B12822, Annual Rept for 1987
| ML20147B321 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Haddam Neck, 05000000 |
| Issue date: | 12/31/1987 |
| From: | Mroczka E CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES |
| To: | |
| References | |
| B12822, NUDOCS 8803020065 | |
| Download: ML20147B321 (494) | |
Text
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General Offices
- Selden Street, Berlin, Connecticut 9 USU.1%l~.~
P.o. Box 270 g ' J
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HARTFORD. CONNECTICUT 06141-0270 (203) ees-sooo February 23, 1988 Docket Nos. 50-213 50-245 50-336 Nf3 B12822 Re: 10CFR50.59 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Vashington, D.C. 20555 Gentlemen:
Haddam Neck Plant Hillstone Nuclear Power Station, Units Nos. 1, 2 and 3 Annual Report Pursuant to the Provisions of 10CFR50.59; Sections 6.9.1.c and 6.9.1.e of Appendix A to DPR-61; Sections 6.9.1.4 and 6.9.1.5 of Appendix A to DPR-21 and DPR-65; and Section 6.9.1.2 of Appendix A to NPF-49, this report is submitted covering operations for the period January 1, 1987 to December 31, 1987. Additionally, this report contains a summary of the challenges to re dated June 10,1980.gf/safetyvalvesascommittedtoinaletter (1) V. G. Counsil letter to D. G. Eisenhut, dated June 10, 1980. l l
Very truly yours, CONNECTICUT YANKEE AT0 HIC POVER COMPANY NORTHEAST NUCLEAR ENERGY COMPANY 8803020065 871231 PDR ADOCK 05000213 l R DCD .
[ &
E. Mjftroczka /
Seni6r Vice President Enclosure ,
cci V. T. Russell, Region I Administrator A. B. Vang, NRC Project Manager, Haddam Neck Plant i H. L. Boyle, NRC Project Manger, Hillstone Unit No. 1 D. H. Jaf fe, NRC Project Manager, Hillstone Unit No. 2 R. L. Ferguson, NRC Project Hanger, Hillstone Unit No. 3 n,u J. T. Shedlosky, Resident Inspector, Haddam Neck Plant Nj l idl V. J. Raymond, Resident Inspector, Hillstone Unit Nos. 1, 2, and 3 d
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CONNECTICUT YANKEE Contents Section Eage Changes Design Changes...............................................................2 Procedure Ch an ge s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 6 Jumper Device Chan ges . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 84 S e tpoi n t Ch an ge s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 90 Tests..................................................................................93 Ex pe rime n ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4 Occupational Radiation Expos ure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 95 Challen ges to Relie f Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 96 J
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t Plant Desien Chance Index PDCR Number Iille 473 Steam Generator Primary Manway Diaphragm Fasteners 544 Tank Heating System Modifications 547 Tank Heating System Modifications - Foundations 576 Emergency Diesel Monitoring 57'8 New Spare Generator Step-Up Transformer 640 Enlarge Service Water Pump Access at Elevation Eight Feet 642 Boric Acid Heat Trace Modification 687 WaterTreatment Piping Modifications 694 Microwave System 697 Hydrazine Storage and Delivery System 699 Replacement of MS-TV-1212 and MS-TV-1213 708 Pressurizer Relief Tank MPX Point 713 Process Computer Replacement - Room Construction 722 Enhancement to Microwave Fields 726 Control Room Habitability - Central Alarm Station Cooler 730 Turbine Deck Mezzanine Office 750 Maintenance Shop Office Module 793 Radwaste Reduction Facility 804 Bracing of Main Control Board Auxiliary Relay Panels 817 Component Cooling Pump Mechanical Seals 818 Diesel Generator Service Water Pilot Solenoid Valve Modification 838 Backup Meteorological Monitoring System 844 Switchgear Room Halon Fire Suppression System Modification 848 Installation of Computer Room Hardware 850 Re-evaluation of Safety-Related Piping 855 Administration Building Addition - Power and Security 862 Service Water Isolation Valves for SW-MOV-5 and SW-MOV-6 863 Lowering of the North Truck Bay and West Sorting Room Radiation Shield Walls in the Radwaste Reduction Facility 2
PDCR Number lg T
864 North Boundary Security Fence for the New Switchgear Building 866 New 4160 Volt Bus 9 Circuit Breaker Cubicle 867 Replacement of Station Service Transformers 484,485,496, and 497 870 Reactor Coolant System Component Support Modifications 875 Installation of Outside Stem and Yoke Isolation Valves for Individual Sprinkler and Hose Supply Riser at Maintenance Shop Manifold 877 Screenwell Building Security Enhancement 881 Additions to Plant Paging System Inside Containment 882 Moisture Separator Reheater Inlet Steam Flow Deflector Plates 886 law Pressure Turbine Replacement 887 Low Temperature Overpmssure Relief Protection System 889 Ixw Pressure Turbine Replacement / Instrument ar.d Control Portion 890 Motor Operated Valve Overload Protection Upgrade for MOVs 23,34,331,160, and 344 891 Rod Cluster Control Assembly Changeout Tool Modification 898 Steam Generator Plugging 899 Turbine Building Jib Crane 913 Containment Manipulator Crane Modifications 915 Installation of Turbine Lube Oil Centrifugal Purifier Connections l
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Plant Desien Chance Number 473 Plant Design Change Number 473, entitled "Steam Generator Primary Manway Diaphragm Fasteners" is complete.
Description of Chance The steam generator primary manway machine screws for diaphragm retention were replaced with 1/4 turn fasteners.
Reason for Chance This change reduces the amount of radiation exposure aceived during diaphragm installation and removal. The time required for installation and removal of diaphragms was also reduced.
Safety Evaluation The 1/4 turn fasteners provide an improved means of holding the steam generator primary manway diaphragm in place while the manway cover is being installed. They are not pressure boundary components.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of ,
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Plant Desien Chance Number 544 Plant Design Change Number 544, entitled "Tank Heating System Modifications" is complete.
Descriotion of Chance Thermosyphon heaters were replaced on the refueling water storage tank, demineralized water storage tank, primary water storage tank and the boron waste storage tanks with Briskheat electric heating blankets on the outside surface of these tanks. The tanks were insulated with one-inch thick isocyranurate insulation panels on the sides and a built-up fiberglass tar roof.
Reason for Chance These changes were made due to the following:
Addressed concerns regarding contamination of house heating systems from radioactive fluids heated by that steam.
Improved plant reliability by eliminating thermosyphon heaters.
Decreased exposure by eliminating need for maintenance on thermosyphon heaters.
Safety Evaluation All piping modifications associated with safety tanks wem done in accordance with appropriate code requirements.
A safety grade electrical supply for the heaters was determined not to be required since a loss of power has minimal impact on the temperature. This change does not addmss transformer and control panel pad requirements. All other structural, mechanical and electrical modifications wem addressed.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Desien Chance Number 547 Plant Design Change Number 547, entitled "Tank Heating System Modifications - l'oundation" is complete.
Descriotion of Chance Concrete support pads, control panel supports and underground conduit were added for electric heat blanket installation on Plant Design Change Number 544. For the refueling water storage tank (RWST), demineralized water storage tank (DWST), boron waste storage tanks (BWST), and primary water storage tank (PWST) power transformer and control panel mounting, all panels and transformers were Hilti mounted to floors using vendor supplied mounting brackets / mounting kits.
Reason for Chance Concrete support pad, control panel supports, and underground conduit mns were installed for the tank heating system modifications project.
Safety Evaluation This change was reviewed with respect to the structural adequacy of the existing facilities. The mounting of the RWST, DWST, and BWST (A and B) transformer and control panel to the existing floors in the auxiliary feedwater building and the primary auxiliary building did not affect the capacity of these floors since no rebar was cut without review and approval from engineering. The additional normal and seismic loading of those items did not exceed the originally designed floor capacity.
- The PWST transformer and control panel foundation and underground conduit in the yard crane area did not affect the structuralintegrity of any plant stmetures.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Desien Chance Number 576 Plant Design Change Number 576, entitled "Emergency Diesel Monitoring" is complete.
Descriotion of Chanze The following changes were made by this modification:
A temperature gauge was installed between the jacket water themiostat and the jacket water inlet to the lube oil cooler.
A temperature gauge was installed on the lube oil discharge from the engine.
Crankcase explosion proof covers were installed on each engine.
A cylinder pyrometer / thermocouple system was installed (21 cimuits on each emergency diesel engine).
The temperature monitoring system is a pyrometer / thermocouple arrangement which provides the operator with the ability to read (OF) individual cylinder temperatures. His information allows one to determine the operating conditions of the engine and enable a more indepth method for trouble shooting the engine. He pyrometer was mounted in the left wing panel of the engine control panel, the thermocouple elements wem installed in the receptacles located on each exhaust elbow.
The crankcase explosion proof covers increased the operational safety factor for the emergency diesel engines. In the event the diesel engine crankcase becomes excessively aressurized, the explosion proof covers willlift thereby releasing the pressure and significantly ret ucing the chances of a crankcase explosion. The crankcase explosion proof covers were a one for one replacement with the current type crankcase cover. Dere were four explosion proof covers installed per engine.
Reason for Change This change improved engine performance monitoring and increased the operational safety aspects of the emergency diesels.
Safety Evaluation i The Emergency Diesel Engines are safety-related components, therefore, an engineering seismic evaluation was conducted. It was determined that the temperature monitoring system and the installation of explosion proof crankcase covers did not have any effects on, or reduce the design '
function of the Emergency Diesels, !
Derefore, this modification does not constitute an unreviewed safety question per the criteria of l 10CFR50.59. l i
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Plant Desien Chance Number 578 Plant Design Change Number 578, entitled "New Spare Generator Step-Up (GSU) Transformer" is complete.
Descriotion of Chance This change provided a circuit to test energize the spare GSU transformer. This circuit consists of a new indoor 4.16 kV air magnetic circuit breaker,5 kV power cable, outdoor disconnect switch,4.16-18.3 kV step-up transformer,23 kV oil circuit breaker, interconnecting bus and fittings, and all necessary controls, indications, protective relaying and alarms.
Reason for Chance This change allows for periodic test energization of the spare GSU to verify reliability. Test energization is the best means for proving the GSU is fit for service. The requirement to test energize is further strengthened by the need to prove the unwanted core ground is inconsequential to safe, reliable operation of the spare GSU.
Safety Evaluation The GSU transfonuer and the majority of the test energizing circuit is located in the yard area adjacent to the west wall of the turbine building. Since it is outside and not near any safety related equipment,
- does not affect plant safety. ,
The control panel for the GSU test circuit is located in the southwest ponion of the ground floor of the turbine building. Conduit for the test energizing circuits was not routed near safety-related equipment and the control panel location is not in the area of any safety-related equipment.
Power to energize the spare GSU comes from a nonsafety-related 4160 volt bus. It does not affect emergency bus load ~ng. Power for the spare GSU transformer fans and pumps comes from 480V bus 7 (safety related). The feeder breaker will be tripped on a loss of normal power. This will trip the spare transformer fan and pump loads so that they will not automatically be picked up by the emergency diesel generator.
The DC power requirements are very low except when actually operating breakers. The load added has no affect on the time the station battery remains available dunng a complete loss of AC power.
Isolation between the nonsafety related DC circuits and the safety-related DC power system is provided by the distribution panel circuit breaker. Circuit overcurrent protection in the nonsafety-related circuits is provided by fuses.
Inside the plant, fire protection was adequately addressed by specifying proper overcurrent protection, the use of flame retardant cable, and by requiring that penetrations be properly rescaled. Outside in the yard area the design includes provisions for the collection of oil. If a violent fault occurred which ruptured the wall of a transformer or oil circuit breaker, the collection system would limit the amount of oil available for combustion. There are no nuclear safety related fire protection concerns.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Desien Chance Number 640 Plant Design Change Number 640, entitled "Enlarge Service Water Pump Access at Elevation Eight Feet" is complete.
Descrintion of Chance The floor sleeves for service water pumps P-37-1 A, P-37-1B, P-37-1C, and P 37-1D in the screenhouse were replaced in order to accept the new larger service water pump bowl.
The modification enlarged the service water pump floor sleeves from 24" to 26" diameter. The previous 24" STD wall pipe sleeves were drilled out and the opening enlarged to 26". A 26" XS pipe was used to replace the old floor sleeves. The 26" XS pipe was used for reinforcing bar anchorage.
The anchorage details eliminate any leak paths between the pipe sleeve and the concrete.
In addition to the new floor sleeves, a new well seal was required. The well seal provides a water j barrier between the floor sleeve and the pump.
Reason for Chance The outside diameter of the new service water pump bowl is 9/16"larger than the old pump bowls.
l The new bowls did not fit through the previous sleeve in the floor.
Safety Evaluation i This modification has been reviewed per the applicable codes, standards and Technical Specifications I
and found acceptable. The modifications to the screenwell structure were made ensuring that the structure was maintained with no changes to the allowable loadings.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of I 10CFR50.59.
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Plant Desien Chance Number 642 l 1
Plant Design Change Number 642, entitled "Boric Acid Heat Trace Modification"is complete. l 1
Descriotion of Chance This change installed ammeters on the primary heat trace circuits. ;
I Reason for Chance !
This change was made for the purpose of monitoring the current drawn from each heat trace circuit.
Safety Evaluation l This modification did not adversely affect any safety-related equipment or circuits and enhanced operator ability to control and monitor undesirable states.
l Therefore, this modification does not constitute an unreviewed safety question per the criteria of j 10CFR50.59.
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1 Phnt Desien Chance Number 687 Plant Design Change Number 687, entitled "Water Treatment Piping Modifications" is complete.
Descriotion of Chance This change. included the installation of piping to and froin the east wall of the maintenance shop for use with a demineralizer, cross connects allowing the ponable demineralizer to run in series with the mixed beds and a cross connect allowing either mixed beds 1 A or 1B to run in series with mixed bed IC.
Additional piping, valves, and fittings were also installed to connect the deoxygenation process skid to the portable demineralizer supply line.
Reason for Chance These changes provide an improved method of removing dissolved oxygen in make-up water. ,
1 Safety Evaluanon This modification did not involve changes to nor affect any safety-related systems, components, or structures. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Chance Number 694 Plant Design Change Number 694, entitled "Microwave System" is complete.
Descriotion of Chance A complete microwave communication system was installed allowing Connecticut Yankee to communicate throughout the Nonheast Utilities Microwave Communications Network.
Reason for Change The past failure of the cable that crosses the Connecticut River left Connecticut Yankee with only one form of direct telephone communication. This change allowed a second contingency in the event of a landline failure.
Safety Evaluation A review was performed for the above installation, which includes a dish antenna on the turbine building and a reflector on the containroent.
The report by URSS. A. Blume and Associates, "Scoping Study for a Limited Structural Evaluation of Plant Structures to Resist Tornado Loading," was used to evaluate possible missile generation by the microwave dish. His repec uses a utility pole as the limiting missile.
Since the dish would have less energy and is a less compact object than the utility pole, it introduces no new missile risk. The reflector weighing less than the disk was not considered.
The utility pole remains the ;;over ng .nissile for the site study. De new installation did not create a i condition that aheady does not exit nom the tornada missile.
Movement of the ref:ector or its deflection during leak rate test or temperature change affecting the I mounting does not appear critical. Proper installation procedures of Hilti bolts were adhered to as not to damage containment integrity and reinforcement which must be maintained.
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. Therefore, this modification does not constitute an unreviewed safety question per the criteria of
! 10CFR50.59.
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Plant Desien Chance Number 697 l Plant Design Change Number 697, entitled "Hydrazine Storage and Delivery System" is complete.
Descriotion of Chance The subject modification involved installation of a closed semi-bulk hydrazine storage and feed system with a nitrogen blanket. This minimizes operator exposure to harmful hydrazine vapors and provides for an efficient means to transfer hydrazine to the hydrazine addition tank.
Reason for Chance t
The fiammability, toxicity, and reactivity characteristics of hydrazine are such that a closed storage and feed system was necessary to minimize operator exposure.
Safety Evaluation i
This modification did not involve changes to nor did it affect any safety-related systems, components i or structures. Herefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. .
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Plant Design Chance Number 699 Plant Design Change Number 699, entitled "Replacement of MS-TV-1212 and MS-TV-1213" is complete. -
Descriotion of Change
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This change replaced MS-TV-1212 and MS-TV-1213 (main steam drain trip valves) with stainless steel valves. Previously, MS-TV-1212 was made of chromemoly and MS-TV-1213 was made of carbon steel.
Reason for Change Vapor, formed by the flashing process in the main steam system, is accelerated by the expansion 3
process to extremely high velocities. The impingement ofliquid droplets entrained in this vapor caused damage to the carbon steel and chromemoly valves.
Safety Evaluation ne modification was evaluated for its possible effects on nfety related systems and components.
Since the valve material was changed to stainless steel, the possibility of a change in weight and thermal expansion characteristics was investigated. It was concluded that given the size of the valves, the seismic and thermal expansion effects on the system was negligible. De change to stainless steel improved valve reliability by reducing the vapor erosion that was taking place.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Desien Chance Number 708 Plant Design Change Number 708, entitled "Pressurizer Relief Tank MPX Point" is complete.
Descriotion of Chance His change added the ability to trend pressurizer relief tank level with the plant process computer. To achieve this objective, a precision resistor was installed in series with the instmmentation loop, and ;
process compater leads were attached to this resistor.
Reason for Chance l This change allows for pressurizer relief tank level trending when miief valves leak. This also prevents the need for temporary instrumentation.
Safety Evaluation 1
Realistic failure modes associated with this modification are the consequences of failure of the l precision resistor, and consequences of failure of the plant process computer. I Use of this precision resistor in this application has a very low failure rate and the two realistic failure mMes would lead to either loss of main control board indication (which is not a safety-related system) or loss of the computer trending ability (no direct consequences).
1 The plant process computer has no intrinsic effect on its sensing points (MPX) upon failure. Again, the worst case would be loss of main control board (MCB) indication. Loss of MCB indication of pressurizer relief tank level, although not desirable, is an acceptable failum mode. Dere exist three MCB indications for the pressurizer relief tank: level, Temperature, and Pressure. There also is qualified indication of pressurizer power operated relief valve position and safety valve actuation (Accustic Valve Monitor).
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Chance Number 713 Plant Design Change Number 713, entitled "Process Computer Replacement - Room Constmetion" is complete.
Descriotion of Chance This change included conversion of the previous computer room and Operations Supervisor's Office to one large, modern facility in order to house the larger computer system and meet rire protection 1 requirements.
He change includes the features listed below:
- 1. Raised floor (18")
- 2. Normal and back-up air conditioning / humidity control -
- 3. Automatic halon fire protection system
- 4. Specializedlighting
- 5. Underfloor conduit and tray system for separate power and computer input / output cabling:
specialized grounding system Reason for Chance The IBhi 1800 computer was replaced with a new hiODCOhiP computer capable of handling plant needs. This new computer is larger than the IBhi 1800 and a larger area was needed for installation.
Safety Evaluation ne condenser weighs 535 pounds and was mounted on the roof of the computer room. De roofis a reinforced concrete slab 22 inches thick with an allowable loading of 500 psf. De addition of the condenser to the roof did not cause the allowable roofloading to be exceeded.
The new raised floor is supponed by an existing concrete slab which has an allowable live load of 125 psf. De weight and layout of the computer equipment is such that the allowable floor loading was not exceeded.
The new wall and ceiling system for the computer system was designed in accordance with standard building codes. Dese additions can withstand all loads associated with normal operating conditions.
In the event of a seismic event, the walls and ceiling are isolated from the control room and thus would have no impact on the control room structure. Herefore, since no safety related equipment is affected in this extreme event, these additions are acceptable.
The penetrations in the nonh and east walls were investigated. De addition of these penetrations in the existing walls did not adversely affect the stmetural integrity of these systems.
A 7% concentration has been designed to exist in the room after discharge. His concentration is in the accepted safety range of code requirements. Derefore, effects of the gas on personnelin the room is not a concern. Since the concentration utilizes a specific amount of halon gas for the designed room volume, the concentration would not rise due to any inadvertent operation or malfunction of the system.
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A predscharge alarm (horn / light) will activate approximately 30 seconds before the discharge will occur. This allows ample time for the occupants of the room to evacuate.
Direct contact with the Halon vaporizing liguid as it discharges from the nozzles does have a chilling effect on objects. However, this rapid coohng is not a concern as the cooling effect is limited to the immediate area of the discharge nozzles because the liquid vaporizes rapidly when mixed with air.
Also, objects are not located in the immediate area of the nozzles because a clear area must exist for unobstructed discharge.
Therefore, this modification does not constitute an unn: viewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 722 Plant Design Change Number 722, entitled "Enhancement to Microwave Fields" is complete.
Description of Change Microwave fields were modified by adding another set of microwave cones. This change extends the height of the intruder detection zones.
Reason for Change These changes were made as a result of a protection appraisal by Connecticut Yankee and an NRC audit.
Safetv Evaluation Use of this microwave detection device can be found throughout the plant security system. Failure of the transmitter willlead to the receiver alamiing. Failure of the receiver will lead to an alarm signal being generated. There is no effect on any safety-related system.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Desien Chance Number 726 Plant Design Change Number 726, entitled "Control Room Habitability - Central Alarm Station (CAS) Cocler" is complete.
Descriotion of Nnce his change installed a 56,000 BTU,2,600 cubic foot per minute (CFM), dedicated self-contained air conditioning system in the CAS and reduced the previous 850 CFM ventilation air, supplied by the Control Room air conditioning system, to 100 CFM. ,
Reason for Chance i
Bis change provided for sufficient cooling to meet room heat loads imposed on the room from the outside environment, personnel, lighting, and a new security computer (not yet installed). This I 1
change also eliminated past room cooling problems and associated security computer failures caused by inadequate ventilation. l l
The cooling and dehumidifying loads on the new CAS air conditioning unit were minimized, while '
still providing enough fmsn air for personnel as required by building codes.
Safety Evaluation The system was sized based on heat load calculations. De reduced ventilation air flow (850 CFM !
reduced to 100 CFM) remains within the recommended flow range delineated in ' Basic Mechanical l Code (1981). This system was manufactured and installed in accordance with the appropriate codes I and standards applicable for air conditioning systems. I Complete failure of the CAS cooler would result in a loss of cooling for the security computer.
Temporary cooling could be attained by opening the door of the CAS. If longer term cooling is desired, the ventilation flow from the Control Room HVAC System could be increased through manual action. In the event that this temporary cooling method is not sufficient (resulting in the shutdown of the security computer), stauon procedure requires the posting of security guards at vital security points. Here will also be a redundant computer not located in CAS. Derefore, a temporary loss of the security computer would not compromise plant security.
A fmon line break in the CAS Cooling System would probably result in the entire freon charge being released into the CAS. However, the size of the unit is small enough that the freon would be
! sufficiently diluted so as not to pose a har.rd to personnel in either the CAS or the Con'.rol Room.
4 A break in the 3/8" cooling water supply line would not produce water leakage significant enough to present an operational problem to either the CAS cooler or the security computer.
The civil / structural portion of this modification involved the installstion of a 400 pound roof top cooling unit on the service building.
He slab, on which the cooling unit was mounted, is safety-related. It was analyzed for the dead weight and cooling unit weight with the SSE vertical seismic factors included and has been proven to be structurally adequate, 19 4
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'Re objective for the electrical ponion of this project is to supply the new CAS cooler with a reliable source of power. The design was done in accordance with applicable standards and satisfactorily considers the issues of material quality, flame retardance, and seismicity. Although the installation of this cooler was considered a nonsafety.related addition, cable and conduit were routed thmugh safety-
- related areas. The design properly addresses these concerns by including a seismic conduit installation in the design and certified flu retardant cabling for fire protection concerns. l Additionally, bus loading was reviewe.s A..sure that the power supply was adequate to supply the ,
additional HVAC load. An edditional review was made to ensure that the addition of the new HVAC load did not adversely affect the undervoltage r,- degahd voltage setpoints.
Therefore, this nulification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. l 4
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Plant Desien Chance Number 730 Plant Design Change Number 730, entitled "Turbine Deck Mezzanine Office" is complete.
Descriotion of Chance A two story modular office building was installed along the north wall of the turbine hall on the turbine deck. This change included the relocation of a fire hose station, removal of a turbine hall heater, removal of a modular building in the southeast corner of the turbine deck.
Reason for Chance This building was installed to provide suitable storage and office area for turbine maintenance crews dring unit overh3ul. The modular building in the southeast corner of the turbine deck was no longer need-d.
Safety Evaluation It was determined that the construction and load of the building did not cause any problems with floor loading. The installation of the building required that a unit space heater be renoved from service.
This heater served to warm the turbine hall m cold weather. It was not related to nor did it affect any plant safety system. Fire hose reel station #134 was also rerouted around the modular building.- It still protects the north end of the turbine hallin its new location. A failure in this line does not compromise any fire protection systems in any safety-related area. Lastly, e sprinkler line and sprinklers were routed from the turbine hall sprinkler loop to protect the new modular building in tiie event of a fitt. This branch line does not compromise any sprinkler system in any safety-related area should the pipe rupture.
This change also removed the sprinkler system for the modular building (southeast corner). These sprinklers and associated piping provided protection for this building only. The removal of this system had no affect on any safety-related system.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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I Plant Desien Channe Number 750 Plant Design Change Number 750, entitled "Maintenance Shop Office Module" is complete.
i Descriotion of Channe l
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A two-story modular office building was installed in the northwest area of the Maintenance shop. 1 Reason for Chance l l
Dis change was made to allow the Maintenance Department foremen better access to the work force -
l and better ability to monitor activities by placing them in the shop area.
Safety Evaluation l
The building itself has no safety function, nor is there any safety related equipment in the vicinity of ,
or connecting to this building. l Therefore, this modification does not constitute an unreviewed safety question per the criteria of -
10CFR50.59. I 1
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Plant Desien Chance Number 795 Plant Design Change Number 793, entitled "Radwaste Reduction Facility (RRF)" is complete.
Descriotion of Chance A facility was constructed under an earlier PDCR (733) to provide handling and storage 'of dry, low level radioactive waste.
This change was limited to electrical, instrumentation and controls, heating, ventilation and air conditioning (HVAC), and fire protection required in the facility.
Reason for Chance This change provided the electrical requirements for the RRF.
Safety Evaluation The RRF is heated utilizing electric unit heaters. Failure of any or all of the unit heaters will not adversely affect the safety of any other building systems, plant personnel, or the public. Here are no pited water systems in this building and then: fore failure of the unit heaters will not result in any frecze problems. In the event of failure, portable heaters can be used if building habitability is required.
Failure of the ventilation system for this building will have no effect on safety. During a total ventilation system failure, the overhead doors can be opened or portable fans brought in for temporary ventilation.
Building ventilation does not require IIEPA filters. This conclusion was based on the insignificant annunt of particulate activity which could be released from this facility compared to normal station ,
releases. This was confirmed by sampling performed, in which no activity v/as detected from the discharge of the compactor and dry cleaning units.
All floor drains are directed to a sump where any spillage would be collected, analyzed for contamination and, if warranted, pumped into drums for removal. All piping, fittings, and joints were leak tested prior to burial.
The installation involved obtaining power from a 480 volt load center, Bus 10 (not safety related) and adding cable, conduit, step down transfo mer, lighting panels, lighting, grounding, and radiation monitors.
Safety precautions for working near the 115KV energized lines including minimum distances were at.idressed. Area radiation monitors wem installed to warn personnel of potentially hazardous conditions. Emergency lighting was provided in the event of a loss of power. A plant paging system was added to adequatety cover the areas in the facility.
Seismic concems with ermored cable being run in the service building were addressed. This cable run was installed to meet two over one concerns, as they run near safety-related cables.
Derefore this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Desien Change Number 804
, Plant Design Change Number 804, entitled "Bracing of Main Control Board (MCB) Auxiliary Relay .
Panels"is complete.
Description of Chance his change was limited to improving the seismic capability of original plant design auxiliary relay mounting panels, with Westinghouse AF relays included, within the MCB.
The auxiliary relay mounting panels were modified by the addition of multiple angle brackets, from the bottom of each panel to the top inside of the MCB.
Reason for Chance Westinghouse AF relays went out of production in the early 1970s, are in short supply, and no longer available as spare parts.
This modification assures that the relay panels are adequately supponed for future relay replacements.
Safety Evaluation These modifications atfected only the structural integrity of the auxiliary relay panels, nese changes in no way influenced the function of any plant systems. Therefore, no new failure modes wem identified.
These brackets did not compromise the seismic integrity of the MCB as established in the SEP program, Topic III-6. Hey greatly improved the seismic integrity of the auxiliary relay panels and in no way altered the seismic qualification of the MCB. Further, these brackets assure that the auxiliary panels provide competent support for future relay additions and/or replacements.
herefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Desien Channe Number 817 Plant Design Change Number 817, entitled "Component Cooling Pump Mechanical Seals" is complete.
Descriotion of CFaugt
) His modification replaced the packing on the three component cooling pumps with mechanical seals.
l Reason for Chance The installation of mechanical seals eliminated the problem of packing leakage. This leakage consisted of chromated water, which is a hazardous material.
Safety Evaluation ne change was evaluated for wssible effects on safety-related systems and components. Since the
, component cooling system, alt.1ough not in itself a safety-related system, supplies safety-related .
l equipment, failure of the mechanical seals was evaluated for adverse effects on this equipment. All known failures resulted in leak rates ofless than 40 ml/ min. This small leak rate allows adequate time for operator response.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Desien Chance Number 818 Plant Design Change Number 818, entitled "Diesel Generator (DG) Service Water Pilot Salenoid Valve Modification" is complete.
Description of Chance Pilot solenoid valves SW SOV-129/130 control the operation of the Diesel Generator A and B Coolant Admission valves SW-FCV-129 and SW-FCV-130 respectively. His change replaced the pilot solenoid valves SW-SOV-129 and SW-SOV-130 with continuous duty, normally closed ASCO solenoid valves. Also, the logic was modified to keep the pilot solenoids normally energized, thus, admitting air to DG coolant admissions valves SW FCV-129/130 to maintain these valves closed while the,DGs are not running. Upon loss of power, the service water valves will fail safe in the open position.
Furthermore, raw water switches were replaced in both DG control panels with 2 position, maintained (OPEN, AUTO) switches. These switches control the operation mode for DG coolant admission valves SW FCV-129/130. De already existing "NOT AUTO" annunciator on the DG alarm panel will be triggered when the raw water switch is in "OPEN" position.
Reason for Chance Previously, on a los of offsite power followed by an interruption of power to the MCC5, the pilot solenoid valves SW SOV-129/130 de-energized, preventing DG coolant admission valves FCV-129/
130 from opening.
This would result in a possible loss of both diesel generators due to overheating. As a temporary remedy, the air supply lines to the air operated service water valves were disconnected, which allowed constant coolant flow to the DG lube oil heat exchangers. This condition was not desirable due to excessive operation oflube oil heaters because of unnecessary heat loss.
Safety Evaluation Credible failures modes associated with this change consist of:
o Loss of power to solenoids o Solenoid malfunction o Loss of air to solenoid valves o Cable open circuit o Cable short circuit All above credible failures will result in de-energizing the solenoid or exhausting air from "air to close" AOVs. These actions will, in turn, open coolant admission valves in a safe position.
This change did not have a potential impact on any design basis accidents.
The removal of the stop position on the raw water switch and modifying service water valves to fail open upon loss of power decreased the probability of operation without diesel generator cooling water.
De perfonnance of the safety system was improved since upon loss of MCC5, cooling water will be immediately admitted to the DG without the delay of repowering MCC5 to open the service water valves. The admission of service water prior to DG startup will have no adverse affects because IX3 lube oil temperatures are continuously contmiled by immersion heaters.
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f He failure modes associated with the thange did not represent a new unanalyzed accident. The locatior, and arrangement of Class IE equipment was unchanged so that the effects of common mode failures, such as those resulting from missiles, fire, flooding, earthquake, and extremes of environmental conditions, were not increased.
Derefore, this modification does not constitute an unreviewed safety question per the criteria of
- ! 10CFR50.59.
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I Plant Desien Chance Number 838 i Plant Design Change Number 838, entitled "Backup Meteorological Monitoring System" is complete. l Description of Chance This change installed a backup meteorological monitoring system (Doppler acoustic sounder) and equipment shelter adjacent to the station Emergency Operauons Facility (EOF). It also provided for electric service to the equipment shelter from a distribution panel in the EOF.
Reason for Chance -
'Ihis change was installed to provide measurements of wind speed, wind direction and atmospheric stability at severallevels that correspond to the heights of potential plant sources of airborne radioactive releases.
This system was required to provide availability of on-site meteorological measurements in the event of a loss of offsite power and/or failure of the onsite meteorological tower.
Safety Evaluation This design change did not involve a modification to or near a safety-related system or component.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. ;
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Plant Desien Chance Number 844 Plant Design Change Number 844, entitled "Switchgear Room Halon Fim Suppression System Modifications"is complete.
Descriotion of Chance This change covered the necessary engineering, design, procumment, quality assurance, and
, construction services required to upgrade the fire suppression capabilities of the Halon suppression system in the Switchgear Room.
Reason for Chance Based on a review of test data, diffemnces between the test results and code requirements for Halon concentrstion were noted. Modifications were, therefore, tequired to satisfy code requirements and ' j fulfill NRC commitments. ,
i Safety Evaluation It was detemiined that the Service Building structural concrete and steel stresses remain within the allowable criteria when subjected to the loadings imposed by the new halon system tank modifications, i
Guidelines for rigging of the tanks in the Switchgear Room were provided to assure the safety of l Category 1 equipment. !
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, ne loading placed on the Service Building strocture due to the following was also addressed: l l
- 1. Relocation of one and installation of three new fire detectors. l
- 2. Removal of nine and installation of twenty-six halon spray nozzles and associated piping.
The loads associated with the above modifications, including scismic loads, were investigated. It was concluded that all concrete and steel stresses remained within the allowable code criteria.
The overall operation and performance of the existing Halon system remained unchanged even though significtnt work was required inside the Switchgear Room.
Herefore, .his nudification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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1 Plant Desien Chance Number 848 Plant Design Change Number 848, entitled "Installation of Computer Room Hardware" is complete.
Description of Chance His change installed, energized, and tested the integrated computer system (ICS) hantware in the computer room including the following:
o Placement of hardwam in the computer room.
o Routing and connection of cables o Temporary hookup of peripheral equipment in the computer room.
o Site acceptance test Reason for Chance This change completed the installation of hardware in the computer room and prepared the ICS to interface with plant systems.
Safety Evaluation The credible failure modes associated with this change involve the power source for the computer systum. The existing plant computer and the new ICS are powered from the same unintermptable power supply (UPS). The following failures are credible at the UPS interface: open circuit on power input, short circuit on power input, and short circuit to ground on power input These are the same failures which might have occurred with the old computer, therefore the change had no effect on this failure mode.
There were no safety systems affected since the ICS was not connected to any plant systems under this change.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59, 1
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Plant Desien Chance Number 850 ,
Plant Design Change Number 850, entitled "Revaluation of Safety Related Piping" is complete.
Description of Chance For the following systems, an upgrade was made of the piping support system by modifying supports or adding new suppons to meet the increased seismic loads u, nposed:
- 1) Service water to the containment air recirculation (CAR) fans.
- 2) Service water to the diesel generators.
- 3) Service water to the screen wash booster pump.
- 4) Residual heat removal in the containment.
- 5) Diesel generator staning air.
- 6) Diesel generator exhaust.
Reason for Chance NRC SEP Topic III 6 required that safety-related systems be cvaluated and modified, if necess
- to ,
ensure integrity and continued operation when subjected to a postulated Safe Shutdown Earthqu e (SSE).
Safety Evaluation All structural steel support members were evaluated ud designed in accordance with AISC Design Requirements (8th Edition) while the anchor bolt acA baseplate designs were performed in accordance with the Baseplate and Anchor Bolt Analysis Criteria. This ensured that all the design guidelines stipulated in I&E Bulletin 79-02 were met for each support design issued. These supports have relatively simple designs which minimized installation time and man-rem exposure. All su? pons were fabricated using carbon steel and/or stainless steel material which is consistent with origina l design requirements.
