ML20012G455
| ML20012G455 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 12/31/1992 |
| From: | YANKEE ATOMIC ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20012G452 | List: |
| References | |
| NUDOCS 9303020475 | |
| Download: ML20012G455 (11) | |
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YANKEE ATOMIC ELECTRIC COMPANY YANKEE NUCLEAR POWER STATION (DOCKET NO 50-29) 1992 ANNUAL REPORT P
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i 9303020475 930226 PDR ADOCK 05000029 R
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1992 ANNUAL REPORT TABLE OF CONTENTS l
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TITLE PAGE l
t Introduction.....................................
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i Changes A.
Engineering Design Changes...............
2 B.
Plant Design Changes.....................
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C.
Non-Nuclear Safety Changes...............
'3 D. Temporary Design Changes.................
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E. Other Changes............................
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INTRODUCTION The Yankee Nuclear Power Station, a pressurized water reactor plant previously rated at 185 MW electrical capacity, has been shut down since October 1, 1991, and the reactor permanently defueled as of February 15, 1992.
The plant has been operating under a Possession Only License (POL) since August 5, 1992.
The Defueled Security Plan was approved by NRC on November 24, 1992 and implemented on December 19, 1992.
The Defueled Emergency Plan was approved by NRC on October 30, 1992 and implemented on December 18, 1992.
This report of changes, tests, and experiments is submitted in accordance with 10CFR50.59 (b) (2).
It also satisfies the requirements for Technical Specification 6.9.2.b, c and d.
6 The changes identified in this report have been reviewed for, and were determined not to constitute, an unreviewed safety question as described in 10CFR50.59 (a) (2).
Technical Specification 6.9.2.b requires reporting any other i
unit-unique reports.
No such reports were applicable in 1992.
Technical Specification 6.9.2.c requires reporting a summary of l
safety valve and relief valve failures and challenges.
There were no challenges to the pressurizer or steam generator safety and relief valves, nor were there any failures of those safety and relief valves required to be operable by the Technical Specifications in 1992.
Technical Specification 6.9.2.d requires reporting of specific activity analyses submitted to identify primary coolant which has
~i exceeded the limits of Technical Specification 3.4.7 No such limits were exceeded in 1992.
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A. Enaineerina Desian Chances (EDCs) o EDC 90-309 - Station Service Upgrade This change installed a new 1,000 kVA unit substation for plant loads and a new 150 kVA pole-top transformer for the Training Center, powered from the existing 13.8 kV feeder from Massachusetts Electric Company (MEC).
This equipment will be used to power selected plant loads which were supplied from the auxiliary power supply.
The loads powered are the Training Center, Gatehouse, and Modular Building with the Training Center and Gatehouse each receiving at least 60%
additional electrical capacity.
Power will also be supplied to the Vapor Container and contractor trailers.
This change was necessary because powering these loads from the auxiliary power supply limited the capability of the Electrical System.
The shifting of the above loads will improve voltages and increase margin for safety-related loads powered from the station service buses.
The nature of the modifications do not directly relate to the operation of plant equipment important to safety and/or used for the mitigation of a postulated accident.
The powered equipment from the MEC is used only for services.
Since the MEC will be physically and electrically isolated from the plant electrical system, no direct or induced failures in one system will affect the other system.
o EDC 92-303 - Security Modifications - IEC l
This design change provided the evaluation and justification for instrument and control changes necessary to implement the Defueled Security Plan, and as such, is considered Safeguards Information.
B.
Plant Desian Chances (PDCs)
O PDC 90-006 - HPSI Seal Water Tubing Modification This change modified the seal water piping for the HPSI pump mechanical seals from 1/2 inch schedule 80 stainless steel piping to 1/2 inch x 0.035 inch stainless steel tubing and swaglock fittings.
The seal I
water coolers were also removed.
Due to the configuration of the HPSI pump seal water piping, the threaded fittings and flanges could not be 2
properly tightened and aligned.
