ML20236D049
| ML20236D049 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 12/31/1988 |
| From: | Creel G BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8903220436 | |
| Download: ML20236D049 (12) | |
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10CFR50.59 Annual Report Enclosure 1 Pace 1 i l
l 79-0110 (Units 1 & 2)
This FCR provided for a spreader beam type rig designed and i supplied by Par (the vendor) to facilitate lifting of the refuel-ing machine trolley vertically up off the bridge to access the lower hoist box region for maintenance purposes. The rig con-sists of a 14 foot long 10WF beam with two chains extending from two lifting lugs on the top of the beam to a common link.where the polar crane is to be attached and four sling chains extending from two lugs at the bottom of the beam to all four corners of the refueling machine trolley. This tool is used during shutdown and refueling modes. Extension cables have been furnished for both the electrical and air supplies so that the control console located on the trolley will remain functional during the lifting process. This will allow for immediate testing of the machine after repairs have been made. This FCR did not constitute an unreviewed safety question or Technical Specification change.
~ 80-0017 (Units 1 & 2)
There were six safety evaluations written under this FCR between 1981 and 1987. The original intent of this FCR was to address erosion problems downstream of 1(2)CV-5209 and -5214. These valves were added by FCR 75-1100 in an unsuccessful attempt to control erosion downstream of the Service Water Heat Exchanger discharge throttle valves, 1(2)CV-5210 and -3212. The basic idea of this FCR was to lock open the newer valves and rubber-line the piping for some distance downstream of the original throttle valves. In later years, the FCR was also used to issue supple-ments allowing rubber-lining of salt water piping and replacement of piping spools elsewhere in the salt water system. Safety analyses 1 and 2 accompanied supplements which allowed locking open 1(2)CV-5209 and -5214 and allowed the rubber lining of piping downstream of these valves. Safety analysis 3 provided a justification for removing all engineered safety features actua-tion signals from these valves as well as the actuatorc, con-trols, and air lines. Safety analysis 4 addressed the removal of the valve disks from 1(2)CV-5209 and -5214 and the modification required to maintain the pressure boundary integrity of the valve bodies, which remained in the piping. Safety analysis 5 provided a detailed, thorough justification for allowing the rubber-lining of any LJ-l Class piping in the salt water system. The original specification for LJ-l piping called for cement-lined pipe. The evaluation covered the following issues: Code conformance; effect on system flow characteristics; effect on the piping stress analysis; equivalence of rubber-lining versus cement as a corrosion / erosion barrier; possibility of rubber-lining peeling off and blocking or fouling downstream components. Safety analysis 6 provided a detailed justification for a '.owing a spool piece made from API-5L steel to be used in some LJ- piping. The original specification for LJ-l called for ASTM A-5a pipe. The evaluation considered factors such as material strength, weldability, and code conformance, f/
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1 l 10CFR50.59 Annual Report Enclosure 1 Pace 2 l I
82-0025 (Units 1 & 21 This modification replaced small vent valves on the main steam ,
piping with blind-flangea connections. The vent valves were l l originally intended to allow venting of the steam generators i during nitrogen purging. Nitrogen purging was not implemented, and the valves became an unnecessary maintenance item. The change to the blind-flanged connections conformed to ANSI B31.1, i the Construction Code for the main steam piping. The FSAR had no description of these vent valves nor of their function, except for the inclusion of these valves in one of the FSAR figures (Figure 10-1). The appearance of those valves on this FSAR figure was incidental; the figure is simply a copy of the BG&E Main Steam System Piping and Instrumentation Drawing, which of course shows all drain and vent valves regardless of signifi-cance. Safety evaluations were written for this modification.
In retrospect, however, this change falls outside the bounds of 10CFR50.59.
l 82-0045 (Units 1 & 21 l l
This FCR deleted the Containment Isolation Signal from the Steam l Generator Blowdown valves. Though this removes the automatic closure of these valves, the valves still possess the capability l of being closed remotely from the control room. These valves I constitute a Type III piping penetration which requires either an automatic, locked closed, or remote normal operation isolation valve. The Steam Generator Blowdown valves continue to meet the I design criteria for Type III penetrations as described in the l FSAR, and Criterion 57 of 10CFR50. l 83-0111 (Units 1 & 2) l This FCR added a manual isolation valve on the pressurizer vapo.c sample line of the Reactor Coolant Sample system. This change prevented the leak-by from the control valve on the pressurizer vapor sample line from contaminating the Reactor Coolant hot leg sample by pressurizer vapor. The safety analysis reviewed the relevant FSAR sections and concluded that the criteria for system operability was not affected. The change complied with the original construction code, piping class sheets. A stress analysis concluded that existing piping and supports remained adequate. RCS chemistry limit requirements were not affected.
