ML20236F893
| ML20236F893 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 10/28/1987 |
| From: | Tucker H DUKE POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8711020331 | |
| Download: ML20236F893 (8) | |
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- Duke POWER GOMPANY
. P.O. DOX 33189
-l CHARLOTTE, N.O. 98942
' Hall 11.TUCKERL retzpues
' YNJS PRESIDENT s-Ob4NS nuos.sas enouvunos 4
- October 28,-1987' L
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U.S yNuclear,,Regula tory Commission'.
(~DecsmeadComiro1[DesQg-NishinsEh~rs V."C.720555
Subject:
~ McGuire Nuclear' Station Docket Nos. '50-369, -370 Changes to Facility Pursuant To 10CFR-50.59-Gentlemen:
Attached is.a summary of changes m'ade to the McGuire Nuclear Station, pursuant to
~10CFR 50.59.
JVery truly yours, l
1 Hal B. Tucker l
SAG /93/j gc j
Attachment xc: ~ Dr.' J. Nelson Grace, Regional Administrator U.S.. Nuclear Regulatory Commission - Region II 101 Marietta Street, Suite 2900 Atlanta, Georgia 30323 1
Mr. Darl Hood, Project Manager Office of Nuclear Reactor Regulation
.U.S. Nuclear Regulatory Commission
'. Washington, D.C.
20555
- Mr. W.T. Orders NRC Resident Inspector:
McGuire Nuclear Station 871102Og 87-ena 5 p M
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-CHANGES TO FACILITY DESCRIPTION
. NUCLEAR STATION MODIFICATIONS 1
(PURSUANT TO 10CFR 50.59) '
_r NSM No.-
(MG-
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11711 DESCRIPTION Class F bypass piping around valves ; 2NM424, 425,: 426 and 427 re-placed with (upgraded to) Class B piping.
SAFETY EVALUATION
SUMMARY
No safety system will be degraded and no functional' change. to. any system will be made as a result of this piping upgrade.
The remain-der ? of the associated piping ' was' correctly installed to Class B criteria..
~20628.
' DESCRIPTION-Various modifications to Spent Fuel Handling ' area - to accommodate element. fuel shipping' cask:
1)
Replace grating overlay in Decon Pit with overlay with. larger opening.
2)' : Add Grating at bottom of. Decon Pit, elevated approximately 3",
. to keep. valves at bottom of cask out of water.
3)
Add lighting to Decon Pit.
4)
Mount -storage stands for new cask handling equipment.
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i SAFETY EVALUATION
SUMMARY
The modifications ' are intended to facilitate. handling of ' the new-casks and do not directly affect x the cask drop analysis.
No new accidents are created by these modifications. The consequences of a cask drop have been analyzed and found to be acceptable.
11863.
DESCRIPTION This modification.will move flow instrumentation (ORNFE ' 6120)~
upstream of control cable: and equipment room air conditioning.
condenser.
' SAFETY EVALUATION
SUMMARY
' This modification will' improve the accuracy of. the flow instru-mentation.
No safety system will - be degraded or safety margin
' decreased.
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1, 1 NSM No.
(MG-
)
'11841
' DESCRIPTION y
Flow' measurement orifice and pressure E aps - for RN Pump. are changed t
from a non-standard corrosion-susceptible. design to stainless' steel.-
4 SAFETY-EVALUATION
SUMMARY
This change will improve. the accuracy of '. the instrumentation and.
will not degrade any safety system.
--11854
' DESCRIPTION
- a. :
Drain connectors are added to RN Heat Exchanger.1B Inlet and Outlet piping,oin~ order to facilitate cleaning and' draining..
SAFETY EVALUATION
SUMMARY
I No safety function is. adversely affected by this modification.
No D
accident probability is created or increased.
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'11862 DESCRIPTION Insta11agion of isolation valves on heater vent lines, to enable change of piping.- The valves.will prevent vacuum loss in the main
- condenser during-changeout.
SAFETY EVALUATION
SUMMARY
.No unresolved safety question is created by this modification.
No system will~ be degraded"and no functional changes will-be made to any system as a result of this modification.
