ML20235N047
| ML20235N047 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 12/31/1988 |
| From: | Marbaugh K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8902280551 | |
| Download: ML20235N047 (12) | |
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ATTACHMENT-4 Consumers Power Company Big Rock Point' Plant Docket 50-155' 1
ANNUAL REPORT OF FACILITY CHANGES, TESTS AND EXPERIMENTS February 24, 1989 t
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~ Facility Changes (FC)
FC-599A: Neutron Monitoring System Power Range Monitor Replacement This change replaced the power and intermediate range monitor channels 1, 2, 3, 4, & 5 in the Neutron Monitoring System with wide range monitor channels 1, 2, & 3 due to the unavailability of parts for the older equipment.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
FC-604: Removal and Relocation of Office Trailers This change consisted of the removal / relocation of various office trailers from/within the protected area.
The safety evaluation concluded that this change did not constitute an unreviewed safety. question.
FC-606A:
Installation of the Spent Fuel Pool Underwater Filtering Unit This change consisted of installation of an underwater filtering unit for the -
spent fuel pool.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
FC-609: Demin Pit Platform / Ladders This change involved the installation of a new extension ladder and modification to an existing platform in the clean-up demin pit to allow personnel safer access and comply with MIOSHA rules regarding the landing platform.
The safety evaluat1on concluded that this change did not constitute an unreviewed safety question.
FC-613: Recire Pump Room Lighting This change consisted of providing an additional power source for an emergency lighting unit to allow lighting in the steam drum enclosure to be turned off during plant power operation. This change was necessary to resolve personnel safety concerns about adequate lighting for initial entries into the area after plant shutdown.
I The safety evaluation concluded that this change did not constitute an
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unreviewed safety question.
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2 FC-614: Remote Pilot Valves This change consisted of replacing the existing pilot valve assembly'and piping on the reactor depressurization valve tops with a flanged pilot valve assembly and two isolation valves to allow change-out of a pilot valve assembly during plant operation.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
Subsequent Technical Specification Amendment No. 93 changed surveillance test requirements associated with this modification.
FC-615: Modify Big Rock Point CCTV System This modification involves changes to the Big Rock Point Closed Circuit TV system. This is a security modification.
The safety evaluation concluded that this change did not constitute an unreviewed. safety question.
FC-617:
East Building Light and New Appliance Circuit This change installed a new pole light by the East Office Building to improve ~
safety and security. This modification also added a new appliance circuit to~~
facilitate a remodeling change in the office building.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
FC-619:
Security Equipment Power Supply This-change involved a security modification.
The safety evaluation concluded that this change did not constitute an i
unreviewed safety question.
FC-621: Removal of ACB 052 - 1A51 This change consisted of removal of an unused breaker (052 - 1A51) in the station power distribution system because the operating handle on the breaker broke during testing.
-The safety evaluation concluded that this change did not constitute an unreviewed safety question.
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3 FC-633: Modification to Room 213 This change involved the relocation of walls and doorways near the radiation protection office..This provided a low background area for the installation i
of two whole body personnel contamination monitors and provided an entrance to the radiation protection office which is not through a radiologically controlled area.
The safety evaluation concluded that this change did not const.itute an unreviewed safety question.
I FC-634: RDS Jib Crane Modification l
This' change consisted of the addition of a telescoping beam assembly to the i
existing Reactor Depressurization System (RDS) jib crane to provide the l
D capability of extending to the most remote RDS valve train area.
The safety evaluation concluded that this change did'not constitut'e an unreviewed safety question.
FC-637: Hand Trip Solenoid Coil Continuity Lamp This change consisted of the installation on an indicating lamp in series with the turbine hand trip solenoid coil to provide continuous monitoring of circuit continuity through the coil.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
FC-638: Additional Conduit This change consisted of installing additional conduit between two junction boxes and two cable trays.co facilitate cable replacement for two valve operators on the Primary Coolant System.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
Specification Charges (SC)
-This change consisted of replacement of the station power system 480V ac circuit breaker 052-1D22 with an equivalent type.
The original breaker failed to meet magnetic trip requirements during routine testing.
The' safety evaluation concluded that this change did not constitute an unreviewed safety question.
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4 SC-87-020 This change replaced the upstream and downstream' nitrogen test connection valves (VRDS-102C 6 VRDS-103C) on the Reactor Depressurization System due to leakage on VRDS-102C and the fact that VRDS-103C had been furmanited and was scheduled to be replaced.
This.also changed the specification to allow the use of either a 1/2" ASME Winter 1974 Class 2 Section III nuclear valve or
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a 3/4" ASME Winter 1973 Class 2 Section III nuclear valve.
The safety evaluation concluded that this change did not constitute an unreviewed. safety question.
i SC-87-023 This change consisted of replacement of the 60 cell, Exide types EHGS-17 and ET-13, station battery and rack with a 60 cell, Exide type EA-13, station battery and battery rack.
