ML20107F560
| ML20107F560 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 12/31/1984 |
| From: | Lundvall A BALTIMORE GAS & ELECTRIC CO. |
| To: | Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| NUDOCS 8502260169 | |
| Download: ML20107F560 (13) | |
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FCR 77-1033
.This FCR allowed the use of new model Sigma indications which (Unit 1, Unit 2) utilize a servomotor driven meter movement as a replacement for the original models which used a standard D'Arsonnal meter movement. The manufacturer, Sigma, had notified BG&E that the original models could not meet any specifications that required the indicator to function during a seismic event. Our review of this l.
replacement model showed that there would be no degradation of i
l inputs or interface with the Reactor Protective System (RPS) or the Engineered Safety Features Actuation System (ESFAS). The original desigii u L; ef tr.=: rf:tm.;
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FCR 79-1035 An auto-start of the Auxiliary Feedwater (AFW) pumps on loss of (Unit 1, Unit 2) -
main feedwater was installed by this FCR. The Feed Line Break j-and Loss of Main Feedwater Analyses were unaffected since the AFW regulating valve controllers were required to be set to zero flow when the system was in standby. A redundant three minute time delay was installed to prevent a return to power which could have been caused by feeding a ruptured steam generator with cold AFW at excessive rates.
AFW regulating valve opening was restricted mechanically to 75% to prevent cavitation of the AFW pumps on auto-start coupled with loss of control air. The system is capable of delivering flow in excess of that required even with the valve opening restricted to 75% and steam generator pressure at the relief valve setpoint.
l FCR 79-1060 This FCR provided for auto-start of the Auxiliary Feedwater (AFW)
(Unit 1, Unit 2) pumps on low steam generator level. The Feed Line Break and loss of Main Feedwater analysis were - unaffected since the AFW regulating valve controllers were required to be set to zero flow when the AFW system was in stand-by. A redundant 3-minute time delay prevents a return to power which might otherwise occur by feeding a ruptured steam generator with cold AFW at excessive rates. Control of flow for pump protection is ensured by modifying the air to the AFW regulating valves so that air is available from the instrument air header or the salt water system air compressors.
FCR 80-0074 This FCR modified the drawings showing the diesel generator (Unit 1, Unit 2) exciter. The control logic for the diesel generator start or trip circuit were unaffected by this drawing update. The change did not constitute an unreviewed safety question since the operation and availability of the diesel generators were unaffected.
FCR 80-0084 This FCR provided for drilling a series of test holes across one side The of the supply air ducts on the control room HVAC system.
purpose was to allow air flow quantity measurements needed to adjust air flow rates to the designed quantitles. The work and tests had no effect on operation of the system and there was no effect on the technical specification.
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y FCR 80-1065 Check valves were installed in the underground drain piping from each of the three diesel generator rooms to prevent flammable liquids from one space being carried to any other space. Generator operations were not affected and there was no impact on the technical specifications.
FCR 82-0005 This FCR allowed an air flow balance test to be performed on the (Unit 1, Unit 2)
Control Room HVAC system in order to restore air flow quantities Lin various spaces to the original system design values. No change to the Technical Specifications or the FSAR was required by this FCR.
FCP. ?24011 This FCR involved the jumpering of an individual cell in Battery (Common)
- 12 and allowed the continued operation of this battery with only fifty seven cells. No unreviewed safety question was created since i
Battery #12 is capable of powering the emergency DC loads even with one cell jumpered.
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FCR 82-0019 This modification installed jumpers on the test pressure and (Unit 1, Unit 2) temperature inputs to the Subcooled Margin Monitor. Combustion-Engineering had reviewed the operating history of the Subcooled Margin Monitors and determined that these pressure - and temperature test inputs allow a charge to buildup on the input -
l component of the test circuit. Eventually, the charge increases to a high enough value to override the actualinputs thereby creating an error. in the calculation. Installation of the jumpers will short the test input thus alleviating any charge buildup. No changes to j-the operation of the Engineered Safety Features Actuation System (ESFAS) or the Reactor Protective System (RPS) resulted from this modification thus no unreviewed safety question was created.