The valves will ' be Linstalled during an outage to enable' a changeout while on line.
i 11827-DESCRIPTION
_j Installation of check valves in the discharge piping of the Instru-ment Air (VI),' to prevent air bleedoff (and resultant Reactor Trip)
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in the event of a VI hose failure.
1 SAFETY EVALUATION
SUMMARY
This modification will reduce the probability of unnecessary reactor
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trip transients by mitigating the effects of a break in a VI line.
No decrease in any safety margin or increase in the probability of any accident'will occur.
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L NSM No.
(MG-
)
01768 DESCRIPTION Rev 2 Modifications to Cold Leg Accumulator (CLA) as a result of Upper Head Injection (UHI) isolation / removal.
The changes include:
1)
Change in discharge line orifice plate hole to ensure system operability.
2)
Increase nitrogen cover pressure to 585 psig.
3)
Relocate CLA level transmitters to monitor new operating range.
SAFETY EVALUATION
SUMMARY
1)
The change in CLA discharge line orifice was analyzed - by Westinghouse and found to be appropriate in conjunction with UHI removal.
2)
Increase of nitrogen cover pressure will cause CLA blowdown to occur sooner in loss of-coolant accident analysis.
3)
As a result of UHI removal, CLA inventory is reduced.
The relocation of the level transmitters is consistent with this change in level.
These changes will not result in any significant increase in acci-dent probability or decrease in cafety margin.
11685 DESCRIPTION Instrument Air (VI) line run to detectors in Count Room to detect ambient Xenon.
The presence of Xenon in the Count Room ambient air has resulted in contamination of samples.
SAFETY EVALUATION
SUMMARY
The routing of VI to detectors does not affect the function of any safety or non-safety system and will not increase the probability of any accident or decrease any safety margin.
01765 DESCRIPTION Installation of 2 thermocouple in Control Room to monitor ambient air temperature. The 2 thermocouple produce an average temperature output which provides an " ALERT" signal at 85'F (increasing) and an
" ALARM" at 90*F (increasing).
SAFETY EVALUATION
SUMMARY
This modification will provide increased assurance of protection of equipment from failures due to overheating. No accident probability will be created or increased, or any safety margin diminished.
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NSM No.
(MG-
)
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'11456 DESCRIPTION 1
This modification adds a permanent low pressure supply of nitrogen to the KR (Recirc Cooling Water) Storage Tank, in order to preclude Oxygen from entering the tank.
i SAFETY EVALUATION
SUMMARY
This modification will assist in the chemistry control of the KR system and will not have an impact on safety.
11426 DESCRIPTION Installation of sightglasses to monitor Reactor Coolant System water
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1evel during maintenance and refueling.
The sightglasses are added i
to Loop C.
The sightglasses will be isolated during normal system operation.
I SAFETY EVALUATION
SUMMARY
i Because the system is isolated during normal operation, no probabil-ities of accident are created or increased, and no safety margins are significantly affected.
11414 DESCRIPTION Replacement of Contaminated Liquid Waste (WL) flow instrumentation l
which will allow a water solid condition to be maintained, thus i
reducing errors caused by evaporation of water in instrumentation legs.
I SAFETY EVALUATION
SUMMARY
I The new instrumentation is expected to be at least as accurate /re-liable as the replaced instrumentation; no unreviewed safety ques-tion will be created.
11396 DESCRIPTION Boric Acid Air Conveyor unit is replaced with a screw-type unit to promote ease of operation, reduce dust creation, and reduce clog-ging.
SAFETY EVALUATION
SUMMARY
The function of Boric Acid Conveyor unit is unchanged. The new unit will improve the efficiency / reliability of the system; thus, no unresolved safety issues are created.
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- NSM No.
(MG-
)
11336 DESCRIPTION Various Carbon Filter units (Auxiliary Building, Control Room, Annulus, Fuel Handling, Containmen t Purge, and Turbine Building Ventilation Units) are susceptible to flooding in the event of a failure of a float ball water feeder.
This modification will provide a valve in each filter unit drain line to isolate the unit form possible flooding and carbon destruction.
SAFETY EVALUATION
SUMMARY
The drains were intended to divert water from a deluge fire protec-tion system.
