ML19310A310
| ML19310A310 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 06/06/1980 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML19310A307 | List: |
| References | |
| NUDOCS 8006110072 | |
| Download: ML19310A310 (19) | |
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FACILITY CHANGES, TESTS AND EXPERIMENTS Facility Change No. 265 This change consisted of the installation of an eight (8) point temperature recorder to monitor and alarm the temperatures of the Liquid Poison System's seven (7) explosive Squib valves and monitor the temperature of the liquid poison tank. An alarm, both audible and visual, will alert Operations when the temperature of a Squib valve reaches or exceeds 120*F. This is to pro-
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tect the explosive Squib valve's primer from extended exposure to high temper-atures which decrease the primer's effective life expectancy. This Facility Change was performed in conjunction with Specification / Field Change SFC 79-018.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
Facility Change No. 382 This change relocated the decontamination laundry from 'the access control area to an area near the turbine. This provided greater radiation protection / access control area and improved che washing facilities.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
Facility Change No. 392 i
This change consisted of the addition of an indicating lamp, control relay and the utilizing of existing low voltage relays to provide Operations with a positive indication of 125 Volt D-C System avai' ability.
l The Safety Evaluation concluded that t.nis change did not constitute an unreviewed i
i safety question.
Facility Change No. 409 l
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This change placed the discharge canal liquid process monitor cable in shielding conduit 'to eliminate induced voltages and spurious alarms.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
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2 Facility Changes. Tests and Experiaments (contd)
Facility Change No. 432 This change updated the Plant's sewage system to meet present day environmental standards. The old chlorination system was removed and a new drainfield was installed.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
Facility Change No. 434 This change consisted of the addition of a monitor relay to the emergency diesel generator auto start circuitry to alarm, via the existing annunciator window, the loss of the emergency power system control voltage or the, lacing of the emergency power auto selector switch in the "off" position. Along with the addition of the alarm relay contacts to the above annunciator window, the contacts of the local emergency diesel generator control switch utilized for alerting the control room operator to non-automatic local control, were relocated from annunciator window labeled " Emergency Generator Engine Trouble" to the annunciator window labeled " Emergency Generator Start / Control Failure".
Specification Field Change 79-007 was performed in conjunction with this Facility Change.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
Facility Change No. 441 This change consisted of adding three new non-Q electrical power distribution panels (lF, 2F, & 2E) to the plant station power 480V AC system as well as re-powering frem another source busses (MCC's) MCC-1D, MCC-2D, MCC-1E, MCC-2C and distribution panel 1P.
Several non-Q components were then moved from Q-listed busses over'to these newly installed busses.
The purpose of the change was to reduce the current being drawn through 480V AC circuit breakers ACB-52-1A and ACB-52-2A and allow them to operate further from their trip setpoints.
l The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
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Facility Changes, Tests and Experiments (contd)
Facility Change No. 457 (A-H)
This change consisted of performing the NRC required modifications to update the plant security system to comply with 10CFR73.55 requirements.
The Safety Evaluation concluded that these changes did not constitute an un-reviewed safety question.
Facility Change No. FC-462-A This change involved replacing some existing doors with approved three-hour rated fire doors; the doors replaced are in walls that serve as fire barriers, The doors are equipped with fusible links to assure automatic closure in the event of a fire. The change was done in accordance with Amendment 25 to the Technical Specificaticus.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
Facility Change No. 462-F This change installed splash panels over the motor control centers, 2400 volt switchgear, and the transformers in the electrical equipment room. The splash panels are to prevent water damage from operation or malfunction of the overhead sprinkler system. The panels are constructed of sheet metal and sealed with silicone foam. The change was done in accordance with Amendment 25 to the Tech-nical Specifications.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
Facility Change No. 464 i
This change replaced the Emergency Core Cooling System ring sparger within the reactor vessel with a sparger having a spray distribution demonstrated to.
meet the requirements of 10CFR50 Appendix K.
The replacement was performed as required by Amendment 15 to the Facility Operating License. Qualification testing of the spray distribution was performed as described in CPCo letters to the NRC dated November 21, 1978 and March 28, 1978.