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FCR 82-0054 This FCR relocated the Main Steam Pressure transmitters 1 & 2-l (Unit 1, Unit 2)
PT-3991 and I & 2-PT-4008 from the MSIV room to the Electrical l
Penetration room to minimize temperature drift effects.
No unreviewed safety questions exist because all new instrument tubing and electrical conduits were installed in accordance with safety related seismic criteria.
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FCR 82-0061 This FCR replaced the slide link terminal blocks to the Metrascope 1
(Unit 1, Unit 2) position cables. The modification additionally improved access and l
ease of troubleshooting for system maintenance.. The safety analysis concluded that no unreviewed safety question since the original design critieria was maintained.
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, FCR 82-0067 This FCR added scribed reference lines to the Unit I and Unit 2 l
(Unit 1, Unit 2) refueling machines. ' The purpose was to improve the operability of the refueling machines. The lines were scribed on small stainless steel plates which were then welded to the refueling bridge and trolley. There is no effect on these structures. The effect of a plate falling on the reactor core during refueling was evaluated and I
found to be insignificant. There was no impact on the technical l
specifications.
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FCR 82-0070 This FCR provided for engineering analysis and repairs to return (Unit 1)
Unit I to service by restoration of reactor vessel closure head stud hote Nos. 4 and 12 which were damaged during disassembly for the Ur.it 1 - 1985 Refueling Outage in the following manners o Stud bolt and hole No. 4 sustained approximately 36% damage to the threads.
o Stud bolt No.12 was cut off.
- The repair work restored full thread engagement to No. 4 stud and all but approximately 1-1/2 threads for full thread engagement to No. 12 stud.
The restored condition exceeds the minimum acceptable condition for return to service.
This FCR also provided for shielding as worker protection to reduce '
exposure during repairs to acceptable limits.
FCR 82-0078
- This FCR relocated power cables for Pressurizer Back-up Heater (Unit -1)
Bank #2 icom Penetration IZEB2 to Penetration IZEB3. Inspection and testing of Penetration IZEB2 indicated the feed thru for the -
Back-up Heater to be overheated and grounded. A spare feed thru in Penetration IZEB3 was then utilized for the Back-up Heaters.
This modification did not affect functionally how the back-up heaters operate, therefore no technical specifications are affected.
FCR 82-0079 This FCR relocated power cables for #11 Containment Cooler from
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Penetration 12EB1 to Penetration IZEA4. Inspection and testing of Penetration :1ZEB1 Indicated _ the feed thru for the Back-up Heater to be overheated. A spare feed thru in Penetration IZEA4 -
was then utilized for the Containment Cooler. This modification did not affect functionally how the back-up heaters operate, therefore no technical specifications are affected.
FCR 82-0080 The FCR covered temporary mounting of an ultrasonic flow (Unit 1) transducer on salt water piping for one of the service water heat exchangers. The transducer only weighed nine pounds, which is insignificant when compared to the mass of the cement-lined steel salt water piping to which it was naunted. Thus the mounting of p
the transducer did not affect the seismic design of the piping. In addition, the' mass of the transducer was small enough so as not to L
cause a hazard to adjacent safety-related piping were it to separate from the pipe during a seismic event.
h FCR 82-0102 This FCR provides for securing one damaged tube in steam (Unit 1) generator No.11 at position L23, R136. The damage occurred as a result of work accomplished to complete FCR 81-76 (perform rim 4
cut on steam generators' partial tube support plates per CE proposal).
The damage was determined by eddy current examination and j
visual examination using fiber optics.
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p The damage tube was secured and plugged in accordance with NCE Engineering Procedure EP-72267_-103, Revision 0 and FCR 82-51 (Steem Generator Tube Plugs), respectively.