Integral attachments match the piping material eliminating bi-metallic welds on the pr ssure boundary thereby minimizing the potential for cracking of the weld. The associated systems were out of service to permit welding of the attachments without compromising the operability of the system. I local pipe wall stresses induced as a result of deadweight, themul and seismic loadings on the pipe lugs were evaluated. Lug dimensions and weld sizes were restricted to account for the effects of I potential thermal gradients between the lug and pipe wall surface.
To mainain allowable stress levels in the piping system during construction activities, installation l instructions on removal and installation sequence along with requirements for temporary suppons were provided. Support loads induced on foundation structures were reviewed.
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Based on the above, it is clear that these supports do not jeopardize the intended functions of the system design, but enhance the structuralintegrity and reliability of the systent It should be noted that these supports are not subject to gross failun: or collapse based on inherent design conservatisms combined with conservative analysis assumptions.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of ,
10CFR50.59. i r
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l Plant Desien Chance Number 855 Plant Design Change Number 855, entitled "Administration Building Addition - Power and Security" is complete.
Description of Chance A new Nuclear Records vault was added on to the Administration Building, which is located outside the Protected Area. His change was to controlinterfaces between the new addition and existing plant equipment. Specifically, this included:
- 1) A power supply for the building from htCC-4 (Position 2MR); including a 75 KVA step down transformer, ending in a distribution panelin the Administration Building.
- 2) Relocation of a security light and security barriers.
Reason for Chance This change was necessary because the old xcords vault did not have the capacity to meet the needs of Connecticut Yankee.
Safety Evaluation The change did not have any direct intedace with nor is it located in close proximity to any safety-related equipment.
There are two postulated failum modes for the new addition whien could have some safety impact.
The first failure mode would be a short in the new power wiring. This shon however would effectively be isolated by the new 480V circuit breaker in MCC-4 (not safety related) before it could have an effect on any safety-related portions of the electrical system.
De second failure mode would be failure of either the building addition or any of the new conduit or transformer suppons during a seicmic event. The failure of the building addinon was not a concern since its size is such that it cannot impact the Turbine Building during a seismic event. Failure of the equipment in the Turbine Hall was also not a concem since there is no safety-related equipurnt in close proximity to the new equipment and the new equipment is too small to damage the Turbine Hill structure significantly.
Herefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Desien Channe Number 862 Plant Design Change Number 862, entitled "Service Water Isolation Valves for SW-MOV-5 and SW-MOV 6"iscomplete.
Descrintion of Channe f This change installed manual butterfly isolation valves in the service water (SW) supply line to the residual heat removal (RHR) heat exchangers. Dese valves were installed downstream of SW-MOV- ;
5 and SW MOV-6. Flush connections were also installed between the MOVs and the new :
downstream isolation valves.
Reason for Channe Dese valves were installed to minimize silt transfer into the component cooling water system when r
SW-MOV-5 and SW MOV-6 are tested.
Safety Evaluation i Installation of this modification was performed within the bounds of the limiting conditions for -
operation as defined for the refueling mode in the Connecticut Yankee Technical Specifications.
j During normal operation, SW-MOV-5 and 6 are maintained in the closed position. De downstream _
isolation valves will be locked open. Maintaining the butterfly valves locked open ensures that service water flow to the RHR heat exchangers will not be challenged during an accident scenario. When ,
SW-MOV-5 and 6 are stroke tested during refueling outages, the butterfly valve will be closed. This i method for testing the MOVs will not be altered in that they will be tested indisidually when only one train is required. ;
ne additional flow resistance caused by the butterfly valves was evaluated. It was determined from !
calculations that the change in service water flow due to the butterfly valves is negligible and does not compromise RHR heat removal capacity.
The new butterfly valves, flush connections, and associated pipe fittings reruit in mMed weight to the I service water system. Additional pipe supports were installed. De pipe supports ensure that the ,
piping system will not be overstressed due to the additional loading. !
Therefore, this modification does not constitute an unreviewed safety question per the criteria of :
10CFR50.59. l I
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Plant Desien Change Number 863 Plant Design Change Number 863, entitled "Lowering of the North Truck Bay and West Sorting Room Radiation Shield Walls in the Radwaste Reduction Facility" is complete.
Descriotion of Chanze This change involved lowering the north truck bay and west sorting room radiation shielding block walls in the Radwaste Reduction Facility (RRF) from 14'0" to 8'0".
Reason for Change ne lowering in height of these two walls was due to industrial safety concerns.
Safety Evaluation This change was accomplished by saw cutting the block with a masonry saw and chipping the block to an 8'0" height. During removal of the block, care was taken to ensure that no damage was incurred to the wall below the 8'0" height. The modified walls are not safe y-related block walls.
The lowering of the high density radiation shielding block walls for the above two areas increased the structural analysis margin of safety since floor slab and foundation loadings were decreased due to the removal of block. Here is no safety-related equipment within the RRF building structure, nerefore, this modification does not constitute an umeviewed safety question per the criteria of 10CFR50.59.
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Plant Desien Chance Number 864 Plant Design Change Number 8M, entitled "North Boundary Security Fence for the New Switchgear Building"is complete.
Descrintion of Chance A security fence and steel support structure were constructed along the primary auxiliary building, waste disposal building, and emergen,y diesel generator building, which tied mto the existing protected area security fence.
The majority of the fence is Il' high (including l' barbed wire section) and constructed at plant grade elevation 21' along the east, west, and south side of the project area. A portion of the east and south fence is 14' high and constructed to extend above the primary auxiliary building (PAB) and radwaste building roofs. To facilitate fence post attachment to the PAB and radwaste buildings, a steel support system was constructed and attached to the north wall of the PAB and west wall of the radwaste building. All other fencing is supponed from posts anchored to new concrete piers. The only underground utilities that could be affected by the concrete pier installation would be the fuel lines for the emergency diesel tanks. Excavation was done by hand digging near the fuel lines.
The steel support system is composed of a network of W6 x 15 beams and columns. Five new fence columns were attached to the PAB building.
Reason for Chance This change was made to facilitate construction of the new switchgear building north of the primary auxiliary building in a non-protected area.
Safety Evaluation ne security fence support system was designed to safely withstand wind loads generated from a 90 l mph wind as required by the Connecticut State Building Code. '
I The fence and suppon system induces wind and seismic loads into the north wall of the PAB, a i safety-related structure. An evaluation concluded that the wind load on the security fence is not a detriment to the structuralintegrity of the PAB.
The maximum seismic accelerations in the steel suppon structure were evaluated, nese reactions are approximately the same as those noted for wind load; therefore, the existing PAB structure is adequate and no further investigation is necessary for Safe Shutdown Earthquake loads. ;
1 The steel suppon system for the fence on the radwaste building is identical to that of the PAB. Since I the fence loads are at nodal points of the building's 2' thick walls and floor slabs, they are minor in !
regards to the load resisting in plane shear capacity of the floors and walls and therefore no further i analysis is necessary.
Herefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR30.59.
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Plant Desien Chance Number 866 Plant Design Change Number 866, entitled "New 4160V Bus 9 Circuit Breaker Cubicle"is complete.
Descriotion of Chance nis change installed a new Class IE,5 kV circuit breaker cubicle at existing 4.16 kV emergency bus 9 adjacent to cubicle 1. In addition,125 V dc control and 120 V ac space heater circuits were wired from existing cubicles 1 and 5 of emergency bus 9 to the new circuit breaker cubicle. This new circuit breaker becomes an installed spare for a future 480 V load center bus 11, which will be located in the new switchgear building.
Reason for Chance His change was ret uired to supply power to the Class IE loads that will be relocated to future load center bus 11 and MCC-1211 (located in the new switchgear building). This new circuit breaker cubicle was provided in order to meet fire protection requirements.
Safety Evaluation The new breaker is categorized as an installed spare. His breaker will eventually supply power to future load center bus 11. While the breaker is an installed spare, it has no impact ori the function of emergency bus 9. A postulated fault on the new cubicle breaker terminals will not damage the emergency bus because the breaker will remain open, and the cubicle and anchorage are seismically designed to preclude any adverse interactions dunng a seismic event. The ratings of the new circuit breaker have been matched to the ratings of the. existing switchgear and to the requirements of future load center bus 11.
Control power for the new breaker is supplied from the 125 V de supply located at cubicle 1 of emergency bus 9. The 125 V de supply has been evaluated and determined to be adequately sized to supply the control power for the new breaker, ne de circuits in the new cubicle are protected from hot shorts by internal fuses. If a short circuit is postulated, these fuses will open and isolate the faulted circuit in the new cubicle from the remainder of the bus 9125 V de control power. De .
remainder of the 125 V de power system would remain unaffected. An oyn circuit in the new cubicle !
control circuit would result in loss of control power for the new breaker, :>ut would not affect the '
remainder of the 125 V de system.
i Because all wiring is enclosed in the breaker cubicles, the increase in combustible loading is '
negligible.
- The new breaker cubicle is designed and installed to seismic Category I requirements.
1 Derefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Desien Chance Number 867 Plant Design Change Number 867, entitled "Replacement of Station Service Transformers 484,485, 496, and 497"is complete.
Dncription of Chance nis change replaced safety related, PCB oil-nlled,1500 kVA,4160-480 V load center transformers (484,485,496, and 497) located in the switchgear room of the service building with dry-type, 1500/2000 kVA AA/FA transformers. A transfomier trouble alarm was provided in the control room to annunciate the following conditions:
- a. Transfomier winding temperature high.
- b. Cooling fans power failure.
This alarm is paralleled at one annunciator window for all four transformers. Additionally, separate indication is available locally at each transformer. Space heaters fed from a separate power supply were provided for each transfomier.
Reason for Chance The old transfomiers were Glled with an insulating oil (Askarel) which contains PCB levels greater than 60,000 ppm and considered a health risk to plant personnel. A transformer fire or a fire reaching the transformers could have aroduced dioxin compounds as a result of the breakdown of the PCB.
These compounds are hazarc ous to plant personnel. Because of dioxin hazanis, evacuation of the switchgear room would be necessary, and difncult decontamination efforts would have to be i implemented. Removal of this 1.azardous material (oil with PCB) from the plant results in improved 1 personnel safety, ne Connecticut Yankee probabilistic safety study concluded that the replacement of the transformers decreased the total core melt frequency due to internal events and Gres by 20.5 percent, a signincant reduction in accordance with company nuclear safety goals.
A transformer trouble alann in the control room provides infom1ation to the operators regarding l transformer cooling systems for corrective action in the event of overload / overheating.
Electrical space heaters are provided in each transformer for use during periods when the transformer is not energized to prevent buildup of condensation and moisture.
Safety Evaluation This modification involved a one-for-one replacement of the four station service transformers. The replacement transformers were designed to power the same equipment as the previous transformers.
De credible failure modes of the replacement transformers aru the same as those of the previous transformers, and the replacement transformers do not use insulating oil. The reolacement transformers are new components that were factory tested and were preoperationa!!y tested before being put into service and are more reliable; thus, a failure of a replacement transformer is less probable than the failure of the old transformer.
The higher transformer impedance of 7.5 percent +/- 5 percent was evaluated and detennined to have no adverse effects on voltage regulation for equipment powered from the four 480 V load center buses.
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The ratings of the replacement transfomiers am 1500 kVA with ambient air (AA) cooling and 2000 kVA with the cooling fans in operation (FA). Loading the new transformers to the 1500 kVA AA limit was evaluated and detemiined to have no adverse effects on the four 4.16 kV buses (1-2,1-3,8, and 9).
He design of the switchgear room HVAC system assumes a heat loss of 20,000 watts for each old transformer at 1500 kVA (full load). The replacement transfomiers are rated at 17,000 watts heat loss at 1500 kVA load (AA). Thus, for the old loading, ;ondition on the transfomiers, which is less than 1500 kVA limit, the operation of the switchgear room IIVAC system is not adversely affected, ne forced air (FA) load limit of the replacement trarsf>rmers is 2000 kVA with the cooling fans in operation. Each old transformer was loaded below tl .> AA limit of 1500 kVA; therefore, cooling fan operation was not required and a cooling fan failum will not result in transformer overload.
Inadvertent cooling fan operation would not affect the functioning of the transfomier or the switchgear room HVAC system because the transformer loading and heat loss would not be affected.
This change replaced the transfomiers with dry-type transformers. Also, the annunciator and space heater circuits were routed in conduit throughout their runs. Therefore, no increase in the combustible loading results.
All new cables were routed in conduits that are seismically supported so as to preclude damat,e to safety related equipment during a seismic event. The replacement transformers are designed to seismic Category I requirements and are qualiGed for the design basis seismic conditions of the switchgear room elevation. The equivalent noor loading from the new transformers is less then that of the old transformers. Therefore, the switchgear room floor slab is adequate to support the weight !
of the replacement transformers.
The annunciator and space heater circuits are nonsafety-related. Because of the electricalisolation and separation requirements between safety- and nonsafety-related equipment, they are not directly interfu.ed with any existing safety system.
The conduits were seismically supported to preclude damage to safety-related equipment dunng a I seismic event. The additional loads are insignificant compared to the member capacity, therefore there is no impact to the stmeturalintegrity of the existing structure.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Desien Chance Number 870 Plant Design Change Number 870, entided "Reactor Coolant System Component Support l Madifications"is complete.
Description of Chance l The following changes were made:
l l a. Replaced spring hanger RC-H-17 on the pressurizer surge line with a rigid sway strut.
- b. Added cover plates to the back-to-back angles which make up the pressurizer truss and repair truss end connections.
- c. Readjusted reactor coolant pump (RCP) spring hangers, as required.
- d. Removed reactor vessel neutron shield tank snubbers.
Reason for Chance
! Under the Systematic Evaluation Program (SEP) Topic III 6, the reactor coolant system was reanalyzed nor normal operating and faulted conditions. The results of these analyses identified three specific areas of the system which would be overloaded during the faulted or Safe Shutdown Eanhquake (SSE) condition. In order to prevent failure of these overloaded components, and possible subsequent failures, these suppons were modified (or adjusted in the case of the RCP springs) to withstand the new loadings, j l
Additionally, these analyses did not take into account the effects of the neutron shield tank snubbers l upon the system, because of their relatively low stiffness. Therefore, they were not required to j maintain the stmeturalintegrity of the reactor coolant system and were removed.
Safety Evaluation 1
l Loadings on the building structure, which have increased as a result of the aforementioned analyses, j l were reviewed and found to be acceptable.
l All analyses and designs meet or exceed original design requirements. These modifications did not I
affect the intended functions of the systems or associated suppons. The margin of safety was increased since the modifications enhanced overall system reliability and structural integrity.
l Therefore, these modifications do not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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I Plant Desien Change Number 875-Plant Design Change Number 875, entitled "Installation of Outside Stem and Yoke (OSY) Isolation Valves for Individual Sprinkler and 11ose Supply Riser at Maintenance Shop Manifold"is complete.
- Description of Change L
i nis change installed two OSY valves needed to isolate individual sprinkler and hose supply risers in the Maintenance Shop manifold.
Reason for Chance This change was necessary in order to allow isolation of the sprinkler / deluge water supply to the Maintenance Shop cable tray, Service Building or Diesel Generator Building without ta ang the entire header out of service for maintenance, modificatica, or in the case of a leak or rupture.
' Safety Evaluation The addition of two OSY gate valves induces negligible additional pressure drop in the sprinkler / deluge systems supply lines. Rus, the water supply to the various sprinkler and deluge systems is not affected.
The added weight of valves, piping, and flanges affects the pipe stress of the as built piping arrangement. It was determined that the existing pi x supports at the manifold are adequate to accommodate the addition of two OSY isolation va ves. Thus, the modification did not affect any safety-related components and/or system.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of :
10CFR50.59. ;
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Plant Desien Chance Number 877 Plant Design Change Number 877, entitled "Screenwell Building Security Enhancement" is complete.
Descriotion of Chance his change involved the attachment of three foot high galvanized floor grating material over the outside face of the Screenwell ventilation louvered openings.
Reason for Chance ,
his change was required by the Nuclear Regulatory Commission to enhance security measures in the Screenwell Building. -
Safety Evaluation The grating and connection is non safety-related since it serves as a security function and its failure does not affect any safety related system.
The 3 foot high steel grating would induce a 12.6 lb per foot dead load on the existing 8 inch bottom sill channel that currently supprts only the small load from the siding and louvers. The deadload on
, the channel from the steel graung was determined to be insignificant.
The effect of the steel grate loading on screenwell house steel superstructure was investigated regarding Safe Shutdown Earthquake (SSE) seismic accelerations, his results in an insignificant load on the existing steel superstructure members.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Desien Chance Number 881 Plant Design Change Number 881, entitled "Additions to Plant Paging System Inside Containment" is complete.
Descrintion of Chance His change involved the addition of speakers, amplifiers and their cab!c connections inside the containment.
Reason for Chance Additions to the plant paging system made it more audible. The paging system previously could not be heard clearly m all parts of the plant. It was determined that the paging system needed to be upgraded.
Safety Evaluation The change involved splicing into the paging system in various areas in the plant via the existing amplifiers. With the additions, the existing amplifiers could be turned down in volume thereby reducing the ec,ho effect of sounds being heard indirectly off walls and ceilings, and thereby providing more even pagmg coverage.
j De design was reviewed with respect to the following:
a) ne paging system is not a Class 1E system, and the additions were connected electrically only to the existing paging system, thereby not effecting any Class IE systems, b) He structural installation was reviewed for seismic two over one concems to ensure that in a seismic event, these additions would not affect any safety related item.
c) De mmored cable used was analyzed and selected so it would not significantly affect comoustible loading in the areas it was used, consequently the cable did not add to intervening combustibles. He cable was purchased to meet IEEE-383 flame test requirements.
d) De paging system is powered from the serri-vital power panel via circuit breaker #17 in panel i #1. A short circuit to the existing paging system or additions to this system will result in a trip of this existing breaker. He additional load is within the rating of this circuit breaker and associated circuit.
l Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Desien Chance Number 882 Plant Design Change Number 882, enNed "htoisture Separator Reheater (hiSR) Inlet Steam Flow Deflector Plates" is complete.
Description of Chance An inlet steam flow deDector plate was installed in each of the four htSR units. He function of the deflector plate is to streamline the flow of cycle steam into the htSR chevron banks. Its installation results in a more uniform steam flow distribution, as well as the elimination of the direct impingement of wet steam on the carbon steel htSR seal end plate. There nuy also be a slight decrease in the cycle steam pmssure drop through the shellside of the htSR, but this affect will most likely be insignificant.
There were no direct changes to the p essure retaining htSR shell.
He selection of 304 stainless steel, as the deflector plate construction material, provides the erosion-corrosion resistance necessary in the high velocity wet steam environment that is present in the htSR inlet chamber. !
In contrast to the original design which had a stainless steel deflector plate welded directly to the carbon steel htSR shell, the new plates possess carbon steel brackets that were fastened to the carbon 1 steel seal end plate. This is an effective design in that it eliminates direct attachment to the heater shell. I In so doing, the thermal stresses imposed on the deflector plate by the expansion / contraction of the htSR shell, due to the variation in coefficients of thermal expansion, are effectively eliminated. In addition, the new deflector plate was designed with adequate free space such that there can be no significant pressure differential across the plate. Brough the minimization of the thermal and pressure stresses imposed on the plate, the new design addressed the failure mechanisms that caused the failure of the original deflector plates, l
Reason for Chance l l
This change streamlines the main steam flow into the htSR chevrons and protects the htSR seal end plate from incurring any degradation due to steam impingement.
Safety Evaluation The only credible failure mode is the failure of the plate attachment welds or tearing / buckling of the deflector plate itself. In either occurrence, the plate could detach itself from its support brackets and possibly be carried over in the cycle steam flow. A small enough piece might pass down the steam chute, but because of the htSR internal design,it could not damage the chevron panels or tube bundle.
This is due to the presence of perforated plates in the steam flow path, prior to entering the chevrons.
Therefore, the only effect of this failure is the potential impedance of steam flow through the steam chute or into the chevron banks.
This might result in an increased cycle steam pressure drop and could potentially shift the steam demand to the other htSR units. Since all the existing units are designed to withstand the effects of Intercept Valve Testing, this increased steam flow would not present a scenario in which funher damage to any of the other htSR units could be incurred. Overall, the flow blockage might result in a slight loss of secondary plant efficiency, caused by the incirased pressure differential, but there would be no effect on any safety-related equipment or its functioning.
Herefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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b Plant Desien Chsnce Number 886 Plant Design Change Number 886, entitled "low Pressure Turbine Replacement" is complete.
Description of Chance his change replaced degraded Low Pressun: (LP) Turbine Units -inner casings and rotors, with two (2) new LP Turbine Units of similar, but improved, design. The new LP Turbine Units provide improved resistance to stress corrosion cracking at disc bore keyway areas, provide a better blade attachment configuration, and improved erosion resistance in steam path areas. New coupling bolts featuring quick connect / disconnect design were used in lieu of original equipment bolts.
To facilitate replacement of the existing inner casings, the extraction lines (9/ unit) needed to be severed. Since the condition of the exnnsion joints and piping adjacent to the casing was degraded and fit-up of the new casings required adjustment in piping lengths, the expansion joint and large sections of pipe on each line were replaced.
Reason for Chance The Westinghouse design LP Turbine Units have exhibited, over the years, disc cracking, blade root cracking, erosion of stationary and rotating blading which has resulted in forced outages, reduction in MWe output, and repeated refurbishment of existing rotors. Installation of a new LP Turbine Unit Design, which does not have a history of these types of problems, reduces these concems.
The coupling bolts have had a history of problems during installation and removal due to the tight clearances that need to be maintained. The new design provides zero clearance, however, it allows for easy assembly and disassembly.
Safety Evaluation The nuclear safety concern, associated with the replacement of the LP turbine rotors is a potential increase in the probability of an external turbine generated missile. External missiles could impact and damage safety equipment and structures (such as the containment structure) and become a safety hazard.
Nuclear safety is assured by maintaining an acceptably low probability of extemal missiles in the following manner:
o Provide an inherently reliable design with suitable materials.
o Perform periodic tests and inspections.
o Provide physical barriers.
In their review of SEP Topic III-4.B. the NRCjudged the turbine generator, and periodic testing to provide adequate assurance that the probability of external missiles is low enough to protect the health and safety of the public. Since the replacement components' characteristics meet or exceed those of the old components, then the replacement components are acceptable.
SEP Topic III-4.B identifies two types of failures leading to external turbine missiles, namely:
- 1) excessive overspeed due to multiple failures within the redundant turbine overspeed protection system, and 2) LP turbine failure at or below design overspeed due to disk defects that initiate and propagate undetected while in service. All other failure modes such as blade failures, bearing failures, shaft failures, etc., do not generate external missiles and pose no threat to safety.
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l Since the existing turbine overspeed protection system remained unchanged by the replacement of the LP turbines, the probability of turbine missile generation due to excessive overspeed remained unchanged.
He periodic LP turbine disk inspection program provides an acceptably high degree of assurance that turbine disks will not fail at speeds up to design overspeed. The inspection program consists of periodic ultrasonic examinations of the LP turbine disks per the manufacturer's recommendations.
The mechanical design of the replacement LP turbines, and the materials employed were reviewed in detail and found to be equal to or superior to that of the original LP mtors. The new LP turbines offer improved reliability and maintainability and funhermore are expected to solve the problems experienced with the original LP turbines, namely blade failures, disk rim cracking, disk bore cracking, and erosion.
The probability of LP disk failures at or below design overspeed is mainly governed by the failure mechanism of disk bore cracking due to stress corrosion cracking (SCC). The new LP rotor design contains many features to reduce or eliminate susceptability to stress corrosion cracking. Also, the disk design has proved to be very successful in extensive operating experience on large European nuclear units since 1972 without a single disk bore crack or other SCC problem detected.
Utility Power Corporation (UPC), supplier of the replacement LP turbines, perfomied a turbine missile analysis which indicated failures would be contained by the heavy inner casings at design overspeed (125 percent of rated speed) and up to 149 percent of rated speed. The design of the UPC turbine inner casings features unusually heavy and rigid construction with "crash rings" to minimize the ris c of turbine missiles.
Since the heavy inner casing design of the replacement LP turbines has eliminated the LP disk failure mode as a safety concern, it can be concluded that this change improved plant safety.
The turbine components were transported over the discharge tunnel. The loads on the discharge tunnel were calculated. The resulting factomd monx nts and shear loads wem below the ultimate strength of the discharge tunnel. The turbine building operating floor was evaluated for equipment laydown loads imposed during construction.
This myiew found all stress and strain levels to be within code allowables. The loads imposed on the standard lifting components (shackles, slings, and turnbuckles) were found to be less than the rated safe working load of these components. The turbine components (rotors, casings, and hood) are supported on a steel turbine base. Analysis results show that stresses in the turbine base are well l below the allowable limits.
The concrete pedestal is a massive reinforced concrete structure that supports the high pressure (HP),
LPs, and generator. This structure includes the pedestal foundation (below 21'6") and extends to the 59'6" elevation. It is composed of large concrete columns and beams (10' deep), and is structually independent of the turbine building structure. It was originally designed with allowable stress design criteria. It was determined that the stmeturalintegrity of the plant was not affected. l Therefox, this modification does not constitute an unreviewed safety question per the criteria of l 10CFR50.59.
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Plant Desien Chance Number 887 Plant Design Change Number 887, entitled "Low Temperature Overpressure Relief Protection (LTOP) System" is complete.
Descriotion of Chance his change replaced terminal block terminations used for LTOP circuits inside containment at the electrical penetrations and in one junction box with qualified splices.
Reason for Chance his change was necessary to qualify the LTOP system for use in an accident environment.
Safety Evaluation The only changes made to the existing installation were the one-for-one changeout of the terminal block terminauon with in-line splices. All splicing materials were purchased as safety-related and the work was performed under safety-related control.
The only credible failure mode is that of an improper connection. Assurance that this did not occur was made by working on only one splice at a ume, maintaining safety-related control for verification, and testing each circuit after splicing.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59, 47
Plant Desien Chance Number 889 Plant Design Change Number 889, entitled, "Low P rssure (LP) Turbine Replacement / Instrument and Control (l&C) Portion" is complete.
Descriotion of Chance This change installed new axial growth and vibration monitors and transducers using the same locations as the old equipment with the exception of the axial growth monitor which was relocated from hiain Control Board "G" to hiain Control Room Cabinet AA. There was a one-for-one cable replacement resulting in no significant weight or conduit loadinJ change. The existing raceway was reused requiring no additional raceway or supports. All new ca ale is IEEE 383 flame retardant.
Conduit runs from the turbine to the hiain Control Room were smoke and fire sealed at both ends.
For rotor axial growth monitoring, the old transducer on the turbine rotor, the monitor on hiain Control Board "G" and interconnecting cable were removed, ne transducer mounting was modified and a new transducer was installed and wired to the new monitorlocation. A new monitor and rack similar to the old vibration equipment was installed on Cabinet AA in the hiain Control Room. De new monitor has controls similar to the old monitor. Annunciation and recording capability were maintained.
For rotor vibration monitoring, new transducers were installed, two per bearing, on new mounts on either side of the rotor 45 from the top. The new monitors are similar in structure to the old monitors and were mounted in the same rack on Cabinet AA. The new monitors have controls similar to the old modules. Existing monitoring and recording functions were maintained. In addition to relative vibration monitoring, this system was designed for seismic and absolute vibration monitoring and is low frequency compensated.
For turbine exhaust pressure monitoring, basket-tip probes with piping forindicators were added around the low pressure turbine. The low pressure turbine has two connected sections. Each section has two exhaust compartments. A pressure probe was added on either side of each exhaust compartment for a total of eight (8) probes and indicators around the perimeter of the low pressure turbine. He probes were bracket mounted inside the condenserjust above the low pressure turbine exhaust / condenser flange area. De pipes were routed through the wall of the condenser and up through shutoff valves mounted along the turbine skirt.
Reason for Chance These changes were made for the following reasons:
o Rotor Differential Expansion hionitoring - Improvement of the reliability and accuracy oflow pressure rotor data sources. Retirement of obsolete equipment. Improved ability to secure replacement parts.
o Rotor Vibration hionitoring - Improvement of the reliability, accuracy and completeness of data sources which improved the ability to balance the turbine rotor at low speeds. This capacity is intended to reduce the (critical path) time for the dynamic balancing operation.
Since the new equipment is identical to existing equipment at other plants, spare parts inventory were reduced, o Turbine Exhaust Pressure Indication - Addition of turbine exhaust basket tip probes pressure measurement provide necessary data to measure low pressure exhaust and verify the exhaust pressure correction curve.
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Safety Evaluation This change did not affect any safety systems. Additionally, the change had no affect on systems designed to reduce challenges to safety systems and so had no impact on the performance of safety systems.
Therefore, this nxxiification does not constitute an unreviewed safety question per the criteria of ~
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Plant Deslan Chance Number 890 i
Plant Design Change Number 890, entitled "Motor Operated Valve (MOV) Overload Protection :
Upgrade for MOVs 23,34, 331,160, and 344" is complete. !
Descriotion of Channe l nermal overload relays and heaters were replaced on the following safety related MOVs: ;
RH MOV-23 FW-MOV 160 ' .
RH-MOV-34 FH-MOV-344 l CH MOV 331 Ambient temperature compensated units were used. -
ne replacement of the overload relays accomplished the following:
- 1. Provides sufficient operating time and safety margin, under full load conditions, to complete valve stroke without tripping. ;
- 2. Maintains three pole protection as recommended by the National Electric Code. .
Es.ason for Changt :
his change was made to conform with the requirements of SEP Topic III 10.A, "Dermal Overload i Protection for Motors of Motor Operated Valves" and provide consistency of application with other i safety related MOVs. ,
F ne use of ambient compensated relays provides better overall MOV overload protection by !
eliminating the effects of ambient temperature at the relay (MCC) location, thereby allowing the relay j to respond to current only.
Saferv Evaluation I i
De only significant change brought about by the relay and heater replacement was a decrease in the tripping time at fullload current. However, it has been maintained beyond four (4) titres the stroke !
time of the valve. Recognizing that a minimum of two (2) times the stroke gnut be allowed by the overload relhy, a safety margin of at least double the valve's stroke time was provided. ,
i All other possible failure effects of the new overload relays remain identical to the old relays and {
heaters. Failure of a relay to trip when called upon to do so will result in damage to the overpowered !
MOV. A relay overtrippmg (tripping too soon) will result in the removal from service of an MOV l which may be operatin events from occumng.g normally. Testing of the new overload relays was geared to p Derefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i Plant Desien Chance Number 82], i j
Plant Design Change Number 891, entitled "Rod Cluster Control Assembly (RCCA) Changeout Tool Modification" is complete.
Descriotion of Chance This change consisted of an 18" overalllength decrease of the RCCA changeout tool. This was accomplished by replacing the 172" su ppon tube with a similar 154" support tube. Further, to I compensate for radiological shielding t:1e 10' extension cable used to matntain adequate water coverage was replaced with a similar sling increased by 18" to 1l'6". ,
This change affected the overall tool length making the motor control panel on top of the tool more accessible from the spent fuel pool bridge platform deck. He length decrease was compensated for j by an identical length increase in the crane attachment sling.
Reaso'1 for Chance The RCCA changeout tool was shonened in order to achieve a normal operating elevation that would
- be consistent with the spent fuel pool bridge.
Safety Evaluation The credible failure modes of the tool associated with the change remain the same as those identified !
with the original tool:
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- 1) Tool dropped on a fuel assembly
! 2) Tool dropped on the poolliner l
- 3) Stuck RCCA
- 4) Tool drop on fuel rack No design basis accidents were identified for which failure trxxles associated with the change could be
, an initiating event. 1 I
J nerefore, this modification does not constitute an unreviewed safety question per the criteria of !
10CFR50.59. j i
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t Plant Desien Chance Number 898 Plant Design Change Number 898, entitled "Steam Generator Plugging" is complete.
Descriotion of Chance Mechanical plugs were installed in both ends of steam generator tubes which were identified by Eddy [
Current Testmg to have unacceptable indications.
Reason for Chance These plugs were installed to comply with the requirements of Connecticut Yankee Safety Technical Specifications.
Safety Evaluation Plugging of steam generator tubes resulted in a reduction in reactor coolant flow, steam generator heat transfer area, and primary coolant inventory. Also, the reactor coolant pump coastdown rate was increased. These factors were reviewed with respect to DNB cvents, overcooling and overheating ,
events, reactivity events, and loss of coolant accidents. In all cases, the effect of the tube plugging ,
was detennined to be acceptable. Installation of plugs into degraded tubes did not significantly alter ,
plant response. Plugging of degraded tubes reduces the probability of a steam generator tube rupture.
Reactor coolant flow testing, as required by Technical Specifications, will confirm adequate flowinte.
. This testing is performed during Mode 1 operation.
t i Therefore, this modification does not constitute an unreviewed safety question per the criteria of l 10CFR50.59.
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Plant Desien Chance Number 899 4
Plant Design Change Number 899, entitled "Turbine Building Jib Crane" is complete.
l Description of Chance i A five tone jib crane with a 20' reach was installed in the Connecticut Yankee Turbine Building. The crane was mounted on column B-3 at the 79.5' elevation,20' above the operating floor. The crane's position is at the nonh end of the loading bay, allowing access to both the loading b,ay and the operating floor. De crane is derated to a maximum payload of two tons and a maximum reach of 10', until such time as additional stiffeners and bracing are installed. Clamps are attached to the jib beam to prevent the tmlley from moving beyond 10'. In addition, an administrative control is imposed to prevent the simultaneous use of the jib crane and the overhead crane when the overhead crane is within one bay of column B-3. The crane is bolted with A325 bolts to the column flange, j Reason for Chance 1
a This crane was installed to maintain the ability to bring smaller items up or down through the cranewell without using the main Turbine Butiding crane.
Safety Evahiation The crane is not seismically qualified, but its effect on the building's seismic analysis was determined to be acceptable. Here is no equipment below the crane that is required for shutdown.
De jib crane with the above restrictions does not adversely affect the strength of the column. The column was evaluated for the worst combinanon of positions (180 rotation). The resulting loads on ;
the column, due to the jib crane and all other loads (roof load / dead loads), are below the allowable limits. In addition Safe Shutdown Earthquake (SSE) seismic loads on the column due to the building i
response andjib crane's res nse are well below the allowables. The hole pattem on the column's flange does not significanti affect the strength of the column.
Therefore, this modification does not constitute an unn: viewed safety question per the criteria of 10CFR50.59.
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Plant Desien Chance Number 913 Plant Design Change Number 913, entitled "Containment Manipulator Crane Modifications" is complete.
Description of Chance This change included the removal of the manipulator crane's upper superstructure to relieve stresses in the end tmcks. One column remains at approximately 17' and was braced to act as a temporary festoon mast. De end trucks, wheels, gear drives and cam rollers were repaired and aligned as necessary to retum the crane to operable service. This change also included removal and reinstallation of electrical and compressed att semees including cabling to an area radiation monitor.
Reason for Chance The manipulator crane's superstmeture was damaged when the Polar Crane collided into it. The stmeturalintegrity of the superstructure was questionable and may have been imposing forces on the manipulator crane bridge. Th:se forces may have been contributing to the misalignment of the wheels and drive gears. Removal of the superstructure was necessary for the alignment repairs and for the safe use of the crane during fuel unloading. Modifications to the northeast column were necessary to act as a festoon mast for electrical and air services. The alignment of wheels and drive gears was necessary to allow full travel of the bridge and provide accurate mast placement (indexing) during the fuel unloading process. A collision of the mampulator crane with the polar crane is no longer possible.
Safety Evaluation The potential failures and consequences associated with this job were: 1) the structural failure of the modified crane during the fuel handling process,2) the failure of any of the reinstalled electical/ air cables, or 3) failure of the crane during the load test.
Regarding structural failure of the modified crane during fuel movement, an evaluation was performed which concluded that the modified crane is well able to support the weight of the combined load of a fuel assembly and RCCA without the crane superstructure, ne loads imparted by the new cable support arm and column are insignificant compared to the design capacity of the crane (6000 lbs.).
The cable support arm was designed not to fail during a seismic event. Regarding the failure of the reinstalled air and electrical cabling, the only design change was the rerouting of the cables to the northeast column (cable support arm). The crane mechanism is designed so that loss of air or power will not release the fuel assembly from the gripper mast. No new failure mechanism was introduced.