Design and installation of -the sparger were completed in accordence with design docume. ts presented in November 21, 1978, January 16, 1979 and April 6, 1979 letters to the NRC.. To minimize radiation exposure during the replacement, the steam baffle to which the sparger is mounted was replaced as well.
4 Eacility Changas. Tesen and evner4, anes (contd)
Facility Change No. 464 (contd)
Installation of the new sparger/ baffle assembly provided the Big Rock Point reactor with two qualified, redundant sources of core spray for mitigating the consequences of a LOCA.
The Safety Evaluation concluded that installation of the new sparger insured the accuracy of loss of coolant analysis and did not constitute an unreviewed safety question.
Related plant modifications include SFC-79-008 - Sparger Ball Joint Modification and SFC 79-013 Sparger Ball Joint Support Tackwelds.
Facility Change No. 470 This change consisted 6f installing an air tap in the service air line within containment to provide an air source to assist with the annual air tests on penetrations and pipes.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
Facility Change No. 474 This change installed emergency lighting in each of four Reactor Depressurization System uninterruptable power supply rooms.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
Facility Change No. 475 This change installed an undervoltage relay on the turbine generator AC emergency bearing and seal oil pump to sense loss of AC power and automatically start the DC emergency turbine generator bearing and seal oil pump.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
Facility Change No. 476-This change relocated the radiation portal monitor ani light gate installed in the front lobby of the service building to access control for use in con--
trolling ingress and egress into contamination zone #1.
This portal monitor and light gate have been replaced by new units in the security building for monitoring those individuals entering and leaving the restricted area.
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5 Facility Changes.' Tests and Exneriments (contd) l Facility Change No. 476 (contd)
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
Facility Change No. 479 4
This change provided a means of measuring flow rate through the shell-side of the core spray heat exchanger. A differential pressure gauge was installed on a section of the shell-side discharge pipe. Numerous tests were run to determine the flow rate, and a restricting orifice in the line was increased in size to allow the proper flow rates through the heat exchanger..
The Safety Evaluation concluded that this change did not constitute and unreviewed safety question.
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6 specification / Field Changes SFC 78-026 This change involved a review of the Cycle 16 core loading and compared its nuclear and thermal hydraulic characteristics with reference analysis associated with the reference Cycle 15.
Included in the core loading were the following Big Rock Poinc fuel designs:
General Electric Reload F - 4 assemblies Exxon Nucicar Reload G - 20 assemblies
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Exxon Nuclear Reload G/U - 20 assemblies Exxon Nuclear Reload G3 - 40 assemblies Each of the reload G are mixed oxide assemblies. All twenty-six fresh assemblies inserted in the core were of the reload G-3 design. Assemblics from four R&D pro-grams were inserted as a part of the core loading: Exxon /EPRI Cladding Ductility Program, 2 reload G assemblies (originally described in CPCo to NRC 7/9/74); GE Corner Red Program, 4 reload F assemblies (a continuation of the program described in CPCo to NRC 3/26/74 and 1/16/78); Exxon /DCE Fuel Performance Improvement Pro-gram; 4 assemblies (See SFC-78-027); and Exxon Extended Burnup Program (See SFC-79-012).
The Safety Evaluation concluded that nuclear characteristics for Cycle 16 were similar to reference Cycle 15, previous transient and accident analysis conservatively bounded core conditions for Cycle 16, existing Technical Specifications would be met during Cycle 16, and the new core loading did not constitute a potentially un-reviewed safety question.
SFC 78-027 This change is a DOE sponsored.R&D effort to study and reduce stresses associated with pellet cladding interaction. The assemblies associated with this program have the following fuel rod types, some of which differ from standard Reload G-3 fuel rods:
1.
Solid pellets 2.
Pellets with chamfered edges 3.
Annular pellets 4.
Packed power fuel 5.
Sphere pac fuel 6.
Pressurized rods 7.
Graphite coated rods, and 8.
combinations of the above.