83-0124 (Units 1 & 21 Sight flow indicators were installed in the common drain line that comes from 1/2-Pl-214, x, y, z (Charging Pump Discharge Pressure). The purpose of the addition is to ensure that the discharge lines and pressure instruments are adequately vented prior to placing the charging pump (s) back in service. The engineering effort included the selection of materials and routing of tubing to ensure compliance with code requirements.
The safety analysis addressed the adequacy of the design (i.e.
materials, tube routing) to ensure it would have no negative
10CFR50.59 Annual Report Enclosure 1 Pace 3 impact on accident prevention or mitigation. Note that the new' indicator and tubing are only SR for seismic II/I considerations.
The SR boundary stops at the normally closed upstream drain valve.
83-0136 (Comaon1 This FCR allowed installation of a local static head pressure gauge to indicate diesel fuel oil day tank level on all three tanks (11, 12 and 21). The local gauges were constructed to seismic, deadweight and thermal criteria contained in M-500, Instrument Installation Details. Appropriate FSAR and Technical Specifications sections were reviewed. The conclusions were that neither document was affected.
83-1003 (Unit 1) i Safety Analysis #1 replaced nine containment isolation solenoid valves with environmentally and seismically qualified valves.
The qualification of equipment was to eliminate the possibility of a common mode failure of equipment under accident conditions which could degrade our ability to maintain the plant in a safe shutdown condition. The valves involved were:
1-SV-6507A -
North Primary Shield, Cntmt Atmosphere H Sample 1-SV-6507B -
South Primary Shield, Cntmt Atmosphere H Sample 1-SV-6507C -
Pzr Compartment, Cntmt Atmosphere H 2 Sample 1-SV-6507D -
East El 135', Cntmt Atmosphere H Sample 1-SV-6507E -
West El 135', Cntmt Atmosphere H Sample 1-SV-6507F -
Dome El 189', Cntmt Atmosphere H Sample 1-SV-6507G -
Cntmt Atmosphere H 2 Sample Return 1-SV-6529 -
Pass RCS Sample Liquid Return 1-SV-6531 -
Pressurizer Quench Tank 0 2 Sample Safety Analysis #2 reviewed the use of Raychem on a cable splice inside a junction box. These safety analyses concluded that these modifications did not constitute unreviewed safety ques-tions.
83-1057 (Units 1 & 2)
The Core Exit Thermocouple (CET) system was installed as part of the Inadequate core Cooling (ICC) modifications required by NUREG-0737 and the Post *"I-2 Action Plan. The CET system consists of 47 fixed in-core detectors which feed primary (non-1E) and backup (1E) systems. The primary system display is the Safety Parameter Display System (SPDS) which is part of the Plant Computer System. The backup display consists of digital indicators with selector handswitches mounted on the main control boards. Multichannel trend recorders record all CET channels.
Some lE cable assemblies remain to be installed during the Unit 2 Spring 1989 Outage. The system design was accomplished within the bounds of criteria detailed in the FSAR and NUREG 0737.
Auxiliary system loads were reviewed to ensure adequate capacity during and after design basis events. Isolators were installed i
,' 10CFR50.59 Annual Rqnort Enclosure 1 Pace 4 !
where 1E and non-1E circuits shared common signals. There was no {
effect on the reactor vessel or piping components. This modifi- j cation did not constitute an unreviewed safety question. j M3-0152 (Units 1 & 2) j This FCR was written to allow a change in the intake sluice gate material to ASTM A439, Type D3 Austenitic Ductile Iron which ic a ;
more corrosion resistant material. The FCR also provided for j It was assumed that-the ;
retainer clips and their installation.
internal threads of the gate thimble may be graphitized similar to the original gate material. Since these threads cannot be vicually inspected it was necessary to provido a redundant means to assure the gate would not fail when required to close. The only safety-related aspect of this FCR involved the installation l of the Hilti Company's stainless steel concrete wedge anchors !
which were installed in accordance with Civil Standard CS-5, and j the stainless steel retainer clips. Civil calculation C-85-16 ;
was prepared in this analysis. Also, the probability of I cccurrence or the consequences of an accident not previously h evaluated in the FSAR is not increased since no changes were made j in the operation or design of the saltwater system.
