency did not provide any means for an increase in the dose from any previously l analyzed accident and did not make any changes to the plant or its design basis. Th:refore, the safety evaluation concluded that the identified records deficiency did not involve an unreviewed safety question.
l l
175
- . - - _ _ .