7 Specification Field, Changes (contd)
SFC 78-027 (contd)
All rod types are designed to reduce cladding strain and have had previous irradiation experience within the industry. Some rods are segmented to accomodate future irradia-tion and testing in a test reactor. From a mechanical, thermal hydraulic and nuclear standpoint, the FPIP assemblies are nearly identical to a standard Reload G-3 assembly.
Minor local peaking increases (on the order of several percent) in the area of the segmented rod connectors are accounted for in thermal hydraulic calculations. All G-3 Technical Specification thermal hydraulic limits apply to the FPIP assemblies.
The Safety Evaluation concluded that as the mechanical, nuclear and thermal-hydraulic characteristics of FPIP assemblies were very similar to reload'G3, and as existing G-3 Technical Specification limits applied to FPIP assemblies, the Fuel Performance Improvement Program did not constitute an unreviewed safety question.
Related Modifications - See SFC 79-026 (C16 Core Loading)
SFC 78-028 This change consisted of replacing an ordinary 1/2" pipe plug on the emergency diesel generator motor cooling water drain with a gate valve and a plug.
The reason for this change was to permit ease of drainage and to provide a means of controlling the drainage of cooling water, when necessary.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
SFC 78-029 This change was to replace Mark #110 1500 globe valve with a 1500# Herma-valve from Rockwell Company on the steam drum level sensing root valves.
The reason for this change was to decrease packing failure since the replacement valve has no packing.
The Safety Evaluation concluded that there was no unreviewed safety question be-cause no piping or valve functions will change. The new valve will not affect the instrumentation capabilities of the line. Nothing is changed or altered that will affect the safety analysis - material, pressure rating, and valve size all remain the same, 1
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8 Specification / Field Changes (contd)
SFC 78-030 i
This change involved replacing pressure controller NC-18 on the control rod drive accumulator charging header with an equivalent never model. BS&B type 72-24-2 pressure controller was replaced with a WKH type 72-14-1 because of the unavailability of repair parts for the BS&B model. The new pressure controller was installed at the same location and serves the same function as the old pressure controller.
The Safety Evaluation concluded that this change did not constitute an unreviewed
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SFC 78-031 This change modified the piping on the turbine stop valve before and af ter seat drain lines to facilitate installation of replacement valves. The original valves leaked through and could not be repaired.
The Safety Evaluation concluded that nothing was changed to alter the original design and that the modification did not constitute an unreviewed safety question.
SFC 79-001 This change was performed to allow replacement of ASCO model 830060R solenoid valves with ASCO model HTX-8300C61RF solenoid valves.
The Safety Evaluation concluded that no unreviewed safety question exists be-cause the replacement solenoid meets or exceeds the requirements of the original valve. The replacement valve is an improved model and will not affect original equipment design functions.
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SFC 79-002 This change replaces Honeywell microswitch B2E6-2RN with BZE6-2RN or BZE-2RN.
The reason for this change is for convenience and either switch may be used.
The Safety Evaluation concluded that there was no unreviewed safety question because the replacement microswitch has the same electrical and operational N thing is changed specifications and physical dimensions as the original.
o that will affect the safety analysis report.
SFC 79-003 This change modified the containment supply ventilation butterfly valve disc.
- The rubber. seat' retaining configuration was changed. to allow the use of the Allis Chalmers "Hemi" rubber seat. This.was performed to improve the sealing reliability of the' butterfly valve.
Specification / Field Changes (contd)
SFC 79-003 (contd)
The Safety Evaluation concluded the modification will improve the sealing reliability of the butterfly. The modification does not pose an unreviewed safety question.
SFC 79-004 This change was to replace a 3/4" drain on the 3" cleanup drain / recycle line to the #1 recirculation pump with a section of new pipe. The reason for this change was to eliminate a bad weld on the dead end drain line.
The Safety Evaluation concluded that there was no unreviewed safety question because the drain is not required for plant operation and the system can be drained, if required, through the reactor recirculation pumps. Nothing will be changed or altered that affects the safety analysis or plant operation.