To assure correct installation, "red line" checks of the cabling and a source check of the area monitor were performed.
Regarding the load test, the manipulator was positioned east of the reactor vessel over an open area of the cavity for this test. Crane failure was unlikely since the design capacity is 6000 lbs. and the test load is approximately 2300 lbs. Nonetheless, a crane failure would drop the test load approximately one foot onto the cavity floor; this would not cause any significant damage to the cavity floor or liner.
The spent fuel pool was isolated from the cavity during this test.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Desien Chance Number 9M Plant Design Change Number 915, entitled "Installation of Turbine Lube Oil Centrifugal Purifier Connections" is complete.
Descriotion of Chance This change provided connections to allow temporary reinstallation of the Delaval centrifugal turbine lube oil purifier which was removed under a previous change.
Reason for Chance .
A previous change replaced the Delaval purifier with a Keene-Bowser lube oil conditioner. While the Keene Bowser is a better purifier during normal conditions, it cannot handle water removal under abnormal conditions such as water inleakage from the lube oil coolers. Recent incidents of this 3roblem have caused excessive lube oil water content which has led to excessive wear to turbine
> earings and control oil system disturbances (itsulting in a turbine trip). Installation of the Delaval allows it to be used to supplement the Keene-Bowser unit during penods of high water intrusion.
Safety Evaluation i
This modification had no effect on any safety-related equipment. The Turbine Lube Oil System serves no safety function directly or indirectly. ,
Potential failure modes include system rupture, which could be no worse than present system rupture and does not affect any safety-related equipment; and fire which is a minimal probability given the ;
one hour fire patrol, existing sprinklers and detectors in the area.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of j 10CFR50.59, 1
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Procedure Chances t
Procedure Number Illit SUR 5.7-118 Senice Water Penetration Check Valves SUR 5.7-119 Inservice Inspection of Fill Header FH hiOV 344 Bypass Valve SUR 5.7-120 Inservice Inspection of Fill Header FH-FCV-295 Valve SUR 5.7-121 Inservice Inspection of Reactor Coolant Pump (RCP) Seal Water Supply Check Valves, CH-CV-305A,305B,305C, and 305D SUR 5.7-126 Inservice Inspection of Charging Valve CH-h10V-298 Chip 8.5165 hiaintenance of Spent Fuel Pit Pump P-21-1B SPL 10.7 268 Operation at 5620F Tave for Pre LP Turbine Replacement Testing SPL 10.7 273 Residual Heat Removal (RHR) Flow Test to Determine FCV-7% Collar Dimensions SPL 10.7 277 Repressurization of the Power Operated Relief Valve (PORV) Accumulator SPL 10.7 292 RHR Charcoal Spray Hydrostatic Test SPL 10.7-307 Installation and Removal of Fuel Ultrasonic Test (UT) Equipment in the Spent Fuel Pit (SFP)
VP 213 Guideline Specification for Structural Concrete Bonding Process VP 273 Chem Nuclear Systems, Inc. hiobile Demineralization Procedure VP-312 Fuel Assembly Inspection Procedure--Failed Fuel Rod Detection System EOP 3.1-8 Complete less of Condenser Vacuum EOP 3.1-23 Loss of Circulating Water EOP 3.1-34 Complete 1 Ass of Control Air Supply
- EOP 3.1-46 Totalless of Semivital Power EOP 3.1-49 Partial Loss of DC l EOP 3.150 loss of hiotor Control Center htCC-5 AOP 3.2 23 hialfunction of Rod Control System AOP 3.2 37 Cable Vault Fire AOP 3.2-39 Containment Fire ES 1.3 Transfer to RHR Recirculation ES1.4 Transfer to Two Path Recirculation f
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Number 31 tic l SUR 5.7-118 Service Water Penetration Check Valves Descriotion of Chance SUR 5.7-118 provides instructions for testing the operability of the service water penetration check :
valves. Normally, the service water pumps are running dunng operation. Guidehnes have been provided to measure the service water flow and pressure outside the containment. This was a new procedure. No change was made to an existing procedure. .
Reason for Chance This procedure allowed for testing the service water penetration check valves.
Safety Evaluation ;
This procedure does not disturb any operating path or condition.
l Therefore, this procedure does not constitute an unreviewed safety question per the criteria of 10CFR50.59. !
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Number Iilla SUR 5.7-119 Inservice Inspection of the Fill Header FH-MOV-344 Bypass Valve i
Description of Change SUR 5.7-119 provides instructions for testing the fill header bypass valve. Guidelines have been ,
provided to measure opening / closing time for FH MOV-344. This was a new procedure. No change was made to an existing procedure.
Beason for Chance l This procedure allows for testing FH MOV-344.
Safety Evaluation The subject test does not disturb any operation or condition of the lines the valve is in, or any other l lines.
l Therefore, this procedure does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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SUR 5.7-120 Inservice Inspection of FillIIcader FII FCV 295 Valve :
Descriotion of Chance SUR 5.7-120 provides instructions for testing fill header FH FCV-295 valve. Guidelines have been provided to nwasure opening / closing time for this valve. This was a new procedum. No change was -
made to an existing procedure.
Reason for Chance f
This procedure provides instructions for testing FH FCV-295. I Safety Evaluation The subject test does not affect any operation or condition of the line the valve is in, or any other line.
Therefore, this procedure does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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SUR 5.7121 Inservice Inspection of Reactor Coolant Pump (RCP) Seal Water Supply l 3
Check Valves CH CV-305A,305B,305C, and 305D -
l Descriotion of Chance a SUR 5.7-121 provides instructions for testing the operability of RCP seal water supply check valves 2 CH-CV-305A,305B,305C, and 305D. Guidelines have been provided to determine and record the flow on the flow indicators FI 101 A, FI 102A, FI 103A, and FI-104A. His was a new procedure, i j No change was made to an existing procedure, j
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l j Reason for Chance his procedure provides instructions for testing the RCP seal water supply check valves.
l Safety Evaluation j De subject test does not affect any operation or condition of the lines the valves are in, or any other l
- lines.
Therefore, this procedure does not constitute not an unreviewed safety question per the criteria of <
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Number Ig]g SUR 5.7126 Inservice Inspection of Charging Valve CH MOV 298 Descriotion of Channe SUR 5.7126 provides instructions for testing CH MOV 298. Guidelines have been provided to measure opening / closing time for CH MOV-298. This was a new procedure. No change was made to an exisung procedure.
Reason for Channe This procedure provides instructions for testing CH MOV-298.
Safety Evaluation The test is performed during plant cold shutdown and does not have any affect on operation of the line the valve is in,or any other hne.
Therefore, this procedure does not constitute an unreviewed safety question per the criteria of 10CFR50.59, 61
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CMP 8.5-165 Maintenance of Spent Fuel Pit Pump P-21-1B !
l Qgsgjption of Channe l
Dis procedum provides instructions for corrective and prevendve maintenance on spent fuel pit i (SFP) pump P 21 1B. This was a new procedure, No change was made to an existing procedure.
l Reason for Change l- Originally, there was only one SFP pump P 21 1 A. P 21-1B was installed several years ago to i l provide a backup to P 21-1 A. The new pump and motor are different than the or einal, therefore l requtnng new procedures to be developed for conective and preventive maintenance.
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! Safety Evaluation i This procedure is essentially the same as the corrective maintenance procedure for P-21 1 A except for l the changes needed in technical content because of the different pump manufacturer, and a different ;
tagging list. j nerefore, this procedure does not constitute an unreviewed safety question per the criteria of 10CFR50.59, l
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Number Tuk ,
SPL 10.7-268 Operation at 5620F Tave for Pre-Low Pressure (LP) Turbine Replacement Testing Description of Change His procedure was originally used to determine optimum Tave. His change allowed for pre.
replacement heat rate testing of the turbine, e
Reason for Change I
his procedure was changed to support testing of the LP turbine rotor at 5620F Tave/ optimal governor valve position for s eight hours.
In the refueling outage following Cycle 14 operation, the stationary and rotating elements of the low pressure turbines were replaced. It is necessary to perform turbine testing before and after the turbine replacement. Also, it is imponant to perform both tests under the same conditions.
Safety Evaluation During the Cycle 14 outage, PDCR 841 was implemented which allowed for an indicated Tavg of 562 F. Previously, the plant operated with a 557 F Tavg.
The plant was operated during Cycle 14 at a Tavg of 5620F for a test period of no more than eight i hours. !
6 No additional failure modes were created due to this test. Based on the test procedure, plant !
conditions during this test were never more limiting than initial conditions assumed in ne design basis analysis.
Operation with a Tave of 5620F does not modify the plant response to the point where it can be considered a new accident. De plant response and accident consequences for loss of coolant accident (LOCA) and non LOCA events were not adversely affected. ;
Herefore, this procedure does not constitute an unreviewed safety question per the criteria of ;
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Number Iglg SPL 10.7 273 Residual Heat Removal (RHR) Flow Test to Detemune FCV-7% Collar Dimensions i Descriotion of Channe !
A Dow test of the RilR system was perfomxd to determine the throttle position of the Dow control valve FCV-796. i Reason for Chance I
'Ite purpose of the now test was to determine the collar sizes, which after installation ensure that the valve provides a pre-detemuned now rate through the now control valve FCV-796. i Safety Evaluation The test was performed with the core off loaded. While perfomung the test, the RCS cooldown rate was limited to < 300F/hr. Also, as a part of this test, the RIIR pump motor current at different pump Dow rates was measured.
None of the design basis accidents were impacted because the test was performed in with the core off-loaded. The failure modes associated with the test do not become initiating events. ,
, The throttled position of valve FCV 7% would be incorrect if the collars were mis-installed following this test or any future use of the RilR system. Such a misinstallation could occurif the collars were inadvertently interchanged. An incorrectly throttled FCV-7% may impact the performance of the RilR system. Ilowever, the collars are chained to the surrounding structures such that the respective .
3 collars cannot be interchanged. Thus,it is concluded that the performance of the RHR system will be i
consistent with the assumptions in accident analysis following the pmposed test and the addition of the collars on FCV 796.
Therefore, this procedure does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Number Ink SPL 10.7 277 Repressurization of the Power Operated Relief Valve (PORV) Accumulator Description of Chance nis procedure provided instructions for repressurization of the PORV accumulator. This was a special procedure. No change was made to an existing procedure.
Reason for Chance This procedure was necessary to ensure meeting plant Technical Specifications which require PORV accumulator pressure be maintained above 118 psig.
Safety Evaluation Three areas of concern were evaluated by this safety evaluation: use of a compressed gas bottle inside the containment, effect of performance of this procedure on PORV accumulator pressure, and consequences of the worst case failure during performance of this test.
l Use of a compressed air bottle in containment was acceptable because of the precautions that were I taken. Appropriate industrial safety precautions for the handling of compressed air bottles were I followed. The gas bottle assembly was assembled and disassembled on the charging floor, and was not stored inside containment. He compressed gas used to pressurize the PORV accumulator was inert, non-flammable gas.
As a result of these precautions, the probability of creating a missile hazard in the containment was l
minimized.
De effect of performance of this procedure on PORV accumulator pressure must also be considered.
l Performance of this procedure was acceptable because PORV accumulator pressure was maintained
! above the technical specifications limit of 118 psig and below the design limit of 145 psig. As a result of this control, the PORV accumulator relief valve will not be challenged during a small break LOCA or a transient that would require feed and bleed to remove decay heat.
De worst case failure during performance of this procedure would be failure of the compressed gas bottle pressure regulator. For this failure to have significant repercussions, additional failures would have to o: cur. De additional failures would be either failure of the regulator output 3ressure gage I
and the PORY pressure indicator PI-9P, or else rapid pressurization that would not a low the operator to shut CA-V-824B, before the PORV relief valve lifts. Additional failures that would have to occur would be either failure of relief valve, CA RV 829, to lift, or the failure of this relief valve to rescat i after opening. Results of these failures are already addressed in the Technical Specifications. '
Increased leakage through the PORV accumulator relief valve could result in PORV accumulator leakage greater than allowed in the design calculations for PORV accumulator leakage during a feed !
and bleed transient. As result of this possibility, the procedure required performance of the system leakage portion of the PORY System Operability Test, SUR 5.7112,in the event that the PORV j accumulator relief valve had been challenged during performance of this procedure. '
Overpressurizing the PORY air supply was unlikely because of procedural controls. Proper operation l of the gas bottle pressure regulator was checked before the bottle was used to pressurize the IORV l accumulator. Cautions were incorporated into the body of the procedure, providing guidance in the )
event of failure of the pressure regulator.
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Potential relief valve failures were either unlikely to occur, or the consequences of a failure were ;
addressed. Because of the simplicity of the design of the relief valve, failure of the relief valve to lift - i
- or failure to rescat was felt to be unlikely. CA RV-829 is a 1/4 inch Nupro relief valve. nis valve is '
also an ASME III Class 3 valve, and is set,x> int tested every 5 years. Increased leakage through the valve after it has lifted is felt to be highly possible. As a result, performance of the PORV system air leakage test would be prudent.
herefore, this procedure does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Number I 111 2 SPL 10.7 292 Residual lleat Removal (RiiR) Charcoal Spray liydrostatic Test Description of Chance his procedure established the boundary (valve lineup) for the performance of a visualinspection during the RllR charcoal spray hydrostatic test. This was a special procedure. No change was made to an existing procedure.
Reason for Chance This procedure was performed for requirements of ash 1E Section XI Ten Year Ilydrostatic Testing.
Safety Evaluation The plant was in hiode 6 prior to inglementing this procedure. The hydrostatic test pressure was 330 psig (+5 psig. -0). Pirssure indicating instrumentation was calibrated to .5% of full scale prior to the test. De test gage (s) had a range at least 1.5 times but not greater than 4 times the test pressure.
Pressure indicating instrumentation was calibrated within the previous two weeks. Relief valves used had a setting of 335 psig. A second operable hydrostatic test pump was available for a backup.
Proper ventmg of the system was insured when filling for the hydrostatic test.
At completion of the test, all lines were drained immediately.
Since the test was performed in hiode 6, there was no plant specific safety concerns. If there had been a failure during the test, there would be no adverse effect on the plant. The failure would be repaired and retested in accordance with the Inservice Inspection Program per Section XI of the AShiE Boiler and Pressure Vessel Code.
He performance of the RIIR Charcoal Spray liydrostatic Test ensured its integrity but did not change nor reduce the reliability of the RiiR Charcoal Spray.
Therefore, this procedure does not constitute an urueviewed safety question per the criteria of 10CFR50.59.
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l Number Eth 4 SPL 10.7 307 Installation and Removal of Fuel Ultrasonic Test (UT) Equipment in the Spent Fuel Pool (SFP)
Description of Change The procedure consisted of the temporary installation and operation of a UT system on SFP racks.
The UT system resided on an empty 4 x 3 array of fuel racks with adjustable pins keyed into rack locations for horizontal stability. All fuel assemblies tested remained fully grappled at all tinws while panially insened into a rack location to receive the ultrasonic probe. De system was removed upon testing completion. This was a special procedure. No change was made to an existing procedure.
Reason for Change Dis procedure is used to identify failed fuel rods.
Safety Evaluation i The UT iris xction system is considered a heavy load, since the load weight exceeds the 1650 lbm.
heavy load imit. De mechanical and criticality aspects for the control of heavy loads and the installation, use and removal of the UT inspection system were evaluated and are acceptable provided the combined weight of the load plus block is demonstrated to be less than 1875 lbm. prior to lifting over the SFP.
1 Design Basis Accidents reviewed for impact by this procedure include: heavy loads-fuel drop event and handling (dropped tool and fuel assembly accident) fuel drop event.
De parameters and systems affected by this procedure did not impact the consequences of these
, accidents and are bounded by the fuel drop event. A safe load area was identified for the installation,
! use and removal of the UT system. The safe load area was surrounded by a buffer zone of at least j seven empty cell locations. Since there was no spent or fresh fuelin the safe load area and the buffer zone, there were no criticality concerns due to heavy loads. The seismic response of the fuel racks ,
was not affected by the presence of the UT system. The impact of dropping the UT system on the ;
liner and additional loads on the Spent Fuel Pit structure were considered negligible.
No design basis accidents were identified for which failure modes associated with the change could be an initiating event. Dropping a fuel assembly or dropping a load on the fuel assembly is the initiating event for the design basis fuel drop event.
The installation, use and removal of the UT sy stem and associated failure modes do not natify the response of the fuel drop event or a seismic event.
l Therefore, this procedure does not constitute an unreviewed safety question per the criteria of i 10CFR50.59, l
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MI Tidt VP 213 Guideline Specification for Structural Concrete Bonding Process Descriotion of Chance This specification defines the work, including materials, equipment and execution of structurally repainng cracks in portland cement concrete by the injection of an epoxy resin adhesive. This is a vendor procedure. There was no change to an existing procedure.
Reason for Chance Groundwater in leakage throu.;h the cable vault floor slab, elevation 5'-6",is a continuing problem. ,
Nine cracks, primarily perpenc icular to the containment wall, developed in the cable vault floor slab.
His floor slab spans 15 feet from the exterior containment wall to the north wall of the cable vault at elevation 5'-6". Water in leakage through these cracks must be continuously monitored and removed.
A significant buildup of water in the cable vault could submerse cables threatening their capability to function properly and would be lazardous for workers in this area.
Safetv Evaluation A calculation was generated to review the design capability of the cable vault walls and floor slab to withstand a 30' mean sea level (MSL) flood elevation, his calculation determined that the design outline of the cable vault structure can withstand a 30' MSL flood elevation without structural failure.
Several steps were taken to assure that the structural margin of safety in the floor slab was not jeopardized due to the construction process. As outlined in Specification VP 213:
o Only concrete bits were used. This assured that the concrete reinforcing steel was not cut or damaged during the installation of entry ports.
o Pressure check tests and mix ratios were periodically monitored. Adequate pressure to fill the existing cracks was required. Overpressure may spall or otherwise damage the slab. Proper mix ratios were required to obtain the required bcnd.
o All entry ports were alugged at the end of each working day. His assured that increased l
leakage through unfi led entry ports was limited.
o Approved specific epoxy adhesive (s) used.
o Work would be suspended upon increased water in leakage pending the review and approval to proceed. Increased water in leakage, although improbable, would be limited to controlled amounts.
Water control equipment was readily available to control any increased water in-leakage. Plant personnel were available to assure that increased water in-leakage could be accomnniated without l threatening to submerge cables, j l
The work order to perform this job referenced the applicable station and maintenance procedures to assure that safe setup, construction, and closcout practices were used.
Herefore, this procedure does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Number 311k VP-273 Chem Nuclear Systems,Inc. (CNSI) Mobile Demineralization Procedures Description of Chance This operating p ocedure is for dewatering bead resin and activated carbon in CNSI 14-195 or smaller liners. He procedure was tested for certification in meet.ng i the Barnwell Site Disposal Criteria and ;
llanford Site Criteria for free-standing liquids. This is a vendor procedure. No change was made to an existing procedure.
Reason for Chance The purpose of this change was to allow replacenwnt of the waste liquid evaporator with a mobile unit to reduce radwaste generation and costs.
Safetv Evaluation l De mobile demineralizer system consists of four ion exchange vessels, two filters, a vendor supplied i booster pump and associated control and sample skids. This equipment is located in the Prinury Auxiliary Building (PAB) drumming room. He mobile demineralizer ties into permanent plant equipment in the middle level of the Waste Disposal Building. Two hoses are run from the Waste Disposal Building to the drumming room through the PAB pipe trench. Iloses are also used in the drumming room to interconnect the CNSI equipment. Spent resin is sluiced to a liigh Intensity Container (lilC) on a trailer outside the drumming room. The IllC is shielded by a decertified shipping cask.
One concem was the potential for a hose rupture in the PAB pipe trench to disable safety related equipment in the pipe trench or in the RiiR pit. The only safety-related electrical equipnxnt in the pipe trench are the solenoid valves and the positioners for the charging pum ) flow control vahts (Cli-FCV 110/110A). The only safety related electrical equipnwnt in the R?iR pit lower elevation are the RIIR pump motors. A calculation was perfomxd to demonstrate that flooding the PAB pipe l trench at 60 gpm expected by a hose rupture will not disable safety related equipment in the PAB pipe I *rench or the RilR ? i t . In addition, precautions were taken to prevent a hose rupture, ne hoses in the PAB pipe trenc 1 are made of a sturdy, non-collapsible rubber. Rese hoses have a design pressure of 150 psig, and are hydrostatically tested to 225 psig. The CNSI equipment operator walks down hoses in the crumming room and the waste disposal building every shift. He hoses in the pipe trench are walked down every week.
, A radiological safety review of this project was perfomrd which concluded that the consequences of l
dumping one ion exchange vessel full of spent resin to the environment was acceptable. In addition, other precautions were incorporated into the mobile demineralizer design and the CNSI operating procedure to prevent this event from occurring. First, the resin slurry hose is made of a sturdy, non.
collapsible rubber. His hose has a design pressure of 150 psig and is hydrestatically tested to 225 psig. The slurry hose is covered with a heavy PVC sleeving. Covers are placed over yard drains in the vicinity of the sluice hose. Sluicing water pressure is controlled, and should not exceed 15 psig.
Proper connection and resin slurry valve line-up are independently checked before resin slurry starts.
Finally, access to the area is restncted during resin slurry operation.
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Another concem was the potential for the uncontrolled release of radioactive water from a postulated hese rupture through the drumming room door or for contamination of the ground floor of the PAB.
A dike was built around the CNSI equipment at ground level in the drumming room. In addition, a calculation was perfonned to demonstrate that no radioactive water would be released to the environment. Fmally, a cover was installed over the grating near the door from the PAB to the drumming room.
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There was some concem about adequate drainage in the drumndng roont CNSI ec uipment skids have drip pans to collect radioactive water. Drip pans were fabricated for the filter lousings. All these dnp pans drain to a 100 gallon drain tank in the drumming pit. Drain water is then pumped to the floor drain outside the drununing room door by a small pump.
Another concern was the potential for release of airborne activity from filter replacement. Filter replacement requires takmg an air sample. If there was evidence of an airborne radiation problem, a llEPA filter would be installed on the vent line Therefore, this procedure does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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a Number Igk VP.312 Fuel Assembly Inspection Procedure Failed Fuel Rod Detection System ;
(FFRDS) ;
i Description of Channe i
nis procedure provides instruction for installing and removing the FFRDS equipment in the spent '!
fuel pool and at the pool side. His is a vendor procedure. There was no change made to an existing l procedure. ;
t Reason for Change ~
nis procedure is used to identify failed fuel rods. i l1)fety Evaluation !
see Safety Evaluation for procedure SPL 10.7 307.
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EOP ?.1-8 Complete Loss of Condenser Vacuum Descrintion of Chance Section 5.0, "Subsequent Operator Actions," was modified. Included in this modification was a change in format. The revised format is consistent with the Emergency Operating Procedure (EOP) procedure upgrade. The steps in the procedure provide instructions for controlling reactor coolant system (RCS) terr.perature, maintaining feed to the steam generators, maintaining demineralized water storage tank (DWST) inventory, informing the Duty Officer, and filing a Plant Information Report (PIR). The actica pertaining to the Duty Officer and PIR was deleted. This action is in ES-0.1 < tep 15 which would be xrformed after this procedure. Therefom, this step is not needed. The other steps are maintainec in the revised procedure. The wording is also consistent with EOP procedure upgrade which reduces the potential for confusion in performing EOP 3.1-8. In addition, the revised procedure steps give additional guidance for handling the event.
E.gason for Chance This procedure was changed due to a procedure upgrade program.
Safety Evaluation The revised procedums only provide instructions for diagnosing and mitigating an event or challenge to the Critical Safety Functions (CSFs). These instructions are used after an initiating event has occurred and therefore cannot be an initiating event.
The only credible failure mode that could be associated with the procedure changes would be if the revised procedures contained inappropriate instructions for diagnosing and mitigating an event or challenge to the CSFs.
The revised procedures provide instructions on the use of safety systems. In addition, they give l instructions for the use of other systems to mitigate an event or challenge to the CSFs. The l
! verification program ensures that applicable generic and plant specific technical infomiation has been i properly incorporated. In addition, the validation program confirms that the actions specified in the l revised procedures can be performed to mitigate an event or challenge to the CSFs. Derefore, the 1 reded procedures did not change the probability of failure of these systems.
Herefore, this procedure does not constitute an unreviewed safety question per the criteria of ,
10CFR50.59. ;
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l EOP 3.1-23 Loss of Circulating Water Description of Chance Immediate Operator Action step 4.1 was expanded and a note prior to this step was added. In addition, the step pertaining to notifying the Duty Officer and filing a Plant Information Report (PIR) was deleted. Tlie expansion of step 4.1 clarified the intent of perfonning the immediate actions in E-0. This reduced the potential for confusion in performing the immediate actions in EOP 3.1-23.
The note added instructs the operator to perform EOP 3.1-23 in parallel with E-0 if the safety injection system is actuated or requireJ Reason for Chance This procedum was changed due to a procedure upgrade program.
Safety Evaluaion See Safety Evaluation for EOP 3.1-8.
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Number TitLq EOP 3.1-34 Complete Loss of Control Air Supply l
Descriotion of Chance 1 l
.Section five of this procedun: was changed to reflect new Emergency Operating Procedun: format.
Reason for Chr.nce This procedure was rewritten to the new format and to correct problems found during simulator validation.
Safety Evaluation The changes to EOP 3.1-34 impact the following systems during the loss of control air event; component cooling water, charging system, auxih,ary feedwater system, and the control air system (restoration o0 The changes involved reordering of steps with no functional change in the use of the systems to mitigate the loss of control air. These changes do not impact these systems during other events.
The changes did not affect the functionalintent of the procedure. The trordering of the steps and the new format were examined and do not increase the potential for operator error.
Therefore, this procedure does not constitute an unreviewed safetv question per the criteria of 10CFR50.59.
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I Number Tg).c EOP 3.1-46 Total Loss of Semi Vital Power Descriotion of Change Sections of this pmcedure were changed to reflect new Emergency Operating Procedure format.
Reason for Chance
- These sections were rewritten to the new format and to correct problems found during simulator !
validation.
Safety Evaluation The changes to EOP 3.1-46 impact the following systems during a loss of semi-vital power: charging system, feedwater system, auxiliary feedwater and the pressurizer heaters.
The changes to EOP 3.1-46 or failure modes associated with the changes do not impact the failure probability of safety systems. De changes involved the reformatting of the procedure steps with no functional change in the use of mitigating systems. The changes did not impact systems for other events.
Therefore, this procedure does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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4 Number Tida EOP 3.1-49 Partial Loss of DC Descriotion of Chance Sections four and five of this procedure were changed to reflect new Emergency Operating Procedure -
format.
Reason for Chance This procedure was rewritten to the new format and to correct problems found during simulator validation.
Safety Evaluation Since there were no hardware modifications associated with the changes, no hardware failure modes are associated with the change. The failure mode examined is the potential to improperly reset Safety Injecdon (SI) or failure to reset the signal due to the changes in the procedure. Other changes in the procedure were examined for potential increases in operator error.
The changes to EOP 3.1-49 did not impact the performance of the SI systems. SI reset using procedure ES-1.1 does not differ from the reset procedure provided in Appendix A of the current procedure. Sufficient cautions are provided to assure safety injection is not terminated when required to mitigate an event.
Therefore, this procedure does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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EOP 3.1-50 Loss of Motor Control Center MCC-5 l l
Descriotion of Chance Sections two, four and five of this procedure wem changed to reflect new Emergency Operating Procedum format. l l
Reason for Chance This procedure was rewritten to the new format and to correct problems found during simulator j validation. 1 Safety Evaluation The changes to EOP 3.1-50 or failure modes associated with the changes did not impact the failure probability of the systems. The changes involve the reordering of several steps with no functional change in the use of mitigatirg systems. The changes also did not impact the systems for other events. )
l The changes in order of actions did not impact the functional intent of this procedure.
Therefore, this procedure does not consetute an unreviewed safety question per the criteria of 10CFR50.59.
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Number - Ig[q AOP 3.2-23 Malfunction of Rod Control System Descriotion of Chance This change removed a refen:nce to procedum EOP 3.1-12. A step was added to emergency borate 200 ppm for each stuck rod.
Reason for Chance Procedure EOP 3.1-12 was cancelled. The step to emergency borate was added due to the cancellation of EOP 3.1-12.
Safety Evaluation See Safety Evaluation for procedure EOP 3.1-8.
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Number Title AOP 3.2-37 Cable Vault Fire Descriotion of Chance -l l
'Ihis change removed a reference to procedure EOP 3.1-12.
Reason for Chance The procedure was replaced by the new Emergency Operating Procedures.
Safety Evaluation See Safety Evaluation for procedure EOP 3.1-8.
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AOP 3.2-39 Containment Fim Descriotion of Chance This procedure was changed to remove incorrect mferences and correct typographical errors.
Reason for Chance The references did not comply with Emergency Response Guidelines.
Safety Evaluati.gn See Safety Evaluation for EOP 3.1-8.
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Number Title ES- 1,3 Transfer to Residual Heat Removal Recirculation Description of Chance his procedure is functionally the same as the pmvious revision. Changes were made to impmve clarity.
Reason for Chance This procedure was changed because it was cluttered and difficult to follow.
Safety Evaluation No new accident was created by the change. He change did not involve any hardware modifications nor did it affect the normal operating conditions of the plant.
The functionally equal new procedure provides for loss of coolant accident (LOCA) mitigation during the transfer to rectreulation and the rectreulation phase which is equal to the old procedures.
Derefore, this procedure does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Number Iille ES- 1.4 Transfer to Two Path Recirculation Description of Chance l
This procedure is functionally the same as the previous revision. Changes were made to improve I clarity.
Reason for Chance This procedure was changed because it was cluttered and difficult to follow. ,
Safety Evaluation See Safety Evaluation for procedure ES-1.3.
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Jumoer-Lifted Lead and Bvoass (J-LL-B) Chance Index J-LL-B Number Tnle 87-004 Nuclear Instrumentation Channel 32 Imad Runback 87-010 Inadequate Core Cooling (ICC) B level Indication 87-016 Temporary Liquid Radioactive Waste Demineralizer 87-025 Manipulator Crane Upender Drive Cimuit 87-026 Motor Control Operator (MCC) 2-4 Feeder Breaker DC Control Power 84 t
L J-LL-B Number 87-004 J LL-B Number 87-004, entitled "Nuclear Instrumentation (NIS) Channel 32 Imad Runback," is complete.
Descriotion of Chance An electrical jumper was installed on NIS Channel 32 load runback circuit. Thejumper has been removed.
Reason for Chance The bypass was necessary because the circuit was producing spurious signals which could have resulted in an undesirable inadvertent turbine load runback.
Safety Evaluation
- The only safety analysis which cmdits the operation of the cimuit being bypassed is the dropped rod. The latest rod drop analysis is documented in Cycle 5 modification of control 1od bank B.
This analysis was performed to detemdne the effect of reducing the size of control rod bank B.
The analysis took credit for the turbine load cutback and rod withdrawal block features, however, those features were assumed to be actuated by the rod bottom bistables. Therefore, this bypass did not affect the assumptions rnade in the safety analysis. -
This bypass did not disable any channels which provide protection for high nuclear flux. It only pmvented one of the four NIS channels from initiating a turbine runback and blocking automatic rod withdrawal. It cannot, of itself, cause a transient. The only event that this bypass could affect is a rod drop, which does not actuate any rod bottom Sistable and is not sensed by any of the other three operable nuclear channels. Because of the temporary natum of this change, this event is considered to be highly unlikely.
Although credited in the dropped rod safety analysis, the dropped rod protective circuits are not part of the Technical Specifications (TS). However, the TS permit one excom detector to be taken out of service. Since taking one excore detector out of service would have the same effect as this bypass on the dropped rod protection, no degradation in dropped rod protection exists beyond that allowed by TS. Since the plant administratively controls removing an excore detector from service :
which would take more than one dropped rod channel out of service, three of the dropped rod channels remain in service.
l Therefore, this bypass does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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l; J-LL-B Number 87-010 L
l J-LL-B Number 87-010, entitled "ICC-B Level Indication," remained open through 1987.
l Descriotion of Chance l
One water level sensor was removed from service in the ICC-B Train probe.
Reason for Change This bypass was necessary because the Number 8 heater shorted to the Number 8 level in the ICC-B Train causing false indication and alarms.
Safety Evaluation Removing one water level sensor from service in the B Train probe does not significantly degrade the ability to track water level inventory in the vessel. In this case, the bottom sensor (#8) was removed from service.
Technical Specifications permit removing up to two sensors from service in the plenum area.
Therefore, with only the #8 sensor out of service, the Train B water level measurement system is considered operable.
Lifted lead / jumper tags were hung on both lifted field leads from the heater and tags were hung on the new leads to the resistor.
Therefore, this bypass did not constitute an unreviewed safety question per the criteria of 10CFR50.59. ,
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J-LL-B Number 87-016 J-LL-B Number 87-016, entitled "Temporary Liquid Radioactive Waste Demineralizer," remained open through 1987.
Descriotion of Chance This bypass was initiated to install a mobile filter /demineralizer system. It is expected to remain in place for one or two operating cycles.
Reason for Chance This system was necessary to augment the existing waste liquid polishing demineralizer, which was unable to achieve the limiting chromate discharge limits on its own.
Safety Evaluation The hose run in the Primary Auxiliary Building (PAB) pipe trench does not mach the vicinity of the only safety-related electrical equipment in the pipe trench (solenoid valves and positioners for valves CH-FCV-110 and 110A). Flooding, due to hose or pipe breaks,is detected by level switches in the dmmming room sump, the east end of the PAB pipe trench, the west end of the PAB pipe trench and the Residual Heat Removal (RHR) pit which annunciate in the control room.
The pipe trench slopes toward the RHR pit. Oxrator action within 30 minutes of the flood alarm can isolate a hose ruptum before the pipe trenc:1 water level reached 8 inches or the RHR pit water level reaches 5 inches. No safety-related electrical equipment would be flooded at these levels.
Flooding in the Waste Disposal Building is not a concern due to the lack of safety related equipment. The outdoor resin slurry line is sleeved and an operator present during use. Yard storm drains in the vicinity of the resin sluny line are plugged during slurry operations.
Radwaste quality assurance components need not be seismic. However, hoses and cauipment are located in seisnue structures and draped over seismic supports. Floor loading (including seismic loads) was evaluated. There is no safety related equipment for mobile system equipment to fall on during a seismic event.
The dose from the worst credible failure associated with this change is approximately one order of magnitude lower than the Waste Gas Incident. Rus, the radiological protective boundary is not compromised.
Offsite dose is not increased for this category of accident. Therefore, the safety limitt are not challenged.
Therefore, this bypass does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i J-LL-B Number 87-025 J-LL-B Number 87-025, entitled "Manipulator Crane Upender Drive Circuit," remained open through 1987.
Description of Change This jumper removed Relay Trolley Centerline contact #3, trolley position limit switch #6, from the upender drive circuit.
Reason for Change t his jumper prevents manipulator crane upender movement which would normally be allowed .
when the trolley is not on the upender center line. i Safety Evaluation This contact allows upender movement when the trolley is not on the upender/ canal center line. By removing the relay contact from the circuit, upender movement is not allowed for any trolley position. Upender movement is only allowed when:
- 1) He manipulator crane is over the com, or
- 2) De gripper tube is up.
This configuration is more restrictive and is therefore safer than the way the system operates as described in design documents.
All other functions of limit switch #6 and relay trolley centerline were not affected by this temporary wiring change.
j Therefore, this bypass does not constitute an unreviewed safety question per the criteria of I a
i 10CFR50.59. i j
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J-LL-B Number 87-026 l J-LL-B Number 87-026, entitled "MCC 2-4 Feeder Breaker DC Control Power," is complete. l l
Descriotion of Changc He MCC 2-4 feeder breaker DC control power fuses wem removed and the breaker was closed. l This jumper has been removed.
Reason for Chance This change was necessary to allow power to MCC 2-4 during the time that the 27Y4 relay coil was burned out for the purpose of allowing spent fuel pool cooling and fuel handling in the spent 1 fuel pool. l l
Safety Evaluation The purpose of this change was to allow manual closure of this breaker with an undervoltage trip Hgnal present. The undervoltage trip signalis not an actual condition. It was pmsent due to failed relay coils in the undervoltage relays 27Y4 and 27Y5.