Exide type EHGS-17 is an obsolete type and was reaching the end of its life.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
'SC-87-024 I
This change consisted of replacement of the lock washers on the Target Rock valves (SV-4984 & SV-4985) on the Reactor Depressurization System with flat washers. This change was necessitated by the fact'that the lock washers cut into the flanges and nuts leading to thread contamination and galling. This i
modification also changed the. specification to allow the use of either 1-1/8" carbon steel lock washer or 1/8" x 1.140" I.D. x 1.875" 0.D. carbon steel flet washer.
The safety evaluation concluded that this change did not constitute an i
unreviewed safety question.
I SC-87-029 This change consisted of replacement of a Reactor Depressurization System pilot valve assembly (SV-4984) internals with that of internals modified by Target Rock Corp. to eliminate / reduce pilot disc leakage. This modification also changed the specification to allow the use of the modified internals in other pilot valve assemblies.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
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"SC-87-033 This change consisted of replacement of two control rod drive pump Crosby, j
sty 1'e JMB-C, relief valves _(RV-5050 & RV-5051) with circle seal, style 3159-S-6MP-1950, relief valves in order to reduce valve seat leakage and related problems.
TheLsafety evaluation concluded that this change did not constituteLan.
unreviewed safety question.
This change provides for the installation of a second seal in series with the existing piston seal ring'in the. control rod drive tube and flange assembly._ The new seal will reinforce.the pressure boundary between the insert and withdraw flow paths. This should reduce internal bypass leakage across the seal-ring and result in improved drive performance.
The' safety evaluation concluded that this change did not constitute an unreviewed safety quistion.
SC-87-035-This. change was to permit the use of either a 3/8" or a 1/2". flow control valve on the_ Control Rod Drive system. This will increase the flow through
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the' control rod drive during insertion and withdrawal and should offset the effects of higher internal seal leak rates and improve control rod performance.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
SC-88-001 1-This change consisted of replacement of three relief valves (RV-5077, RV-5078
& RV-5082) on the Post Incident System from a Kunkle model no. 20-1 to a Circle Seal model 5132-T-6MP in order to reduce valve seat leakage and related problems.
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The safety evaluation concluded that this change did not constitute an-
.unreviewed safety question.
I SC-88-003-This change involved the replacement of K-386B, General Electric relay, type
'HEA51, model 12HEA51B237 with an Electro-Switch, (LOR series 24) relay,
~#7331-7806E on the turbine generator system due to binding in the hand reset switch mechanism which resulted in coil failure.
The safety evaluation concluded that this change did not constitute an unreviewed safety. question.
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'This' change, consisted of replacement of the recirculation pump heat exchanger S
'(HX+21B)- due to tube bundle. leakage. The original Borg-Warner model SS-7421 is no longer manufactured and was replaced with a Borg-Warner model NX-0750-HT-FW.
The safety. evaluation concluded that this change did not constitute an unreviewed safety question.
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SC-88-005 This change' consisted of replacing three control rod drive selector and
. solenoid valves:(CV-NC19C2, CV-NC19C3, CV-NC19C4, SV-NC19C2, SV-NC19C3 and SV-NC19_C4). This' modification also changed the original specification to allow the use of.either the Hydromatics model 612D697G1, style C-715 or the
. Worcester model 4 for selector valve replacement. This change also allows the
.use of either the Bellows-Valvair solenoid valve model APCSK-15B or the Skinner solenoid valve model PM3B5080 for solenoid valve replacement on the control rod system.
The safety, evaluation concluded that this change did not constitute an unreviewed safety question.
SC-88-006 ThisEchange involved changing the cycle 17/ cycle 22 reload fuel design package to also establish a control rod insertion-sequence which allows.
-running control rods in after the turbine generator has been taken off line.
The safety evaluation' concluded that this change did not constitute an unreviewed safety question.
SC-88-008 This ' change consisted of installation of a front shaf t journal sleeve and replacement of the existing rear shaf t journal sleeve, with one of greater length, on the clean-up pump (P-5) on the reactor clean-up system. This
. change was necessary to restore surface condition to the pump rotor shaft due oto pitting and wearing at the front and rear journal.
.The safety evaluation concluded that this change did not constitute an unreviewed safety question.
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L SC-88-010,88-011, & 88-012 These changes involved the cycle 23 reactor core reload. The new fuel falls within previously docketed designs.
Reconstituted fuel assemblies are in
-accordance with Consumers Power Company guidelines and are consistent with previous low power design.
~The safety evaluations concluded that.these changes did not constitute an unreviewed safety question.
SC-88-016-This change consisted of replacement of a liquid poison tank Crosby style JMB-S relief valve (RV-5049) with a Lonergan model 40/S4 relief valve due to failure of the Crosby relief valve. This change also allows for the use of either model valve if future replacement is required.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
I SC-88-017-This' change consisted of replacement of an Agastat relay model 2432 ANN (K-0127-5) with that of an Agastat relay model 7032ABC on the station power system due to failure of'the timing relay to function within acceptable time frame during routine testing.