SFC 79-005 This change involved a vendor design change of the scram valve internals. The mating surfaces within the valve remained unchanged. The mod'ification was to hold parts in better alignment and prevent the cold flow of the Teflon seat.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
SFC 79-006 This change document.ed a di=ension change by the vendor for the pilot disc of the Reactor Depressurization System depressurization valves. The disc was machined to a slightly larger diameter (for tighter clearance) to prevent tilting of the dise when being seated. Also, an additional washer was in-serted above the pilot disc spring to increase the preload on the disc.
This change was approved by the vendor in order to prevent foreign material from contaminating the mating surface.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
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i Specification / Field Changes (contd) 10 SFC 007 This change involved the substitution of a number twelve (12) AUG, Type SISS, 600 volt, switchboard wire for a number fourteen (14) AWG, Type SISS, 600 volt switchboard wire due to the lack of the number fourteen (14) wire. The number fourteen (14) AWC wire was of adequate size for the installation, therefore, number twelve (12) AWG wire is more than adequate. The insulation character-istics are identical for both wires. This Specification Field Change was-
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performed in conjunction with Facility Change FC 434.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
SFC 79-008 This change replaced tackwelds designed to capture the nut attaching the inlet pipe to the sparger with a 1/4" anti-rotation pin. The new core cooling sparger was installed as described under FC-464. The inlet pipe from the fire system is attached to the sparger by a ball joint, free to rotate in order to ease installation. Tackweld shrinkage froze the balljoint in plac.e, hampering installation. The tackwelds were removed and replaced by a 1/4" pin capturing the ba11 joint and inlet pipe during installation.
The Safety Evaluation concluded that replacement of the tackwelds with a pin designed to capture the ba11 joint nut fulfilled the original intent of the tackwelds. The pin was subjected to the same proof load required for' tha tackwelds. No potentially unreviewed safety question was created.
Related documents - FC-464 Sparger Replacement SFC 79-009 This change consisted of replacement of control circuit wiring for the Emergency Condenser System motor operated valves located in the steam drum enclosure. The original wiring, examined during routine maintenance and inspection of the _ valve operators, had begun to show' signs of' deterioration.
The wire size was not changed, only the type of insulation. The replacement wiring exceeded the original requirements.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
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Specification / Field Changes (contd) 11 SFC 79-010 This change involved the replacement of the 480V AC, MCC 72-33 and MCC 72-34 breakers utilized for supplying power to MO-7053 emergency condenser outlet valve (72-33) and MO-7063 emergency condenser outlet valve (72-34). The ori-ginal breakers, CE type TEF 124C5020 were found to be defective during routine breaker maintenance. Breaker type TED 126020 was used as a direct replacement following technical evaluation by both Consumers Power Company Systen Protection and Laboratory Services Department and the v'endors; the original type breaker is obsolete and can no longer be purchased.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
1 SFC 79-011 This change involved the replacement of the 480V AC Mcc 52-2B26 breaker utilized for supplying power to electric fire pump. The original breaker, GE type TJ236C6225 was found to be defective during routine breaker maintenance.
Breaker type TJ236C5225 was used as a direct replacement following technical evaluation by both Consumers Power Company System Protection and Laboratory Services Department and the vendor; the original type breaker is obsolete and can no longer be purchased.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
1 SFC 79-012 This change involved the insertion of high exposure Reload G fuel rods in newer host assemblies to extend the in-core life of 64 fuel rods. This activity is in association with the Big Rock Point Extended Burnup Demonstration Program -
Phase I.
The program is a DOE sponsored project to demonstrate the feasibility of extending commercial boiling water reactor fuel beyond current typical levels i
(frome-27 MWD /MT to 735 MWD /MT).
1 The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
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Specification / Field Changes (contd) 12 SFC 79-013 This change modified the method of capturing the sparger ball joint support as described in a letter to the Directorate of Licensing dated April 6,1979.
Sparger ball joint support nut tack welds could not be proof loaded as specified in the sparger design report. Multiple tack welds were used to capture the joint support using equipment and methods previously proven to make a sound weld.