This procedure provided guidance for removing a portion of the breached tube within the tube bundled and stiffening the breached section to avoid fatigue failure (tube whip) during future operation. A tube plug was installed in each end of the cut tube.
FCR 82-0107 This FCR added capacitors across the hot and cold leg temperature
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transmitters. The modification reduced the noise generated by the transmitter power supplies to the Reactor Protection System. The addition of the capacitors increased the transmitter response time but Jthe total channel response remained within the original criteria.
FCR 82-0114 This FCR modified the Main Steam Isolation Valve Vapor Barrier (Unit 1)_
-section facilitates easier rigging of the MSIV's for maintenance on for Unit I to provide a bolted, removable section. This removable the valves and their operators.
Bolted construction is used throughout the vapor barrier and is an accepted practice. The vapor barrier retains all its design functions. There is no impact on any technical specifications.
FCR 82-0115 This FCR modified the Main Steam Isolation Valve Vapor Barrier (Unit 2) for Unit 2 to provide a bolted, removable section. This removable section facilitates easier rigging of the MSIV's for maintenance on the valves and their operators.
Bolted construction is used throughout the vapor barrier and is an accepted practice. The vapor barrier retains all its design functions. There is no impact on any technical specifications.
FCR 82-0142 The Main Steam Isolation Valve actuator cylinder drain tubing (Unit 1,-Unit 2) material and wall thickness were changed. The tube was changed from two-inch OD x.065 wall thickness, ASTM-A269 seamless type 304 stainless steel to two-inch OD x.134 wall thickness, ASTM-A192 seamless carbon steel.
The change was initiated after
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numerous instances of the tubing blowing out of the compression fittings and was based on the fitting manufacturer's recommendations. Thicker wailed tubing of a softer metal provides a more reliable connection at t_he fitting. This change does not adversely affect the original design and operation of the actuator system.
4 FCR 82-0145 This FCR added a Safety Injection Actuation Signal (SIAS) to the (Unit 1, Unit 2)
Hydrogen Purge Exhaust MOV's (1 & 2 MOV-6900 and I & 2-MOV-6901)'to allow the valves to be opened at power in order to vent containment atmosphere and maintain containment pressure below the technical specification limit nf 1.8 psig. All design criteria for this modification were consistent with original design criteria for containment isolation valves and a technical specification change was requested to allow opening the valves at power.
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7 FCR 82-0184 This FCR provided for the purchase and installation of an under-(Unit 2) water inspection periscope in the Unit 2 Spent Fuel Pool. The periscope when used in conjunction with an existing fuel bundle elevator allowed underwater inspection of spent fuel bundles before insertion in the fuel racks. All penetrations and attachments to the spent fuel pool liner were seal welded in order to maintain the
-watertight integrity of the pool liner.
Additionally, the supports for the: periscope were. designed seismically.
The FSAR was changed to include a description of the instrument and its function.
FCR 82-0185 -
This FCR removed the internals of 1/2-SW-196 and 1/2-SW-197.
(Unit 1, Unit 2)
The modification did not adversely. affect the seismic design of the valves or the piping in which they were installed since the removal of the valve internals decreasef the valve mass. An arrangement was designed and installed to plug the shaft hole for each valve in order to maintain the pressure boundary integrity of the salt water system.
There was no impact on any technical specifications.
These valves were not required for plant operation; in fact their existance in the common salt water discharge lines for the component cooling and service water heat exchangers was actually detrimental to reliability of the salt water system.
FCR 82-0186 This FCR relocated the Emergency Control Center (ECC) from the 4
(Common)
South Service building to the Technical Support Center Annex.
Dedicated telephone instruments and the control unit for the emergency radio were moved.
A new telephone cabinet was installed in the Cable Spreading Room.
This cabinet and the associated conduit were mounted seismically in accordance with approved BG&E standards.