This change allowed re-energization of MCC 2 in order to re-establish spent fuel pool cooling.
With the DC fuses removed, the MCC feeder breaker did not have load shed shunt trip signal available. This was acceptable due to plant conditions:
- 1) All fuel was removed from the core.
- 2) The safety injection system was removed from service and '.mavailable.
Since the undervoltage protection system's purpose is to assure acceptable starting voltage for the safety injection pump and motors, then undervoltage tripping of safeguard loads was not required at this time.
Removal of the DC control power fuses also defeated the feeder bre&er overload protection trip.
His was acceptable based on the following:
- 1) Loading was minimized since only the spent fuel cooling system and spent fuel crane was re-energized from the MCC.
- 2) Any significant electrical malfunctions would be identified at the main contml board, at which point an operator would have tripped the bus breaker 4841,
- 3) ne period of time this bypass was in effect was only a matter of a few hours.
Herefore, this bypass did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setooint Change Index SCR Number Igig 87-P-006 Incorporate Design Data on Safety Valves Into Master Setpoint List 87-T-025 Incmase Torque Switch Setting on DH-MOV-310 1
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Setooint Chance Number 87-P-006 Setpoint Change Number 87-P-006, entitled "Incorporate Design Data on Safety Valves Into Master Setpoint List," is complete.
Descriotion of Chance The setpoints on MS-SV-1216A and MS-SV-1216B were changed. This change also added the following valves to the Master Setpoint List due to a change in Category I status: CA-RV-829, CA-RV-838A, CA-RV-838B, PU-RV-1844, DH-RV-1847, MS-SV-1216A, and MS-SV-1216B.
Reason for Chance This setpoint change request incorporated design data of safety valves CA-RV-829, CA RV-838 A, CA-RV-838B, PU-RV-1844, DH-RV-1847, MS SV-1216A, and MS-SV-1216B into the Master Setpoint List. !
Safety Evaluation All of these valves were determined to be safety-related, and incorporated into the Material, Equipment, and Parts List. These serpoints and tolerances are the original design setpoints, with the exception of the serpoint tolerance of MS SV-1216A/B and do not mpresent a change to plant design.
The setpoint tolerances for MS SV-1216A/B are in accordance with ASME III and ASME XI and are more conservative than the original design.
Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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1 Setooint Chanze Number 87-T-025 Setpoint Change Number 87-T-025, entitled "Increase Toque Switch Setting on DH-MOV-310," is complete. l Descriotion of Chance I
The torque switch setting on DH-MOV-310 was increased from 2.0 to 4.5.
1 Reason for Chance The change was necessary to allow the valve to achieve the proper seating force when electrically operated.
Safety Evaluation l l
It was determined that an incirase in the torque switch setting from 2.0 to a value of 4.5 is sufficient to seat the valve properly before the valve motor torques out. The force on the valve stem was calculated to be increased from the value of 3443 lbs, to a value of 8678 lbs. The structural evaluation of this valve determined that a stem force of 8800 lbs. would not overstress the valve. The change in torque switch setting therefom does not produce any mechanical damage to the valve, or reduce its .
seismic requirement. !
Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59. i l
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There were no tests performed under the provisions of Title 10, Code of Federal Regulations, Section 50.59 during 1987.
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Experiments There were no experiments performed under the provisions of Title 10, Code of Federal Regulations, Section 50.59 during 1987.
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- Challenges to Relief Valves In accordance with the commitment made under Item II.K.3.3 of NUREG 0737 (TMI Action Plan) in the W. G. Counsil letter to D. G. Eisenhut, dated June 10,1980, the following is a report of challenges to Relief / Safety Valves during 1987.
Here were no challenges made to the Primarf Relief/ Safety Valves in 1987.
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ri l) l MILLSTONE UNIT 1 1
U CONTENTS SECTION ,PAGE Changes 1
Design Changes .
Design Change Evaluations . . . . . . 120 Procedure Changes . . . . . . . . . 125 Jumper-Lifted Lead-Bypass Changes . . . 126 Setpoint Changes . . . . . . . . . 133 i Tests . . . . . . . . . . . . . . 159 Experiments . . . . . . . . . . . . 162 ,
i i Occupational Radiation Exposure . . . . . . 163 Challenges to Relief Valves . . . . . . . 164 1
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' PLANT DESIGN CHANGE INDEX i
PDCR Number -Title 1-074-82 Replacement.of Reactor Building Solenoid l Valves 1-008-83 Automatic Pressure Relief / Safety Relief f Temperature Indication Transfer 1-022-83 Core Spray Auto Restart Switch Removal ;
1-055-33 Replacement of Gas Turbine Governor Motor ;
1-086-83 Instrument and Station Air Compressor Oil .
Pressure Alarms
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1-007-84 Revision 1 Process Computer Facility'- Conversion Part 1 1-045-84 CR 2820 Relay Replacement !
1-048-84 0xygen Analyzer - Lover Level Heating and !
- Ventilation Room 1-088-84 Feedvater Heater Hi-Hi Level Protection l 1-016-85 Power Feed to MCC EF-7 Changed from 480V 12E ;
to 12F l
- 1-037-85 Turbine Building Equipment Drain Sump Pump j
- Counter l 1-041-85 1-RR-31 Valve Removal
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- 1-067-85 Revision 1 Solid Radvaste Building Ventilation and Roof ,
Modifications -
l 1-069-83 Eliminate RBCCV Heat Exchanger Vent Floats I
- 1-079-65 EEQ MOV Modifications i 1-097-85 Diesel Air Start Systen Modification 1-111-85 Service' Vater Piping Replacement i 1-123-85 15G-31S Bus Insulation 1-134-85 Unit 1 Turbine Deck Security Alarm i
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PLANT DESIGN CHANGE INDEX (Continued) q PDCR Number Title 1
1-137-85 Emergency Gas Turbine Vibration Monitor Replacement 1-148-85 Isolation Transfer Svitches and Post Fire Shutdovn capability 1-149-85 MP1 Roof Replacement 1-009-86 Blockhouse Feeder Upgrade 1-011-86 Reactor Building Crane Mode 2 Lcgic Change 1-013-86 Vacuum Breaker Alarm Switch Modification .
1-015-86 Revision 1 'B' Service Water Pump Lubricating Vater Modification 1-022-86 Revision 1 Sullair Control System Modification 1-023-86 MP1 Reactor Protection Trip System 1-024-86 Reactor Building Equipment Drain Tank Pump Running Time Counter 1-026-86 MP1 Appendix R Control Room Fire Isolation ~
Switch Panel i 1-030-86 Airvash System Piping Replacement 1-032-86 Replace Isolation Condenser LeJ'el Transmitter 1-043-86 Level Indicator For Spent Resin Task 1-047-86 PAM 101 (H 220 Analyzer) Vacuum Breaker 1-050-86 Revision 1 MP1 Maintenance Veld Shop Trolley / Monorail '
1-053-86 Revision 1 LLRT Connection For 1-CU-2, 2a 1-056-86 Seal Vater to 'C' and 'D' Service Vater Pump i Modifications 1-058-86 Process Computer Replacement - Nonvital Point 1 Transfer {
1-064-86 MP1 Reactor Vater Cleanup System Setpoint l Reduction 1-065-86 House Heating Boiler Stack Emissions Monitor Replacement PLANT DESIGN CHANGE INDEX (Continued)
PDCR Number Title 1-066-86 Xenon-Krypton System Two Dryer Operation 1-067-86 Monorail and Lifting Padeyes - Emergency Diesel Enclosure 1-068-86 Hydrogen Vater Chemistry Sample Tap 1-069-86 Millstone Unit 1/2 Appendix R Backfeed, Pre-Outage Vork 1-070-86 Sprinkler Design for Maintenance Shop Prefab Building and Relocation of Hose Station Supply Line 1-071-86 1-AS-13A/13B Replacement of 10" Rockwell Butterfly Valves 1-075-86 Head Vent Isolation Valve 1-076-86 Installation of Telephone and Maintenance Jacks at North Vall of Turbine Building 1-077-86 Maintenance Building Air Conditioner Pover Supply 1-078-86 Stack Flov Honitor Replacement 1-081-86 HVAC-4, Switch Gear Area Ventilation A/C Unit 1-082-86 PEECo Flov Svitch Removal - RBCCV to Dryvell Sump 1-084-86 Diesel Fire Pump Cooling Syster Modification 1-085-86 Main Steam Line Trap Bypass Valve Indication 1-087-86 CRD Pressure /Flov Instrumentation Piping Modification 1-088-86 Process Computer Replacement - Vital Point Transfer j 1-089-86 Dryvell Nitrogen Compressor Dryer Tover i Svitching Failure Alarm l l
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PLANT DESIGN CHANGE INDEX (Continued)
PDCR Number Title 1-090-86 Gas Turbine Fuel Shutoff Valve Logic Modification 1-092-86 Process Computer Replacement - New (SPDS)
Input Signals 1-094-86 Process Computer Replacement - RPV Vater Level Instrument 1-095-86 Radvaste Piping Modification to Support Chem-Nuclear FTDS System
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1-096-86 Contingency Fuel Storage Rack 1-098-86 Millstone Unit 1/2 Appendix R Backfeed Outage Vork 1-099-86 Lighting and Re aptacles for I6C Hot Shop 1-100-86 Solid Radvaste Building Ventilation and Roof Modifications 1-101-86 Installation of Grease Relief Kits to Limitorque Operators 1-102-86 Control Roo:a Halon System 1-103-86 Main Transformer and Deluge System Replacement 1-104-86 Dryvell Compressor Instrument Filters 1-105-86 High Radiation Area Gate Alarms and Varning Lights 1-106-86 Rod Vorth Minimizer Replacement 1-108-86 Appendix R Outage Related Emergency Lighting 1-002-87 Main Transformer Remote Annunciation i 1-004-87 Scram Discharge Vent Line Modification -
South Header 1-005-87 CST Fill Line Modification 1-006-87 Jet Pump Instrumentation Nozzle Assembly j Replacement l
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I PLANT DESIGN CHANGE INDEX (Continued)
PDCR Number Title 1-020-87 Revision 1 Process Computer Replacement Hardware Installation :
1-021-87 Stack Radiation Monitor - Millstone Unit 2 Interface 1-022-87 Hydrogen Vater Chemistry Recircul'ation Vater Sample Tubing 1-023-87 CRD Pump Piping Modification for Alternate Cooling
~
1-026-87 Millstone Unit l' App. R and SEP FVCI Area Ventilation Modifications 1-029-87 Replace LPCI Containment Cooling Timers Logic Time Delay Relays 1-030-87 Replace Overcurrent Trip Devices - Vital 480V Breakers I 1-032-87 MP1 Dryvell Temperature Monitoring System 1-033-87 N2 Compressor Flovmeter Relocation 1-037-87 Replace Gas Turbine Generator Neutral Grounding Transformer 1-038-87 Power Cold Shutdown Equipment Following :
1 Appendix "R" Fires 1-040-87 Appendix R Firecoating of Structural Steel 1-041-87 Millstone Unit No. 1 Appendix "R" Cable Fire f Vrap Program and MI Cable Substitution-and ,
Reroute :
1-042-87 Installation of Excess Flov Valve / Bypass -
l Valve Assembly in Hydrogen Supply System 1-047-87 Replacement of Core Spray Valves CS 21A and --
- 21B 4
1-049-87 Hydrogen Vater Chemistry Preimplementation Test 1-051-87 Appendix R, Alternate Cooling of Shutdown
, Cooling /Dryvell Cooling
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PLANT DESIGN CHANGE INDEX (Continued)
PDCR Number Title 1-052-87 Chemical Decontamination of the Recirculation-and'RVCU Systems 1-053-87 RBCCV Essential. Header Stop Valve 1-RC-15 1-054-87 GEZIP Demonstration Test 1-055-87 Revision 1 Millstone Unit 1 Lov Pressure B Rotor Replacement 1-056-87 Emergency Diesel Generator Fuel Oil Piping Modification 1-058-87 MP1 - Isolation Condenser Steam Trap Reroute - - -
1-060-87 Conduit Duct Bank for Flanders Line 1-064-87 Reload 11 Core 1-066-87 Extraction Steam Piping Replacement 1-067-87 Blade Guide Storage Racks 1-068-87 EEQ MOV Modifications-1-073-87 Appendix R - Auxiliary. Boiler Blast Modifications 1-074-87 Diesel Generator Room Structural Steel Firecoating 1-075-87 LPCI and Core Spray Pump Anchorage Modification 1-076-87 Emergency Diesel Generator, Air Roll-Modification and Indicator Valve Installation 1-077-87 Suppression Chamber Air Temperature Indication 1-080-87 Main Steam Line Support Beam Modification 1-081-87 SRV Pressure Switch Replacement 1-084-87 Electrical Pressure Regulator Operator Interface Modifications 1-086-87 House Heating Boiler Stack Replacement e
i Plant Design Change Number 1-074-82 l
Plant Design Change Number 1-074-82 entitled "Replacement of Reactor-
-Building Solenoid Valves" is complete..
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~ Description of Change The solenoids on twenty-nine valves located in the Reactor Building, H i Steam Tunnel and Dryvell vere' replaced with seismically and environ-mentally qualified solenoids.
Reason for Change The original equipment was found to have insufficient documentation to-ensure seismic and environmental qualification. The solenoids vere
] replaced in order to comply with Equipment Environmental, Qualification.
4 submittals.- ~
t i .m-i Safety Evaluation ,
J The solenoids were a direct one-for-one replacement. The margin of ,
safety was increased due to the fact that the_new components are 4 seismically and environmentally qualified thus reducing the probabil-1 ity of malfunction in the event of exposure to a harsh environment or '
a seismic event. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. i 1
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Plant Design Change Number 1-008 Plant De' sign Change Number 1-008-83 entitled "Automatic Pressure Relief / Safety Relief Temperature Indication Transfer" is complete.
Description of Change The tailpipe temperature recording. points, for the automatic pressure relief and safety relief valves, vere relocated to the-front panel.in the Control Room adjacent to the control switches and high temperature alarms for the relief valves.
Reason for Change The modification was implemented to-facilitate operational monitoring by having the recorders, alarms, and controls together on the same panel. The modification satisfies human factors criteria.
Safety Evaluation The recorder used for this modification does not provide any automatic functions to initiate or otherwise affect safety-related systems.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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l Plant Design Change Number 1-022-83 Plant Design Change Number 1-022-83 entitled "Core Spray Auto Restart Switch Removal" is complete.
Description of Change The Core Spray auto restart push button switches were removed from Control Room Panel 903.
Reason for Change The switches vere redundant to the Core Spray start /stop switches.
Safety Evaluation The switches vere previously disconnected from the Core Spray system logic and performed no function. Therefore, this modification does not constitute an unreviewed safety question per the criteria of -
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Plant Design Change Number 1-055-83
- Plant Design Change Number 1-055-83 entitled "Replacement of Gas Turbine Governor' Motor" is complete.
Description of Change 14 new gas turbine governor motor was installed and seismically mounted with fuse protection added.
Reason for Change The old motor, which was not seismically mounted, had failed due to ;
high currents.
Safety Evaluation The new motor was an exact replacement in the critical parameters such as voltage, horsepover, and speed. The new motor was mounted in accordance with approved seismic standards and independent motor fusing vas added to prevent overcurrent conditions. Therefore, this modification does not constitute ar. unreviewed safety question per the i
criteria of 10CFR50.59. '
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Plant Design Change Number 1-086-83 Plant Design Change Number 1-086-83 entitled "Instrument and Station Air Compressor Oil Pressure Alarms" is complete.
Description of Change The lov oil pressure alarm circuitry on the instrument air compressor ,
and the station air compressor, was modified so that it vill only'
' annunciate when the compressor is running. The additional stage, on ,
the control switch, blocks the alarm when the compressor is in standby.
Reason for Change The modification vas implemented to eliminate the constant lov oil pressure alarm and annunciation when the compressor was in standby, not running. l Safety Evaluation The modification of the air compressor oil pressure alarm circuitry does not adversely impact any reactor safety functions or safety-related components. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-007-84 Revision 1 Plant Design Change Number 1-007-84 Revision 1 entitled "Process Computer Facility - Conversion Part 1" is complete.
Description of Change A new vall and raised floor system including associated conduits,'
j pipes, ducts, vall boxes, panel boards and imbedded grounding equipment was installed in the Computer Room. Also a new Heating
. Ventilation and Air Conditioning system and a Fire Protection System
) etre installed.
4 Reason for Change The modification was implemented to support the replacement of the process computer and the implementation of the Safety Parameter Display System. ,
a Safety Evaluation The power supply and signal cables vere installed to comply with 10CFR50, Appendix R requirements and the new valls vere constructed to >
meet seismic criteria. There vere no connections to safety-related
- power supplies. Therefore, this modification does not constitute an i unrevieved safety question per the criteria of 10CFR50.59. ;
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, Plant Design Change Number 1-045-84 Plant Design Change Number 1-045-84 entitled "CR 2820 Relay Replace-ment" is complete. ,
Description of Change j The time delay portion of the General' Electric CR 2820 125 VDC time .
-delay pick-up relays in the Core Spray, Auto Blovdown, Diesel, and Lov I Pressure Coolant Injection systems-vere replaced.
Reason for Change *
- The time-delay portion of the relays vere replaced with more reliable components because the old relays, when deenergized for more than a month,.could vary in the time delay by more than the accuracy speci. !
fication.
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Safety Evaluation ,
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The new relays are not subject to excessive timer setpoint drift after prolonged deenergization so consequently enhance-the reliability of ,
the systems. In addition they were mounted in accordance with seismic l requirements. Therefore, this modification does not constitute an i unreviewed safety question per the criteria of 10CFR50.59. l i
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l Plant Design Change Number 1-048-84 Plant Design Change Number 1-048-84 entitled'"0xygen Analyzer - Lover I
Level Heating and Ventilation Room" is complete.
Description of Change i
An oxygen analyzer consisting of a sensor mounted near the air con-ditioning compressor units, indicating panel mounted at the top of the stairs and a tie in to the master alarm in the-control Room, was ,
installed in the lover level heating and ventilation room.
Reason for Change The modification was implemented to varn' station personnel, prior to entering, if oxygen level is lov An the room.
Safety Evaluation The oxygen analyzer enhances the personnel safety in the area of the air conditioning units and does not affect any safety-related equipment. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i Plant Design Change Number 1-088-84 Plant Design Change Number 1-088-84 entitled "Feedvater Heater Hi-Hi Level Protection" is complete.
Description of Change Three-vay solenoid valves were installed on the instrument air lines between the level controllers and valve positioners'for the high pressure, high intermediate pressure and intermediate pressure normal drain valves. The solenoids are normally energized to allow normal control air to the valve positioners. On Hi-Hi level, the solenoid vill deenergize to block control air and supply approximately twenty pounds of instrument air pressute to close the normal drain valve.
Reason for Change The modification was implemented to provide increased protection --
against turbine water induction, by isolating cascading drain flov to a heater which is experiencing Hi-Hi level.
Safety Evaluation The modification does not adversely affect normal system operation or any safety-related systems. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-016-85 !
Plant Design Change Number 1-016-85 entitled "Pover Feed to MCC EF-7 Changed from 480V 12E to 12F" is complete.
I Description of Change The incoming power feed to motor control center EF-7 was changed from 480 volt bus 12E to 480 volt bus 12F.
Reason for Change This modification increases the reliability of the diesel generator in the event of gas turbine failure of the automatic bus transfer mechanism.
i Safety Evaluation 3
The' ability of the diesel generator to supply its own vital loads, '
during a loss of normal power, increases the reliability of the
- system. Since the diesel generator is required to accept loads within ten seconds of a start signal, the modification prevents a failure of the automatic bus transfer switch from causing the auxiliaries to remain deenergized until the gas turbine generator is ready to accept the load. Two separate power supplies remain available, following a loss of normal power, to accept the loads associated with the diesel ,
generator. Therefore, this modification does not constitute an
] unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-037-85 Plant Design Change Number 1-037-85 entitled "Turbine Building Equipment Drain Sump Pump Counter" is complete.
Description of Change An elapsed time meter was installed on'the Turbine Building equipment !
drain sump pump and connected to the running, red indicator light on f 1 the radvaste control panel.
Reason for Change The modification was implemented to provide an effective means'of 4
measuring the running time hours of the pump which could be an ,
indication of excessive leakage.
I Safety Evaluation The modification does not. adversely affect any safety-related systems .
or_ components. Therefore, this modification does not constitute an !
unreviewed safety question per the criteria of 10CFR50.59. ;
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Plant Design Change Number 1-041-85 Plant Design Change Number 1-041-85 entitled "1-RR-31 Valve Removal" is complete.
Description of Change A temporarily installed, one inch valve (1-RR-31) vas removed from an instrument line in the dryvell and replaced with stainless steel pipe and couplings.
Reason for Change The valve was removed to eliminate a possible leak source in the dryvell if the valve should fail.
Safety Evaluation Any failure of the new spool piece vill be contained within the -
dryvell and has been previously analyzed. The removal of a lumped weight from the system reduced pipe stresses and support loads in regards to dead veight and seismic conditions. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
Plant Design Change Number 1-067-85 Revision 1
- Plant Design Change Number 1-067-85 Revision 1 entitled "Solid Radvaste Building Ventilation and Roof Modifications" is complete.
Description of Change A corrugated steel building was constructed over the solid radvaste ,
building roof to form a veather tight enclosure around the roof .
exhaust fans.
Reason for Change The modification was implemented to prevent the potential for an unmonitored radioactive release from the contaminated roof and to reduce the strain on the liquid radvaste system that has to process the large volumes of rainvater.
Safety Evaluation The new structure is not safety related and its only interface is with the plant storm drain system. It is not located in, around, or over any safety-related components. Therefore, this modification does not constitute an unrevieved safety question per the criteria of ,
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k j Plant Design Change Number 1-069-85 Plant Design Change Number 1-069-85 entitled "Eliminate RBCCV Heat Exchanger Vent Floats" is complete. ,
Description of Change i i
The float traps in the ccustant vent line from the Reactor Building ;
Closed-Cooling Vater heat exchangers were removed and-replaced with a l globe valve and a rotometer to provide manual adjustment. i Reason for Change The modification was implemented to increase the heat exchanger flow and heat capacity by eliminating the need to partially throttle the service water discharge valve in order to keep the heat exchanger
! tubes covered.
Safety Evaluation i
i The modification does not adversely affect any safety-related systems- i or components. Therefore, this modification does not constitute an i unreviewed safety question per the criteria of 10CFR50.59. i t
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Plant Design Change Number 1-079-85 ,
Plant Design Change Number 1-079-85 entitled "EEQ MOV Modifications"- ;
is complete.
Description of Change The motor operators were replaced on valves 1-RR-2A and 1-RR-25, which are installed on the discharge lines of the Recirculation Pumps.
Reason for Change The original motor operators were not environmentally qualified by !
10CFR50.49.G standards. The new equipment ensures harsh environment l operability of the valves needed to mitigate the consequences of a
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loss of coolant accident.
Safety Evaluation I
In addition to being seismically and environmentally qualified, the l new motor operators vere reviewed to ensure that no potential for i pressure boundary breach exists. All design criteria such as required valve closing times, structural integrity, pressure and temperature -
- vere addressed. Therefore, this modification does not constitute an ;
unreviewed safety question per the criteria of 10CFR50.59. {
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f Plant Design Change Number 1-097-85 ;
Plant Design Change Number 1-097-85 entitled "Diesel Air Start System Modification" is complete.
Description of Change The solenoid operated main air start valves, on the emergency diesel .
generator, were replaced with Grove Flexflow relay valves and solenoid j actuated control systems. .
Reason for Change The new equipment was installed to enhance the reliability of the- .
emergency diesel generator unit by minimizing the effect of corrosion ;
i products in the air start system. ;
t Safety Evaluation The modification does not affect the intended function of the main air i start system and it vill activate via the some diesel generator start signal and operate at the same air pressure. -Therefore, this modifi-cation does not constitute an unrevieved safety question per the ;
criteria of 10CFR50.59. !
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Plant Design Change Number 1-111-85 :
Plant Design Change Number 1-111-85 entitled "Service Vater Piping Replacement" is complete.
Description of Change !
f The deteriorated piping on the service vater blovdown lines, emergency ;
4 service water keeptill lines, and service water to chlorine booster ;
. pump-line vas-removed and replaced. The service water isolation valve ,
to the chlorine booster pump was relocated..
l j Reason for Change <
l The modification was implemented to increase Service Water system reliability by replacing the corroded piping and by providing a means i of isolating the service water to chlorine booster pump piping during l normal operation. -
Safety Evaluation j 1 The piping replacement offers improved corrosion resistance, satisfies ;
seismic considerations, and does not change the intended function or i jeopardize the design basis of the service water system. Therefore, i
this modification does not constitute-an unreviewed safety question !
- per the criteria of 10CFR50.59.
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Plant Design Change Number 1-128-85 Plant Design Change Number 1-128-85 entitled "15G-31S Bus Insulation" is complete.
Description of Change Heat shrinkable bus bar insulation tubing vas installed on the exposed bus bars on the lov side of the Emergency Station Service transformer.
Reason for Change The modification was implemented to eliminate a potential source for electrical faults which could render the transformer inoperable.
Safety Evaluation Ultimately, the modification prevents the loss of an independent-offsite power source to the plant. The modification increases -
transformer reliability and system availability and does not adversely affect any safety-related system or components. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
Plant Design Change Number 1-134-85 Plant Design Change Number 1-134-85 entitled "Unit 1 Turbine Deck Security Alarm" is complete.
Description of Change -l A security alarm was installed on the turbine deck.
Reason for Change .
f The alarm switches used in this modification are a proven method to meet the requirements of 10CFR73.55 and reduce the labor cost of- i posting security guards.
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l' acceptable. Therefore, this modification does not constitute an ,
unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-137-85 Plant Design Change Number 1-137-85 entitled "Emergency Gas Turbine Vibration Monitor Replacement" is complete.
Description of Change l
The vibration monitors and velocity transducers vere replaced with a new power supply, relay module and three individual vibration monitors. The new vibration monitoring system was seismically mounted in the Gas Turbine Control House.
Reason for Change l
l The new system vas installed to increase the reliability, diagnostic i
capability and maintainability of the Emergency Gas Turbine vibration l
system by replacing the obsolete system for which spare parts were unavailable.
Safety Evaluation The new system retains the previously established alarm and tripping l
i limits, and annunciation and control interfaces. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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! Plant Design Change Number 1-148-85 entitled "Isolation Transfer
- Switches and Post Fire Shutdown capability" is complete.
Description of Change [
l Redundant fuses were installed on the Control Room Isolation Switches l l-IC-1, 1-10-2, and 1-IC-4. -;
F Reason for_ Change j The redundant fuses were installed to enable operation of the switches j should the normal fuses blov due to external damage. This change was implemented to comply with the requirements of 10CFR50, Appendix R.
Safety Evaluation h
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j control circuits under normal conditions and does not alter the i l configuration or functional logic of the systems or components. The i j modification increases the reliability of the affected circuits under .
post-fire conditions. Therefore, this modification does not consti-1 tute an unreviewed safety question per the criteria of 10CFR50.59. ;
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Plant Design Change Number 1-149-85 entitled "MP1 Roof Replacement" is j complete. ,
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Description of Change The built-up roofing system on the control Room, Turbine Building and Maintenance Shop vas removed and replaced with a new single ply, fully >
4 adhered, ethylene propylene diene monomer (EPDM) membrane roofing system. '
i Reason for Change
' i The modification was implemented because the old roofing system had ;
deteriorated to the point where water had permeated through and soaked ,
the insulation, Safety Evaluation ,
The ncy roofing system is lighter in veight than the built-up system ,
that was removed so it does not have an adverse affect on the '
- structural or seismic integrity of the roof deck. Local failure of
- the roof would not result in loss of protection to safety-related l equipment or capability of safe shutdown. Therefore, this modifi- ,
. cation does not constitute an unreviewed safety question per the !
criteria of 10CFR50.59. !
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- Plant Design Change Number 1-009 86 i
Plant Design Change Number 1-009-86 entitled "Blockhouse Feeder Upgrade" is complete.
Description of Change The size of the electrical supply to a fabrication shop vas increased from 200 amps to 350 amps.
Reason for Change The modification was implemtnted to accommodate increased power demands in the fabrication shop.
Safety Evaluation The building involved in the modification is not related to any plant safety systems or equipment. The switchgear panel, breaker sizing and -
cables were analyzed to be adequate for the additional current.
Therefore, this modification does not constitute an unteviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-011-86 Plant Design Change Number 1-011-86 entitled "Reactor Building Crane i j Mode 2 Logic Change" is complete. ;
)> Description of Change
! -The Reactor Building crane, Mode 2, logic circuitry was modified to !
l allow the crane and trolley to move in the direction opposite that of
- l. a tripped restricted zone limit switch.
a Reason for Change h (,
The modification was implemented to eliminate the need to place the !
, mode switch in Mode 1 in order to clear a tripped limit switch which . !
- defeats the purpose of the Mode 2 restricted zone feature. -
l Safety Evaluation The modification does not' defeat the features of restricted zone f j movement and does not adversely affect safety equipment. Therefore, '
i this modification does not constitute an unreviewed safety question '
) per the criteria of 10CFR50.59. !
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, Description of Change l
The adjustment setscrews on'the vacuum breaker alarm microsvitches, l vere replaced with setserevs that have a flat surface on the contact
.l end and double nuts were installed to eliminate the possibility of ;
setpoint drift.
l j Reason for Change This modification was implemented to increase the reliability and i reproductivity of the calibration data. !
Safety Evaluation '
r The new setscrews and nuts were installed to meet all the system f design requirements. Therefore, this modification does not constitute- i-an unreviewed safety question per the criteria of 10CTR50.59. f r
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Plant Design Change Number 1-015-86 Revision 1 Plant Design Change Number-1-015-86 Revision 1 entitled "'B' Service Vater Pump Lubricating Vater Modification" is complete.
Description of Change The service water pump lubricating vater system was modified by increasing the size of the upper bearings' supply port / fitting and adding a second supply port / fitting to the upper bearings. Individual flovmeters and throttle valves vere added to the upper and lover bearing supply lines. Revision 1 of this design change was implemented to add pump A modifications to the scope.
Reason for Change
-The modification was implemented to minimize the potential for loss of.
lubricating vater due to fouling of the supply lines and to provide more accurate distribution of lubricating water to the upper and lover -
bearings.
Safety Evaluation The modification enhances the reliability of the service vater pumps and does not affect any safety-related systems or structutes.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-022-86 Revision 1 Plant Design Change Number 1-022-86 Revision 1 entitled "Sullair Control System Modification" is complete.
Description of Change The pressure switch on the station backup air compressor was replaced with two mercoid switches. One of the nev~ switches operates the start feature of the compressor while in the standby mode and the other switch operates the blovdown valve. A new time delay dropout relay was installed and vired to allow the start feature from the standby mode to be sealed in.
Reason for Change The modification was implemented to provide the backup compressor with a standby feature which allows it to automatically start if the air header pressure drops below a set point. -
Safety Evaluation The backup station air compressor control system is not safety related and the modification does not impact any safety related system or components. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. ,
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I Plant Design Change Number 1-023-86 Plant Design Change Number 1-023-86 entitled "HP1 Reactor Protection )
Trip System" is complete. j Description of Change The Automatic Depressurization System timers were replaced with a new solid state timer. The sensor that provides the low reactor _ pressure permissive as part of the Emergency Core Cooling System pump start :
I logic was removed.
i Reason for Ch'ange The timers were replaced to reduce setpoint drift problems. The low reactor pressure permissives were removed to allow the initiation of the Emergency Core Cooling System pumps whenever a Hi Dryvell Pressure or Low-Lov Vater Level condition occurs. .
Safety Evaluation ,
The one-for-one replacement of_the Automatic Depressurization System timers with new, more accurate timers has no impact on the design basis analyses since the function is identical.
The change to allow Emergency Core Cooling System pumps to start on either Lov-Lov Vater Level or High Dryvell Pressure mskes the pump start logic consistent with the Automatic Depressurization System Logic. This change decreases the probability of a malfunction of equipment important to safety since there is one less component that could have a setpoint drift or fail. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-024-86 Plant Design Change Number 1-024-86 entitled "Reactor Building
' Equipment-Drain-Tank-Pump < Running Time Counter" is complete.
Description of Change -
An elapsed time meter was installed on the radvaste control panel and connected to the running indicator light for-the Reactor Building equipment drain tank pump.
Reason for Change The modification was implemented to provide a method of monitoring and recording the running hours of the pump as a means of identifying abnormal leakage.
Safety Evaluation _.
The modification does not interact with any safety-related equipment.
The modification does not affect-any structural or seismic-require-ments. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-026-86 Plant Design Change Number 1-026-86 entitled "MP1 Appendix R Control Room Fire-Isolation Switch Panel" is complete.
Description of Change l
Control switches to provide isolation capability for the four in-board l Main Steam Isolation Valves, the six Automatic Depressurization System-Valves and the Clean Up System Pressure Control Valve vere installed in a new switch enclosure in the Control Room. l Reason for Change This modification was implemented to provide a means of overriding spurious operation of the valves during a Control Room fire thereby ensuring valve closures.
Safety Evaluation -
Since the design capacity of the vall and floor is greater than the design force to be resisted, the installation of the switch enclosure is not a detriment to the structural integrity of the Control Room.
The switches vill allow the operator to isolate the reactor vessel by preventing spurious opening of valves. The modification does not adversely affect the capability of the' Main Steam Isolation Valves closure or the Reactor Vater' Clean Up system isolation function to provide containment isolation. The modification does not affect the pressure control capability of the Safety Relief Valves. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
Plant Design Change Number 1-030-86 Plant Design Change Number 1-030-86 entitled "Airvash System Piping Replacement" is complete.
Deceription of Change The Airvash System piping and tank liners were removed and replaced with stainless steel piping and fittings.
Reason for Change The modification was implemented to restore-system integrity. The piping that was removed was corroded to the point that the system was rendered inoperable.
Safety Evaluation The modification exceeds the original design criteria and enhances'the -
system integrity. The airvash system is not safety related and its failure vill not affect any safety-related equipment. Therefore,-this modification does not constitute an unreviewed safety question per the.
criteria of 10CFR50.59.
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Plant Design Change Number 1-032-86' Plant Design Change Number 1-032-86 entitled "Replace Isolation Condenser Level Transmitter" is complete.
Description of Change The Isolation Condenser Level transmitter was-replaced with a differential pressure cell transmitter with remote diaphragm seals.
The Isolation Condenser level gauge was replaced with a differential pressure level gauge.
Reason for Change The modification was implemented to provide accurate level indication during all modes of operation.
Safety Evaluation The nev transmitter has the same power input and output ratings and-the level gauge meets 10CFR50 Appendix R requirements. The modifica-tion improves the accuracy of the level management and does not alter the system deriC n criteria. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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l Plant Design Change Number 1-043-86 l
Plant Design Change Number 1-043-86 entitled "Level Indicator For Spent Resin Tank" is complete. .i Description of Change The level indication switch and transmitter for the spent resin tank was removed and replaced with a microprocessor and a capacitance probe.
Reason for Change The modification was implemented to provide a more accurate and reliable level indicating system with annunciation capability for high tank level.
Safety Evaluation The modification does not jeopardize the intended _ system function since it provides the same features as the original design. The modification does not adversely affect safety-related systems or components. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i Plant Design Change Number 1-047-86 Plant Design Change Number 1-047-86 entitled "PAM 101 (H 220 Analyzer)
Vacuum Breaker"'is complete.
Description of Change
.A bypass check valve was installed in the Hydrogen /0xygen Analyzer sample tubing.
Reason for Change The modification was implemented to prevent damage to the analyzer's internal pressure regulators resulting from-a plant scram and subsequent Group II isolation logic actuation.
Safety Evaluation The valve was installed outboard of the containment isolation-valves, so containment isolation capability is not degraded. The addition of the valve increases the reliability of the analyzer and does not change the intended function. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-050-86 Revision'l Plant Design Change Number 1-050-86 Revision 1 entitled "MP1 Maintenance Veld Shop Trolley / Monorail" is complete.
Description of Change A monorail with an electric vinch, capable of lifting two tons, was installed in the veld area of the Maintenance Shop.