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The Safety Evaluation concluded multiple tack welds of previously proven 4
strength would satisfactorily capture the ball joint support and no potentially unreviewed safety question existed.
Related documents - FC 464 (Sparger Replacement)
SFC 79-014 This change consisted of modifying extension tubes and rear brackets of Pacific Scientific snubbers used on the Reactor Depressurization System, to include spherical ball bearings. The bolts attaching the brackets / extension tubes to the snubber were secured via lockwire or locknuts as applicable. The modifica-tion was necessary to insure proper clearances of the installation. Improper clearances could result in improper operation during a seismic event.
The Safety Evaluation concluded that the modification would enhance equipment operation thus upgrading equipment integrity. The modification does not constitute an unreviewed safety question.
SFC 79-016
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This change was to replace the original teflon coated metal "O" rings (large ones) on the control rod drive flanges with a rubber "0"-ring and a metal locating ring to assure orientation.
The Safety Evaluation concluded that there was no unreviewed safety question because scramming functions will not be affected. Nothing is changed that will affect the safety analysis report.
1 SFC 79-018 This change involved the addition of an alarm interrupter switch and an alarm circuit by-pasa switch into the existing alarm circuitry of a Bristol eight (8)
Point-temperature recorder. to eliminate the alarm function on point number eight (8).
l This Specification / Field Change was performed in conjunction with Facility Change No. 265 (included in this report).
13 Specification / Field Changes (contd)
SFC 79-018 (contd)
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
SFC 79-019 This change was to strengthen and maintain the shape of the small metal ring which
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holds the rubber "O" ring in place on the CRD flanges. This was accomplished by thickening the ring.
NOTE: This SFC was not used or utilized.
It was proposed and approved but the result was to continue using the thin metal ring as approved in SFC 792016.
The Safety Evaluation concluded 'there was no unreviewed safety question because the design and function of the drive itself remains unchanged. This change was to reduce leakage. Nothing is changed from the original safety analysis.
SFC 79-020 This change consisted of the repairs to control rod drive housing CRD-F2.
A leak had developed at the weld between the housing and the reactor vessel.
I The flaw actually was in the stub tube to reactor vessel weld. The leakage l
showed up during a routine hydro-static test of the primary system. Inves-tigation of the failure mechanism revealed an apparent slag inclusion in the weld during original fabrication. Repairs were made by rolling (or swelling) the control rod drive housing more tightly into the reactor vessel wall material. Subsequent tests following repairs showed no further leakage.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
SFC 79-021 This change increased the length of one thermal shield top seal housing nut by about three-quarters of an inch. This change was required when, during reinstallation of a segment of the thermal _ shield top seal housing, one of the hold-down bolts failed to fully engage. Because of this, credit could no longer be taken for this bolt as far as masking plate or top seal suppc ct
--is concerned. A stress analysis revealed the remaining forty-three hold-down bolts provided the required support without overstressing any of the bolts.
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4 14 Specification / Field Changes (contd)
SFC 79-021 (contd)
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As the bolt was bound up and could not be turned, no further attempts were made to remove it.
The stud-nut-keeper configuration was such that the nut had to be lengthened to allow the nut to housing and keeper to nut tack welds to be made.
i The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
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SFC 79-022 This change was to remove the metal support ring when a rubber."0" ring is used 1
j on the inner ring of the control rod drive flanges. The use of the metal ring j
is not needed with rubber "0" rings and the reason for removal is to prevent improper installation and to facilitate changing "O" rings.
j NOTE: The SFC was approved but was not implemented. The metal ring was still used as approved in SFC 79-016.
The Safety Evaluation concluded that there was no unreviewed safety question because scramming functions should not be affected. Nothing different is added or created that could cause a different type of accident than previously J
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analyzed.
SFC 79-023 This change modified the thermal insulation on the reactor vessel bottom head to allow examination of the CRD F2 penetration, which was repaired for leakage on SFC 79-020. The modification allowed the insulation to be removed locally, leaving the remainder in place.