85-1050 (Common) l This FCR was written to replace the intrusion detection system !
along the intake structure with one using a different operating {
principle. The current system had a high rate of false alarms. l The detection equipment itself is non-safety related. The )
bracket anchors are safety-related because they are mounted more j than two inches deep in the face of the intake wall. The brack- J ets were designed to use Hilti brand stainless steel concrete j wedge anchors. These anchors were coated with a waterproof epoxy sealer to prevent corrosion of the rebar and installed in accor-dance with CS-5. None of the rebar in the intake wall was cut.
The structural analysis of the brackets included additional loads for seismic effects.
86-0022 (Unit 1)
This FCR was written to modify the piping on the discharge, suction, and vent lines for No. 12 Charging Pump to install the replacement charging pump block. The deadweight, seismic, and thermal stresses for the piping system have been analyzed and do not exceed the maximum allowable stresses as a result of this modification. This modification does not constitute an unreviewed safety question.
7 Enclosure 1 Pace U
, 30CFR50c59AnnualReDort 86-0023 (Unit 11 This FCR was written to redesign a base plate on the main steam isolation valve monorail due to the failure of two bolts to pass
! a load test. Due to existing' interferences it was necessary to modify the base plate to maintain its original design capacity.
The safety analysis addressed the fact that the modified basa
) plate was designed using the required design criteria and main-tained the original margins of safety.
g5-0119 (Units 1 & 2)
This FCR deleted eight (8) snubbers in the Plant based on a re-analysis of the piping systems using ASME Code Case N-411.
Six (6) snubbers in Unit 1 have b?en deleted which required the addition of one rigid restraint and the modification of three supports due to new loads. Two (2) snubbers in Unit 2 have been deleted without requiring any modification to existing supports.
The snubbers deleted are:
System Snubber Installed on Snubber No. Location and ElevatioD l-55-3 Component Cooling, Ctnmt (Area 22), El 88'-6" l-69-1 Aux Steam, Aux Bldg (Area 17) El 33'-5" l-71-1 Waste Gas & Misc Waste Process, Aux Bldg (Area 17)
El 83'-4" l-71-2 Waste Gas & Misc Waste Process, Aux Bldg (Area 17)
El 71'-0" 1-71-2A Same as Snubber No. 1-71-2 l-71-3 Reactor Clnt Wste, Aux Bldg (Arca 17), El 39'-3" 2-71-3 Reactor Clut Wste, Ctnmt (Area 25), El 20'-4" 2-71-5 Reactor Cint Wste, Ctnmt (Area 25), El 12'-6" There were no unreviewed safety questions or Technical Specifica-tion changes.
86-0183 (Units 1 & 2)
The gain of one input coefficient was doubled while the voltage across the set resistor was halved in order to provide optimiza-tion of the TM/LP curve fit and provide a gain in margin for DNB LSSS limits over the new 24-month refueling cycle. This circuit- i ry change did not change any reactor coolant system setpoints or exjsting input voltage values but provided the ability to adjust the TM/LP input over the range expected by the new 24-month fuel cycle. No unraviewed safety question was caused nor did this '
modification require a change in the Technical Specifications.
86-0189 (Units 1 & 2) I This FCR replaced or rebuilt fuel injection pumps for the emer-gency diesel generators (EDG), with pumps that have an erosion sleeve. The change was for the three EDGs (11, 12 and 21) to enhance their operability, while decreasing the chance of
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a 10CFR50.59 Annual Reo'rt o ' Enclosure 1 Pace'6'
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through-wall wear of the pump. The safety analysis reviewed.the )
effect on pump operability, method of repair, and possibility of-a new type of failure. .The conclusion was reached'that, after-review of applicable FSAR and Technical Specifications' sections, l an unreviewed safety question <did not exist.
86-0207 (Units 1 &-2) , ]q The FCR provided th+ engineering! design-to m6dify'the bolt holes j
. on each ' of the three. (3) femergency diesel generator (EDG) -
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Missile Resistant-Panels Nosi 11, 12 and:21. .Each panel has eighteen.(18). bolt' holes symmetrically located acrosssthe top and down . each side ' of its: heavy " (18 : WF 114) - steel frame. The six,(6) bolt holes along the.12p' face will be slotted to 1.1/4 x 2 1/2 inches;. the six (6) bolt -holes along each Lgjda face will be- I slotted to 1 1/4 x 2 1/2, inches. Each slot will be cut 1down the vertical axis of'the existing hole. ' Elongation of the bolt holes will facilitate installation andfremoval of the panels Mnd reduce- j the possibility of damaging-the one inch' diameter expansion
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l anchors:by providing greater bolt. hole, tolerance for-fit on=the expansion anchors.' There wasino effect on' seismic design org j structural integrity of the' panels as ai result of.this change.' )
There were no unreviewed safety questions'or Technical Specifica- H I
tion changes.
