ML19289D342
| ML19289D342 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 02/27/1979 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML19289D341 | List: |
| References | |
| NUDOCS 7903050355 | |
| Download: ML19289D342 (12) | |
Text
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FACILITY CHA!!GES, TESTS AIID EXPERIME:!TS Facility Channes (FC)
FC-301 This change consisted of fusing the plant public address systen viring s
This change was made to preclude loss of feed to reactor containment.
the public address systen external to cantainment during a LOCA.
The safety evaluation ccncluded that this change did not constitute an unreviewed safety question.
FC L52 This change removed the Steam Drun Level High detection from the Reactor High level in the stean Depressurization System auto-test circuitry.
annunciation drum is appropriately covered by both conventional plant Drum Level High Bistable in and annunciated tripping of the Steam Coverage by the auto-test cir-the Reactor Depressurization Systen.
feature en and inhibited use of the auto-test cuitry was redundant other important parameters during periods when the steam drum level startup, cool-was intentionally carried at high level; such as plant down, shutdown, etc.
The design of the aute-test circuitry is suci' that feature to " lock up" until a detected fault vill cause the auto-test the fault is corrected.
cenatitute an The safety evaluatica concluded that this change d 3 not unreviewed safety question.
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790305035E
Facility Chances, Tests or Exterinents (contd) r e-4, ol
-n This change consisted of enlarging the cite counting room to inprove envirennental conditions for garna spectroscopy equipment and provide additional space to house all counting equipment in one area.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
ro _.'t V1, This change involved the addition Of improved heating and ventilating '
m components to the service building addition. Components added consisted of the addition and relocation of thermostatic controls and rebalancinc, of the office area forced air flow.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
cCi,og This change involved the additicr of three lighting fixtures and minor relocation of existing fixtures in the plant stockroc to provide safe lighting levels in the storage bin areas.
The safety evaluation concluded that this change did not constitute an unreviewed safety cuestion.
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f This change involved the 2dditicn of control p1nel viring to enable ace cf high temperature alar r..it:hes in the Centainment 3alldin; Temperature /revroint Terperature ?ecorder. The alar cvitches uere vired to an unusued innunciator circuit to provide 20nitoring of selected ege 4
Facility Chances, Tests or Experinents (contd)
FC h71 (contd) ambient and dewpoint temperatures within the containment building ventilation flev paths. Previously, the alarm switches were unuced.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
Functional Equivalent Substitutions (FES)
!!OTE : Effective February 1,1978, the iesignstion for this type of minor modification was changed to Specification / Field Change (SFC).
FES-77-11 This change involved the replacement of the original plant telephone connections external to the plant perimeter fence to a new location within the fenced area.
The safety evaluation concluded thet this change did not constitute I
an unreviewed safety question FES-77-35 This change involved the replacement of cne (1) obsolete turbine-t i
l generator system gland seal exhauster with a replacement unit frc=
l the same manufacturer. Physical sising of the blower was slightly I
different; electrical input requirenents to the blower motor were I
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Functional Eauivalent Substitutions (FES)
FES-78-01 This change involved the replacement of the position Indicating switches on the Reactor Depressurination System isolation valres with LCCA qualified switches.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
FES-78-02 This change involved the replacement of a solenoid valve coil cn the ;urcine air ejector system with a vendor-supplied coil rated for higher anbient temper-ature operation.
The safety evaluation concluded that this chance did not 00nctitute an unrevieved safety question.
Specification / Field Chances (SFC)
SFC-78-C01 This change consisted of replacing a relay in the manual reactor control system with a relay of identical electrical ani physical characteristics encept for a clip to support an external housing.
In this application, the external housing is not used.
".'e safety evaluation concluded that this change dic not constitute an unreviewed safety question.
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Stecification/Fieli Chanzes (FF7)
SFC-73-002 This change consisted of replacing the 3cactor Protection System steam drum level sencor covers with LOCA qualified covers of LEIU naterial.
The safety evaluation concluded that this change did not const'itute an unreviewed stfety question.
FFC-73-00h This change involved mincr modifications to *,he outboard surface of the packing gland thrcat bushing of the control rod drive pumps to allov 2 relief for packing leakage. Tais miner packing leakage provides additional lubric1 tion for the plunger during pump operstions.
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SFC-73-009 This change consisted of replacing the solenoic enclosure for the cler.up decinerslicer resin sluice isolation valves (containment iso-lation sycten) with an enclosure entircnnentally qualified to teet post-L CA conditions.
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SFC-78-006 This change consisted of replacing the packing-type seal for the cooling vater pump on the Energency Diesel Generator with a nechanical seal recon = ended by the punp renaer to eliminate failures associated with.the packing-type seal.
constitute The safety evaluation concluded that this change did not an unreviewed safety question.
S?C-78-012 This change involved repairing and replacing the upper one-half of the containment building insulating material (Insulnastic) with urethane foam and protective coating.
The safct/ -raluation concluded that this change did not constitute an unreviewed safety question.
SFC-7a-013 This change consisted of replacing the plant telephone system incoming telephone lines frcn 50 pair to 150 pair. The addition of the 100 extr1 pair was required to 1) provide expansion of the telephone systen as all lines were in use 2) eliminate splices in the existing cable which had been in use for vl8 years and 3) eliminate "1 coping" of circuits to provide service to outbuildings within the plant perimeter.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
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Specification / Field Chances (centd)
SFC-78-Clk This change involved the replacceen; of the control rod drive "O"-ring spacer plate with a plate approximately.010" thicher to provide optimum compression of the "0"-rings.
The thicker spacer plates were recenmended by the reactor vessel / control rod drive supplier.