Reason for Change The modification was implemented to provide a means of handling heavy -
material in the velding shop. There vere no provisions for this previously.
Safety Evaluation The monorail is not attached to any safety-related structures and does not affect the structural or seismic capability of any safety-related component. Therefore, this modification does not constitute an unreviewed safety question per the c*;iteria of 10CFR50.59.
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l Plant Design Change Number 1-053-86 Revision 1 Plant Design Change Number 1-053-86 Revision 1 entitled "LLRT Connection ~For 1-CU-2, 2a" is complete.
Description-of Change A test connection was installed in'the clean Up System.-
Reason for Change The modification was implemented to provide an accurate means of performing a Local Leak Rate test on the Clean Up System valves.
Safety Evaluation The valves and associated piping vere seismically mounted and any leakage due to the test connection failure vill be contained inside the dryvell. -Therefore, this modification does not constitute an --
unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-056-86 l l
Plant Design Change Number 1-056-86 entitled "Seal ~ Vater to 'C' and !
'D'. Service Water Pump.Hodifications" is complete. l
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Description of Change A ball valve and associated piping and fittings was installed to form an equalization line around the duplex strainers.
Reason for Char.ge The modification was implemented to reduce the vear experienced by the plug valve, due to high differential pressure, when the strainer baskets are removed from service for cleaning.
Safety Evaluation A failure to the pressure boundary of the equalizer valves vill not . - -
affect service vater pump operation. The installation was analyzed to be seismically acceptable and does not jeopacdize the intended function of the system. The reliability of the service water. system is enhanced. Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59, 1
Plant Design Change Number 1-058-86 Plant Design Change Number 1-058-86 entitled "Process Computer l Replacement - Nonvital Point Transfer" is complete, j Description of Change A total of three-hundred twenty-seven (327) nonvital computer points, both analog and digital, vere removed from the old process computer and transferred to the new Integrated Computer System. Early point transfer was performed to minimize the number of points to be transferred during the outage and to gain experience at transferring computer inputs.
Reason for Change The modification was implemented to support the computer _ replacement during the refueling outage, and the early nonvital point transfer helped reduce congestion vithin the input cabinets and cable trays, ,
thus expediting the outage related work.
Safety Evaluation All the points transferred were reviewed and found to be nonvital to the operation of the plant. The points originated from nonsafety-related instrument loops and are not teeuired for automatic control of plant systems. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-064-86 Plant Design Change Number 1-064-86 entitled "MP1 Reactor Vater Cleanup System Setpoint Reduction" is complete.
Description of Change The Reactor Vater Cleanup system isolation setpoint was changed from '
the low reactor water level signal to.the low-low reactor water level signal.
Reason for Change Subsequent to all plant trips, the Reactor Vater Cleanup system inadvertently isolates due to collapsing voids. The setpoint modification was implemented to make the Reactor Vater Cleanup system available to the operators following plant trips, thereby making plant recovery less difficult. _
Safety Evaluation The modification involved only reviring within the Control Room panels and did not alter the normal state of existing relays. Failure.of any '
single sensor vill not prevent system isolation. The setpoint change does not affect Reactor Vater Cleanup system or Primary Containment Isolation system function or operation. Therefore, this modification does not constitute an unreviewed safety question per the criteria of L 10CFR50.59. '
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Plant Design Change Number 1-065-86 .
Plant Design Change Number 1-065-86 entitled "House Heating Boiler Stack Emissions Monitor ~ Replacement" is complete.
s-Description of Change The house heating boiler visible emissions monitor was removed and replaced with a new monitor that is microprocessor based and digitally processed with a recorder output.
Reason for Change Connecticut regulations require visible emissions monitoring of fuel burning equipment. The new system was installcl to meet the' intent of the regulations with a more reliable unit.
Safety Evaluati,on _
The modification performs the same function as the old system and utilizes the same power supply. There are no failure modes which could affect safety-related equipment and no changes to Category 1 equipment. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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1 Plant Design Change Number 1-066-86 Plant Design Change Number 1-066-86 entitled "Xenon-Krypton System Two
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Dryer' Operation" is complete.
Description'of Change The Xenon-Krypton primary system switch was changed from a.three position _ switch to a four position switch to allow operation of either or both of'the cyclic dryer trains from the Control Room.
Reason for Change The modification was implemented to allow single dryer operation under normal conditions, while allowing dual dryer operation during periods of increased offgas flov.
Safety Evaluation
~The modification allows processing of increased offgas flov vithout adverse effects on safety-related equipment. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-067-86 Plant Design Change Number 1-067-86 entitled "Honorail and Lifting Padeyes - Emergency Diesel Enclosure" is complete.
Description of Change Sections of monorail and padeyes vere installed-to the overhead beaa structure in the diesel generator room.
Reason for Change The modification was implemented to facilitate the removal and temporary storage of the crankshaft and other maintenance related activities on the diesel.
Safety Evaluation All the fireproofing material was replaced as found and additional --
fireproofing material was applied to the new monorail beams. The-structural and seismic characteristics of the enclosure vere not adversely affected by the addition of the monorails. Therefore,'this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i Plant Design Change Number 1-068-86 Plant Design Change Number 1-068-86 entitled "Hydrogen Vater Chemistry-Sample Tap" is complete.-
Description of Change
-A pipe tee and globe valve were installed downstream of the isolation valve on the reactor recirculation vater sample sine.
Reason for Change s The modification was implemented to provide a source of reactor recirculation Vater, at-normal operating pressure and temperature, to perform the Hydrogen Vater Chemistry Pre-Implementation test. ,
Safety Evaluation The modification was evaluated and found to meet the desiga bases --
relative to materials compatibility and veldability. The modification ,
does not change the intended function of the system. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-069-86 Plant Design Change Number 1-069-86 entitle'd "Millstone Unit 1/2 Appendix R Backfeed, Pre-Outage Work" is complete.
Description of Change A 5 kV, 600 amp transfer switch and a 4160/480V, 300 kVA transformer was installed in the Reactor Building along with two vall mounted circuit breaker enclosures. Cable and conduit runs vere installed from the outdoor switchgear to the Reactor Building and Control Room.
The backfeed cable and conduit supports vere rodified to meet seismic criteria and 24F bus flood vall and pad vere enlarged to accommodate the new vacuum breaker switchgear.
Reason for Change
- The modifications were implemented to provide alternative shutdown .
capability for Millstone L' nit 1 and Hillstone Unit 2 as required by 10CFR50 Appendix R.
Safety Evaluation The modified flood valls are not required for safe shutdown. The floor loadings for the involved areas were analyzed and found to be '
adequete to resist the additional live loads. All the new equipment inside the Reactor and Turbine Buildings was seismically mounted. The new cables are normally deenergized and perform no safety function.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
Plant Design Change Number 1-070-86 Plant Design Change Number 1-070-86 entitled "Sprinkler Design for Maintenance Shop Prefab Building and Relocation of Hose Station Supply Line" is' complete.
Description of' Change A fire protection sprinkler system was. installed inside a new prefab .
.r icated building to be used as Maintenance Shop offices, and a hose station supply line was relocated.
Reason'for Change The modification was implemented to meet National Fire Protection Association standards.
F Safety Evaluation The sprinkler system in the Maintenance Shop is not required for safe shutdown and its failure vould not adversely affect the operation of safety-related equipment. Therefore, this modification does not t constitute an unreviewed safety question per the criteria.of 10CFR50.59. l 9
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Plant Design Change Number 1-071-86 entitled "1-AS-13A/13B Replacement
-of 10" Rockwell Butterfly Valves" is. complete. l Description of Change i
The steam packing exhauster blover discharge valves and valve y
actuators were removed and replaced with improved valve / actuator assemblies.. The nev valve seat material is more suitable for use in i
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Reason for Change The modification was implemented to provide equipment that-is reliable and serviceable. The equipment that vas removed is discontinued by I the manufacturer and spare parts are not available.
Safety Evaluation The Air Extraction System is not a safety-related system and failure 1 of the new valves vould not adversely affect any safety telated >
systems. 'Therefore, this modification-does not constitute an .
unreviewed safety question per the criteria of 10CFR50.59. I P
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Plant Design Change Number 1-075-86 Plant Disign Change Number 1-075-86 entitled "Head Vent Isolation Valve" is complete.
Description of Change A manual isolation valve was installed on the upstream side of'the solenoid operated, vessel head vent valves.
R,eason for Change The modification was implemented to provide a means of performing-maintenance on the vessel head vent and vessel head vent backup valves while at operating pressure.
Safety Evaluation The additional veight of the valve was analyzed and found to have no -
adverse impact on the seismic qualification of the line or.its associated supports. The valve does not affect the operation of the hvad vent because the valve is normally open. Any, leakage due to valve failure vill be contained inside the dryvell. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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l Plant Design Change Number 1-076-86 Plant Design Change Number 1-076-86 entitled "Installation of Telephone and Maintenance Jacks at North Vall~of Turbine Building" is ,
complete.
Description of Change Two telephone jacks, one of which is the Control Room extension, and two communication maintenance jacks were installed in the Turbine ;
Building. i Reason for Change The modification was implemented to comply with Off Normal Procedures which require the ability to shutdown outside.the Control Room.
Safety Evaluation The modification does not involve changes to safety-related equipment or systems. Therefore, this modification does not constitute an -
unreviewed safety questiori per the criteria of 10CFR50.59.
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Plant Design Change Number ~ 1-077-86 l
Plant Design Change Number 1-077-86 entitled "Maintenance Building Air. l Conditioner Power Supply" is complete.
Description of Change An electrical power supply was installed to feed a roof mounted air conditioner for the Maintenance offices.
Reason for Change This modification vas. installed to handle the additional air condi-tioning requirements for the new offices. ,
Safety Evaluation The addition of the 480V 40 amp feed from the nonsafety-related motor -
control center does not exceed design ratings or compromise any safety-related equipment. Therefore, this modification does not constitute an unreviewed safety question per.the criteria of '
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Plant Design Change Number 1-078-86 J
Plant Design Change Number 1-078-86 entitled "Stack Flow Monitor Replacement" is complete.
Description of Change The Stack Flov Monitor System which is designed to measure stack flov and indicate total stack flov via a recorder and a totalizer located in the control Room, was replaced with an improved, physically and electronically compatible system. l Reason for Change The modification was implemented to provide a system that is both '
reliable and easy to maintain. Spare parts and documentation are not available for the system that was replaced. ,
Safety Evaluation l
The new system provides the same function as the old system and is not- I safety related. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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l Plant Design Change Number 1-081-86 l Plant Design Change Number 1-081-86 entitled "HVAC-4, Svitch Gear Area Ventilation A/C Unit" is complete.
Description of Change A spare set of normally open auxiliary contacts on.each air condition-
- ing supply fan motor starter for the switchgear are.t was connected in parallel. The parallel combination was connected ir...eries with.the chilled water flow switch, the Turbine Building Closed Cooling Vater flov svitch and the parallel combination of the chilled water pumps interlocks. .
Reason for Change The modification was implemented to prevent freeze up of the chillers associated with the svitch gear area ventilation unit when the air :
conditioning unit is running and both supply fans are secured.- -
Safety Evaluation The modification increases the reliability of the heating, ventilation and air conditioning system and does not adversely affect any safety-related systems. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i Plant Design Change Number 1-082-86 Plant Design Change Number 1-062-86 entitled "PEECo Flow Switch i
Removal - RBCCW to Dryvell Sump" is complete.
- Description of Change The flov'svitch from the Reactor Building Closed Cooling Vater system [
l to the Dryvell Equipner.t Sump Cooler was removed and replaced vith. -
Pi Pe. [
Reason for Change .
The switch had been retired in place from a previous plant change. [
The switch was removed to eliminate a potential leak path ~1n the :
dryvell.
Safety Evaluation i
1 Any leak from the pipe, where the switch was removed, vill be
! contained in the dryvell and has already been analyzed. Theref9re, ,
- this modification does not constitute an unreviewed safety question -
I per the criteria of 10CFR50.59.
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Plant Design Change Number 1-084-86 Plant Design. Change Number 1-084-86 entitled "Diesel Fire Pump Cooling System Modification" is complete.
Description of Change An alternate route to discharge vaste water from the die.e4 ngine heat exchanger was installed to discharge outside the tire pump house.
The original piping discharges to a floor drain scupper and remains as the primary passage for vaste water under normal plant conditions. -
Peason for Change The modification was implemented to provide an alternate discharge route to prevent fire pump house flooding during natural occurrences such as hurricanes when the floor drain scupper is required to be plugged. ,
Safety Evaluation The consequences of the additional, alternate discharge piping failure are identical to failures of the primary discharge piping. Therefore, .
this modification does not constitute an unreviewed safety question '
per the criteria of 10CFR50.59.
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i Plant Design Change Number 1-085-86 Plant Design Change Number 1-085-86 entitled "Main Steam Line Trap Bypass Valve Indication" is complete.
Description of Change p
The position indication microswitches on the main steam line trap bypass valves were revired so that they indicate closed position rather than open position via illumination'in the Control Room.
Reason for Change -
The modification was implemented to prevent a loss of power and valves-failing open situation from going unnoticed by the Control-Room operator because the valves are closed during normal plant operation.
Safety Evaluation -
The modification reduces the probability of erroneous valve position indication in the Control Room and does not adversely affect any safety-related equipment or components.- Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50 59.
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Plant Design Change Nember 1-087-86 Plant Design Change Number 1-087-86 entitled "CRD Pressure /Flot Instrumentation Piping Modification" is complete.
Description of Change Additional isolation valves, test connections and associated fittings were installed between lhe Control Rod Drive header isolation valves-and the flov instruments.
i Reason-for Change The modification was implemented to facilitate calibration and servicing of the instruments and timely replacement of the valves, if required.
Safety Evaluation The instrumentation piping that was modified is in the nonsafety-related section of the Control Rod Drive' Hydraulic system. Any loss of pressure, as a result of a leak, can be isolated via the isolation valves. Therefore, this modification does not constitute an unre-viewed safety question per the criteria of 10CFR50.59.
-l Plant Design Change Number 1-088-86 Plant. Design Change Number 1-088-86 entitled "Process Computer.
Replacement - Vital Point Transfer" is complete.
I Description of Change The vital computer inputs from the old plant computer vere transferred-to the' Data Acquisition System. Cabinets of the new plant computer.
Reason for Change i
The modification was implemented to provide the nonsafety-related ,
signals, required for plant operation, to the.new Integrated Computer- l System. ,
1 Safety Evaluation [
P All'the inputs transferred were nonsafety related and do not affect -
the operation of any safety-related signals. The plant computer and all associated inputs are isolated from safety-related instruments and are not utilized in the design basis accident analysis. Therefore, .
this. modification does not constitute an unreviewed safety question I per the criteria of 10CFR50.59.
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i Plant Design Change Number 1-089-86 l l
Plant Design Change Number 1-089-86 entitled "Dryvell Nitrogen !
Compressor Dryer. Tower Switching Failure Alarm" is complete. 1 Description of Change The dryvell nitrogen compressor dryer tower switching failure timers-and pressure switches were replaced with more dependable equipment and relocated in a larger control panel enclosure.
Reason for Change
-The modification was implemented to facilitate calibration of the new equipment and to reduce the frequency of failures due to the poor quality of the equipment that was removed.
Safety Evaluation
'The new timers.are comparable in ratings and more reliable than the old timers, and the original purpose and design vere not compromised.
Therefore, this modification does not constitute an unreviewed safety-question per the criteria of 10CFR50.59.
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I Plant Design Change Number 1-090-86 Plant Design Change Number-1-090-86 entitled "Gas Turbine Fuel Shutoff Valve Logic Modification" is complete.
Description of Change The valve logic was modified so that one of the shutoff valves vill open when the Gas Turbine is started, supplying lubrication and-cooling to the fuel pump. The other shutoff valve logic remains the same to admit fuel to the Gas Turbine when the engine speed reachra '
light-off speed.
Reason for Change The modification was implemented to prevent damage to the fuel pump, due to lubricant leakage, prior to reaching engine light-off speed.
Safety Evaluation The performance of the Gas Turbine is not affected by the modification and total leakage due to valve failure is not increased. Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-092-86 '
Plant Design Change Number 1-092-86 entitled "Process Computer Replacement - New (SPDS) Input Signals" is complete. I Description of Change New cables, conduits, cabinets and electronic isolation devices were installed to support the supply of additional computer inputs to the
-Data Acquisition System.
Reason for Change The modification was implemented to provide a means of supplying [
additional process computer' inputs necessary to support the t requirements of the Safety Parameter Display System portion of the nev plant Integrated Computer System. '
Safety Evaluation I The modification does not affect-the operation of existing plant 5 safety-related instrumentation because the inputs are isolated with ;
qualified devices. The modification enhances the performance of the i
- plant operators by allowing use of-the information as part of the Safety Parameter Display System. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 1
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Plant Design Change Number 1-094-86 Plant Design Change Number 1-094-86 entitled "Process Computer Replacement - RPV Vater Level Instrument" is complete.
Description of Change A new transmitter and associated signal conditioning equipment to monitor the reactor pressure vessel vater level vas installed to provide a computer input and Control Room indication.
Reason for Change The modification was implemented to provide an additional input to the Safety Parameter Display System portion of the new plant Integrated Computer System.
Safety Evaluation The environmentally qualified transmitter, associated process tubing cables and conduits vere seismically mounted. Failure of the new transmitter vould not adversely impact any safety systems because it is redundant to the existing transmitters. Manual globe valves vere installed in series with excess flow check valves on the instrument lines to provide adequate isolation of the protective boundaries. <
Therefore, this modification does not constitute an unreviewed safety-question per the criteria of 10CFR50.59.
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' Plant Design Change Number 1-095-86 ;
Plant Design Change Number.1-095-86 entitled "Radvaste Piping l Modification to Support Chem-Nuclear FTDS-System" is complete.
Description of Change Radvaste piping was rerouted from the concentrator to support the installation of a Fluidized Transfer Demineralization System.
Reason for Change ,
The modification was implemented to provide an alternative method of processing low level vaste streams that' collect in the floor drains and vaste collector tanks.
I Safety Evaluation ;
The assemblies were hydrostatically tested to assure system tightness, ,
- however any leaks vould collect in radvaste floor drains. The procedures for sampling and evaluating effluents, prior to discharge, remain unchanged. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. ;
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Plant Design Change Number 1-096-86 Plant Design Change Number 1-096-86 entitled "Contingency Fuel Storage Rack" is: complete.
Description of Change A new spent fuel storage rack and installation procedures, has been received and stored at the site. The rack is designed for installation in the designated spent fuel pool cask laydown area. The rack vill only be installed to facilitate an emergency full core offload of the reactor core. The new rack is a temporary rack, and once the reactor core was reloaded, vould be removed from the pool and stored until needed again.
Reason for Chango The contingency plan was implemented to provide full core offload storage capability due to a lack of spent fuel storage space in the current spent fuel pool.
Safety Evaluation The new fuel racks have been designed to maintain the fuel geometry in a suberitical configuration. The spent fuel pool cooling system is adequate to remove all expected additional heat loads. A structural analysis determined that all stresses for deadveight, liveload, thermal, seismic and stuck assembly loading conditions are within allovable codes. The reserve rack storage area vas selected such that it does not jeopardize any safety-related components. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
Plant Design Change Number 1-098-86 Plant Design Change Number 1-098-86 entitled "Millstone Unit 1/2 Appendix R Backfeed Outage Vork" is complete.
Description of Change The new vacuum breaker switchgear was installed and the backfeed capability was placed into service for Hillstone Unit I and Unit 2.
(Also see Plant Design Change Number 1-069-86 for pre-outage work).
Reason for Change The modification was implemented to meet 10CFR50 Appendix R requirements to provide alternative shutdown capability for Millstone Unit I and Unit 2.
Safety Evaluation The floor loadings for the involved areas vere analyzed and found to be adequate to resist the additional live loads. All the new equipment was seismically mounted. The new cables are normally deenergized and perform no safety function. Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
i Plant Design Change Number 1-099-86 Plant Design Change Number 1-099-86 entitled "Lighting and Receptacles :
for I&C Hot Shop" is complete. -
Description of Change An electrical power source to supply fluorescent lighting and receptacles was installed on the fourth floor of the Reactor Building.
Reason for Change The modification was implemented to provide power for the Instrument ,
and Control Department Radioactive Materials Workshop which was moved from the third to the fourth floor of the Reactor Building. ,
Safety Evaluation i The power was supplied from a nonsafety-related utility panel.
i Therefore, this modification does not constitute an unreviewed safety t question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-100-86 Plant Design Change Number 1-100-06 entitled "Solid Radvaste Building Ventilation and Roof Modifications" is complete.
Description of Change The ventilation system that services the tool decontamination facility '
of the Solid Radvaste Building was upgraded by adding a dedicated exhaust fan, filter bank and exhaust hoods. Permanent lighting and an external steel stairway was also installed for the roof enclosure of the building.
Reason for Change ,
L The modifications vere implemented to assure that chemical levels do not reach concentrations which could endanger personnel health and to obtain a certificate of occupancy for the roof enclosure. _
Safety Evaluation The modifications do not, in any way, affect any equipment or systems that are safety related. Therefore, these modifications do not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-101 l Plant Design Change Number 1-101-86 entitled "Installation of Grease :
Relief Kits to Limitorque Operators" is complete.
Description of Change Grease relief kits were installed on the Limitorque valve actuators that did not have grease relief designed into.them. {
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- Reason for Change '
4 The modification was implemented to prevent grease from being forced -
from the main gearbox into the spring cartridge enclosure which would interfere with the normal operation of the spring assembly. ,
Safety Evaluation '
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The grease relief kits reduce the potential for valve operator failure l and do not adversely affect the intended. function or operation of the L i valves. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-102-86 Plant Design Change Number 1-102-86 entitled "Control Room Halon System" is complete.
Description of Change An Automatic "Total flooding" Halon 1301 Fire Suppression System was installed in the Cor. trol Room.
Reason for Change The modification was implemented to provide a means of detecting and extinguishing a fire in the Control Room before damage to safe shutdown components occurs.
Safety Evaluation The smoke removal fan circuitry is designed to allow fan operation after Halon system operation for smoke removal or venting. The nominal Halon concentration after complete discharge vill be well below toxic limits. The system is povered by an internal battery so a loss of station power vill not prevent the system from operating as intended. The piping and conduit are supported to withstand a seismic event so that any safety-related equipment located under the Halon piping and conduit vill not be damaged during a seismic event.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
't Plant Design Change Number 1-103-86 Plant Design Change Number 1-103-86 entitled "Main Transformer and .
Deluge System Replacement" is complete. I Description of Change r
The main transformer that connects the main' generator to the
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switchyard and steps up the generator's output to the transmission system was replaced. In addition, the Fire Protection Deluge System vas constructed to suit the replacement transformer dimensions.-
, Reason for Change ,
Catastrophic failure of the transformer caused the plant to be
- shutdown. The modification was implemented in order to return the plant to operation.
Safety Evaluation I
The replacement unit, as a spare, has been used in this position previously. No relay settings vere changed and the switchgear ;
momentary rating was not exceeded. Generator voltage regulator !
setting and the high side tap setting did not need to be changed. The !
transformer and associated equipment does not perform a safety-related i:
- function nor does it interface or adjoin safety-related systems.
Therefore, this modification does not constitute an unreviewed safety I question per the criteria of 10CFR50.59.
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E Plant Design Change Number 1-104-86 Plant Design Change Number 1-104-86 entitled "Dryvell Compressor i Instrument Filters" is complete.
I Description of Change Filters vere installed on the dryvell compressor system instrument i lines between the dryers and the pressure indicators.
Reason for Change The modification was implemented to remove any desiccant in the lines before~it reaches and damages the instruments.
Safety Evaluation The modification increases the reliability of the dryvell compressor system and does not affect any safety-related systems. Therefore,
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this modification does not constitute an unreviewed safety question .
per the criteria of 10CFRSO.59. !
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Plant Design Change Number 1-105-86 Plant Design Change Number 1-105-86 entitled "High Radiation Area Gate Alarms and Varning I.ights" is complete.
Description of Change Devices to provide nonclosure alarms and varning lights were installed at high radiation area access gates that vere not previously equipped with the devices.
Reason for Change The modifications were implemented to comply with 10CFR20 and to provide reliable nonclosure warning which is compatible with normal personnel traffic and material transfer requirements.
Safety Evaluation The installation of the alarms did not introduce any seismic concerns and the utility lighting panels used to supply power are not safety related. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
F Plant Design Change Number 1-106-86 Plant Design Change Number 1-106-86 entitled "Rod Vorth Minimizer Replacement" is complete.
Description of Change The rod vorth minimizer, which monitors the control rod withdrawal sequence and provides a rod block signal if an out of sequence rod is i selected for movement vas replaced with a Nuclear Heasurement Analysis and Control rod worth minimizer standalone computer.
Reason for Change The modification was implemented to provide a functional replacement of the obsolete computer hardware as well as reducing computational load on the plant computer, reducing the potential for human error and providing a positive indication of rods Full-In condition before scram reset.
Safety Evaluation The modification was a direct one for one replacement which utilized I the came power source. The new chassis was seismically mounted in the Control Room and does not change the function of the Reactor Manual
- Control System or the Rod Position Information System. Therefore, the :
,~ modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. i I
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Plant Design Change Number 1-108-86 Plant Design Change Number 1-108-86 entitled "Appendix R Outage Related Emergency Lighting" is complete.
Description of Change An Emergency Lighting System, consisting of remote fluorescent lamps over the control panels and incandescent lamps for access and egress points, was installed in the Control Room.
Reason for Change The modification was implemented to comply with the requirements of 10CFR50, Appendix R, Section III.J.
Safety Evaluation The charging circuit is povered by nonvital pover, so that an electrical failure within the lighting units vill not jeopardize safety-related circuits.
The emergency lighting units located near safety-related equipment are
, seismically mounted to preclude any potential for safety-related equipment damage from missile hazards caused by failure of these lighting units.
The cable used to provide the connection to an outlet is from controlled stock and is qualified to flame retardancy requirements.
Based on the fact the new units have been addressed from a saismic otandpoint and connected to nonvital AC circuits, safety-re.=ted equipment operation vill not be jeopardized. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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l Plant Design Change Number 1-002-87 Plant Design Change Number 1-002-87 entitled "Main Transformer Remote ,
Annunciation" is complete.
Description of Change i
5 A relay vas added to the main transformer control cabinet to allov direct current and high temperature to be monitored from the Control ;
Room.
Reason for Change The modification vas. implemented to allow the operator to determine if a main transformer trouble alarm is initiated due to high temperature i or loss of direct current control power.
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Safety Evaluation
> The addition of the relay only increases the ability to detect abnormal main transformer operating conditions and does not adversely affect the design basis for the electrical system. Therefore, this l
, modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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s Plant Design Change Number- 1-004-87
, Plant Design Change Number 1-004-87 entitled "Scram Discharge Vent Line Modification - South Header" is complete. [
Description of Change
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- The Control Rod Drive vent / drain line and the Head Spray Lov Point ,
5 drain line vere modified at the point where they dump into the drain :
- scupper. t Reason for Change 4 The modification was implemented to prevent contaminated water from ,
splashing over the scupper and onto the floor when the vent isolation ;
- valve opens during the Control Rod Drive header draining and venting
- evolution following a reactor scram.
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I Safety Evaluation ;
. The vent line reroute does not prevent or affect the performance of i l the scram circuitry or support systems because the vent is closed upon '
scram initiation and is opened for use only after the scram is reset. i i Therefore, this modification does not constitute an unreviewed safety :
) question per the criteria of 10CTR50.59. !
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i Plant Design Change Number 1-005-87 Plant Design Change Number 1-005-87 entitled "CST Fill Line Hodification" is complete.
Description of Change The vaste sample tank to condensate storage tank fill line vas modified by replacing the butt-velded piping with flanged spoolpieces.
Also the carbon steel pipe slide supports vere replaced with nonmetallic pipe slides.
Reason for Change The modification was implemented to restore integrity to the system piping and to facilitate future piping replacement. The pipe supports vere replaced to minimize galvanic corrosion of the aluminum piping.
l Safety Evaluation The new spoolpieces meet or exceed the system design parameters with no degradation of the system function or vater quality. Any con-taminated vater released by a piping failure vill be contained in the pipe trench which drains to the liquid radvaste facility. Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-006-87 Plant Design Change Number 1-006-87 entitled "Jet Pump Instrumentation i Nozzle Assembly Replacement" is complete. :,
Description of Change i
The jet pump fr.strument nozzle assemblies and associated drain lines !
] and supports vere removed and replaced with penetration seals.
R Reason for Change l !
The modification was implemented to restore the.struc'. ural integrity ,
l of the reactor pressure vessel pressure boundary and mitigate the !
I effects of stress corrosion cracking. I i Safety Evaluation The penetration seals utilize improved materials and do not alter the +
j configuration of the reactor vessel nozzles. The seals also meet or [
i exceed the original design, fabrication and installation requirements. t Therefore, this modification does not constitute an unreviewed safety i
] question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-020-87 Revision 1 Plant Design Change Number 1-020-87 Revision 1 entitled "Process Computer Replacement - Hardvare Installation" is complete.
- Description of Change The hardware, peripheral cables and interface connections associated with the Integrated Computer System vere installed in the operational support center, computer room, control room and cable vault.
Reason for Change The modifications vere implemented to prepare for the replacement of the obsolete process computer vith the new Integrated Computer System which provides a fast and convenient method for monitoring core performance, equipment status and routine record keeping chores and calculations.
Safety Evaluation The plant process computer provides data acquisition and information output functions. Since these features provide no safety-related functions there is no impact on any design basis accident analysis.
Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-021-87 Plant Design Change Number 1-021-87 entitled "Stack Radiation Monitor Millstone Unit 2 Interface" is complete.
Des _cription of Change A signal input from the Millstone Unit 1 Stack Radiation monitor was installed to the Millstone Unit 2 Integrated Computer System.
Reason for Change The modification was implemented to allow the Millstone Unit 2, as well as the Unit 1 computer, the capability to continuously monitor, indicate and record the discharge of radioactive gases from the plant.
Safety Evaluation No safety-related circuits or functions vere affected by the modifi-cation and no channel separation requirements were compromised.
Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
Plant Design Change Number 1-022-87 Plant Design Change Number 1-022-87 entitled "Hydrogen Vater Chemistry Recirculation Vater Sample Tubing" is complete. !
Description of Change >
Stainless steel tubing was temporarily run from a valve in the Recirculation Vater Sample line to a valve dovnstream of the Cleanup System vent valve. The tubing passed through an autoclave and sample module and associated sample coolers. The sample tubing vas operated for approximately one month prior to the performance of the Hydrogen Vater Chemistry Preimplementation Test to allov an oxide layer to form on the inner surface of the tubing. Preconditioning was necessary to ensure that any changes in the recirculation water chemistry vas a result of the added hydrogen and not the oxide layer formation.
Reason for Change The modification vas implemented to facilitate the Hydrogen Vater Chemistry Preimplementation test. (Also see Plant Design Change Number 1-049-87).
Safety Evaluation 1
Operation of the tubing was prohibited until Reactor Building Closed Cooling Vater was provided. The test equipment is nonseismic and nonsafety related. The preconstruction installation of this equipment did not affect any plant components. Therefore, this modification did
. not constitute an unreviewed safety question per the criteria of 10CTR50.59.
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Plant Design Change Number 1 023-87 Plant Design Change Number 1-023-87 entitled "CRD Pump Piping Modification for Alternate Cooling" is complete.
Description of Change The Control Rod Drive pumps vere modified to provide self-cooling to ensure that the bearings and oil coolers are available during a fire.
The modification consisted of piping and valves between the suction / discharge line of the pumps and the Turbine Building Secondary closed cooling Vater System cooling line to the Control Rod Drive pumps.
Reason for Change The modification was implemented to comply with 10CFR50 Appendix R vhich requires Control Rod Drive pump operability for reactor vessel mockup following fires in the control room, cable vault, turbine building, shutdown pump cubicle, gas turbine or screenvell house.
Safety Evaluation The modifications are manually initiated following a fire and involve a nonsafety-related system. The integrity or function of the affected system were not degraded. The modification has no impact on a design basis accident. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-026-87 .
Plant Design Change Number 1-026-87 entitled "Millstone Unit 1 App. R !
and SEP FVCI Area Ventilation Modifications" is complete. I Description of Change The area coolers required to provide cooling of the feedvater, i condensate, and condensate booster pump areas in the event of a l i
Reactor Building fire vere connected to a source of emergency power to automatically restart during a Loss of Normal Pover. l i
Reason for Change ;
The modification was implemented to comply with 10CFR50 Appendix R I requirements by preventing accelerated aging of the Feedvater Coolant i Injection pump motor insulation systems that vould result from -
excessive ambient temperatures.
Safety Evaluation f The electrical power and control cables were routed in non-Class 1E [
cable trays and conduit to maintain the separation from Class 1E i equipment. The ability of the coolers to restart during a Loss of l Normal Power does not af fect any Class 1E equipment or systems. t Therefore, this mod *2ication does not constitute an unrevieved safety l question per the criteria of 10CFR50.59. !
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Plant Design Change Number 1-029-87 Plant Design Change Number 1-029-87 entitled "Replace LPCI Containment Cooling Timers Logic Time Delay Relays" is complete.
Description of Change The push-on, appliance type, connections used in the Lov Pressure Coolant Injection system time delay circuit relays were replaced with terminal screv type relays which accept ring tongue connectors.
Reason for Change The modification was implemented to reduce the potential for inadvertent disconnection and failure due to poor electrical connections.
Safety Evaluation ,
The improved relays were seismically mounted, the delay settings and functions remain the same, and they do not affect the basis of the Lov Pressure Coolant Injection system logic. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-030-87 Plant Design Change Number 1-030-87 entitled "Replace Overcurrent Trip Devices - Vital 480V Breakers" is complete.
Description of Change The electromechanical overcurrent trip devices that sense an over-current condition and initiate a breaker trip, if_ warranted, of the 480V power circuit breakers, vere replaced with more reliable, solid state devices.
Reason for Change The electromechanical devices were reaching the end of normal lifetime as evidenced by deterioration and occasional failures. The solid state replacements provide greater reliability and accuracy.
Safety Evaluation ..
The solid state trip devices are environmentally and seismically qualified and replaced the electromechanical devices on a one-for-one basis. The solid state design performs the same function as the electromechanical design with greater reliability and does not adversely affect the intended purpose of the circuit breakers.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-032-87 Plant Design Change Number 1-032-87 entitled "MP1 Dryvell Temperature Monitoring System" is complete.
Description of Change A qualified Dryvell Bulk Temperature Monitoring System, consisting of thermocouples, cable and ccnduits, processing cards, a recorder, indicators and alarms, was installed to provide the Control Room operators with continuous indication of bulk temperature as well as recording all sixteen discrete temperature points.
Reason for Change The modification was implemented to satis'y the requirements for a system that is fully qualified for post accident operation.
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Safety Evaluation The components in this stand alone system are environmentally and seismically qualified and do not interface with any safety-related system. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-033-87 Plant Design Change. Number 1-033-87 entitled "N 2 Compressor Flovmeter
. Relocation" is complete.
Description of Change The dryer purge flow meter was relocated from downstream of the dryers to upstream of the dryers between the purge flov valve and the purge flow orifice.
Reason for Change The modification was implemented to prevent excessive cycling of the flow meter when the dryer towers are cycled.
Safety Evaluation The function of the flov meter remains the same although the reliabil-ity of the system is increased. The system is nonsafety related and the Dryvell Compressor skid has isolation valves that close on a Containment Isolation signal. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-037-87_ entitled "Replace Gas Turbine Generator Neutral _ Grounding Transformer" is complete. l Description of Change The gas turbine generator neutral grounding transformer, which was filled with insulating oil containing polychlorinated biphenyls, was replaced with a dry type transformer of the same electrical rating.
Reason for Change The modification was implemented to eliminate the risk to the environ-ment posed by a spill or fire involving highly concentrated polychlorinated biphenyl fluids.
Safety Evaluation-The transformer does not perform a safety-related function and does not trip the generator. The design basis is unchanged because the replacement transformer performs the same function as the old one.