In order to prevent air from escaping from one filter unit to another through the common drain header, a 12-inch vertical water column is maintained in the header by means of a water float feeder.
Should this feeder fail open, the carbon filter elements would be ruined, as has happened in the past, given the low proba-bility of a fire (the treated carbon will not support combustion),
and the inadequacy of the drain system relative to the volume of water to be delivered by the deluge system, the drains may be isolated with no significant impact.
11276 DESCRIPTION This item consists of a two floor addition to the West side of the Service Building to provide 8400 sq. f t. of office and assembly area for Contract Services.
In adoition, some modifications will be made to the existing Security Office area and the PAP.
SAFETY EVALUATION
SUMMARY
This addition will have no detrimental effect upon,any safety system and no accident probability will be created or increased, or safety margin decreased.
No functional change will be made to any struc-ture or system.
No failure of the addition will have an effect on nuclear safety.
11272 DESCRIPTION This item consists of converting the existing QA Warehouse into two levels of shop and office space (9600 sq. f t.) for the I6E Section.
SAFETY EVALUATION
SUMMARY
Th1s modification consists of a conversion of warehouse space into offices and work areas; largely cosmetic, superficial, or non-load-bearing construction.
No safety system is impacted and no unre-viewed safety question is involved.
n I NSM.No.
(MG-
)
l 11271 DESCRIPTION A pre-engineered metal building will be erected to provide warehouse and of fice space.
SAFETY EVALUATION
SUMMARY
No safety system will be affected by this modification.
No col-lapse, fire, or ether event involving the building will result in a
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condition having nuclear safety significance.
10436 DESCRIPTION A check valve
'.s installed in each Reactor Coolant Pump number one seal _ leakoff line, to prevent backflow during periods of low RCS l
pressure.
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i SAFETY EVALUATION
SUMMARY
This modification will prevent particulate carried by backflow through seal leakoff lines from degrading RCP seals. No function of j
the RCP or other component will be affected, no accident probability i
will be created or increased, and no safety margin will be de-l creased.
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i 10660 DESCRIPTION This change involved tha reanalysis of concentrates piping in the chemical and volume concrol system (CVCS) to the Boric Acid Tank (BAT).
The reanalysis wac necessary based upon actual operating 3
temperatures which requiced an upgrade of the piping from 200*F to 250*F.
The analysis cot.:1uded that no hardware modifications were necessary for the upgrado.
SAFETY EVALUATION
SUMMARY
This change ensured system safety and reliability by assuring that the piping was capable of performing its design function under anticipated conditions.
CHANGES TO PROCEDURES 1.
PT/0/A/4150/28 -
Criticality following a change in Core Nuclear Characteristics DESCRIPTION OF CHANGE:
This represents a new procedure for Initial Criticality, different from the existing (FSAR Ref: Table 14.1.4-1, Page 15 of 35) procedure.
The change provides for closer monitoring (via ICRR Instrumentation) of Boron concentration in Mode 3 during startup.
SAFETY EVALUATION
SUMMARY
The probability of a boron dilution accident is reduced because the ICRR is monitored more closely to ensure that criticality occurs in a safe and orderly manner.
The consequences of any accident remain as previously evaluated into the FSAR.
2.
EP/1-2/A/5000/1-14 OP/1/A/6100/02 OP/2/A/6200/10 DESCRIPTION OF CHANGE:
Each procedure was changed to delete reference to Upper Head Injection, which is no longer in place.
SAFETY EVALUATION
SUMMARY
The removal of UHI has been extensively analyzed and reviewed, and has been approved by the NRC Staff.
Changes to procedures to reflect UHI removal will not affect the probability or consequences of any accident beyond what was previously analyzed.
3.
IP/0/A/3250/39 DESCRIPTION OF CHANGE:
This change reflects that the scales of hydrogen analyzers in containment are 0-9% rather than 0-30% as specified in FSAR.
SAFETY EVALUATION
SUMMARY
The concentration of hydrogen in containment will not exceed 4 percent, even if no measures are taken to control hydrogen concentration.
The reduced scale of the analyzers will not result in a change in the probability or consequences of any accident.
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