As a result, these connections were leak paths for borated water.
Replacing the piping with tubing and swaglock fittings resulted in leak free connections, and provided for easier disassembly, reassembly, and handling of the mechanical seals.
To eliminate the threaded fittings, the seal water coolers were removed.
These coolers were supplied as original equipment by the pump manufacturer, but since there is no cooling water connected or required, they served no useful purpose.
O PDC 91-004 - Auxiliary Boiler Modification This change replaced the Auxiliary Boiler fire control system and burner units, improving the efficiency and firing rate.
The original control system is no longer manufactured and the ability to rebuild it no longer exists.
Additionally, the fuel oil supply system has been modified as required by the new burner units.
The dual controls on the Auxiliary Boilers have also been removed.
This dual control function was installed for the LOCA event that would cause high radiation levels in the Auxiliary Boiler room.
In the present plant defueled configuration, this dual control function is no longer needed.
C.
Non-Nuclear Safety Chances (NNS Chances) o NNB Change 92-002 - Security Modifications-Mechanical This design change provided the evaluation and justification for mechanical modifications necessary to implement the Defueled Security Plan, and as such, is considered Safeguards Information.
o NNS Change 92-005 - Modifications to the Evaporator Level Transmitter This change installed a condensate pot downstream of isolation valve WD-V-893 in the reference leg of the evaporator level transmitter, WD-LT-302A, converting this leg to a liquid filled leg.
Changing to a wet leg precludes erroneous level indications caused by moisture in the dry reference leg due to condensation of steam from the evaporator.
The change improves the level control for the evaporator and lessens the possibility of overflowing the evaporator or running it dry.
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D.
Temporary Desian Chances o
TCRs #92-17 & 92 EDG Starter Contactor Replacement The starter contactor on each of the plant's Emergency Diesel Generators (EDG) operates on a 125 V de control signal through the contacts of Relay R2.
Investigations into excessive arcing of the R2 relay contacts revealed that both main contacts and coils in these contactors were rated for ac applications instead of dc. (See LER 91-005).
These TCRs removed the ac contactors and installed new ABB dc-rated contactors.
The replacement of the ac-rated contactors with dc-rated contactors does not adversely affect the function or operational characteristics of the EDGs or their controls.
Since the new contactor de coils will reduce the burden on the R2 relay contacts and increase the contactor contact made/ break capability, the installation of the new contactors enhances the performance of the EDG cranking circuits.
o TCR #92 Waste Disposal Evaporator This TCR allows processing of the Activity Dilution Decay (ADD) and Waste Holdup (WH) tanks in a manner consistent with Class II waste processing (directly to i
the evaporator), and eliminates an unnecessary transfer to the gravity drain tank in the process.
This is acceptable because all liquid waste in the ADD and WH tanks are now Class II wastes since the waste gas system has been purged and is vented to the atmosphere.
There is negligible hydrogen in these systems.
o TCR #92-237 - Blank Flanges for Safe Shutdown System Lay-up This TCR provided the means to blank-off the Safe Shutdown System piping, which is no longer required due to the plant shutdown and is in lay-up, in order to maintain the pressure boundary of the Fire Water Storage Tank, TK-55.
The integrity of the tank was l
maintained by removing the flexible metal hoses and l
installing blank flanges on the nozzle flanges.
This change eliminates the potential for a break or failure in the flexible metal hoses or underground piping upstream of the isolation valves.
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E. Other Chances j
o Component, Equipment and System Classification i
The YNPS has transitioned to a permanently defueled condition.
Safety classification changes reflect the change in plant operating status to a defueled I
condition in which the majority of the original expected plant accidents and transient conditions are no longer applicable.
Without fuel in the reactor vessel the probabilities of most of the design basis accidents are reduced to zero.
Two accident analyses t
previously analyzed within the SAR remain credible during the defueled condition:
a sudden failure of the waste gas system and a fuel handling accident.