The new transformer was seismically mounted and weighs the same as the old one, so floor loading was not a concern. Therefore, this modifi-cation does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
Plant Design Change Number 1-038-87 Plant Design Change Number 1-38-87 entitled "Pover. Cold Shutdown Equipment Following Appendix "R" Fires" is complete.
Description of Change The Isolation Condenser valve control cables, dryvell temperature and pressure indication cables and Isolation Condenser valve position indication cables vere replaced with one hour fire rated cables.
Reason for Change The modification was implemented to comply with the requirements of 10CFR50 Appendix R by ensuring Isolation Condenser operability following a cable vault fire and dryvell temperature and pressure indication following a Reactor Building fire.
Safety Evaluation All the equipment installed for the modification was seismically mounted and environmentally qualified. The dryvell pressure gauge can 1 be isolated so as not to affect other safety-related components in the event of gauge failure. Therefore, this modification does not con-stitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant' Design Change Number 1-040-87 ent'itled "Appendix R Firecoating~ j of Structural Steel" is complete. O
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-l Description of Change
- A three hour fire; resistant coating was applied'to:the structural
~ d steel framing members in the Turbine Building in the area of the Unit 1/ Unit 2 Backfeed' Cables.
-Reason for Change The modification was implemented to comply with 10CFR50, Appendix R, ,
by providing a means of preventing the spread of fire to safe-shutdown areas of the plant from surrounding areas.
Safety Evaluation The applicationaof the' fireproof coating does not adversely' impact the !
structural capability or seismic integrity of the structural steel.
Therefore, this modification does not constitute an unreviewed safety- .
question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-041-87
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Plant' Design Change Number 1-041-87 entitled "Hillstone Unit No. 1 Appendix."R" Cable Fire Vrap: Program and MI-Cable. Substitution and Reroute" is complete.
Description of Change .
A_three-hour, fire-rated cable wrap system was installed on those cables and conduits which must remain functional during.a fire. The Mineral Insulated cables.eere installed to prevent a. fire in the
- Reactor Building from affecting redundant trains of reactor instrumentation.
Reason for Change The modification vas. implemented to comply with the requirement of 10CFR50 Appendix R by providing a three-hour rated ~ fire barrier from safe shutdown electrical cables. ,
Safety Evaluation The new cables and wrap system are environmentally qualified and vere installed to meet seismic criteria. .The modification clearly results ,
in an improvement in overall plant safety. Therefore, this modifi-2 cation does not constitute an unreviewed safety question per the 1 criteria of 10CFR50.59. <
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l Plant Design Change Number 1-042-87 Plant Design Change Number 1-042-87 entitled "Installation of Excess ,
l Flov Valve / Bypass Valve Assembly in Hydrogen Supply System" is complete.
Description of Change An excess flov valve / bypass _ valve vas installed in the hydrogen supply system to the main generator.
Reason for Change The excess flov valve was installed to prevent the leakage of hydrogen into the Turbine Building following a hydrogen supply pipe break. The bypass valve was installed to allow for filling the generator with hydrogen without causing the excess flov valve to close.
Safety Evaluation The Hydrogen Supply System is a non-Category 1 system and does not impact plant operation following design basis accidents. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
I Plant Design Change Numuer 1-047-87 Plant Design Change Number 1-047-87 entitled "Replacement of Core Spray Valves CS 21A and 21B" is. complete.
Description of Change The motor operated valves, which are used for throttling during the surveillance testing of the core spray pumps, vere removed and !
replaced with environmentally and seismically. qualified valves. ,
Reason for Change ;
i The modification was implemented to address the operational problems i of frequent breakdovn of the operator due to motor burn out and the i problem of obtaining qualified replacement parts from the ;
manufacturers. ,
Safety Evaluation
- The design basis for the valves was not jeopardized.and the control [
cables, thermal overload relays and torque switch settings vere '
evaluated as being adequate for proper functioning of the valves. The ;
control circuitry and valve logic remains unchanged. Therefore, this ,
modification does not constitute an unreviewed safety question per the !
criteria of 10CFR50.59. I, i
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Plant Design Change Number 1-049-87 Plant Design Change Number 1-049-87 entitled "Hydrogen Vater Chemistry Preimplementation Test" is complete.
Description of Change A three-day test, consisting of. injecting hydrogen into, and moni-toring the oxygen level in the reactor feedvater, was performed to reduce the corrosion potential of the water to a point where it no longer facilitates intergrannular stress corrosion cracking.
Reason for Change The preimplementation test was performed to measure the effects and determine the suitability for a Hydrogen Vater Chemistry program at Hillstone Unit 1.
Safety Evaluation The seismic qualifications of the valves used to conduct the test vere not adversely affected. Manual isolation valves were installed on the sample lines as a means of preventing leakage in the event of a sample line break. Containment was inerted during the test so there was a negligible impact on post accident hydrogen control. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
l Plant Design Change Number 1-051-87 Plant Design Change Number 1-051-87 entitled "Appendix R, Alternate Cooling.of Shutdown Cooling /Dryvell Cooling" is complete, j i
Description of Change 2
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Connections were' installed in the Reac' tor Building Closed Cooling Vater (RBCCV). piping and fire protection piping solthat. Fire Vater may be. introduced to the RBCCV-system and used to cool the shutdown i cooling system heat exchangers and pump coolers. Connections vere also installed to discharge the fire water to the Service Vater discharge tunnel.
Reason for Change The modification _vas implemented to enable the plant to achieve a cold shutdown condition even after a loss of Reactor Building Closed
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Cooling Vater and/or Service Water as postulated by an Appendix R fire. The modification also allows dryvell cooling to occur even after the loss of the dryvell coolers and/or numerous other-fire related equipment failures as postulated by an Appendix R fire. ,
Safety Evaluation The modification helps to ensure that_both hot and cold shutdown can i be achieved after fire in the plant, accompanied by a loss of off-site ,
power. The changes increase the overall margin of safety of the plant and do not adversely affect the function or operability of the Low Pressure Coolant Injection system, the Reactor Building Closed Cooling Vater system, or the Fire Vater system during normal or post design basis accident situations. Therefore, this modification-does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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l Plant Design Change Number 1-052 Plant Design Change Number 1-052-87' entitled "Chemical Decontamination.
of.the Recirculation and RWCU Systems" is complete.
Description of Change
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s Temporary decontamination equipment was installed on the recirculation system piping-and the cleanup system piping to rencva the' oxide corrosion layer and radioactive contaminants from the recirculation suction and discharge piping and the tube side of-the reactor water '
cleanup system including the regenerative and nonregenerative heat exchangers. The dissolved metal ions and. chemical reagent vere removed via ion exchange resins. ,
Reason for Change The chemical decontamination was performed to reduce radiation; fields [
of the recirculation and reactor vater _ cleanup piping in support of
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inservice inspection and maintenance activities.
Safety Evaluation The temporary connections and electrical loads did not-compromise the -
design basis of the affected systems, all the related equipment was removed, and all plant systems were restored to. original condition.
The reactor was defueled throughout the application of the decon-tamination process. Therefore,'these modifications did not constitute an unreviewed safety question per the criteria of.10CFR50.59.
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Plant Design Change Number 1-053-87 Plant Design Change Number 1-053-87 entitled "RBCCV Essential Header Stop Valve 1-RC-15" is complete.
Description of Change The opening and closing circuitry of the essential header outlet valve was modified-to allow the valve to throttle in the open or closed directions.
Reason for Change The modification was implemented to provide a method of throttling Reactor Building Closed Cooling Vater flow thrcugh the essential header following a total loss of Reactor Building Closed Cooling Vater t
to prevent thermal shock of the recirculation pump seals.
Safety Evaluation The modification does not jeopardize the intended function of the '
system design because the valve can still be remotely operated from the Control Room but with the ability to position the valve at a desired position instead of fully open or fully closed. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-054-87 Plant Design Change Number 1-054-87 entitled "GEZIP Demonstration Test" is complete.
Description of Change A :emporary skid containing the necessary equipment to inject zine into the normally closed, high point vent valves of the :eedvater system suction line was installed to support the General Electric Zinc Injection Passivation test.
Reason for Change Operating boiling vater reactors with a continuous zine concentration in the reactor water exhibit lever radiation fields and thinner oxide layers on the surfaces of their recirculation piping systems. The modification was implemented to determine the effects of continuous zine injection on a middle aged boiling water reactor.
Safety Evaluation -
Spill pans were installed under the equipment skid and splash valls l vere installed around the equipment to protect surrounding equipment.
The floor slab was analyzed and found to be adequate to support the ;
additional veight of the equipment. A check valve vas installed to protect against a line break. The modification did not affect any system operation or performance. Therefore, this modification does ;
not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-055-87 Revision 1 Plant Design Change Number 1-055-87 Revision 1 entitled "Millstone Unit 1 Lov Pressure B Rotor Replacement" is complete.
Description of Change The steam turbine low pressure rotor, which had shrunk-on wheels, was replaced with a new rotor of a monoblock design. The coupling bolts vere also replaced with new design bolts.
Reason for Change The rotor was replaced because it contained cracks in the shrunk-on wheels which jeopardized safe operation. The coupling bolts were replaced to reduce radiation exposure and the potential to seize in the rotor coupling holes. ;
Safety Evaluation The replacement of the rotor does not affect any safety-related systems; however, the new monoblock design reduces the wheel missile probability and the source locations from which stress corrosion cracking can initiate. Advances in testing techniques ensure that the quality of the new rotor meets or exceeds that of the original equipment. Therefore, this modification does not constitute an 1 unreviewed safety question per the criteria of 10CFR50.59. l l
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Plant Design Change Number 1-056-87 Plant Design Change Number 1-056-87 entitled "Emergency Diesel Generator Fuel Oil Piping Modification" is complete. ,
Description of Change The thermally fused check valvcs which were origiaally installed to isolate the supply of fuel to the diesel in the event of a fire in the Day Tank Roou vere removed and replaced with seismically qualified gate valves. The fuel oil piping in the diesel generator enclosure was modified to permit the removal and re-assembly of the in-line strainers.
Reason for Change The modifications were implemented to reduce the probability of a common cause seismic induced failure and to reduce the probability of diesel generator failure due to degraded fuel.
Safety Evaluation The modifications vere made when the diesel generator was allowed to be inoperable while the plant was in cold shutdown during a refueling outage. The change does not adversely impact the diesel generator fuel supply because the replacement valves are seismically qualified.
The ability to perform timely preventive maintenance on the fuel strainers enhances the diesel generator performance. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-058-87 Plant Design Change Number 1-058-87 entitled "MP1 - Isolation Condenser Steam Trap Reroute" is complete.
Description of Change The piping that carries condensate discharge from the Isolation Condenser. supply line steam traps to the torus was rerouted and nev valves were installed.
Reason for Change The modification vas implemented to meet the conservative interpreta-tion of the' design rules and to eliminate valves requiring excessive maintenance.
Safety Evaluation ,
The modification only changed piping geometry and not component or system function. There is no impact on safe shutdown analysis. .
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-060-87 ,
Plant Design Change Number 1-060-87 entitled "Conduit Duct' Bank for Flanders Line" is complete.
Description of Change An underground concrete duct bank was installed, with personnel. access covers, to house the incoming power supply cables from the Flanders substation to the safeguard buses of the plant.
Reason for Change The modification was implemented as a measure to protect the incoming '
power supply from exposure to salt air contamination and high vind conditions.
Safety Evaluation The modification increases the reliability of the power supply line by protecting it from the elements and does not change its purpose or function. The modification does not adversely affect any safety-related equipment or systems. Therefore, this modification does not constitute an vnreviewed safety question per the criteria of 10CFR50.59.
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A Plant Design Change Number 1-064-87 Plant Design Change Number 1-064-87 entitled "Reload 11 Core" is '
complete.
Description of Change One hundred ninety six, barrier clad, prepressurized retrofit fuel assemblies with axially zoned gadolinia were inserted into the !
Hillstone Unit 1 reactor.
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Reason for Change !
The modification was implemented to replace the discharged fuel assemblies in support of cycle twelve operation. [
I Safety Evaluation The insertion of the new fuel assemblies does not impact fuel ,
performance. Operational limits are set such that performance, as ;
measured by fuel failures, is equal to or better than previous j operation. Therefore, this modification does not constitute an >
, unreviewed safety question per the criteria of 10CFR50.59. i t
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Plant Design Change Number 1-066-87 Plant Design Change Number 1-066-87 entitled "Extraction Steam Piping Replacement" is complete.
Description of Change The stages of the extraction steam piping inside the condenser that run from the expansion bellows located off the base of the turbine downstream to the condenser vall vere removed and replaced with piping that has better corrosion / erosion resistant properties.
Reason for Change The modification was implemented to restore system integrity and reduce safety and reliability concerns due to numerous pipe failures as a result of vet steam erosion.
Safety Evaluation The modification was a one for one replacement with material that has better corrosion / erosion properties, equal or higher allovable stress '
limits and results in equal or lover thermal loads. The system is not safety related, and there are no failures postulated which could impact safety-re ated equipment. Therefore, this modification does not constitute a unrevieved safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-067-87 Plant Design Change Number 1-067-87 entitled "Blade Guide Storage Racks" is complete.
Description of Change Blade guide storage racks were installed on the vest vall of the spent fuel pool so that the blade guides can be placed in the racks by using the refueling bridge. The racks also provide an adequate measure of protection against physical damage to the blade guides.
Reason for Change The modification was implemented to provide an alternate storage location for the blade guides. The blade guides had previously been stored in the spent fuel pool racks, but due to space limitations, the spent fuel racks are required for discharged fuel assemblies.
Safety Evaluation Seismic analysis and physical placement of the racks assures that they vill not interact or adversely affect the spent fuel racks, and fuel within. Blade guides are refueling tools and are not safety related.
There are no postulated failures which could impact safety-related equipment. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-068-87 Plant Design Change Number 1-068-87 entitled "EEQ HOV Modifications" is complete.
Description of Change The Teledyne motor operator was replaced on the Isolation Condenser supply line isolation valve with a Limitorque motor operator.
Reason for Change The modification was implemented to ensure harsh environment operability of a motor operated valve whose operation is needed to mitigate the consequences of a loss of coolant accident, as required by 10CFR50.49.G.
Safety Evaluation The replacement motor operator meets or exceeds all the design requirements of the original equipment. The motor operator was qualified to meet or exceed seismic, electrical and environmental requirements. The modification does not adversely affect the intended function of the system or any other safety-related system. Therefore, this modification does not constitute an unrevieved safety question i per the criteria of 10CFR50.59. I i
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a Plant Design Change Number 1-073-87 Plant Design Change Number 1-073-87 entitled "Appendix R - Auxiliary Boiler Blast Modifications" is complete.
Description of Change Additional lov vater level cutoff switches were installed on the Unit I house heating boiler and the Unit 2 auxiliary steam boiler and each new level switch was surrounded by a one hour rated fire enclosure. A redundant downstream fuel cutoff solenoid valve was installed on the boiler that did not have one previously and all the downstream fuel cutoff valves were revired to the new, respective level svitches.
Reason for Change The modifications were implemented.to eliminate the possibility of a boiler explosion resulting from fire induced failures of the boiler --
control system. The modifications ensure that safe shutdown can be achieved following a fire in the auxiliary boiler room, consistent with 10CFR50 Appendix R requirements.
Safety Evaluation The modification and its failure modes have no direct impact on any system, except the short term loss of House Heating Boilers which are not safety related. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-074-87 Plant Design Change Number 1-074-87 entitled "Diesel Generator Room .
Structural Steel Firecoating" is complete.
Description of Change The f'reproofing blanket material on the structural steel framing members that support the floor above the Emergency Diesel Generator Room was removed and replaced with a fire resistant coating with a three-hour rati.-'-
Reason for Change The modification was implemented because the previous firecoating material vas inadequate. The ceiling supports both trains of the '
station battery rooms and is required to maintain its integrity during a postulated fire in the Emergency Diesel Generator Room.
Safety Evaluation The firecoating is desigr.ed for two over one seismic criteria and the additional veight of the firecoating does not adversely impact the structural capability of the structural steel. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR5u.59.
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Plant Design Change liumber 1-075-87 Plant Design Change Number 1-075-87 entitled "LPCI and Core Spray Pump Anchorage Modification" is complete.
Description of Change The structural anchc. rage for the Lov Pressure Coolant Injection and Core Spray pumps' vere modified by adding anchor bolts and steel plates which connect the pump bowl to the masonry pedestal and provide a nev load path from the pump / motor assembly to the building structure.
Reason for Change The modification van implemented to provide adequate anchorage for the equipment to assure the operability during a seismic event. The modification takes :Into consideration new loadings which have greater i magnitude as a result of piping re-analysis and support modifications.
Safety Evaluation The modification provides the anchorage required for all loading cases ,
including seismic inertia loads as well as loads from the attached i piping. The modification clearly enhances the seismic integrity of j the pumps and does not adversely affect any safety-related systems. J Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-076-87 Plant Design Change Number 1-076-87 entitled "Emergency Diesel Generator, Air Roll Modification and Indicator Valve Installation" is complete.
Description of Change Two ball valves vere installed in the starting air supply tubing on the Emergency Diesel Generator engine. One of the valves was installed just upstream of the governor oil booster and the other valve was installed just upstream of the main bearing oil booster.
The modification allows "air rolling" of the engine after a surveillance run is performed. Also, indicator / relief valve adapters and indicator valve assemblies were installed to each engine cylinder liner.
i Reason for Change The modification was implemented to provide a method of "rolling" the engine after maintenance, troubleshooting or surveillance testing without having to isolate the air start system and rotate the crankshaft mechanically to minimize the presence of residual oil in the exhaust piping. The indicator / relief valve installation vill allow maintenance personnel to periodically obtain firing pressures for long term trending of the engine's performance.
Safety Evaluation Analysis of the modification revealed that the additional veight of the valves result in no seismic rf,k to the diesel. The indicator valves are designed to fail in the closed position and if they should fail, the engine vould not fail to function. The modification has a net positive impact on the performance of the Emergency Diesel Generator. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-077-87 i Plant Design Change Number 1-077-87 entitled "Suppression Chamber Air Temperature Indication" is complete. -
Description of Change A dual temperature indicator to monitor individual air temperature in the suppression chamber was installed and the signals from the air temperature and Vater temperature sensors were rerouted to the front panel in the control Room.
Reason for Change The modification was implemented to satisfy human factors criteria because an operator vill not be required to obtain temperature readings from the back panel in the Control Room. Also, by utilizing spare terminal points, the modification allovs for the Safety Parameters Display System portion of the new plant computer to monitor i
l the suppression chamber air temperature.
Safety Evaluation !
j The modification implements all design inputs with no adverse effects I on the overall system design. The modification does not degrade any '
safety-related components while improving the method of monitoring !
plant parameters. Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59, i 4
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1 Plant Design Change Number 1-080-87 Plant Design Change Number 1-080-87 entitled "Hain Steam Line Support Beam Modification" is complete.
Description of Change .
The main steam line support beam in the condenser bay area of the Turbine Building was modified by velding supplemental steel to it, to stabilize it in the lateral direction, and adding additional supports to stiffen it in the vertical direction.
Reason for Change The modification was implemented to prevent excessive beam deflection due to piping loads.
Safety Evaluation The modification enhances the structural integrity of the beam without jeopardizing its seismic capabilities. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 1-081-87 l
'1 Plant Design Change Number 1-081-87 entitled "SRV Pressure Switch Replacement" is complete.
Description of Change The safety relief valve pressure switches, which are mounted downstream of each safety relief valve and alert the operator of safety relief valve actuation or leakage, vere removed and replaced '
with qualified switches.
Reason for Change The modification was implemented to provide the Control Room operator :
vith a qualified method of monitoring safety relief valve tallpipe i pressure.
Safety Evaluation The modification does not jeopardize the intended function or system operation because the new switches function in the exact same manner i as the old switches. All the Control Room indications remain the same _,
and the new switches were seismically mounted in the same locations as those removed. Therefore, this modification does not constitute an 1 unreviewed safety question per the criteria of 10CFR50.59. l J
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Plant Design Change Number 1-084-87 ,
Plant Design Change Number 1-084-87 entitled "Electrical Pressure Regulator Operator Interface Modifications" is complete.
Description of Change The modifications consisted of relocating the Electric Pressure Regulator and Mechanical Pressure Regulator setpoint indicators adjacent to each other, providing indication of the Electric Pressure Regulators amplifier condition and providing an annunciator to indicate when the Electric Pressure Regulator is not in control of j reactor pressure. Also, all the servo indicators were changed so that servo position vill travel in the same direction as regulator 4
setpoint.
Reason for Change The modification was implemented to enhance reactor pressure control and improve operator interface between the Electric Pressure Regulator and Mechanical Pressure Regulator.
Safety Evaluation '
i The new servo indicators are electrically identical replacements and do not introduce any new failure modes. Failure of the new relay that provides annunciation vill not result in any adverse effect on '
the pressure control system functions. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number'l-086-87 entitled "House Heating Boiler ,
Stack Rep 10 cement" is complete. '
Description of Change The exhaust stack of the house heating boiler was removed and replaced ;
-vith a new double-vall insulated stack at the same location over a modified foundation.
Reason for Change The modification was implemented to restore the structural integrity and capacity to resist vinds of the house heating boiler exhaust stack.
Safety Evaluation seismic and structural integrity was restored to the exhaust stack and there was no impact on any safety-related equipment or systems.
Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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PLANT DESIGN CHANGE EVALUATION INDEX J t 3 PDCE Number Title !
MP-1-87-027 Dryvell Penetration X-11A
- MP-1-87-030 Mechanical Pressure Regulator Oil Supply Line MP-1-87-034 Fuel Storage Tank Additives MP-1-87-061 GEZIP Injection Pump Flush Lines j .
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Plant Design Change Evaluation HP-1-87-027 Plant Design Change Evaluation' Number HP-1-87-027 entitled "Dryvell Penetration X-11A" is complete.
Description of Change Lead shielding was installed inside the penetration directly above the velded joint that attaches the penetration to the dryvell liner.
Reason for Change The change was implemented to reduce the gamma dose rate and neutron dose rate to a level belov that which requires the area to be locked.
Safety Evaluation The installation of the lead shielding does not reduce the structural integrity of the dryvell liner or adversely affect the seismic integrity. The change does not jeopardize the intended function of the design. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i Plant Design Change Evaluation HP-1-87-030 Plant Design Change Evaluation Number HP-1-87-030 entitled "Mechanical 1
Pressure Regulator Oil Supply Line" is complete.'
l Description of Change The nonfiltered oil supply to the Mechanical Pressure Regulator System vas rerouted through an existing, manually operated, filtering system that was not being used.
Reason for Change The modification was implemented to reduce the amount of lubricating oil impurities which might pass to the Mechanical Pressure Regulator and interfere with system operation.
! Safety Evaluation i A malfunction of the Mechanical Pressure Regulator due to impurities i in the oil, which results in simultaneous closure of both turbine l 1 control and bypass valves, is not a new type failure. The increase in ;
! the reliability of operation of the Mechanical Pressure Regulator [
System that was gained as a result of the filter addition more than 1 offsets the unlikely possibility of failure. Therefore, this modifi-J cation does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Evaluation HP-1-87-034
. Plant Design Change Evaluation Number HP-1-87-034 entitled "Fuel [
Storage Tank Additives" is complete.
Description of Change A program was implemented to add an antioxident/blocide mixture to the Diesel Fuel Oil and the Gas Turbine Jet Fuel storage tanks.
Reason for Change .
The program was implemented to reduce the possibility of water and microbiological growth in the storage tanks that could adversely affect the operability of the Emergency Diesel Generator and the Gas
- Turbine.
Safety Evaluation _
The chemical composition of the additive is carbon, hydrogen and a
oxygen. There is no sulfur or sodium in the mixture which could cause a degradation of the turbine blades or erosion of the diesel compo-
, nents. The modification does not adversely affect the operation of i the Emergency Diesel Generator or Gas Turbine. Therefore, this modi- -
fication does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Evaluation HP-1-87-061 Plant Design Change Evaluation Number HP-1-87-061 entitled "GEZIP Injection Pump Flush Lines" is complete.
Description of Change Stainless steel tubing, valves, and a tank vere installed on the General Electric Zine Injection Passivation (GEZIP) skid.
Reason for Change The Hodification was implemented to allow the GEZIP injection pumps to be flushed with demineralized vater, when they are taken out of service.
Safety Evaluation The modification does not jeopardize the intended function of the system and the reliability of the GEZIP system is increased by eliminating the possibility of pump plugging. Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i PROCEDURE CHANGES There vere no changes to procedures as described in the Final Safety Analysis Report (FSAR) during 1987.
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JUMPER-LIFTED LEAD-BYPASS (J-LL-B) CHANGES INDEX J-LL-B Number J-LL-B Title 1-87-008 Simulated Signal to APRM Channel 6 1-87-014 Diesel Day Tank Fuel Oil Supply 1-87-026 Scram Pilot Air Header Lov Pressure Scram-1-87-037 Reactor Protection System Motor Generator Power 1-87-047 Off Gas Radiation Monitor 1-87-048 1-AC-10 Interlock to HVE 5A and SB
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Jumper-Lifted Lead-Bypass Change Number 1-87-008 Jumper Lifted Lead Bypass Number 1-87-008 entitled "Simulated Signal to APRM Channel 6" is complete.
Description of Change A jumper vas installed to provide a simulated signal to Local Power Range Monitor (LPRM)36-45A in Average Power Range Monitor (APRM)
Channel 6 that was equal to the real LPRM reading at full power, equilibrium conditions. The jumper has been removed.
Reason for Change ,
The simulated signal was input to facilitate troubleshooting the LPRM which experienced previous signal spiking.
Safety Evaluation The constant signal vould cause the APRM to indicate greater than actual power in the unanticipated event of a pover reduction and the Reactor Protection System would not be compromised. The LPRM vas not located in a limiting core location so fuel performance monitoring was not compromised. Therefore, this bypass jumper did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Jumper-Lifted Lead-Bypass Change Number 1-87-014 Jumper Lifted Lead Bypass Number 1-87-014 entitled "Diesel Day Tank Fuel Oil Supply" is complete.
Description of Change The thermally fused, in-line fire check valves on the Emergency Diesel Generator fuel supply line vere locked open by installing stainless steel lockvire from the valve handle to the valve body. The locks vere removed when the valve internals were removed.
Reason for Change The change vas implemented to ensure Emergency Diesel operability during a fire or seismic event which could cause the valves to close and shut off the fuel supply.
Safety Evaluation The valves are not required and did not serve any useful purpose as installed. There vas no change in the operation of the fuel oil system vith the jumper-bypass installed. The change ensured that the Emergency Diesel continued to have a fuel supply. Therefore, this bypass jumper did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Jumper-Lifted Lead-Bypass Change Number 1-87-026 >
Jumper Lifted Lead Bypass Number 1-87-026 entitled "Scram Pilot Air Header Lov Pressure Scram" is complete. :
Description of Change
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The reactor scram associated with lov scram header pressure was bypassed while the reactor was shutdown for refueling. The bypass was ,
removed upon reactor startup.
l Reason for Change
. The Technical Specification Order for Modification of License Concerning BWR Scram Discharge Systems, does not require the reactor scram to be operable in the Refuel or Shutdown mode. . :
i Safety Evaluation i The bypass of this function, during reactor shutdown, was addressed and evaluated in the original design and safety evaluation process.
Therefore, this bypass jumper did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Jumper-Lifted Lead-Bypass Change Number 1-87-037 Jumper Lifted Lead Bypass Number 1-87-037 entitled "Reactor Protection -
System Motor Generator Power" is complete.
Description of Change A temporary jumper was installed to provide control power to the Reactor Protection System Motor Generators while a station battery was taken out of service for preventive maintenance and discharge testing.
The jumper vas removed af ter the maintenance and testing was complete.
Reason for Change 1
2 The change vas implemented to satisfy the Technical Specification !
requirement for two Reactor Protection System electric power '
monitoring channels for an inservice Reactor Protection System Motor Generator Set.
j Safety Evaluation The intended function to prevent long term exposure of Reactor
, Protection System components to degraded voltage and frequency conditions, vas not adversely affected by the temporary jumper.
Therefore, this bypass jumper did not constitute an unreviewed safety :
question per the criteria of 10CFR50.59. t 1
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-Jumper-Lifted Lead-Bypass Change Number 1-87-047 }
i Jumper Lif ted Lead Bypass Number 1-87-047 entitled "Off Gas Radiation Monitor" is complete. l l
t Description of Change 1
-A jumper was installed to connect the off gas radiation monitor using-the idle flux-tilt radiation monitor' cable. At the time of this report, the: jumper had not been removed.
t Reason for Change
't The off gas radiation monitor cable was found to be degraded and the flux-tilt radiatica monitor cable was available for use and elec-
. trically identical.
I Safety Evaluation !
The jumper provides adequate repair of the identified deficiency ,
without impacting in-service equipment. Therefore, this bypass jumper - f does not constitute an unreviewed safety question per the criteria of 10CFR50.59. ,
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Jumper-Lifted Lead-Bypass Change Number 1-87-048 Jumper Lifted Lead Bypass Number 1-87-048 entitled "l-AC-10 Interlock to HVE 5A and 5B" is complete.
Description of Change A jumper was installed to bypass the interlock which prevents the exhaust to standby gas valve from being open concurrently with the standby gas treatment inlet from Reactor Building exhaust valves. The jumper has been removed.
Reason for Change The interlock between the valves was bypassed to eliminate the possibility of a single failure which could result in the inoperabil-ity of both trains of a safety-related system.
Safety Evaluation The Reactor Building ventilation system is continuously monitored and isolates if high radiation levels are detected and the normal venti-lation system vill isolate if signals indicative of a Loss of Coolant Accident are received. The jumpers did not impact the ability to maintain offsite doses within required limits. Therefore, this bypass jumper did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i SETPOINT CHANGE INDEX l SCR Number Title 1-86-002 Floor Drain Collector Tank 1-86-005 Condensate Jemineralizer Spray Indicator 1-86-006 Condensate Demineralizer Flov Indicator 1-86-008 Stetor Hydrogen Purity 1-86-013 Main Steam Line Lov Pressure Containment Isolation 1-87-001 PI-131, All CRD Module Gauges 1-87-003 Torus Temperature Recorder 1-87-004 Reactor Flange and Shell Temperature 1-87-005 Local Power Range Monitor Voltage 1-87-007 Condensate Storage Tank Alarm 1-87-010 Off Gas Recombiner Instrumentation Loop 1-87-012 Area Radiation Monitor Tolerance 1-87-014 Torque and Limit Switch, 1-CN-69 1-87-015 Torque and Limit Svitch, 1-FV-4A 1-87-016 Torque and Limit Switch, 1-FV-4B 1-87-017 Torque and Limit Switch, 1-IC-2 ,
1-87-018 Torque and Limit Svitch, 1-IC-3 1-87-019 Torque and Limit Svitch, 1-IC-4 1-87-020 Torque and Limit Svitch, 1-IC-10 1-87-022 Turbine First Stage Pressure Sensor 1-87-023 Augmented of f-Gas Xenon Krypton Alarm
) 1-87-024 Dryvell High Range Area Radiation Monitor 1-87-025 Main Steam Line Radiation Monitor 1-87-026 Steam Tunnel Ventilation Radiation Monitor
! 1-87-028 Fuel Pool Gate Drain i
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i i Setpoint Change Number 1-86-002
- Setpoint Change Number 1-86-002 entitled "Floor Drain Collector Tank" is complete.
{
Description of Change j The level range for the Floor Drain Collector tank level transmitter I
was changed from 6.5 inches - 169 inches to 17 inches - 177.5 inches. .
C Reason for Change J The change vas implemented to. utilize field measured level rnnge which l l 1s more accurate and vill enhance proper radvaste processing.
Safety Evaluation l
l The change provides an accurate level indication of the tank and does i
not affect any safety-related system. Therefore, this change does not ;
constitute an unrevieved safety question per the criteria of i 10CFR50.59. ;
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Setpoint Change Number 1-86-005 Setpoint Change Number 1-86-005 entitled "Condensate Demineralizer Spray Indicator" is complete.
Description of Change The calibration tolerance of the condensate demineralizer spray and sluice water flov indicator vas changed from 1 percent to 2 percent.
Reason for Change The change was implemented to eliminate the difficulty in obtaining repeatability of 1 percent of an inch of water differential pressure calibration.
Safety Evaluation The instrument provides water flov indication only and does not prcvide any automatic or safety-related functions. The change does not adversely affect the proper operation of the condensate demin-eralizer system. Therefore, this change does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 1-86-006 Setpoint Change Number 1-86-006 entitled "Condensate Demineralizer Flov Indicator" is complete.
Description of Change The calibration tolerance of the condensate domineralizer recirculation flov indicator was changed from 1 percent to 2 percent.
Reason for Change The change vas implemented to eliminate the difficulty in obtaining repeatability of 1 percent of an inch of water differential pressure calibration.
Safety Evaluation The change has no adverse impact on the proper operation of the condensate demineralizer, and the indicator does not perform any safety-related functions. Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 1-86-008 Setpoint Change Number 1-86-008 entitled "Stator Hydrogen Purity" is complete.
Description of Change A tolerance requirement for the generator stator hydrogen purity instrumentation loop vas established to be within 2 percent of full scale.
Reason for Change The change was implemented to formalize stator hydrogen purity, indication, alarm, calibration and acceptance criteria in the surveil-lance procedures where a tolerance had not been specified.
Safety Evaluation The instrumentation loop provides indication and alarm functions only and does not provide any automatic trips or isolations. The instru-mests are not safety related, nor do they interface with safety-related components. Therefore, this change does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 1-86-013 l
, Setpoint Change Number 1-86-013 entitled "Main Steam Line Lov Pressure !
Containment Isolation" is complete.
Description of Change .
The setpoint of the lov pressure main steam line isolation signal was :
increased from 837 psig to 850 t.12 psig. i Reason for Change The change was implemented to provide a larger margin to the Technical Specifications Limit and to minimize the consequences of instrument ;
drifts. !
Safety Evaluation j i-Analysis has shown that the resultant reactor level char.ge is minor -
, and reasonable setpoint drifts down from 825 psig would have no effect f 4
on analyzed transients. The setpoint change vould result in less t I
severe transients. Therefore, this change does not constitute an !
4 unreviewed safety question per the criteria of 10CFR50.59. l t
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Setpoint Change Number 1-87-001 Setpoint change Number 1-87-001 entitled "PI-131, All CRD Nodule Gauges" is complete.
Description of Change 1
l The calibration accuracy requirement for the hydeaulic control unit I accumulator pressure indicators was decreased from 0.5 percent of full l
scale to 1.0 percent of full scale.
l Reason for Change l
l The change was implemented te reduce calibretion time required to I achieve the previous set point which was difficult to read on the
! instrument face. The reduced time results in reduced radiation l exposure to personnel.
Safety Evaluation Operational procedures limit high accumulator pressure vell belov that j vhich could cause drive damage and lov pressure is monitored by a i
separate svitch. Therefore, the change in instrument accuracy does not increase the potential for drive damage or a slov scram time.
l Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 1-87-003 Setpoint Change Number 1-87-003 entitled "Torus Temperature Recorder" is complete.
Description of Change The setpoint on the torus temperature recorder was changed from 89 i 0.5 degrees Fahrenheit to 87 2 2.0 degrees Fahrenheit.
Reason for Change The recorders have 2 degree Fahrenheit divisions and reading and calibrating the instrument to 0.5 degrees was difficult. The Technical Specification Limit is 90 degrees Fahrenheit and remains unchanged.
Safety Evaluation The modification increases the repeatability of the recorder and reduces the possibility of exceeding the limit. The recorder alarms only and does not perform any control function. Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 1-87-004 -l Setpoint Change Number 1-87-004 entitled "Reactor Flange and Shell ;
i
! Temperature" is complete.
Description of Change ,
I The calibration tolerance on the reactor flange and shell temperature r
i recorder was changed.
Reason for Change f
The tolerance vas changed from 3 degrees to i5 degrees to facilitate <
precise calibration because the recorder scale has 10 degree increments. l t
- Safety Evaluation The Technical Specification Limit for the shell flange to shell i
temperature-differential was not changed and readability of the device by control Room operators was not affected. Therefore, this change ,
does not constitute an unreviewed safety question per the criteria of i 10CFR50.59. j i
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i Setpoint Change Number 1-87-005 Setpoint Change Number 1-87-005 entitled "Local Power Range Monitor i Voltage" is complete. :
1 Description of Change :
The voltage supplied to the nev. type of Local Power Range Monitors was I
decreased from 130 volts to 100 volts. I Reason for Change The change was implemented to ensure detector operation on the linear !
j plateau region of the Current versus Voltage curve. !