FCR 82-1001 This FCR provides that cables for 2-CV-505,2-CV-516 and 2-CV-(Unit 2) _
(ASP) located in the Unit 2 - 45 ft. Switchgear Room.
Key 5464 be rerouted through the Unit 2 Auxiliary Shutdown Panel operated switches on the ASP will allow power to be removed to these valves' solenoids to cause the valves to shut to maintain RCS-2 inventory. Work performed under this FCR is as recommended in i
the interactive ' cable analysis '(ICA) for Calvert Cliffs Nuclear n
Power Plant Unit 2, dated February 1,1982. The ICA resulted l
from new regulations (10 CFR 50 Appendix R) which imposed new l
design criteria on independence of equipment.
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' FCR 82-1002 This FCR reroutes the power cables to LPSI pump #21, #22 and (Unit 2) adds a control by-pass switch at the switchgear. Work performed under this FCR is as recommended in the interactive cable analysis L
(ICA) for Calvert Cliffs Nuclear - Power Plant Unit 2, dated February 1,1982. The ICA resulted from new regulations (10 CFR 50 Appendix R) which imposed new design criteria on independence of equipment.
FCR 82-1003 This FCR adds control by-pass switches at the switchgear for (Unit 2)
Saltwater pumps #21, #22, and #23. Work performed under this j'
i FCR is as recommended in the interactive cable analysis (ICA) for Calvert Cliffs Nuclear Power Plant Unit 2, dated February 1, l
1982. The ICA resulted from new regulations (10 CFR 50 Appendix 4
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FCR 82-1004 This FCR reroutes the power feeds to charging pumps #22 and #23, (Unit 2) add a second power cable to charging pump #23 and two manuai 480 volt disconnects to charging pump #23. - Work performed under this FCR is as recommended in the interactive cable analysis (ICA) for Calvert Cliffs Nuclear Power Plant Unit 2, dated February 1, 1982. The ICA resulted from new regulations (10 CFR 50 Appendix R) which imposed new design criteria on independence of equipment.
FCR 82-1006 This FCR adds control by-pass switches at the switchgear for (Unit 1)
Service Water Pumps #11, #12, and #13. Work performed under this FCR is as recommended in the interactive cable analysis (ICA) for Calvert Cliffs Nuclear Power Plant Unit 1, dated October 1, 1981. The ICA resulted from new regulations (10 CFR 50 Appendix
.R) which imposed new design criteria on independence of equipment.
FCR 82-1007 This FCR adds control by-pass switches at the switchgear for (Unit 2)
Service Water Pumps #21, #22 and #23. Work performed under this FCR is as recommended in the interactive cable analysis (ICA) for Calvert Cliffs Nuclear Power Plant Unit 2, dated February 1, 1982.' The ICA resulted from new regulations (10 CFR 50 Appendix R) which imposed new design criteria on independence of equipment.
FCR 82-1008 This FCR reroutes power cables for 2-ERV-402 and 2-ERV-404 in (Unit 2) separate conduits with no other cables. Work performed under this FCR is as recommended in the interactive cable analysis (ICA) for Calvert Cliffs Nuclear Power Plant Unit 2, dated February 1, 1932. The ICA resulted from new regulations (10 CFR 50 Appendix R) which imposed new design criteria on independence of equipment.
FCR 82-1011 This FCR provides ZA facility 125 VDC power from Unit 1 Cable (Unit 2)
Spreading Room to Unit 2 Switchgear Room without going thru the Unit 2 Cable Spreading Room. Work performed under this FCR is as recommended in the interactive cable analysis (ICA) for Calvert Cliffs Nuclear Power Plant Unit 2, dated February 1,1982. The ICA resulted from new regulations (10 CFR 50 Appendix R) which
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imposed new design criteria on independence of equipment.
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FCR 82-1031 This FCR performed in-situ low level vibration testing of various (Unit 1) electrical and mechanical equipment while the plant was in MODE 6.