The waste gas system has been purged and vented and therefore, a release of its entire inventory would not result in an accident which would release radioactive constituents.
Therefore, only a fuel handling accident remains possible.
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Spent Fuel Pit is not increased.
Safety classification l
changes made with respect to SFP operation do not l
l affect the Technical Specification limiting conditions for operation or surveillances related to the SFP
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operation.
Furthermore, the consequences of a fuel handling accident have been reevaluated.
The results j
of this design basis evaluation indicate that the t
radiological consequences are greater than five orders of magnitude below the exposure guidelines of 10CFR100 and are well below the EPA Protective Action Guidelines (PAGs).
The construction of YNPS was pre-ANSI N18.2 (1973),
i ANSI N18.2a, ANSI /ANS-51.1 (1983), and ASME Section III.
The classification of equipment at YNPS is based upon compliance with both the overall intent of ANSI N18.2 (1973) and the fact that the plant is shutdown and in a permanently defueled condition.
i At YNPS, systems, components, and structures have been classified, with respect to functional requirements.
l The full spectrum of normal and off-normal conditions i
are identified in accordance with their anticipated frequency of occurrence and consequences.
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Based on the current plant defueled operating i
condition, there are four conditions or transients which are applicable and have been evaluated in j
accordance with their anticipated frequency of occurrence and consequences in determining appropriate i
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system classification requirements:
5 Loss of Spent Fuel Pit Inventory Complete Loss of All Station AC Power Fuel Handling Accident Loss of Spent Fuel Pit Cooling l
A review of YNPS system functional requirements during I
these four conditions has been performed to address the required safety classification of equipment and components.
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Technical Specification Applicability for a Permanently l
Defueled condition j
Many of the YNPS Technical Specifications are not applicable if the reactor vessel is permanently defueled.
The evaluation documents a clarification of the YNPS Technical Specifications to identify the specifications that remain applicable to a permanently i
defueled condition.
Six operational modes corresponding to operation when fuel is stored in the reactor vessel are defined in the l
Technical Specifications.
These modes are referenced in the applicability statements of most technical requirements.
When there is no fuel in the reactor vessel, the six operational modes defined do not apply.
All technical specifications referencing these modes are considered not to be applicable with the exception of TS 3/4.4.9, Structural Integrity, which shall remain applicable at all times in the permanently defueled condition.
O Fire Brigade Membership The fire brigade has been reduced from five members to three.
This reduction is possible since the plant has been permanently shutdown, fuel has been transferred to the spent fuel pit (SFP), systems required for safe shutdown of.the plant are no longer required to be operable, and the defueled safety analysis demonstrates that 10CFR Part 100 limits cannot be exceeded and exposures are well below the EPA Protective Action i
Guidelines (PAGs).
The three member brigade will provide a level of fire protection in the defueled mode commensurate with the level provided while the plant l
was operating.
The Fire Protection Plan is based on defense in depth.
A fire prevention program will be maintained.
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detection, automatic suppression and manual fire fighting capability (both on site and off site) will assure that fires are promptly detected and controlled.
Operability of required fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires.
This capability is required in order to detect and locate fires in their early stages.
Operability of the fire suppression systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in protected facilities.
These features of the defense in depth program remain in effect in the defueled mode.
When the plant is in a defueled mode (1) systems required to achieve and maintain safe shutdown are not required to be operable and the requirements to provide separation of redundant systems from the effects of fire are not applicable, and (2) applicable accident analyses are not affected by any type of fire.
Ignition sources have becn reduced commensurate with the reduction of plant maintenance and modification activities.
A significant amount of equipment has been de-energized or no longer operates at elevated temperatures.
Transient combustible loading has been greatly reduced resulting in a commensurate reduction in the magnitude of possible fires.
In-situ fire loading (mestly electrical cables) does not support the rapid development or spread of fires.