86-0223 (Common)
A system to detect and annunciate grounds on 125 VDC buses ll and 21 was installed. The equipment continuously monitors six (6)- i distribution' panels on each bus and annunciated in the Control Room when leakage current to ground exceeds 10mA. The sensors for this system are constructed very much like/a current trans--
former and have no connections to the D.C. buses and are not.
capable of degrading the performance of the D.C.. system. A resistor network installed in the Control Room creates the .;
reference to around by center-tapping.a four thousand' OHM '
resistor to ground. A resistor network is installed across the terminals of each of the two batteries. The' reference resisters i are large enough'that the ability of the battery to operate with 4 a single ground will be unimpaired. There were no unreviewed safety questions or Technical Specification changes.
0 87-0021 (Units 1-& 2)
This FCR was written to meet NRC security requirements. Tha - ;
original screens' covering the MSIV room vents were1 replaced with l heavy duty grating Velded to the vent' covers. The safety analy- ;
sis addressed the effects of placing the grating on " .-MSIV. room' !
vents. Calculations were performed to determine th effect.of the additional-weight of the grating on'the vent, the;effect of the grating on the vent flow characteristics and the ability-of the. grating to remain in place in.the event of a steam release.
The analysis determined that the addition of grating had ncC adverse effect on the vent and did not pose un unreviewed safety question.
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10CFR50.'59 Annual ReDort En51osure 1 Pace 1 11-0073 (Unit 2) ,
This FCR involved ~(1) drilling-aLsmall, threaded hole in the-packing gland area of the valve bonnet in ordor to accept an injection fitting, and (2) injecting scalant to stop the packing leak. The safety evaluation considered two effects. First, it looked at the addition of the injection f3tting to the' valve bonnet. This was explained to be covered by provisions'in ANSI B16.34 which allows " auxiliary connections" to valves within q stated guidelines. Second, the effects of sealant which might )
possibly extrude into the piping were' discussed. It was deter- ]
mined that the small amount of sealant wculd not be able.to 1 damate any Secondary System components. The package which issued l the safety evaluation included special instructions to ensure j that the operability of the MOV was maintained following the J injection of the sealant (i.e.,' that the sealant injection would i not cause binding of the valve utem).
87-0099 (Unit 1) j I
The steam gencrator feed pump speed controllers wero replaced j with similar model controllers made by the same manufacturer. ,
I l This equipment is not credited in anv safety analysis and does not interface with any other safety mystem. The equipment is mounted in a control room' panel where electrical separation criteria and seismic design-criteria were maintained. The system .
continues to operate as described in the FSAR. l l
87-0109 (Units 1 & 2)
This FCR replaced a compression / pipe male branch tee with a full compression tube union tee en the high and low flow rate sample lines on the wide-range gas monitor sample conditioning skid.
Replacement of the tee connector with an equisa.ent fitting which has a compression type outlet rather than a tl<r,aded outlet eliminates problems with threads galling and csnsequent difficul-ty in removing or resealing the purge outlet cap. The safety analysis pointed out that the replacement tee does not change the function of the WRNG monitor or system operability. Th6 new tee l is identical to those presently in use except for the compression '
l fitting outlet.
l l 87-0128 (Units 1 & 2)
The solenoid valves that control the service water supply control valves for the Emergency Diesel Generators were changed to a different model. The new solenoid valve is viewed as an enhance-ment since it meets more stringent qualifications (lE, EQ) and it is a universal type valve as opposed to a normally closed valve which will allow it to more reliably shut off the air supply when de-energized. The new solenoid valves were seismically mounted and meet the control valve stroking timeirequirements. No new l system operation or failure mode was introduced.
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10CFR50.59 Annual Report Enclosure 1 Pace 8 87-0143 (Units 1 & 2) l The steam supply check valves to the auxiliary feed pumps were replaced usi.ng Anchor Darling 900# class check valves. The i original valves were 600#, cast carbon steel A-216 Gr. WCB. The i replacement valves are 900# class, cast SA-216 Gr. WCB with a l valve bonnet made of SA-515 Gr. 70 instead of A-216 Gr. WCB.