The Safety Evaluation concluded that this change did not constitute an unreviewed I
safety question.
l SFC 79-025 This change involved the replacement / repair of the contact collector assembly for the reactor crane. The present equipment was obsolete and parts were no longer available for repairs. The improved replacement parts were used, allowing the system to be updated.
- The Safety Evaluation concluded that this change did not constitute an unrevie ed v
safety question.'
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15 Specification / Field Changes (contd)
SFC 79-026 This change is part of Facilicy Change #462D, Fire Barriers and Fire Stops.
Under this change, a shield of metal siding was installed next to the diesel engine driven fire pump. The shield is placed so that a diesel engine fire will not damage the electrical fire pump or the service water pumps. The change was done in accordance with Amendment 25 to the Technical Specifications.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
SFC 79-027 This change removed the welded joint in the containmeat ventilation' exhaust duct and replaced it with a mechanical seal. This allowed easy access to the exhaust valve seats for cleaning and maintenance.
The Safety Evaluation concluded that the change does not constitute an un-reviewed safety question.
SFC 79-028 This change involved the replacement of the 480V AC MCC 52-2A57 breaker utilized for supplying power to the computer. The original breaker, CE type TE136C5020 was found to be defective during routing breaker maintenance. Breaker type TED136020 was used as a direct replacement following technical evaluation by both Consumers Power Company System Protection and Laboratory Servichs Department and the vendor; the original type breaker is obsolete and can no longer be pur-chased.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
SFC 79-029 This change involved the replacement of the 480 V AC MCC 52-1C13A breaker utilized for supplying power to the screenhouse heaters. The original breaker, CE type TE136C5050 was found to be defective during routine breaker maintenance. Breaker type TED 136050 was used as a direct replacement following technical evaluation by both Consumers Power Company System Protection and Laboratory Services Department and the vendor; the original type breaker is obsolete and can no longer be purchased.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
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Specification / Field Changes (contd)
SFC 79-030 By Licensee Event Report 79-20, CPCO reported loose recirculation loop inlet t
diffusers in the bottom of the reactor vessel. To provide access to repair and repisce these diffusers (see SFC 79-031) flow baffles covering these l
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diffusers were removed from the vessel. While replacement of the inlet diffusers was in progress, modifications were made to the large flow baffles j
to simplify their reinstallation and permit access to the repaired diffusers
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r in future examinations. These modifications included cutting viewing holes ~
in the baffle to allow access to'the new diffusers, trimming the edges of the f
baffle to provide more clearance and replacement of bolts with larger, stronger bolts made of XM-19 material.
The Safety Evaluation concluded the modifications made to the flow baffle affected only installation capabilities and to permit future examinations of the replacement diffusers. The baffle still performs its function to divert water below the core l
support place before entering the core. Nc potentially unreviewed safety question
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r exists.
Related documents SFC 79-031 (Diffuser Replacement)
SFC 79-032 (Misc. Design and Installation Specification Changes)
SFC 79-031 This change replaced recirculation loop inlet diffusers which had come loose l
(ref LER-79-20) with new diffusers. Design changes were made to permit installation in the bottom of the vessel and to prevent the diffusers from coming loose in the future. The diffuser plate was made thicker and fabricated of 304L stainless steel. Bolt fasteners were larger in diameter and made of XM-19, an austinetic steel with superior tensil properties and high stress corrosion resistance. Bolts were also secured such that they could not spin and wear due to flow from the recirc loop.
The Safety Evaluation concluded the replacement diffusers were functionally equivalent to the original diffusers but incorporated design changes to
. resist loosening from the vessel wall. No potentially unreviNed safety questions existed.
- Related Documents' - SFC 79-030 (Baffle Modifications).
- SFC 79-032 (Misc. Design and Installation Specifications) r y
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17 Specification / Field Changes (contd)
SFC 79-032 This change described changes made to General Electric's Design and Installation Specifications to reflect work activities as they were actually performed during replacement of the recirculation loop inlet diffusers. No modifications were made to the diffusers as a result of these changes.