Included was investigative data collection as part of the work conducted by the utility group to determine NRC action on USI-A46. The vibration testing was conducted on those components not required for operation during MODE 6, thus no unreviewed safety question existed.
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FCR 82-1033 This FCR involved the replacement of current sensors on a number (Unit 1, Unit 2) of 480 volt load center breakers. This FCR did not constitute an unreviewed safety question because the new current sensors allowed an improvement in the sensitivity of protection for 480 volt motors. The new level of protection was consistent with the original design philosophy.
FCR 82-1037 NCRs 2198 and 2690 identified the presence of caustic liquid (Unit 1, Unit 2) solution in the concrete embedded plastic conduit duct runs connecting the Auxiliary Feedwater Pump Rooms and the Intake Structure. This FCR replaced the existing silicone rubber / asbestos braid cables routed through these ducts with cables Jacketed with thermosetting cross-linked polyethylene. The new cables are more suitable for use in wet environments. As the new cables enhanced the reliability / availability of the components and systems they serve, no unreviewed safety question was created.
FCR 82-1039 This FCR allowed the replacement of the existing ten cubic inch (Unit 1, Unit 2) nitrogen charged bladder accumulators found on the Main Steam Isolation Valve actuator system pumping skid with a new one quart model. The existing accumulator is no longer manufactured and the one quart model was found to be a suitable replacement. The substitution does not alter the design or operation of the actuator system. Supports have been modified to provide secure mounting of the larger size accumulator.
FCR 82-1057 This FCR provided for modifications in the Letdown Heat (Unit 1, Unit 2)
Exchanger Room so that the plant configuration would comply with the description in FSAR 10A.7.
The modifications include removing plates installed in piping blockouts between the West Piping Penetration Room and the Letdown Heat Exchanger Room; and providing a solid door from the Letdown Hec.t Exchanger Room to the Hallway El. 27'-0".
The purpose of the change is to allow pressure resulting from a break in the letdown line occurring in either the West Penetration Room or the Letdown Heat Exchanger Room to be sensed in both rooms at once. This will allow the CVCS isolation system to isolate a postulated break as described in FSAR 10A.7.17 and 10A.7.20.
FCR 83-0043 Valves 2-MOV-5250 and -5251 had suffered from seat deterioration (Unit 2) to the extent that they were no longer suitable for their intended function of isolating the Circulating Water Pump Room Coolers on SIAS. Valve delivery considerations prompted a request for use of an alternate shaft material and a valve without a spacer piece which allowed packing adjustment without removing the operator.
It was determined that the alternate shaft material was suitable for use at least until the next refueling outage. The lack of the spacer actually improved the seismic design of the piping in which the valves were installed.
No technical specifications were affected. The valves were replaced with valves conforming to the original design as to shaft material and operator spacer piece, during the Spring 1984 outage.
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' FCR 83-0045 Allowed the temporary repair of a leaking two-inch manual valve in (Unit 2) the steam generator chemical addition lay-up ~ system. The repair involved the injection of a sealing compound into the valve internals via a passage drilled into the bonnet. The system, and
. therefore, the valve were not required to operate during the life of the temporary repair. The valve was replaced with an identical model during the following refueling outage.
FCR 83-0067 One of the damper operating arms on the ECCS pump room (Unit --1,, Unit 2) ventilation system loosened where the shaft bushing was press fitted into the damper arm. This FCR provided for repair by tack welding the bushing to the arm.
No change to the FSAR or technical specifications was required.
' FCR 83-0087 The FCR was for replacement of the class A fire door between No.
(Common)
- 11 and No.12 diesel generator rooms. Deterioration of the mortar joint between the door frame and reinforced concrete wall reduced the fire rated capability of the room separation. The door was replaced with a slightly wider class A, 3-hour fire rated door in order to permit removal of large equipment. The safety analysis addressed the preservation of the fire rating and that the door was Installed seismically as originally designed.
Therefore, the technical specifications are unaffected.