Both valves e.re tilting disc check valves. The differences between the valves were evaluated and were determined to : 7t adversely affect the valve design. Therefore, an unrevier ad ;
cafety question did not exist. l 1
88-0023 (Units 1 6.11 1 This design activity engineered a one-for-one change out of l
obsolete I/P transducers on 1(2)-CV-4511/4512. The safety analysis described the use of the new Masonellan 8005N I/P's l which have a longer design life (10 years) than the obsolete l 8005A units (5.4 years). )
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_88-3000-1 (Unit 1) j l
This FCR supplement provided for the replacement of four fuel pins with stainless steel pins in the reload fuel of Unit 1 Cycle l 10. Stainless steel pins had been used and analyzed previously. i
, Thay are neutronically inert and have little effect on the i l performance of the core. All existing Unit 1 Cycle 10 safety
! analyses were maintained. There was no unreviewed safety ques-tion and no change to the Technical Specifications.
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AkCFR50.59 Aqnual Report Enclocure 2 Pace 1 j TempRrpry Lifted Lead 85-424 (Unit 1)
This temporary modification added a valve and tubing for tempo-rary sample points off of the steam generator level transmitter t variable legs in order to obtain chemistry samples of the steam i generator water with the normal semple system out of service. In l fact, with the unit on shutdown cooling, the entire secondary system is out of service and is not required for any core heat.
removal function, therefore, the installation could not cause an l accident or mitigate the consequences of an accident. Since the ;
plant was in shutdown cooling, no Technical Specification on the i secondary system applies since the unit was in Mode 4, 5 or 6 and the safety analysis required normal configuration to be estab- ,
lished prior to shutdown cooling being terminated. #
i Temporary Lifted Lead 85-425 (Unit 1)
This temporary modification involved the installation of tubing i for sight glasses in the tubing lines for steam generator level l transmitters LT-llll and LT-ll21 to provide level indication'of )
l the steam generator with the normal level transmitters out of i service and the unit in shutdown cooling. Steam generator level is not required by the Technical Specifications with the unit in j shutdown cooling. The steam generator performed no core heat 1 removal function with the unit in shutdown cooling, therefore, any break in the tubing would not cause an accident or mitigate the consequences of an accident. The request required that the normal configuration be restored prior to termination of shutdown cooling. !
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l Ippocrary Lifted Leads 88-40. 1-88-41, 1-88-83 (Unit 1) l l The safety analysis was written to qualify defeating the inter-I locks on the shutdown (S/D) cooling return headers isolation valves, 1(2)-SI-651, 652. The interlocks are used to prevent the S/D cooling system from being overpressurized by the RCS. The I safety analysis addressed the function of the interlocks and '
requirements, should they be defeated. To prevent S/D cooling from being overpressurized (pressurized above design pressure) a minimum square inches of openings in the RCS had to be main- l l tained. In conclusion, the safety analysis does not change the l bases of the Technical Specifications nor the FSAR. This modifi-l cation was temporary and the system has been returned to normal operation.
Irmocrary Lifted Lead 88-55 (Unit 1)
This electrical jumper and lifted lead modified channel A of the Reactor Vessel Level Monitoring System (RVLMS) of Unit 1 by installing jumpers from the reference thermocouple of sensor four to the inputs to the signal conditioning panel for sensor three.
The alumel wire common to both the reference thermocouple and the heated thermocouple of sensor three has an open circuit which cannot be repaired or replaced. This modification results in all eight sensors of channel A of RVLMS'being operable.
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10CFR50.59 Annupl Pecort Enclosure 2 _
Pace 2
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Temporary Lifted Lead 88-5S (Unit 2)
This electrical jumper and lifted lead modified channel A of the Reactor Vessel Leveling Monitoring System (RVLMS) of Unit 2 by installing jumpers from the thermocouple of sensor five to the j inputs to the signal conditioning panel for sensor six and !
installing a resistor in place of the open heater Clement of sensor six. The heater element for sensor six has an open circuit which cannot be repaired or replaced. This open circuit results in the loss of the four even numbers sensors since the heaters are wired in series. The installed resistor simulates j the heater element and eliminates the open circuit. The loss of !