The safety evaluatien concluded tha this change did not constitute an unreviewed safety question.
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This change involved the renoval of a metal "O" ring in the reactor depressurization valve pilot 2:semblies. Originally, this cc=ponent was utill:cd to minimize laahage around an internal sleeve; present valve design provides a pressed fit and the vendor indicated the "0" ring is no lenger required and could be rencved ithout degrading ym,.,..-
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Scecification/ Field Chances (SFC)
SFC-78-020 This change involved the repair / replacement of anchor bolts in a Reactor Depressurization System Constant support hanger and rotation of the hanger base plate by 90* co allev for better movement of the piping in the north-south direction, The above changes were performed following consultatica and design input from the original piping support design firm.
constitute an The safety evaluation concluded that this change did not r.revictei safety Tc.estion.
SFC-78-021 This change improved the change made in SFC-013 by the addition of a second idler bushing to allev adjustment of the chain tension and provide greater chain contact with the sprockets en the crane safety trip vinch.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
3FC-78-022 This ch1nge involved the fabrication of a spacer ring to provide additional sesling area on the prescure seal of one Reacter Depressurizaticn Systen isolation valve. The spacer was fabricated (follouing censultation with the valve vendor) frc the pressure seal gasket previously recoved to assure material compatibility. A repair of this nature wris required to correct "vire cutting" on the valve bcdy.
The safety evaluation concluded that this change dia not constitute an unreviewed safety question.
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Scecification/ Field Chanaes (SFC)
SFC-78-023 This ch1nge involved the replacenent of teflon coatel "O" rings for the ecoling water ports on the centrol rod drive nating flang2s with "0" rings of Buna-U rubber. Use of the "O" rings was reconnended by the USSS vendor for a period not to exceed six nonths, based on present engineering data and experience.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
SFC-78-02k This change consisted of replacrent of an unusued 3/' " vent connection and as00ciated piping in a 3" return line frca the clean-up demineralizer to the "o. 2 reactor recirculation pump vith a straight length of 3" piping.
The vent was used during plant censtruction and startup and now is not required for general plant operation.
Repair of the section was required as the 3/h" piping stub had cracked at the base of the nipple. The node of failure was tentatively deter-nined to be fatigue, induced by vibraticn of the stub / valving during oceration.
The safety evaluation concluded tha; this change did not constitute an unreviewed safety question.
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n on-This change involved the sealing of all containment elcetrical penetratien splices for equipment requ2 red to operate durinc a 1CCA with envirencentall/
qualified tape.
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Specification / Field Changes (3FC)
SFC-76-025 contd Originally, " weatherproofing" splice connectors were used at the penetration splices. Since the plant was built, other required nodifications resulted in splice connectors that could not be identified as LCCA qualified while some exhibited poor mechanical quality.
All penetration splices both inside and cutside of containment required for LOCA operation vere inspected for mechanical integrity and water-tightness. As a result of the inspection,100 splices inside containment and 75 splices outside containment were repaired with qualified splice connectors (utilizing a ratchet-type compression tool of the type initially used for installing th.e original splice connectors).
In addition, all the splice connectors (312) inside containment were covered with a two-tape LOCA qualified wrap to assure water-tightnecs during LOCA conditions; the seventy-five splices outside containment were also wrapped to acsure water-tightness against flooding during fire suppression by water sprays.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
Srecial Site Tests (SST)
SST-12 Investication of CED C1 and E3 Inadvertent Scr1m In October 1977, a SCRAM of CRD Sh was initiated at power and CRDs C1 and E3 inserted as well. CRD Bh was selected at the time of the SCRAM and the suspected cause of C1 and E3 insertien was selector valve leakage.
In January 1978, a special test was performed while shut devn to test the integrity of the selector valves for C1 and E3 The insertien as noted in October 1977 could not be repected and selecto: valve leak 1ge appeared to be =inimal.
The safety evaluaticn concluded that this test did not consti+ute an unreviewed safety questien.
SST-13 Rine Scareer Nozzle Measurements An exe ption from 10 CFR 50.h6 was granted for the course of Cycle 15 due to unce'rtainty in the adequacy of the spray distribution of the core spray ring sparge /.
In January 1978, tools and a special test were de-veloped to locate pre cisely the locaticn of the core grid at which eaw-no::le of the ring sparger was aired.
The tool consisted of a laser =cunted on an air-operated collet attached to an actuator pele. The air-cperated collet vculd be placed over a given no::le in the vessel and retracted fixing the laser en the no::le. The laser would be excited and the locatien of the beam picked off a target grid located en tcp of the core. '4hile the tecl verked reasonably well, precise location of the aiming targets was not possible as the collet used Special Site Tests (SST) (Contd)
SST-13 (Contd)
Ring Sparrer No :le Measurements (Contd) in the assembly fit icosely over the ac::les resulting in incensistent results. The safety evaluation concluded that this test did not ccnsti-tute an unreviewed safety question.
SST-lh Irradiated Fuel Examinatien by Los Ala=cs Scientific Laboratorv (LASL)
Research Team 31g Ecek Point Nuclear Plant was one of several power reactor sites visited by the LASL Research Team to perform a special nuclear materials safeguard study for the US Department of Energy. This study will provide a theoretical basis for the ncndestructive assay of irradiated fuel assen-blies. The application of this study will create a method of independent verification, by a ec=pliance inspector, of irradiated fuel assembly special nuclear raterial centent. This independent verification vill provide a check to insure that no special nuclear raterial has been diverted. The safety evaluation concluded that this test did not censitute an unreviewed safety questien.
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