The Safety Evaluation concluded as the function and integrity of the diffusers were not affected by the changes to the design and installation specifications,
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no potentially unreviewed safety question existed.
SFC 79-033 This change involved replacement of the ventilation system multipoint temperature indicator (TI-9338) located within containment. This unit was a motor driven, analog, Honeywell Electronik 17 temperature indicator, range 0-200 F.
It was replaced with a digital Slimline Series II temperature indicator, range 0-750 F.
The new temperature indicator is installed in the same location and serves the same function as the old temperature indicator.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
SFC 79-034 This change involved the replacement of the 480V AC MCC52-2B22 breaker utilized for supplying power to the personnel lock. The original breaker, GE type TE136C5020 was found to be defective during routine breaker maintenance.
Breaker type TED 136020 was used as a direct replacement following technical evaluation by both Consumers Power Company System Protection and Laboratory Services Department and the vendor; the original type breaker is obsolete and can no longer be purchased.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
SFC 79-035 This change modified three " neutron windows" prior to installation into the reactor vessel to provide ease of handling. The new " neutron windows" were installed because the old " neutron windows" had become filled with water.
The Safety Evaluation concluded the increased neutron leakage.will not affect core physics, restore neutron windows to the original design parameters and that the modifications do not constitute an unreviewed safety question.
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18 Specification / Field Changes (contd)
SFC 79-036 This change modified the existing Yarway constant head level chambers utilized i
for steam drum and reactor vessel level systems. The change consisted of removing the temperature compensating jee.ket, thus reducing the temperature on the reference leg of each level sensor to an average temperature of less than 250*F to prevent flashing of the leg during depressurization of the primary system. Appropriate setpoint changes were performed, following modifications, so the required engineered.
safety feature response will occur before or as designed.
t The Safety Evaluation concluded that the margin of safety as defined in the basis 1
for the Technical Specification concerning level instrumentation was not reduced.
This was addressed in Amendment 31 to the Plant Technical Specifications dated November 2, 1979.
SFC 79-037 This change increased the clearances associated with the steam baffle door latching mechanism after the mechanism seized up due to galling. Additional clearances were to allow more freedom of movement with less interference between sliding parts, yet still keep the steam baffle doors closed during power operation.
The Safety Evaluation concluded as the latching mechanism still provided assurance that the steam baffle doors would remain closed during operation, no unreviewed safety question existed.
SFC 79-038 This change modified the containment supply ventilation butterfly valve' operator.
The reason for the change was to allow the option of an intermediate valve position rather than full open or full closed. This modification was made in conjunction with the NRC concerns of butterfly valve performance during a design basis accident.
The Safety Evaluation concluded the modification to the valve operator will allow more flexibility in valve disc position but not sacrifice the margin of safety.
There exists no unreviewed safety question involve'.: W.tk this modification.
SFC 79-039 This change modified the cantainment exhaust 7ent11ation butterfly valve operator.
The reason for the change was to allow the option of an inte.rmediate valve position rather than full open or full closed. This modification was made in conjunction with the NRC concerns of butterfly valve performance during a design basis accident.
Soccification/ Field Changes (contd)
SFC 79-039 (contd)
The Safety Evaluation concluded the modification to the valve operator will allow more flexibility in valve disc position but not sacrifice the margin of safety.
There exists no unreviewed safety question involved with this modification.
SFC 79-040 This ch;nge was to inject sealant inside the incore housing to prevent the metal "O" ring and tapered seat from leaking.
The Safety Evaluation concluded that there was no unreviewed safety question because ASME design pressure boundaries, components, etc. will remain the same and sealant techniques will affect reactor operations or jeopardize safety related actions or materials. Nothing is altered or changed.
SFC 79-041 This change replaced the solenoid operated control valves on the turbine bypass valve hydraulic power unit. The original valves were worn out and were obsolete, with no replacement parts available. The valves were replaced by the power unit manufacturer with equivalent valves of a different brand than the criginal.
The Safety Evaluation concluded that this change did not constitute an unreviewed safety question.
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