a heater element results in the loss of that sensor. The inputs ;
to the signal conditioning panel for sensor six are supplied from the sencor five thermocouple. This modification results in six of the eight sensors of channel A of RVLMS being operable (sensor ,
four heater element previously had failed) while all eight l sensors of channel B are operable. ]
Temporary Lifted Lead 88-141 (Unit 11 The "Tcold" input from 1TT122CB was removed from the Reactor Protective System (RPS) by jumpering because this RTD was supply-ing an erratic signal. Because only one of two "Tcold" signals )
is selected by the RPS this jutapering only forced the RPS to select the signal from the other loop. Because one "Tcold" input was still supplied to the RPS for use in the calculations of Pvar and delta T power no loss of channel redun' lance resulted. This electrical jumper and lifted wire did not result in an unreviewed safety question. )
Igpoorary Lifted Leads88-157 and 1-88-169 (Unit 1)
The "Tcold" input from 1TE122CB was removed, by jumpering at the slide links, due to the failure of the RTD. No loss of system function was caused by the loss of this "Tcold" signal because the circuits using "Tcold" as an input either auctioneer the redundant inputs or display each separately. With the one "Tcold" input operational no commitment for required channels was violated because no equipment was inoperable. No unreviewed safety question resulted due to this electrical jumper and lifted wire request.
Iempordrv Lifted Lead 88 .170 (Unit 11 j
This electrical jumper and lifted lead modified channel B of the Reactor Vessel Level Monitoring System (RVLMS) of Unit 1 by installing jumpers from the reference thermocouple of sensor six to the inputs to the signal processing panel for sensor five, The reference thermocouple junction of sensor five has an open circuit which cannot be repaired or replaced. This modification results in all eight sensors of channel B of RVLMS being opora-ble.
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s' 10CFR50059 Annual ReDort 'Enclosurq % __
Pace 3 Temocrary Lifted Lead 88-190 (Unit 1) i "Ihis safety analysis approved the gagging open of MOV-6615 for all modes of operation until replacement parts for the valve were obtained. MOV-6615-is a normally opened 1" valve located down-stream of DRl8. DR18' allows the transfer of moisture, pipe scale, sediment, etc. .from the Main Steam' System. piping to the .
main condensers. Given a steam generator tube rupture event, !
MOV-6615 should close'but analysis showed that its being 1 aft opened does not prevent.the mitigation.of the accident nor does it significantly contribute to the 10CFR100 limits. Note: A significant contribution would require a change to those values l listed in the.FSAR.. This was not the case.
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's- d. : - i 4- i BALTIMORE 1 GAS AND- '
' . ELECTRIC
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CtlARLES CENTER . P.O. BOX 1475 '. BALTIMORE, MARYLAND 21203 March 14, 1989 U.S. Nuclear Regulatory'Comntission .s Washingt'on, D.C. 20555 j ATTENTION: Document Control'D6sk O
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SUBJECT:
Calvert Cliffs Nuclear Power Plant 1 Unit Nos. 1&27 Docket Nos. 50-3.17 &~50-318-
, fl L Renoit of CAg)1g % Tests and Experiments ;
1
REFERENCE:
(a) 10 CFR 50, Paragraph 50.59(b)+
Gentlemen: -
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As required by;the above reference,l attached is a report l containing discussions of th4 Changes, Tests;and Experiments i completed on Calvert Cliffs Unit.1 and/or 2.under the provisions of 10 CFR 50.59(a), _ including a.summaryLof the safety evaluation of each.
l j Items in this report are referred'to by " Facility Change Request" -(FCR) (Enclosure 1) and " Temporary Lifted Lead" number (Enclosure 2). The latter comprisu111ftedtl'eads,and-jumpers and temporary mechanical devices' deemed to be'-
changes within the scope of 10 CFR 50459.
This report covers the period ,fromiJanuary 1, 1988, through j December 31, 1988, except that some MTemporary Lifted Lead" !
items had been evaluated. pursuant to 10 CFR 50.59 prior to 1988, but wore not included in prior years' annual' surma- ]j ries. In the future, temporary modifications yhich require-safety evaluations will be summari.:ed on an annual basis.- l Very ru y s ,,
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/sf< k Geopge . ree 1 Vice Presiden Nuclear Energy GCC/ENM/lw
- h. Enclosure j cc: D. A. Brune, Esq. '
- l. J. E. Silberg, Esq. j R. A. Capra S. A. McNeil j W.' T. Russell V.; L. Pritchett T. -Matjet.te // .
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