.The safety evaluation concluded that this change did not constitute an unreviewed safety question.
SC-88-018-This change consisted of installation of of a new disc in a check valve (VPI-303) in the post incident system fabricated from SA-105 (forged carbon steel) rather than from ASTM A296 CA15 (cast stainless steel) as listed in the original specification and allows the use of either in the future. The manufacturer was unsble to supply a disc that met the original specification on an expedited basis and recommended the use of SA-105.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
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'This change consisted of replacement of an Electric Machinery Mfg. Company voltage regulator model #R-9806-27 with that of a Basler Electric Company voltage regulator model KR2FF with Basler Model BE10317 isolation transformer on the emergency diesel generator. The regulator was replaced because it had exhibited several failures to provide excitation to the generator exciter over a period of years.
.The safety evaluation concluded that this change did not constitute an unreviewed safety question.
.This change consisted of changing one relief valve (RV-5032) on the reactor cooling water system from a Crosby, model H3-J0-25, with a Crosby model H3-J05-15, which is the current replacement model for that valve. This change also allows the use of either model valve on two other relief vales (RV-5028 &
RV-5029) on the reactor shutdown cooling system.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
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This change modified the control rod withdrawal sequence found in the Cycle 23 final. physics package from notch 80 through notch 112.
This was done to eliminate a high notch worth at the point of criticality for beginning of i
cycle cold conditiens and provide for a smoother transition into the heating
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range.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
This change consisted of changing the input sensitivity for the new de wide j
range monitors in the Neutron Monitoring System to match the input detector current from the compensated ion chambers (Reference FC-599A).
The' safety evaluation concluded that this change did not constitute an 1
unreviewed safety question.
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9 SC-88-025 This change involved a part number change for the Reactor Depressurization System UPS battery charger oscillator board by the manufacturer. The original board part number is obsolete and an updated board serves as the replacement part.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
SC-88-026 This change consisted of expanding the detector sensitivity adjustment and the rated power adjustment for the de Wide Range Monitor on the Neutron Monitoring System. This change was necessary to provide enough latitude to calibrate the monitors to various plant power levels (Reference FC-599A).
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
SC-88-027 This change consisted of replacement of two control rod drive pump Circle
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Seal, style,3159-S-6MP-1950, relief valves (RV-5050 & RV-5051) with Crosby, style JMB-C, relief valves due to worst than expected valve seat leakage with the Circle Seal relief valves installed under S0-87-033 on this report.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
SC-88-028 This change consisted of installation of a stainless steel sleeve on the motor shaft of the #2 air compressor motor (M-49B) which had been found galled and undersized.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
SC-88-029 This change consisted of replacement of the ion chamber assembly (part no. 847B-1) with a new detector and preamp (part no. 847C-1) in the high range noble gas monitor due to failure of the original assembly.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
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SPC-88-001 1
This chinge consisted of adjusting the float voltage level on battery charger l
(Bd-1 and BC-1A) as recommended by the manufacturer to ensure that tne proper float voltage.is impressed across the new station battery bank.
(Reference SC-87-023)
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
SPC-88-002 This change consisted of raising the set point on three relief valves (RV-5077, RV-5078 and RV-5082) on the Post Incident System.
The previous set point was too close to the system operating pressure.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
SPC-88-003 This set point change temporarily raised the equalize voltage on the alternate shutdown battery charger to enhance the raising of the specific gravity values of the alternate shutdown battery bank during performance of j
procedure TR-65-ASD, Alternate Shutdown Battery Service Test.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
SPC-88-005 This change temporarily increased the reactor building high temperature alarm set point due to the hot weather.
This change allowed the alarm to remain in service to monitor further temperature increase.
The safety evaluation concluded that this change did not constitute an unreviewed safety question.
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@ POWERING Consumers POW 81 MKNIGAN'S PROGRESS Genersi Offices: 1945 West Parnall Road, Jackson, MI 49201 * (517) 788-0550 February 24, 1989' Neelear Regulatory Commission Document Control' Desk Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT -
1988 ANNUAL REPORT OF FACILITY CHANGES, TESTS AND EXPERIMENTS Attached is Consumers Power Company's Annual Report describing the Facility Changes (FC), Specification Changes (SC),. Set Point Changes (SPC), tests.and
- experiments performed at the Big Rock Point Plant in 1988. There were no
- changes to procedures in 1988 that are reportable to NRC. This report is submitted in accordance with the provisions of 10CFR50.59(b).
0 Kenneth E Marbaugh Plant Licensing Engineer CC Administrator, Region III, NRC NRC Resident Inspector - Big Rock Point Attachment f
8 OC0289-0050-NLO4
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