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! Safety Evaluation t t
The adjustment of the power supplies does not affect the sensitivity I l or life of the detectors or their associated electronic signal !
conditicsing equipment. Therefore, this change does not constitute an l unreviewed safety question per the criteria of 10CFR50.59. l t
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Setpoint Change Number 1-87-007 l 2 !
Setpoint Change Number 1-87-007 entitled "Condensate Storage Tank
- Alarm" is complete. !
I Description of Change
~
The lov ve.ter level alarm'for the Condensate Storage Tank vas tempo- ;
rarily increased from 62.3 percent level to 92 percent level. I i
Reason for Change i
- The temporary setpoint was implemented to provide an alarm to ensure i that the Condensate Storage Tank inventory remained above the required i 1 minimum level while the torus was drained during the refueling outage. L 1 f' Safety Evaluation L +
l The change only affected the level alarm function and had no affect on l vater level indication. Prior to startup, the lov vater level alarm l j setpoint was restored to its original value. Therefore, this change !
does not constitute an unreviewed safety question per the criteria of t 10CFR50.59.
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Setpoint Change Number 1-87-010 Setpoint Change Number 1-87-010 entitled "Off Gas Recombiner Instrumentation Loop"'is complete.
Description of Change The calibration tolerances for the pressure transmitter and the pressure indicator on the off gas recombiner preheater inlet pressure instrumentation loop were changed to meet the manufacturer's specifications.
Reason for Change The change vas implemented to allow the instrument technician to obtain a meaningful calibration without having to make numerous adjustments to obtain an accuracy that is greater than the manufacturer specifies. ')
Safety Evaluation The resolution on the indicator face is such that the tolerance change-is not detectable so the Control Room operator's ability to read the instrument is not affected. The system performs no safety-related function. Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 1-87-012 Setpoint LChange Number;1-87-012 entitled "Area Radiation Monitor
~ Tolerance" is complete. -
Description of Change The calibrat ion accuracy for.the dryvell high. range area radiation monitor ve; changed to a more realistic tolerance. l Reason'ior Change The change was implemented to facilitate a more realistic calibration accuraty that takes into consideration background radiation, internal source, and calibration source errors.
Safety Evaluation The change only applies to the source calibration so the' tolerance for ,
the electronics calibration remains unchanged. The instruments do not' perform any automatic or safety-related functions. Therefore, this change does not constitute an unreviewed safety question per the' criteria of 10CFR50.59. .
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- - - , - . - - , . . . . - . - , - , - . . , . , . . - - . . i,,-. . , , , . . - . . - , . , - , - - .
Setpoint Change Number 1-87-014 Setpoint Change Number 1-87-014 entitled "Torque and Limit Svitch, 1-CN-69" is complete.
Description of Change The torque svitch setpoints and limit switch setpoints were adjusted to allow enough torque to open or close the valve, but not enough to cause damage.to_the valve.
- - Reason for Change The change was implemented to ensure design basis operability ofLthe.
motor operated valve.
Safety Evaluaticn The setpoint change does not interfere with the ability of the valve to perform its safety-related function. Therefore, this change does ,
not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 1-87-015 Setpoint Change Number 1-87-015 entitled "Torque and Limit Switch, 1-FV-4A" is complete.
Description of Change The torque switch setpoints and limit switch setpoints were adjusted-to allow enough torque to open or close the valve, but not enough to cause damage to the valve.
Reason for Change The change was implemented :o ensure design basis operability of the motor operated valve.
Safety Evaluation The setpoint change does not interfere with the ability of the valve to perform its safety-related function. Therefore, this-change does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 1-87-016 Setpoint Change Number 1-87-016 entitled "Torque and Limit Svitch, 1-FV-4B" is complete.
Description of Change The torque switch setpoints and limit switch setpoints were adjusted to allov enough torque to open or close the valve, but not enough to cause damage to the valve.
Reason for Change The change was implemented to ensure design basis operability of the motor operated valve.
Safety Evaluation The setpoint change does not interfere with the ability of the valve to perform its safety-related function. Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 1-87-017 Setpoint Change Number 1-87-017 entitled "Torque and Limit Switch, 1-IC-2" is complete.
Description of Change The torque switch setpoints and limit switch setpoints vere adjusted to allow enough torque to open or close the valve, but not enough to-cause damage to the valve.
Reason for Change The change was implemented to ensure design basis operability of the motor operated valve.
Safety Evaluation The setpoint change does not interfere with the ability of the valve ~
to perform its safety-related function. Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 1-87-018 Setpoint Change Number 1-87-018 enti..ed "Torque and Limit Switch, 1-1C-3" is complete.
Description of Change The torque switch setpoints and limit switch setpoints were adjusted to allow enough torque to open or close the valve,.but not enough to cause damage to the valve.
Reason for Change The change was implemented to ensure design basis operability of the motor operated valve.
Safety Evaluation s
The setpoint change does not interfere with the ability of the valve to perform its safety-related function. Therefore, this change does not constitute an unreviewed safety question-per the criteria of 10CFR50.59.
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L Setpoint Change Number 1-87-019 Setpoint Change Number 1-87-019 entitled "Torque and Limit Switch, 1-IC-4" is complete.
Description of-Change The torque svitch setpoints and limit switch setpoints were adjusted to allow enough torque to open or_close the valve, but not enough to cause damage to the valve.
Reason for Change The change was implemented to ensure design basis operability of the motor operated valve.
Safety Evaluation The setpoint change does not interfere with the ability of the valve to perform its safety-related function. Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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l Setpoint Change Number 1-87-020 Setpoint Change Number 1-87-020 entitled "Torque and Limit Svitch, 1-IC-10" is complete, j
~ Description of Change The torque switch setpoints and limit switch setpoints were adjusted-to allow enough torque to open or close the valve, but not enough to cause damage to the valve.
l Reason for Change The change was implemented to ensure design basis operability of the motor operated valve. !
Safety Evaluation The setpoint change does not interfere with the ability of the valve to perform its safety-related. function. Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint change Number 1-87-022 Setpoint Change Number 1-87-022 entitled "Turbine First Stage Pressure Sensor" is complete.
Description of Change The setpoint for the pressure switches that sense turbine first stage pressure to provide for bypass of the reactor scram due to turbine stop valve closure and turbine control valve fast closure was decreased.
Reason for Change The change was implemented to allov for instrument drift and ensure that the Technical Specifications Limit is not exceeded.
Safety Evaluation The reduction of.the setpoint is in the conservative direction which results in an increase in the margin to the Technical Specification Limit. Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 1-87-023 entitled "Augmented Off-Gas Xenon Krypton Alarm" is complete.
Description of Change The Augmented Off-Gas Xenon Krypton gas cooler outlet temperature alarm setpoint was raised from 10 degrees Fahrenheit to 15 degrees Fahrenheit.
Reason for Change The change was implemented to allov for the changes in efficiency of the Reactor Building Closed Cooling Water or Glycol cooling system through seasonal changes when the outlet temperature operates between 8 and 12 degrees Fahrenheit.
Safety Evaluation The as-built efficiency of the off gas system far exceeds the require-ments of the plant due to new fuel designs and improved cladding. If the off gas system should fail, off gases can be routed to the off gas delay line. The setpoint change is for the alarm function only and does not affect any automatic control systems. Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i Setpoint change Number 1-87-024 i Setpoint Change Number 1-87-024 .atitled "Dryvell High Range Area Radiation Monitor" is complete.
Description of Change The calibration accuracy for the dryvell high range area radiation monitor was changed to-be within a tolerance of a factor of two.
Reason for Change The calibration accuracy for the dryvell high range area radiation monitor was changed to comply with Regulatory Guide 1.97, Revision 3.
Safety Evaluation The change only applies to the source calibration so the tolerance for the electronics calibration remains unchanged. The instruments do not
~
perform any automatic or safety-related functions. Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 1-87-025 Setpoint Change Number 1-87-025 entitled "Main Steam Line Radiation Monitor" is complete.
Description of Change The Main Steam Line Radiation Monitor alarm and trip setpoints were temporarily changed from three times and seven times normal full power background respectively, to tventy-four times and fifty-six times, respectively. The setpoints have been returned to their original levels.
Reason for Change The temporary change was implemented to facilitate the Hydrogen Vater Chemistry Pre-Implementation test, during which the Main Steam Line Radiation Monitors were monitoring higher background radiation levels.
Safety Evaluation The only accident which takes credit for these monitors is'the Control Rod Drop Accident, and then only at power levels less than 10 percent.
This change only allowed setpoint changes above 20 percent power.
Therefore, this change did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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l Setpoint Change Number 1-87-026 Setpoint Change Number 1-87-026 entitled "Steam Tunnel Ventilation Radiation Monitor" is complete.
Description of Change The steam tunnel ventilation radiation monitors setpoint was l permanently increased from twelve millirem per hour to fifty millirem per hour. Also, the trip setpoint was further increased on a temporary basis up to one hundred millirem per hour to facilitate the Hydrogen Vater Chemistry Pre-Implementation test.
Reason for Change The change was implemented to accommodate the higher background radiation levels observed during the Hydrogen Vater Chemistry Pre-Implementation test.
Safety Evaluation This setpoint is based on preventing allovable release limits to be exceeded. A recalculation of those limits and their relation to this setpoint shows that a fifty millirem per hour limit is still con-servative.
The only accident analysis that takes credit for these monitors is the Control Rod Drop Accident, and then only at power levels less than 10 percent. The temporary change only allowed setpoint changes above 20 percent power. Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
- 157 -
Setpoint Change Number 1-87-028 Setpoint Change Number 1-87-028 entitled "Fuel Pool Gate Drain" is complete.
Description of Change The setpoint for the switch that detects leakage from the spent fuel pool was changed to alarm at 7.5 gallons per minute instead of 10 gallons per minute.
Reason for Change The change was implemented to alert the operator of a leak in a much shorter time. Due to the small size of the sensing line, the 10 gallons per minute flow rate was barely reached.
Safety Evaluation The setpoint change is in a conservative direction and has a bene-ficial effect on safety. The change does not adversely affect any safety-related systems. Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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TESTS INDEX Number Title T-87-1-02 APRM Scram and Annunciator Relay Sequence T-87-1-14 HP1 Torsional Test i
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- 159 -
i Number Title T-87-1-02 APRM Scram and Annunciator Relay Sequence Description of Test A relay sequence test of the Reactor Protection System. scram and annunciator circuit for Average Power Range Monitor channels one and six was conducted to verify the existence of a time delay between the circuits.
Rf ason for Test The test was conducted to demonstrate the proper trip sequence and confirm the integrity of the Average Power Range Monitor trip logic to annunciate and actuate the Reactor Protection System scram relays on Hi-Hi flux conditions.
Safety Evaluation The test procedure required testing each channel individually and tripping the channel before testing which is similar to normal surveillance testing. No additional equipment was added which could introduce a new failure mode. Therefore, the test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i Number Title T-87-1-14 MP1 Torsional Test Description of Test A torsional test was performed consisting of a speed ramp from 1350 to 1980 rpm with 8 percent negative phase sequence current applied and dwells performed at various speeds.
Reason for Test The test was performed to determine if high order resonant torsional frequencies are present around that which could create a reliability problem for the last stage buckets of the main turbine.
Safety Evaluation Reactor power was maintained below the point at which a turbine trip causes a reactor trip. The Turbine Generator was operated within safa limits during the test. Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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EXPERIMENTS There were no experiments performed under the provisions of Title 10, Code of Federal Regulations, Section 50.59 during 1987.
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i CHALLENGES TO RELIEF VALVES In accordance with the commitment made under Item II.K.3.3 of NUREG 0737 (THI Action Plan) in the V. G. Counsil letter to D. G. Eisenhut, dated June 10, 1980, the following is a report of challenges.to relief / safety valves during 1987.
There were no challenges to relief / safety valves during 1987.
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I MILLSTONE UNIT 2 CONTENTS SECTION PAGE Changes Design Changes . . . . . . . . . . 1 ;
Design Change Evaluations . . . . . . 30 ;
Procedure Changes . . . . . . . . . . 34 Jumper-Lifted Lead-Bypass Changes . . . 43 Setpoint Changes . . . . . . . . . 56 Tests . . . . . . . . . . . . . . 66 Experiments . . . . . . . . . . . . 92 -[
Occupational Radiation Exposure . . . . . . 93 4
Challenges to Relief Valves . . . . . . . 94
- Primary Coolant Iodine Spiking . . . . . . 95 Steam Generator Tube ISI . . . . . . . . 96 i
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PLANT DESIGN CHANGE INDEX PDCR Number Title 2-039-85 Replacement of Piping and Adding Flanges to the Service Vater Discharge to the Reactor Building Closed Cooling Water (RBCCV) System Heat Exchangers 2-046-85 Process' Computer Replacement - Control Room Operator Console Replacement 2-071-85 Vital Direct Current (DC) Switchgear Chiller Vater System 2-085-85 Spent Fuel Pool Rerack Preliminary Work 2-001-86 Update Design of the Reactor Coolant Pump (RCP) Motors Upper Guide Bearing 2-017-86 Pressurizer Pressure Control Channel Deviation Alarm 2-029-86 Process Computer Replacement - Nonvital Point Transfer 2-037-86 Automatic Svitch Company (ASCO) Solenoid Valve Change-out 2-065-86 Lifting Eye Near the Blovdown Quench Tank 2-068-86 Reactor Building Closed Cooling Vater (RBCCV)
Pump Motor Base Plate Modifications 2-094-86 Check Valve 2-SV-13A and 2-SV-13B Replacement 2-100-86 Replacement of Diesel Generator Differential Relays 2-106-86 Change the Auxiliary Feed Pump Room Sumps Power Supply to Vital 2-002-87 Seat Ring Haterial Change 2-003-87 Process Computer Replacement - IBM Removal 2-004-87 High Radiation Area Gate Alarms and Varning ,
i Lights i 2-005-87 Steam Cenerator Tube Stabilizers l
2-007-87 Fuel Consolidation Demonstration Program l
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PLANT DESIGN CHANGE INDEX (Continued) 1 PDCR Number Title 2-008-87 Installation of A Manual / Air Actuator For Appendix R on 2-SI-657 2-011-87 Process Computer Replacement - Heteorological Data Input 2-013-87 Main Exhaust Fans Suction Isolation Dampers 2-016-87 Condensate Polishing Facility Improvements -
Vaste Neutralization Sump Discharge System 2-017-87 Microvave Antenna Attachment to Millstone .
Unit 2 Condensate Polishing Facility (CPF)
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Plant Design Change Number 2-039-85
~ Plant Design Change Number 2-039-85 entitled "Replacement of Piping
,and Adding Flanges to the Service Vater Discharge to the Reactor !
Building Closed Cooling Water (RBCCV) System Heat Exchangers" is ,
complete. ,
Description of Change
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Portions of the' cooling water (Service Water) discharge piping from the RBCCV heat exchangers and the service water supply to the emer-gency diesel generators were replaced with stainless steel piping and additional flanges were installed. An existing support was also modi-fled as a result of this change.
Reason for Change The existing carbon steel piping was replaced with type 316 stainless steel because of severe corrosion / erosion problems. The stainless steel piping provides greater resistance to corrosion and eliminates the effects of erosion. The nev flanges vere installed to provide ease of maintenance and to provide a transition point'from stainless to carbon steel. The support was modified to accommodate the increased seismic loading from the weight of the nev' flanges. l Safety Evaluation These piping replacement and support modifications have been designed i in accordance with ANSI B31.1 1973 Edition and with the AISC Manual for Steel Construction 8th Edition. These codes and standards are consistent with the original design requirements of the Millstone Unit 2 Final Safety Analysis Report Section 6.0.
The existing system is comprised of ASTM A-53 GR.B carbon steel pipe.-
Those portions of the systems that' vere modified used ASTM-A-312 TYP !
316L stainless steel pipe. This type of pipe has a higher minimum tensile strength and a greater allovable_ stress when compared to the :
existing carbon steel and therefore exceeds the original system design ;
requirements. It should be noted that the new flanges installed were i' carbon steel. The carbon steel flanges are consistent with the orig-inal design requirements and the installation of these items did not !
compromise the system integrity. In addition, the postulated failure l i
of any one section of piping vill not preclude the intended function '
of the system because of its two header design. ,
i Also, these modifications did not require any dissimilar metal l velding, and all transitions from stainless to carbon steel vere made ;
vith the use of insulating flanges to prevent galvanic reactions from l occurring.
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fj Plant Design Change Number 2-039-85 (Continued)
Safety Evaluation (Continued)
Engineering reviewed the proposed replacement pipe material (316L) for acceptability in the service vater environment. This material is superior to the current material in the high flow areas, but may suffer pitting attack in. stagnant or lov flov areas. Through routine monitoring of service water pipe performance, any leakage or degrada- ,
tion vill be detected and repairs implerented. However, based on this, stainless steel was considered an acceptable replacement material.
Therefore, this modification does not conetitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 2-046-85 Plant Design Change Number 2-046-85 entitled "Process Computer Replacement - Control Room Operator Console Replacement" is complete.
Description of Change ,
This change replaced the Control Room operators' console with a new console, added a new telephone console, installed two new printers associated with the new computer system, relocated four existing printers associated with the new computer system, and installed two new monitors and keyboards on the new operators' console.
Reason for Change The new operators' console was compatible with the new process computer replacement system Control Room peripherals. The printers ,
vere required to support the increased output of.the new computer.
The new telephone console was needed to accommodate the communications changes / additions.
Safety Evaluation None of the equipment listed in this change performs a safety-related function. However, because the operator console and the two new printers vere located in the vicinity of the main control boards, !
their mounting was seismic Class II/I controlled. The installation of ,
the other equipment was not safety related. j l l The conduit and supports that were installed in the cable vault vere l installed using Category I criteria. The failure of this equipment from an electrical standpoint, or the possible electrical degradation of the existing equipment was evaluated and was found to be satis-factory. l l
1 Reviews were made regarding the installation of the safety-rela'.eu '
equipment and they demonstrated that these modifications met all the mechanical and seismic design bases.
Therefore, this modification does not constitute an unreviewed safety ;
question per the criteria of 10CFR50.59. l l
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Plant Design Change Number 2-071-85 Plant Design Change Number 2-071-85 entitled "Vital Direct Current (DC) Switchgear Chilled Vater System" is complete.
Description of Change The two service water to freon heat exchangers for the vital Direct '
Current (DC) switchgear chilled vater system were replaced. Piping, supports, int' connections were modified to accommodate the new heat exchangers. A third heat exchanger was retired in place.
Reason for Ch.tnge i The two previous vital heat exchangers were rendered inoperable due to corrosion damage. The third, nonvital unit, was retired because the new units provided adequate cooling during accident conditions and the auxiliary chilled voter system provided cooling during normal opera-tion.
Safety Evaluati,on This change vas a one for one replacement of the existing heat >
exchangers with all the original design requirements being met. The replacement units were designed, tested, inspected, and stamped in accordance with the ASME Boiler and Pressure Vessel Code Section VIII, ;
Division I. These h?at exchangers as well as all the necessary fit-up piping and valves, have been seismically qualified. The new heat exchangers vere improved by increasing their corrosion resistance.
They were sized to accommodate 100 percent of the post accident heat load.
Therefore, this modification does not constitute an unrevieved safety l l
question per the criteria of 10CFR50.59.
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Plant Design Change Number 2-085-65_
Plant Design Change Number 2-085-85 entitled "Spent Fuel Pool Rerack Preliminary Vork" is complete.
Description of Change A number of modifications were required 1. order 'o accomplish the ;
rerack of the spent fuel pool. The tempor m modirications included installation of a temporary spent fuel rack in the cask laydown area, installation of a temporary fixture to hold fuel skeletons, removal of the drain line piping in the cask laydown area, removal of the spent fuel skimmer suction line, removal of the pump in the cask laydovn area, removal of the fuel inspection periscope mounting bracket, and relocation of the new fuel inspection stand. The permanent modifi-cations included removal of a lighting bracket in the cask laydown area, modification of the spent fuel pool platform crane overhead hoist south side hard travel stop, and removal of the spent fuel inspection machine.
Reason for Change These activities vere prerequisite to the reracking of the spent fuel pool.
Safety Evaluation The temporary removal of the cask laydown area piping / pump and the skimmer line did not compromise the ability to supply cooling to the spent fuel system. The cask laydown area drain piping does not have a connection to any cooling system and runs from the cask laydown area to the spent fuel pool. Removal of the piping eliminated the poten-tial of any capillary action when the laydown area was drained. The overall filtration of the spent fuel pool system was not affected.
This was due to the fact that the skimmer lines were disconnected at the ports and the . system was still functional. In addition, the primary means of filtering the pool was through the spent fuel cooling system. The piping was reinstalled in accordance with the original requirements of ANSI B31.1 Code, 1973 Edition and ASME Section III, f Subsection ND - 1973 Edition.
l Also, the consequences of a failure of the cask laydown area gate vere revieved and was found acceptable.
The temporary removal of the nev fuel inspection machine, the tem-porary removal of the fuel pool inspection periscope mount and the
, permanent removal of the lighting bracket vere deemed acceptable l because these components were not considered part of a functioning i
system. In essence, these items vere trals that vere used in the storage and receipt of fuel and their relocation had no impact on any plant system.
l Plant Design Change Number 2-085-85 Safety Evaluation (Continued)
The temporary installation of a spent fuel storage rack in the cask laydown area was considered acceptable provided no fuel was moved into the rack. The installation of this rack vas similar-to placing a cask in the laydovn area. This modification was within the bounds of pre-viously analyzed accidents.
The installation of the temporary rack to hold fuel skeletons had no impact on the spent fuel pool and was designed in accordance with the AISC nalual for steel construction.
In order to provide access to fuel storage locations after the rerack, the spent fuel pool platform crane overhead hard travel stops were extended approximately four inches. This modification did not com-promise the existing analysis of the platform crane and was consistent with the original plant design critcria for this component.
The spent fuel inspection machine was structurally attached to the pool liner and was removed with remote tooling. The removal of this component did not prevent the pool or other safety-related equipment from performing its intended function. Procedural and administrative controls were in place to preclude damage to the liner, while this work vas being performed. In addition, any damage to the liner causing a leak would be accounted for by system design which incor-porates detection and isolation capabilities to prevent a loss of pool ,
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Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 2-001-86 Plant Design Chr.nge Number 2-001-86 entitled "Update Design of the i
Reactor Coolant Pump (RCP) Motors Upper Guide Bearing" is complete.
Description of Change The C Reactor Coolant Pump (RCP) motor upper guide bearing (a plain cylindrical type) vas replaced with a segmental shoe (tilt pad) type bearing incorporating eight shoes. Hodifications have been made to the lube oil system to suit the new bearing. Two of the eight shoes have incorporated built-in temperature detectors (one operational and one spare).
l Reason for Change l The old plain cylindrical journal bearing design was susceptible to an oil whirl instability. All of the RCP motors had exhibited oil whirl characteristics to some extent and due to the configuration of each motor installation, each vibrated at different amplitudes. Thu oil vhirl caused a high potential for accelerated degradation of '.he upper guide bearing during start-up and normal operation. Replacraient of these bearings provided an improvement in the units' overs 11 reli-ability.
l l Safety Evaluation l The system affected by this modification was the RCP. The design basis accidents which vere affected by this change vere the loss of Reactor Coolant System flov and RCP seized rotor events. The new bearing design did not affect operability of the RCPs in a harsh environment caused by a high energy line break or by control rod l ejection. .
I Test data has shown that the coastdown rate with the new bearing was l infinitesimally slover. This had an equally small, w .:rtheless bene-ficial, effect to the loss of Reactor Coolant System flov transient.
The only credible failure mode associated with the change was inade- '
l l quate bearing lubrication and subsequent pump coastdown. The pump l
coastdown resulted in a loss of Reactor Collant System flov event, l vhich was already analyzed as a design basis event.
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1 The lubrication lines which feed oil from the lov pressure header to the bearing vere seismically designed with a higher safety factor of l
I 6.20 (versus 4.0 in the existing design). Therefore, the probability I
of a abrication line failing and causing inadequate bear $ng lubri-cation h small.
l l This change ind no ef fect on the results of the RCP seized rotor event as analyzed. This was because the analysis assumed that the core flow decreased instantaneously to its three pump value, which remains the limiting scenario regardless of RCP pump or motor bearing design.
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Plant Desi gn Change Number 2-001-86 (Contiw ed)
Safety Evaluation (Continued)
Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 2-017-86 Plant Design Change Number 2-017-86 entitled "Pressurizer Pressure Control Channel Deviation Alarm" is complete.
Description of Change A deviation alarm circuit was added to the Pressurizer-Pressure !
control loop.
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, Reason for Change i
J This alarm provided the Control Room operator with early indication of i- pressure deviation from setpoint for pressurizer pressure control.
Safety Evaluation The alarm circuit provided alarm function only. All components ~
utilized for this change were IEEE 323-1974 and IEEE 344-1975 >
1 qualified. The new circuitry did not have a failure mode which could ,
j impact safoty-related equipment. There was no single failure mode i i
which could impact the operation of the pressure control loops. The
! change had no impact on plant response to operational parameters. ,
There were no credible failure modes which could become initiating events resulting in the degradation of the. protective boundaries. ;
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 2-029-86 Plant Design Change Number 2-029-86 entitled "Process Computer Replacement - Nonvital Point Transfer" is complete.
Description of Change This change added a new cable tray within the cable vault to the analog input cabinet and transferred 510 of the nonvital computer inputs from the existing plant computer to the new Integrated Computer System Input Cabinets.- Some existing field cables vould not extend into the new cabinets and an inline splice was made to obtain the proper length.
Reason for Change Transfer nonvital cables to the new computer prior to plant shutdown without affecting the operation of the safety features of the old computer.
Safety Evaluation The transferred cables consist of a single analog, digital, or pulse input to the plant process computer. The points transferred vere reviewed to insure they would not affect the safety operation of the existing computer. They were chosen because they originated from nonsafety-related instrument loops, did not provide an input to any special application programs, and vere not required for automatic control of plant operations.
Since all the cables transferred vere from nonsafety-related instrument loops, they were not utilized in the determination of any design basis accident analysis.
Any failure of a transfer vill only affect the instrument it is connected to. Since these points were nonsafety related and vere not required for plant operation, the failure modes associated with this change vould not represent a new unanalyzed accident. l Therefore, this modification does not constitute an unreviewed safety questien per the criteria of 10CFR50.59.
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Plant Design Change Number 2-037-86 .
I Plant Design Change Number 2-037-86 entitled "Automatic Switch Company (ASCO). Solenoid Valve Change-Out" is complete. ,
Description of Change This change replaced six nonqualified solenoid valves with environ-mentally qualified solenoid valves. The model VPHT 8300 series was replaced with the model 206-380 series.
t Reason for Change ,
The solenoid valves replaced were on valves 2-AC-12, 2-AC-15, 2-AC-20, 2-AC-47, 2-EB-92, and 2-EB-99. These valves were originally installed with the splice connecting the valve to the power supply with the coil ,
housing. Temperatures within the coil housing vere higher than the- ,
design temperature for the splice. Therefore, with the new valves, the splice was made outside the coil housing in a lov temperature area
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negating the possibility of splice failure due to higher than design temperatures.
Safety Evaluation The replacement of the solenoid valves was one-for-one and the design function of the solenoids was not changed by this plant design change.
The operation of the valves remained the same. No additional supports or viring changes were required. ;
Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59. 3 1
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Plant Design Change Number 2-065-86 Plant Design Change' Number 2-065-86 entitled "Lifting Eye Near the Blovdown Quench Tank" is complete. i Description of Change This change installed a lifting eye on the Enclosure Building structural steel that can be used for attachment of a lifting device.
The capacity of the lifting eye is 1000 pounds.
Reason for Change Installation of this lifting ~ eye provided a safe method to lift equipment for repair / replacement work at the blovdown quench tank.
Safety Evaluation.
The lifting eye was designed out of standard structural steel and was velded to the underside of an existing structural beam. It has an 8:1 safety factor. The added 1000 pound load was reviewed from a static and seistic stai.dpoint and was found to be acceptable. Equipment in the local area was reviewed with respect to damage if a load being lifted vere to sving or fall. This review indicated that any equip-ment that could be made inoperable vould not cause an unanalyzed j accident.
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Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. ,
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l Plant Design Change Number 2-068-86 Plant Design Change Number 2-068-86 entitled "Reactor Building Closed Cooling Vater (RBCCV) Pump Hotor Base Plate Modifications" is complete.
Description of Change
- Replace the steel base plates at the Reactor Building Closed Cooling 1 Vater (RBCCV) motor feet with nev. base plates that would permit the mounting of the existing General Electric motors or a nev Vestinghouse.
motor.
Reason for Change The Appendix R program required that a spare motor be available for replacement of an RBCCV pump motor. No General Electric motors that vere Class lE qualified vere available in the motor frame size neces-sary for this application. A Vestinghouse motor was procured as a replacement. The Vestinghouse motor was an equivalent electrical replacement but its installation required the base plate to be changed. The base plate was changed in order to accommodate both types of motors. i Safety Evaluation i
The installation of the new base plates and the effect of the extra -
veight of the Vestinghouse motor was evaluated and was found to be
- acceptable. The operation of the new motor was not different than the original's therefore, existing plant procedures were used and there ;
vas no impact on the safety of the plant.
i This change did not have any impact to the correct functioning of Class lE equipment and systems.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 2-094-86 Plant Design Change Number 2-094-86 entitled "Check Valve 2-SV-13A and 2-SV-13B Replacement" is complete.
Description of Change This change replaced two 8-inch check valves located on the outlet side of the service water cooling to the diesel generators with check valves having a rubber lining.
Reason for Change The old check valves were damaged due to erosion / corrosion.
Safety Evaluation Failure of the new check valves did not have a different effect on safety or plant operation than the old check valves. The pressure rating of the nev valves was compatible with the system design pressure. The new valves did not affect flov of service water from the diesel generators. The neoprene rubber liner was compatible with the sea vater floving through it and would prevent damage to the valve in the future. The replacement did not affect any seismic analysis since the veight of the new valves was approximately the same as the old.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 2-100-86 Plant Design Change Number 2-100-86 entitled "Replecement of Diesel Generator Differential Relays" is complete.
Description of Change This design change replaced the differential relays on diesel generator feeder breakers A312 and A401.
Reason for Change The old relays vere General Electric 12CFD22B1A. It was determined that these relays vere not seismically qualified above the 14'-6" elevation. The relays are located on the 31'-6" elevation, therefore, they were changed to General Electric 12IJD52A11A. These new relays meet all seismic criteria for Hillstone Unit 2 and IEEE 322-1975.
Safety Evaluation The relays vere a one-for-one replacement. No mounting modification or viring changes vere required. The operation of the diesel generator feeder breakers remained the same. An analysis was per-formed that determined that the old relays were not qualified at any elevation above 14'-6". The new relays were chosen based on the seismic criteria developed in this analysis. Failure of these relays vould not change the response of the plant as previously analyzed.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR30.59.
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Plant Design Change Number 2-106-86 Plant Design Change Number 2-106-86 entitled "Change ~the Auxiliary Feed Pump Room Sumps Power Supply to Vital" is complete.
Description of Change This change installed a vital power supply to two existing sump pumps (P125 and P72B), one in each of the tvo auxiliary feed pump rooms.
Reason for Change !
This change improved the reliability of the sump pumps. The sump pumps vere previously povered from a nonvital supply. When this supply was out of service due to a storm, or for any other reason, vater vould accumulate on the floor in the pump rooms. Temporary pumps vere used to remove this Vater but their installation violated the watertight integrity of the steam driven auxiliary feed pump room. ,
This violation was deemed unacceptable and the power supply was -
changed so that the water could be removed utilizing the existing sump pumps.
Safety Evaluation New cable was installed between the vital Motor control Center (Mcc),
MCC B61, and the sump pumps. Its installation required use of existing cable trays and new conduit. Separation criteria was veri-fled as acceptable. Cable tray loading was verified acceptable. New ,
conduit was sized and installed as required for this seismic, safety-related application. The nonvital MCC cubicles were upgraded to safety related and reused in the vital MCC.
The installation of vital power to these two pumps did not impact any design basis accidents as listed in the Millstone Unit 2 Final Safety '
Analysis Report. This change did not impact operation of MCC B61 as J
the pumps were isolated by a breaker if a short was encountered at the pump (s). If a loss of normal power occurred, power vould be available for the sump pumps as soon as the diesel generators vere loaded as the sump pumps vere permanently attached loads. Addition of these loads to the vital bus did not create a nr.v accident as the new loads on the diesel generator vere reviewed and found acceptable.
Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Plant Design Change Number 2-002-87 Plant Design Change Number 2-002-87 entitled "Seat Ring Material Change" is complete.
Description of Change This change allowed the use of an alternate material for seat rings on some Fisher valves in the Charging and Safety Injection Systems. The original seat ring material was ASTM A276, grade 316, hard faced, and alloy #6 casting. The alternate material is ASME SA479, grade 316, hard faced.
Reason for Change The original material type was no longer available from the vendor.
But some spare parts that vere made out of the old material vere still available therefore the old material type was also left applicable. ;
Safety Evaluation The chemical and mechanical properties of the existing and alternate material was revieved. The results shoved that they are equivalent in these attributes. All these materials are high temperature service stainless steels. No material incompatibility existed for the condi-tions of service or with the other valve components. The seismic characteristics of the valve were not affected and the change did not adversely affect any safety-related systems.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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l Plant Design Change Number 2-003-87 Plant Design Change Number 2-003-87 entitled "Process Computer Replacement - IBM Removal" is complete.
Description of Change This change supplied three new inputs to the new computer from the Hillstone Unit 2 steam line radiation monitors (R4299A,B,6C), supplied an input to the nev computer from the Hillstone Unit 1 stack radiation monitor (R1705-79), supplied power to the computer resistance tempera-ture detector (RTD) cabinet, removed the old IBM computer, and installed a permanent stairway into the new computer room.
Reason for Change Thir, change completed the new computer input terminations and facility modifications.
Safety Evaluation These electrical changes did not have any adverse affect or increase the probability of a failure of any safety system. The existing instrument loops were reviewed with regard to circuit loading, separation, and isolation.
The failure modes associated with these electrical changes were reviewed and found not to have an impact on any safety-related system.
Isolation devices were tested with all the credible failures applied directly to the output, including fault voltages to 140 VAC and 140 VDC. This testing ensured that any credible failure of the nonsafety-related system did not affect the Class lE circuit. Based on this information, this change did not modify the plant response to the point where it was considered a new accident. Since all these points vere isolated from any Class lE circuits, the failure modes identified did not represent a new unanalyzed accident.
The addition of the three inputs from the Millstone Unit 2 steam line radiation monitors vere required for the Safety Parameter Display System. This additional display did not have any effect on the safety system but did enhance the instruments performance and met the opera-tional and licensing needs of the plant.
The removal of the old IBM computer and the installation of the stairs j into the new computer room did not affect plant operation because the ;
old computer had been disconnected and the stairs had been determined to not be a concern.
Therefore, this modification does not constitute an unreviewed safety l question per the criteria of 10CFR50.59. i l
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J Plant Design Change Number 2-004-87 ;
Plan't Design Change Number 2-004-87 entitled "High' Radiation Area Gate l Alaras and Varning Lights" is complete.
Description of Change This_ change provided reliable nonclosure alaras and warning lights at high radiation area access gates-which vere not equipped with these i
devices. l Reason for Change-High radiation area gates are normally locked and can be opened only vith a special key in order to control access. These alarms and .
- lights are used as a back-up and they provide assurance that the gates
- are closed when they are supposed to be. ,
Safety Evaluation -
i 4 The additional 350 vatts of electrical load that was added to the l
- local lighting branch circuits was reviewed and was found to be !
acceptable. The installation did not affect any safety-related l j systems, '
a j Therefore, this modification does not constitute an unreviewed safety' question per the criteria of 10CFR50.59.
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Plant Design Change Number 2-005-87 Plant Design Change Number 2-005-87 entitled "Steam Generator Tube Stabilizers" is complete.