FCR 83-0089 The waste handling wash-down sump was connected to the yard (Common) drainage systems through a valve. The FCR allowed removal of the valve and a blank flange to be installed to isolate the sump drain from the yard drains. The purpose was to remove any chance of contaminated water from being released to the environment through the yard drains.
FCR 83-0098 This FCR eliminates the aluminum multi-purpose terminals on (Unit 1, Unit 2) molded case breakers (Gould),480V Motor Control Centers. This modification is due to the low torque values of aluminum connectors and screws, creating a high resistance condition. Some cable terminations #4 to 4/0 had heating and some failed on power feeders to motors 20 HP and above. The use of the ilnned-copper certified connector compressed on a copper wire, bolted to the molded case breaker copper post with a steel screw, reduces the interface resistance and heating. This all copper termination has higher torque values and is mechanically stronger therefore increases reliability.
This modification did not affect the functional operation of safety related equipment. There was no impact on the technical specifications.
FCR 83-0109 This FCR was for the structural review of certain existing pipe (Common) hangers and several pieces of equipment to withstand the loading created by the placement of temporary shielding. ~ Calculations were performed for each hanger affected. An analysis was made to verify the capability of the ICI Removal Bridge to sustain the added weight of lead sheets. Combustion Engineering reviewed the effects of adding shielding over the No.11 steam generator tube bundle, and Bechtel Power Corp. utilized piping stress analysis 8
r techniques to review shielding of the pressurizer spray lines from reactor coolant pumps Nos.11 A and 11B cold legs. Since the reactor was in mode 6 and some of the equipment disassembled, there was no accident possible which was not previously addressed.
All supports were structurally adequate ano no technical specifications were affected.
FCR 83-0117 This FCR allowed the use of a through-bolt and nut on the #11 (Unit 1)
Emergency Diesel Generator exhaust pipe. The original bolt whicn was tapped into the flange had broken and the threaded hole was damaged. The through-bolt was substituted instead of having to redrill and re-tap existing holes. The evaluation concluded that the proposed design was equivalent to the original design.
FCR 83-0120 This FCR provided an alternate means of terminating Containment (Unit 1)
Air Cooler Motor leads. Motor removal for maintenance required compression lugs (connectors) to be cut off the ends of the leads.
Each removal decreased the leads length. The Burndy Scruing type KPA can be removed without cutting motor leads. The substitution was suggested by Joy Manufacturing Co. These lugs had been qualified for containment use. These lugs were installed according to specifications and are equal to or better than those supplied with original equipment.
There is no impact on t!ie technical specifications and this modification did not affect the functional operation of the equipment.
FCR 83-0137 This FCR increased the original shaft diameter tolerances at the (Unit 1) thrust and radial bearings of the containment spray pump shafts.
The original tolerances were too low resulting in loose bearing fits and wear on the pump shaft. The new diameter tolerances will result in a slight interference to press fit the bearing to the shaf t.
The new tolerances are in accordance with those recommended by the Anti-Friction Bearing Manufacturer's Association. The change does not alter the design basis of the pump and improves the pump's reliability.
FCR 83-1001 This FCR replaced the fuses in the vital 120 volt AC panels. This (Unit 1, Unit 2) work did not constitute an unreviewed safety question because the fuses were selected to properly protect each load while improving the electrical coordination in the 120 volt vital AC system.
l FCR 83-1019 This FCR provides all necessary engineering for plant modifications (Unit 2) to acccmodate construction of the new Access Control Area, which consists of a new one story structure totalling approximately 10,300 sq. ft. constructed above the Unit 2 Heater Bay Area.
The new structure consists of a steel framework, reinforced concrete f!oor slab, insulated metal siding for the enclosure, and four-ply asphalt built-up roofing.
The structural steel is an extension of the existing steel column framework in the Heater Bay Area which is supported primarily from the Turbine Building base slab at El.12'-0".