Nearly all of the in-situ fire loading is protected by automatic fire suppression systems.
o Fire Protection Technical Requirements The safety evaluation provides the basis and justification for replacing the existing fire protection Technical Specifications with a new manual comprising the " Fire Protection Technical Requirements."
These requirements incorporated the original fire protection Technical Specifications, National Fire Protection Association and American Nuclear Insurers recommendations that the plant adopted.
The requirements maintain all of the original Technical Specification systems.
The requirements were changed to indicate commitments that were more conservative than the Technical Specifications.
o vital Buses No. 1 and 2 Permanent Bypass to Normal AC Power The Technical Specification requirements associated 7
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to the current plant status.
Following a loss of ac power, both vital buses.will be repowered from the EDGs in a timely manner.
Moreover, a detailed review of the operational requirements for the loads powered from the vital buses determined that none of the loads would be needed immediately following a loss of ac power.
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o ISI/IST Requirements for the Plant Predictive Maintenance Program i
Submission and implementation of the new ten year ISI program and continuation of the YAEC IST program is no i
longer required due to the POL status of the Plant.
i However, to ensure reliability of operation, certain i
ISI/IST related testing is still deemed appropriate.
This testing will be performed under the Plant
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Predictive Maintenance Program and will be controlled-by new and existing plant procedures.
Inservice l
3 inspection and testing requirements for the Spent Fuel 1
Pit Cooling System, portions of the Shutdown Cooling System, the Service Water System, and the Component j-Cooling System have been defined to ensure system l
integrity and reliability.
O Removal of Potassium Chromate and Sodium Hydroxide from the Component Cooling Water System These chemicals were listed in the FSAR as corrosion inhibitors for the Component Cooling System.
Removal l
of potassium chromate is desirable because this chemical is listed as a hazardous material, and
-j therefore eliminates the potential for leakage to the environment or the generation of mixed waste.
The removal of the corrosion inhibitor has no impact on the component cooling system in the defueled condition.
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I has no impact on the fuel handling accident which is i
the only remaining credible accident.
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j The original Appendix R requirements called for the 1
illumination of all areas needed for the operation of safe shutdown equipment and illumination of the interior access and egress paths.
Since the plant has l
been permanently shutdown, systems required for safa 4
shutdown are no longer required to be operable.
Since the plant has received a POL, the site is much more an industrial facility and can be considered a facility to which NFPA 101' applies.
With the POL, the emergency lighting system is required to provide lighting for operation, safe access and safe egress in the event of
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i a fire.
Therefore, the safe shutdown emergency lighting requirements of Appendix R no longer apply to YNPS.
Consistent with the POL, the emergency lighting requirements for the lighting credited under Appendix R have been reduced from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> in accordance with NFPA 101, " Code for Safety to Life from Fire in Buildings and Structures," 1991 Edition.
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Leakage Reduction Program The systems formerly addressed by the leakage reduction program do not include the Spent Fuel Pit Cooling or Ventilation systems.
The fuel in the Spent Fuel Pit is t
incapable of contaminating the fluids in those systems i
to the point where they would be characterized as highly radioactive.
Following the permanent shutdown t
and defueling of the YNPS reactor, and with the current accident analysis, there is no reason to include them now.
Because the systems formerly addressed in the program j
no longer have the reactor as a source of highly i
radioactive fluids, and because there is no new source of highly radioactive fluids from the Spent Fuel Pit, the need for an active, proceduralized program has ceased to exist at YNPS.
The intent of the program, protecting the public from a Part 100 release, has been permanently achieved via the shutdown and defueling of YNPS.
o Control Rod Removal from the Spent Fuel Pit The acceptability of moving the cask liners, with the HRP inner liners inserted, and the crusher / cutter south of the cask hatch opening (approximately 15 inches) was evaluated.
The shipping cask liners and the crusher / cutter will be prevented from travelling over fuel assemblies by administrative control and by physical restraints.
An accidental drop of either one
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of these two components will not impact spent fuel from above.
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