Floor steel for the new structure is crected above the existing Heater Bay Roof El. 69'-0" making the new floor elevation 72'-0".
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The Plant modifications required to support this construction consisted of:
o Extending 14 vent pipes passing through Unit 2 Heater Bay Roof to a point above the roof the new Access Control Building.
o Providing 14 new support designs for the new pipe and support modifications to existing pipes; modifying building siding where the supports tie in.
o Providing reinforcement (cover plating) of 8 columns along column row 3 below evaluation 27'-0".
o Providing 12 additional steel beams along column line H for tie in from the new floor at Elevation 72'-0" and roof at Elevation 8 6'-6".
All modifications to existing structures and the addition of new structures were designed to be consistent with original design standards.
FCR 83-1010 Added a pipe support on the auxiliary high pressure safety injection (Unit 2) header relief valve to ensure structural integrity during a seismic event. Based on a field inspection and engineering evaluation, it was determined that a support was required on the one-and-one-half inch relief piping. The additional support ensures that the header will not exceed its design stress limits due to all postulated loading conditions.
FCR 84-0070 This FCR relocated power cables for Pressurizer Back-up Heater Bank #4 from Penetration 2ZEB4 to Penetration 2ZEB3.
Penetration 2ZEB4 terminal #4 was broken off (termination stud) necessitating the switch to 2ZEB3.
This modification did not affect functionally how the back-up heaters operate, therefore no technical specifications are af fected.
FCR 84-1015 This FCR was written to provide for new civil standard (s) to incorporate standard designs for baseplates, corebores and cut-outs.
This would allow for more efficient installation while avoiding the cutting of reinforcing steel.
By avoiding and minimizing the cutting of rebar, the structural integrity of the various plant structures can be maintained. There was no impact on the technical specifications.
FCR 84-1025 The purpose of this FCR was to provide a Technical Requirement Document (TRD) to supersede the existing specification A-24
" Painting and Special Coatings" which was designed for a plant under construction and was dif ficult to work with. All coatings were evaluated for their suitability for the design bases accident conditions.
Thus, the coating systems will not degrade the Engineering Safety Features system by delaminating, flaking or peeling. There was no impact on the Technical Specifications.
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O FCR 84-1029 The manufacturer of Mk 7M3 and 283M3 valves changed
.e type of material used for certain internal valve parts.
The material substitutions were evaluated for their suitability in their respective operating environments. The safety analysis concluded that the materials substitutions were adequate and had no effect on the operation of the valves.
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bec: Messrs. A. E. Lundvall, Jr.
C. H. Poindexter C. H. Cruse -
- 3. A. Tiernan R. F. Ash / File L. B. Russell
. R. M. Douglass R. C. L. Olson M. 3. Miernicki 1
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J BALTIM ORE GAS AND ELECTRIC CHARLES CENTER. P. O. BOX 1475. BALTIMORE, MARYLAND 21203 AnTHun E. LUNDVALL. JR.
V4CC PRESsDENT SUPPLY Mr. Thomas E. Murley Regional Administrator, Region I Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406
Dear Mr. Murley:
Subject:
Calvert Cliffs Nuclear Power Plant Units Nos.1 Ac 2, Docket Nos. 50-317 & 50-318 Report of Changes, Tests, and Experiments
References:
10 CFR Part 50, Paragraph 50.59(b)
As required by the above reference, attached is a report containing a brief description of the changes, tests, and experiments completed on Calvert Cliffs Units I and/or 2 under the provisions of 10 CFR 50.59(a), including a summary of the safety evaluation of each. This report covers the period from January 1,1984 through December 31,1984.
Items in the attached report as referred to by " Facility Change Request (FCR)" number.
Very truly yo xdwdQ AEL/ERZ/vf Attachment cc: Mr. R. C. DeYoung - NRC D. A. Brune, Esq.
G. F. Trowbridge, Esq.
Mr. D. H. Jaffe - NRC
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