Description of Change This change permitted the use of stabilize.s in the number one steam generator hot leg. Stabilizers were installed in five steam generator tubes prior to plugging the tubes. A stabilizer was a steel rod which fit closely within the steam generator tube. Its length varied from 35 to 55 inches with a 0.600 inch to 0.654 inch outside diameter. If the stabilizer vas loose, a 0.645 inch outside diameter by .25 inch high spacer was provided to preclude vear on the tube plug below the stabilizer.
Reason for Change Stabilizing is a preventative measure for steam generator tubes. It restricts tube displacement in the event the tube becomes severed. In this vay, the potential for additional tube failures due to wear, fatigue, or loose parts was minimized.
Safety Evaluation Since the stabilizer was intended to minimize the potential for tube failures, it was found to be consistent with the Final Safety Analysis Reviev (FSAR) and Technical Specifications. It has been shown that the bounding stabilizer design was acceptable in accordance with the ASME code. The stabilizer vould not affect the pressure boundary (the plug belov it) under normal operating, test, or accident condition?.
It limits tube fatigue failures in the severed tube. In addition, the stabilizer itself does not fail under maximum displacement load condi-tions. The stabilizer met or exceeded the applicable plant design bases.
Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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i Plant Design Change Number 2-007-87 Plant Design Change Number 2-007-87 entitled "Fuel Consolidation Demonstration Program" is complete.
Description of Change This demonstration was conducted under the provisions of 10CFR50.59-Special Tests and Experiments. The demonstration performed a coordi-nated disassembly of two spent fuel assemblies and subsequent recon-figuration and repackaging of the 352 fuel rods into a consolidated spent fuel storage box. The nonfuel bearing components (skeletons) were volume reduced and packaged into a vaste container. The scope of the demonstration was to consolidate a total of ten spent fuel assem-blies into five consolidated storage boxes for long-term storage in the spent fuel pool.
Reason for Change In the present spent fuel pool configuration, Millstone Unit 2 vill experience diminished storage capacity as early as 1994. Since current federal policy stipulates that nuclear utilities are respon-sible for interim storage of spent fuel until longer term disposition .
provisions are available, Millstone Unit 2 investigated this means to increase spent fuel storage capacity. Spent fuel consolidation has the potential to extend the current storage capacity of the existing spent fuel tacks to the Millstone Unit 2 license expiration date
- (2010).
i Safety Evaluation i Credible failure modes in the following areas were reviewed:
- 1. Failures that could result in a loss of suberitical margin.
- 2. Failures that could result in loss of cooling to the consolidated fuel. !
- 3. Failures that could result in heavy objects being dropped in the
- cask laydown area or in the pool.
] 4. Failures that result in dropped or damaged fuel assemblies.
The design basis accidant reviewed for potential impact due to the Spent Fuel Consolidation Program vas the Fuel Handling Incident (in i the spent fuel pool) - Final Safety Analysis Report (FSAR) Section -
14.19.
The offsite dose consequences due to the failure modes associated with !
dropping or damaging a fuel assembly were revieved and vere found to be bounded by existing analysis. ]
1 The overall probability of a Fuel Handling Incident was lov. Based on ;
the training, procedures, equipment design, and the fact that spent !
J fuel consolidation was not performed continuously, the probability of l a Fuel Handling Incident was effectively unchanged.
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P) ant Design Change Number 2-007-87 (Continued)
Safety Evaluation (Continued) .
Impact of the change and failure modes vere reviewed and it was found that the plant response has not been changed to the point where it can be considered a new accident.
The failures that affect cooling of the consolidated fuel had been evaluated for the potential for cteating a nev, unanalyzed accident and they were found acceptable.
The protective boundaries vere reviewed and the results indicated that '
the structural integrity of the pool was maintained with no signifi-cant pool vater inventory loss.
The Technical Specification requirements regarding K effective vere revieved and could not have been violated due to any failure mode related to the process.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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- i i Plant Design Change Number 2-008-87 j Plant Design Change Number 2-008-87 entitled "Installation of A Manual / Air Actuator For Appendix R on 2-SI-657" is complete, j Description of Change ;
This change installed a replacement pneumatic actuator / manual operator which allowed for manual local operation of the shutdown cooling heat l exchanger flow control _ valve (2-SI-657) in the event of a fire upon
- loss of instrument air and/or electrical power.
i Reason for Change
- i This actuator change was required to comply with the Appendix R requirements for cold reactor shutdown following a fire.
Safety Evaluation i
This change did not affect the automatic operation of the valve since .
the handvheel is disengaged when not in use. The automatic features '
i of the actuator did not change from the actuator it replaced. ,
1 l l This change did not have any effect'or have any potential impact upon !
- any of the design basis type accidents. 1 i !
i Failure modes associated with this change did not result in an ini-l tlating event that would result in a design basis accident.
, This change did not change or adversely impact the probability of a failure of any safety system.
i This change did not impact any of the protective boundaries as the i pressure boundary of the valve was not changed.
I Therefore, this modification does not constitute an unreviewed safety l
f question per 10CFR50.59. l E l 1
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Plant Design Change Number 2-011-87 Plant Design Change Number 2-011-87 entitled "Process Computer Replacement - Meteorological Data Input" is complete.
Description of Change This change provided seven new inputs from the meteorological tovet is the plant computer system.
Reason for Change This change supplied all the necessary meteorological data to the Hillstone Unit 2 computer in order to meet the requirements of the Hillstone Unit 2 Technical Specifications. The installation of these inputs permitted the removal of the meteorological data recorder from the Hillstone Unit 1 Control Room which had-been used by Hillstone Unit 2 in the past. ,
Safety Evaluation The seven parameters are nonsafety related and vere not utilized in the design basis accident analysis. Since this instrumentation continues to operate as it did in the Unit 1 Control Room, this change did not have any impact on the design basis accidents revieved.
These instruments are isolated from any safety-related computer inputs.
Therefore, this modification does not constitute an unreviewed safety question per 10CFR50.59.
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l Plant Design Change Number 2-013-87 ,
Plant Design Change Number 2-013-87 entitled "Main Exhaust Fans Suction Isolation Dampers" is complete.
Description of Change
! This change installed isolation dampers on the suction side of the main exhaust fans, F-34A, B, and C. ]
Reason for Change l
These)dampersaidthesafemaintenanceofthefanunitsbyprovidinga positive means to protect the fan (s) from backdraft conditions. ,
Safety Evaluation ;
Installation of these dampers was in a nonsafety-related system and .
t this system was not required to operate during accident conditions.
The fans are protected from damage by a lov flov trip mechanism which ,
vould activate if the damper (s) failed closed during routine opera-tion. This installation did not affect the normal on-line operation I
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of the fan assemblies.
l Therefore, this modification does not constitute an unreviewed safety l
- question per 10CFR50.59. ;
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Plant Design Change Number 2-016-87 Plant Design Change Number 2-016-87 entitled "Condensate Polishing Facility Improvements - Vaste Neutralization Sump Discharge System" is complete.
Description of Change The old pump and unpressurized filter system vere replaced with pressurized filters and higher head pumps. In addition, the old sandpiper pumps were removed. This required repiping the system and adding new instrumentation and a local alarm panel with associated viring and cabling. The discharge piping from radiation monitor 2CND-RE-245 was also re-routed.
l Reason for Change These modifications were necessary to increase the frequency with which the condensate polishers can be cleaned using the ultrasonic resin cleaner or regenerated with acid and caustic. Vhen the sumps are full, further bed regenerations or ultrasonic cleanings must vait until the sumps are emptied in order to be completed. These changes greatly increased the rate which the vaste neutralization sumps vere processed. The radiation monitor was re-routed to eliminate a high point in the sample flow path and increase the sample flow rate.
Safety Evaluation Inadvertent opening of a sump discharge isolation valve during sump recirculation vill have no off site dose consequences as there is a second discharge isolation valve downstream of the sump discharge valves. These valves are in series and preclude inadvertent discharges.
l System leakage or pipe break did not affect safety-related equipment since no safety-related equipment was located near the system that was modified. In addition, the new equipment was located in a diked, drained area or in an area drained by floor drains within a seismic enclosure. All vater leakage vill be collected.
1 No modifications made under this project impact the existing safety ;
analysis or any safety-related equipment. l l
Therefore, this modification does not constitute an unreviewed safety question per 10CFR50.59.
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t j I Plant Design Change Number 2-017-87
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Plant Design Change Number 2-017-87 entitled "Microvave Antenna i 4
Attachment to Millstone Ucit 2 Condensate Polishing Facility (CPF)
Building" is complete. ,
Description of Change This change installed an eight foot diameter microvave antenna to the '
Condensate Polishing Facility (CPF) Building.
Reason for Change This change provided the Northeast Utilities microvave system with an interconnection between the Millstone Nuclear Power Plant Site and the Montville Station.
Safety Evaluation The installation of this antenna and the antenna itself vere designed to vithstand a 125 mile per hour vind with one inch of radial ice.
The CPF Building is not a safety-related structure.
- If this antenna vere to break free during a hurricane or similar !
event, it could become a missile. The surrounding structures that are i important to safety are missile protected for missiles with higher ;
energy than the antenna therefore this is not a concern. Also, loss ;
of the transformers due to this missile vould not compromise safety i since other backup systems are available to take over. This instal-1 lation has no effect on any safety-related system. i
} Therefore, this modification does not constitute an unreviewed safety l question per 10CFR50.59. ,
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PLANT DESIGN CHANGE EVALUATION INDEX PDCE Number Title ,
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t MP2-87-051 Turbine and Auxiliary Building Supply Fans !
Inlet Screen Installation.
MP2-87-073 Installation of a New Containment Temporary i Equipment Hatch MP2-87-074 Control Room Air Conditioning (CRAC) System d
, Test Connections _
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Plant Design Change Evaluation R'37-051 Plant Design Change Evaluation HP2-87-05: entitlerl "Turbine and Auxiliary Building Supply Fans Inlet Screen Installation" is complete.
i-Description of Change Debris screens vere installed on the inlets to supply fans for the Turbine (F-101A through G) +nd &>xiliary Building fans (F-52 and F-17). F-17 supplies air to ibt cable vault and east 480V load center room. The fan is povered from a nonvital bus and has a redundant vital supplied backup. F-52 is the vital povered supply far. for the east 480V svitchgear room.
Reason for Change The screens mitigated transportation of foreign material to the Turbine and Auxiliary Buildings.
Safety Evaluation The Turbine Building fans serve no safety-related function. The screen installed on F-52 did not affect the ability of the fan to start ar.d run on temperature demand during any analyzed events. The fans flow capacity .as verified during post installation retest. The installation did not create any accidents not previously analyzed in that any debris which was trapped by the screens vould have been deposited on the supply registers and/or indigenous components.
Therefore, this modification does not constitute an unrevieved safety question per 10CFR50.59.
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Plant Design Change Evaluation HP2-87-073 i Plant Design Change Evaluation HP2-87-073 entitled "Installation of a New Containment Temporary Equipment Hatch" is complete.
Description of Change A new temporary containment equipment br,tch was redesigned and constructed for use during refuel outager Reason for Change The temporary hatch allowed outage cabling and process fluids to enter and exit containment uninterrupted during periods of required con-tainment integrity.
Safety Evaluation The new hatch was designed to equal or exceed the requirement of the original hatch. The same design assumptions were used with respect to pressure loading, cable / piping loading, dead veight. Accident bases loads were also used for the new design. All penetrations vould be sealed with foam if not supplied with a flanged cover. An additional feature was added which allowed the door to open full height without removing any portion of the hatch. In addition, the new design !
provided easier emergency access utilizing two hinged and latched doors. Additional penetration liner velded bolt connections were required and vere analyzed to be satisfactory.
Therefore, this modification does not constitute an unreviewed safety question per 10CFR50.59.
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Plant Design Change Evaluation MP2-87-074 Plant Design Change Evaluation MP2-87-074 entitled "Control Room Air Conditioning (CRAC) System Test Connectitos" is complete.
Description of Change Two flanged sheet metal test connections were installed in the Control Room Air Conditioning (CRAC) System. One connection was installed on the supply and one on the exhaust ductvork.
Reason for Change These connections vete required for conducting Control Roon leak testing. .
Safety Evaluation 1
The material, joint design, and sealing vere consistent with the existing ventilation specifications and construction standards. There was no impact on the seismic response of the system due to the insig-nificant veight of the connections. The change did not significantly increase ventilation system leakage.
Therefore, this modification does not constitute an unreviewed safety question per 10CFR50.59.
PROCEDURE CHANGE INDEX Procedure Number Title E0P 2525 Standard Post Trip Actions Rev. 2 E0P 2534 Steam Generator Tube Rupture Rev. 4 E0P 2536 Excess Steam Demand Rev. 3 E0P 2540 Functional Recovery Rev. 4 1
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o NUMBER TITLE E0P 2525 Standard Post Trip Actions Rev. 2 Description of Change This procedure was changed to reflect the Reactor Coolant Pump (RCP) trip 2/ leave 2 scheme. .This scheme involves the operater tripping two RCPs (one in each loop), in lieu of all four, when the Rea_ tor Caolant System (RCS) decreases to 1600 psia. The remaining two RCPs vould be tripped either when subcooling, based on the hot leg temperature, decreases to 30 degrees Fahrenheit or when the operator has diagnosed the accident to be a Loss of Cooling Accident (LOCA).
Reason for Change The reactor plant cauld be controlled easier if two of the RCPs were operating. Also, during a Steam Generator Tube Rupture Accident, with two RCPs tripped, a reduced heat transfer in the affected steam gen-erator would be realized resulting in less off-site dose than the case with four RCPs running.
Safety Evaluation The Loss of Coolant Accident (Small Breaks), Steam Generator Tube Rupture (SGTR) Accident, Steam Line Break (SLB) Accident, and Loss of-Coolant Accident (Large Breaks) were identified for potential impact by this procedure change.
SGTR - the consequences of a SGTR in the old licensing analysis were calculated assuming all four RCPs were running. Vith only two RCPs running, a reduced heat transfer in the affected steam generator vould be realized and therefore less off-site dose. _Therefore, the conse-quences of a SGTR'as previously analyzed did not increase due to this chaNe.
SLB - the old analysis assumed all four RCPs vere running. Vith only two RCPs running, the remaining two would be tripped only if at least one hot leg subcooling decreased belov 30 degrees Fahrenheit. The consequences of a SLB vith two pumps tripped at 1600 psia vould remain bounded by the old licensing analysis.
LOCA (Large Break) - in this case the RCS pressure and hot leg sub-cooling would drop belov 1600 psia and 30 degrees Fahrenheit almost immediately and the operator vould trip all four RCPs simultaneously.
It was concluded that the probability of the operator failing to trip all four RCPs did not increase due to this procedure change. Hence, the consequences were not affected by the change.
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i NUMBER TITLE E0P 2525 Standard Post Trip Actions (Continued)
Rev. 2 Safety Evaluation (Conti.:Jed)
LOCA (Small Break) - the reduction in time it took before all four RCPs vere tripped was reviewed. The reduction in operator action time from 338 seconds to 300 seconds was considered ~small and therefore, was deemed acceptable. These times were based on the limiting small break, which would be readily diagnosed because of its relatively large size (4 inch diameter break). For smaller breaks, the time available to take action vould be much longer therefore, more time vould be available for the operators to diagnose this condition.
Therefore, it v:- concluded that this change did not increase the consequences of c Jmall break LOCA.
The LOCA analysis assumed that the RCPs would be tripped when_the hot leg looses subcooling. The instrument uncertainties vere considered to be less than 30 degrees Fahrenheit. A trip setpoint of 30 degrees Fahrenheit in the procedure vould ensure that the RCP trip criteria was met before the hot leg saturates.
The probability of .'n operator tripping two pumps in the same loop vas reviewed as a potenti.'l for creating a new unanalyzed event. It was concluded that the procability of this event was very lov and not considered credible for a design basis event.
Therefore, this procedure change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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NUMBER TITLE E0P 2534 Steam Generator Tube Rupture Rev. 4 Description of Change This procedure was changed to reflect the Reactor Coolant Pump (RCP) trip 2/ leave 2 scheme. This acheme involves the operator tripping two RCPs (one '.n each -loop), in lieu of all four, when the Reactor Ceci.nt System (RCS). decreases to 1600 psia. The remaining two RCPs vould be tripped either when subcooling, based on the hot leg temperature, decreases to 30 degrees Fahrenheit or when the operator has diagnosed the accident to be a Loss of Cooling Accident (LOCA).
Reason for Change The reactor plant could be controlled easier if two of the RCPs were operating. . Also, during a Steam Generator Tube Rupture Accident, with
'two RCPs tripped, a reduced heat transfer in the affected steam gen-erator vould be realized resulting in less off-site dose than the case with four RCPs running.
Safety Evaluation The Loss of Coolant Accident (Small Breaks), Steam Generator Tube Rupture (SGTR) Accident, Steam Line Break (SLB) Accident, and Loss of Coolant Accident (Large Braaks) were identified for potential impact by this procedure change.
SGTR - the consequences of a SGTR in the old licensing analysis were calculated assuming all four RCPs were running. Vith only two RCPs running, a reduced heat transfer in the affected steam generator would be realized and therefore less off-site dose. Therefore, the conse-quences of a SGTR as previously analyzed did not increase due to this change.
SLB - the old analysis assumed all four RCPs vere running. Vith only two RCPs running, the remaining two would be tripped only if at least one hot leg subcooling decreased belov 30 degrees Fahrenheit. The consequences of a SLB vith two pumps tripped at 1600 psia vould remain bounded by the old licensing analysis.
LOCA (Large Break) - in this case the RCS pressure and hot leg subcooling would drop below 1600 psia and 30 degrees Fahrenheit almost immediately and the operator vould trip all four RCPs simultaneously.
It was concluded that the probability of the operator failing to trip all four RCPs did not increase due to this procedure change. Hence, the consequences were not affected by the change.
NUMBER TITLE E0P 2534 Steam Generator Tube Rupture (Continued)
Rev. 4 Safety Evaluation (Continued)
LOCA (Small Break) - the reduction in time it took before all four RCPs vere tripped was reviewed. The reduction in operator action time from 338 seconds to 300 seconds vas considered small and therefore, was deemed acceptable. These times were based on the limiting small break, which would be readily diagnosed because of its relatively large size (4 inch diameter break). For smaller breaks, the time available to take action vould be much longer therefore, more time vould be available for the operators to diagnose this condition.
Therefore, it was concluded that this change did not increase the consequences.of a small break LOCA.
The LOCA analysis assumed that the RCPs vould be tripped when the hot leg looses subcooling. The instrument uncertainties vere considered to be less than 30 degrees Fahrenheit. A trip setpoint of 30 degrees Fahrenheit in the procedure vould ensure that the RCP trip criteria was met before the hot leg saturates.
The probability of an operator tripping two pumps-in the same loop vas reviewed as a potential for creating a new unanalyzed event. It was concluded that the probability of this event was very lov and not considered credible for a design basis event.
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Therefore, this procedure change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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NUMBER TITLE E0P 2536 Excess Steam Demand Rev. 3 Description of Change This procedure was changed to reflect the Reactor Coolant Pump (RCP) trip 2/ leave 2 scheme. This scheme involves the operator tripping two RCPs (one in each loop), in lieu of all four, when the Reactor Coolant System (RCS) decreasts to 1600 psia. The remaining two RCPs vould be tripped either when subcooling, based on the hot leg temperature, decreases to 30 degrees Fahrenheit or when the operator has diagnosed the accident to be a Loss of Cooling Accident (LOCA).
Reason for Change The reactor plant could be controlled easier if two of the RCPs were operating. Also, during a Steam Generator Tube Rupture Accident, with
'tvo RCPs tripped, a reduced heat tiensfer in the affected steam gen-erator vould be realized resulting in less off-site dose than the case with four RCPs running.
Safety Evaluation The Loss of Coelant Accident (Small Breaks), Steam Generator Tube Rupture (SGTR) Accident, Steam Line Break (SLB) Accident, nd Loss of C0olant Accident (Large Breaks) vere identified for potential impact by this procedure chaage.
SGTR - the consequentes of a SGTR in the old licensing analysis vere calculated assuming all four RCPs vere running. With only two RCPs running, a reduced heat transfer in the affected steam generator vould be realized and therefore less off-site dose. Therefore, the conse-quences of a SGTR as previously analyzed did not increase due to this change.
SLB - the old analysis assumed all four RCPs were running. Vith only two RCPs running, the remaining two would be tripped only if at least one hot leg subcooling decreased belov 30 degrees Fahrenheit. The consequences of a SLB vith two pumps tripped at 1600 psia vould remain bounded by the current licensing analysis.
LOCA (Large Break) - in this case the RCS pressure and hot leg subcooling vould drop belov 1600 psia and 30 degrees Fahrenheit almost immediately and the operator would trip all four RCPs simultaneously.
It was concluded that the probability of the operator failing to trip all four RCPs did not increase due to this procedure change. Hence, !
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E0P 2536 Excess Steam Demand (Continued) j Rev. 3 '
Safety Evaluation (Continued) l LOCA (Small Break) - the reduction in time it took before all four .)
RCPs vere tripped was reviewed. The reduction in operator action time from 338 seconds to 300 seconds vas considered small and therefore, was deemed acceptable.- These times vere based on the limiting small~
break, which vould be readily diagnosed because of its relatively large size (4 inch diameter break). For smaller breaks, the time available to take action would be much longer therefore, more time vovld be available for the operators to diagnose this condition.
Therefore, it was concluded that this change did not increase the '
consequences of a small break LOCA.
The LOCA analysis assumed that the RCPs vould be tripped when the hot leg looses subcooling. The instrument uncertainties were considered to be less than 30 degrees Fahrenheit. A trip setpoint of 30 degrees Fahrenheit in the procedure vould ensure that the RCP trip criteria was met before the hot leg saturates.
The probability of an operator tripping two pumps in the same loop vas reviewed as a potential for creating a new unanalyzed event. It was concluded that the probability of this event was very lov and not considered credible for a design basis event.
Therefore, this procedure change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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NUMBER TITLE E0P 2540 Functional Recovery Rev. 4 Description of Change This procedure was changed to reflect the Reactor Coolant Pump (RCP) trip 2/ leave 2 scheme. This scheme involves the operator tripping two RCPs (one in each loop), in lieu of all four, when the Reactor Coolant System (RCS) decreases to 1600 psia. The remaining two RCPs vould be tripped either when subecoling, based on the hot leg temperature, decreases to 30 degrees Fahrenheit or when the operator has diagnosed the accident to be a Loss of Cooling Accident (LOCA).
Reason for Change The reactor plant could be controlled easier if two of the RCPs were operating. Also, during a Steam Generator Tube Rupture Accident, with two RCPs-tripped, a reduced heat transfer in the affected steam gen-erator vould be realized resulting in less off-site dose than the case with four RCPs running.
Safety Evaluation The Loss of Coolant Accident (Small Breaks), Steam Generator Tube Rupture (SGTR) Accident, Steam Line Break (SLB) Accident, and Loss of Coolant Accident (Large Breaks) vere identified for potential impact by this procedure change.
SGTR - the consequences of a SGTR in the old licensing analysis vere calculated assuming all four RCPs were running. Vith only two RCPs running, a reduced heat transfer in the affected steam generator would be realized and therefore less off-site dose. Therefore, the conse-quences of a SGTR as previously analyzed did not increase due to this change.
SLB - the old analysis assumed all four RCPs vere running. Vith only two RCPs running, the remaining two would be tripped only if at least one hot leg subcooling decreased belov 30 degrees Fahrenheit. The consequences of a SLB vith two pumps tripped at 1600 psia "ould remain bounded by the current licensing analysis.
LOCA (Large Break) - in this case the RCS pressure and hot leg subcooling would drop below 1600 psia and 30 degrees Fahrenheit almcst immediately and the operator vould trip all four RCPs simultaneously.
It was concluded that the probability of the operator failing to trip all four RCPs did not increase due to this procedure change. Hence, the consequences were not affected by the change.
NUMBER TITLE E0P 2540 Functional Recovery (Continued)
Rev. 4 Safety Evaluation (Continued)
LOCA (Small Break) - the reduction in time it took before all four RCPs vere tripped was reviewed. The reduction in operator action time from 338 seconds to 300 seconds was considered small and therefore, vas deemed acceptable. These times were based on the limiting small break, which would be readily diagnosed because of its relatively large size (4 inch diameter break). For smaller breaks, the time available to take action vould be much longer therefore, more time vould be available'for the operators to diagnose this condition.
Therefore, it was concluded that this change did not increase the consequences of a small break LOCA.
The LOCA analysis assumed that-the RCPs vould be tripped when the hot leg loses subcooling. The instrument uncertainties vere considered to be less than 30 degrees Fahrenheit. A trip setpoint of 30 degrees Fahrenheit in the procedure vould ensure that the RCP trip criteria was met before the hot leg saturates. ;
The probability of an operator tripping two pumps in the same loop was reviewed as a potential for creating a new unanalyzed event. It was concluded that the probability of this event was very lov and not '
considered credible for a design basis event.
Therefore, this procedure change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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JUMPER-LIFTED LEAD-BYPASS (J-LL-B) CHANGES INDEX ;
J-LL-B Number J-LL-B Title 2-87-005 Vide Range Channel 'B' Ex-Core Detector Cable Removal 2-87-009 Spent Fuel Pool (SFP) Upender Limit Switches 2-87-011 Spent Fuel Pool (SFP) Platform Crane North /
South Hoist Trolley Limit Switch 2-87-012 Spent Fuel Pool (SFP) Platform Crane Staging 2-87-013 Spent Fuel Pool (SFP) Upender and Platform Crane Hovement 2-87-020 Operator Shaft Clamp for the_'B' Service Vater (SV) Header Cross-Tie Valve ,
2-87-028 Spent Fuel Fool (SFP) Upender and Platform Crane Hovement 2-87-029 Boric Acid Heat Trace Alarms 2-87-030 Spent Fuel Pool (SFP) Level Switch Rack Removal 2-87-035 Caustic Bulk Storage. Tank Pump Lov. Level Shutoff 2-87-047 Actuator Change-Out For Lov Pressure Safety
, Injection (LPSI) Boundary Valve
'A' Diesel _ Generator Starting Air System 2-87-048 !
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Jumper-Lifted Lead-Bypass Change Number 2-87-005 Jumper. Lifted Lead Bypass Change Number 2-87-005 entitled "Vide Range Channel 'B' Ex-Core Detector Cable Removal" is installed.
Description of Change The signal cable for the ex-core vide range nuclear instrument flux monitor for channel 'B' that inputs to the preamplifier was dis-connected. This was-a temporary change to the system pending a per-manent repair. This jumper is still installed.
Reason for Change This cable was disconnected to eliminate an overcurrent condition in the preamplifier. The cause of the overcurrent was not known but was isolated to between the detector and the preamplifier unit.
Safety Evaluation At normal power levels, two detectors are not needed so removal of this signal did not affect indication. The second detector signal
- increases the sensitivity in the very low power level regions during shutdown operation. However, since only channel 'B' is affected, the minimum number of channels was still satisfied by other inservice units.
Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Jumper-Lifted Lead-Bypass Change Number 2-87-009 Jumper Lifted Lead Bypass Change Number 2-87-009 entitled "Spent Fuel Fool (SFP) Upender Limit Svitches" is complete.
Description of Change The Spent Fuel Pool (SFP) upender vertical limit switch (2LS-BV) and the transfer carriage upender home limit switch (2LS-BVH) were electrically bypassed. This jumper has been removed.
Reason for Change The upender was required to be horizontal with the platform crane over the transfer canal area for the fuel reconstitution vork that was performed in the transfer canal. The existing limit switch configu-ration did not permit the upender to be horizontal when the platform crane entered the transfer canal area.
Safety Evaluation This jumper closed the contacts to make the upender appear to be vertical to the platform crane. This interlock feature is described in the Hillstone Unit 2 Final Safety Analysis Report (FSAR). This interlock is designed to ensure that a fuel assembly cannot be inserted into the upender unless the upender is completely vertical.
Since there was no intent to place any fuel assemblies into the upender during fuel reconstitution, the upender could be maneuvered in this manner. Normal fuel handling procedures were utilized and any potential fuel handling accidents in the spent fuel pool / transfer canal vere bounded by the fuel handling accident analysis of the FSAR.
Therefore, this change did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Jumper-Lifted Lead-Bypass Change Number 2-87-011 Jumper Lifted Lead Bypass Change Number 2-87-011 entitled "Spent Fuel Pool (SFP) Platform Crane North / South Holst Trolley Limit Switch" is installed.
Description of Change The' function of the north / south trolley limit switch on the Spent Fuel Pool (SFP) platform crane was disabled. This jumper is still installed.
Reason for Change This limit switch was removed from service to allow the fuel handling tool to access the northernmost rows of the spent fuel storage locations in order to access spent fuel for the fuel reconstitution project.
Safety Evaluation This limit svitch restricted the SFP platform crane from accessing the last two rows of storage locations on the north and south sides of the SFP. This limit switch is described in the Hillstone Unit 2 Final Safety Analysis Report (FSAR), "to prevent the crane from running a fuel assembly into the SFP valls." Normal fuel handling procedures, spent fuel pool criticality requirement procedures, and special nuclear material controls procedures vere utilized, and potential fuel handling accidents in the SFP are bounded by the fuel handling accident analysis in the FSAR.
Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Jumper-Lifted Lead-Bypass Change Number 2-87-012 Jumper Lifted Lead Bypass Change Number 2-87-012 entitled "Spent Fuel Pool (SFP) Platform Crane Staging" is complete.
Description of Change A scaffold, fifteen feet high by six feet long by four feet vide was constructed on the floor of the platform crane. This jumper has been removed.
Reason for Change Due to the configuration of the equipment used for fuel reconstitu-tion, the working elevation of the manually operated grapple mechanism was positioned approximately fifteen feet above the floor of the platform crane. In order to assure the capability to grapple and ungrapple the fuel assembly, it was necessary to construct this st y,ing.
Safety Evaluation The scaffolding was of tubular steel design and resulted in a conser-vative weight of less than 1500 pounds. This load is less than the heavy loads criteria of 1800 pounds. Therefore, the bounding heavy loads analysis remained unchanged. The scaffolding structure was rigidly attached to the platform structure to prevent the scaffolding from falling in the direction of the fuel handling tool. The extra veight compared to the weight of the bridge changed the center of gravity. Since the center of gravity shift was insignificant, the structural evaluation of the platform crane remained unchanged.
Furthermore, the location of the scaffold...g on the platform crane floor did not interfere with the operation of the crane. Finally, while the extra veight did affect the response of the crane during a seismic event, the impact was considered insignificant.
Therefore, this change did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Jumper-Lifted Lead-Bypass Change Number 2-87-013 Jumper Lifted Lead Bypass Change Number 2-87-013 entitled "Spent Fuel Pool (SFP) Upender and Platform Crane Hovement" is complete.
Description of Change The Spent Fuel Pool (SFP) upender relay (2CR-BV) contact was elec-trically bypassed in order to provide indication to the platform crane that the upender was in the vertical position. This jumper has been removed.
Reason for Change The upender was required to be horizontal with the platform crane over the transfer canal area for the fuel reconstitution vork that was performed in the transfer canal. The existing interlock configuration did not permit the upender to be horizontal when the platform crane entered the transfer canal area.
Safety Evaluation This jumper simulated closing the contact of the relay to make the upender appear to be vertical to the platform crane. This interlock feature is described in the Hillstone Unit 2 Final Safety Analysis Report (FSAR). This interlock is designed to ensure that a fuel assembly cannot be inserted into the upender unless the upender is completely vertical. Since there was no intent to place any fuel l assemblies into the upender during fuel reconstitution the upender l could be maneuvered in this manner. Normal fuel handling procedures j vere utilized and potential fuel handling accidents in the spent fuel j l
pool / transfer canal vere bounded by the fuel handling accident.
analysis of the FSAR.
Therefore, this change did not constitute an unreviewed safety question par the criteria of 10CFR50.59.
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1 Jumper-Lifted Lead-Bypass Change Number 2-87-020 Jumper Lifted Lead Bypass Change Number- 2-87-020 entitled "Operator Shaft Clamp'for the 'B' Service Water (SV) Header Cross-Tie Valve" is complete.
Description of Change i A shaft clamp vas installed on the operator shaft.of the 'B' Service Water (SV) header cross-tie valve (2-SV-97B). This jumper has been removed.
Reason for Change The shaft clamp held the valve in a set position which allowed the operator shaft keyvays to be repaired.
Safety Evaluation
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Clamping this' valve'did not deviate from the way the system operates as described in the Final Safety Analysis Report (FSAR), but did affect the redundancy built into the system, in that the 'B' pump vas
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not available for operation on both headers. The 'B' pump was only available for operation, as a sving pump, on one header. Technical. . ,
Specification 3/4.7.4 allows one header to be inoperable for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. In the event that the header without the sving pump experi-ences a failure of its pumps, this clamped valve could be_ repositioned to restore the second header to operation within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (probably within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />).
Therefore, this change did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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I Jumper-Lifted Lead-Bypass Change Number 2-87-028 Jumper Lifted Lead Bypass Change Number 2-87-028 entitled "Spent Fuel Pool (SFP) Upender and Platform Crane Movement" is complete.
Description of Change The Spent Fuel Pool (SFP) upender relay (2CR-BV) contact was elec-trically bypassed in order to provide indication to the platform crane that the upender was in the vertical position. This jumper has'been removed.
Reason for Change The upender was required to be horizontal with the platform crane over the transfer canal area for the fuel reconstitution vork that was performed in the transfer canal. The existing interlock configuration did not permit the upender to be horizontal when the platform crane entered the transfer canal area.
Safety Evaluation This jumper simulated closing the contact of the relay to make the upender appear to be vertical to the platform crane. This-interlock feature is described in the Millstone Unit 2 Final Safety Analysis Report (FSAR). This interlock is designed to ensure that a fuel assembly cannot be inserted into the upender unless the upender is completely vertical. Since there was no intent to place any fuel assemblies into the upender during fuel reconstitution the upender could be maneuvered in this manner. Normal fuel handling procedures vere utilized and potential fuel handling accidents in the spent fuel pool / transfer canal vere bounded by the fuel handling accident analysis of the FSAR. ;
I Therefore, this change did not constitute an unreviewed safety l question per the criteria of 10CFR50.59. l Jumper-Lifted Lead-Bypass Change Number 2-87-029 Jumper Lifted Lead Bypass Change Number 2-87-029 entitled "Boric Acid Heat Trace Alarms" is complete.
Description of Change The alarm card for the boric acid heat tracing high/lov. temperature alarms at control console (RC-22) was removed. This jumper has been removed.
Reason for Change As the temperature recorder reads each circuit, there vere momentary alarms which energize and then clear. When this happened the Control Room boric acid heat tracing high/lov temperature alarms did not lock in causing numerous nuisance alarms.
Safety Evaluation Removal of the alarm card at RC-22 instead of putting the heat tracing recorders in alarm-defeat, allowed any valid temperature alarm to still indicate at the local panel. The local panel was checked under normal Plant Equipment Operator rounds on a shiftly basis. Any legitimate alarms vere identified during these rounds. The Technical Specification requirement to perform temperature checks once every seven days was satisfied.
Therefore, this change did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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1 Jumper-Lifted Lead-Bypass Change Number 2-87-030 Jumper Lifted Lead. Bypass Change Number 2-87-030 entitled "Spent Fuel Fool (SFP) Level-Switch Rack Removal" is complete.
Description of Change The leads ~ vere lif ted for the Spent Fuel' Pool (SFP) Low Level Alarm-(LS7424), the SFP High Level Alarm (LS7031), and-the SFP Primary Make-Up Vater (PNV) Automatic Make-Up Level Switch which caused the make-up-vater valve to fail shut. This jumper has been removed.
Reason for Change These instruments vere disconnected.in order to temporarily remove the level switch station, so that fuel assemblies located under the switch station can be moved.
Safety Evaluation Vith the leads lifted, the SFP High Level Alarm, the SFP Lov Level Alarm, and the Auto PMV make-up control to valve 2-RV-137 on SFP High Level were not available. Even with the PMV make-up not available, there were three other sources of make-up water available that could be manually initiated; these vere auxiliary feedvater, fire water, and the refueling water storage tank. Personnel continuously monitored the SFP vater level during the time that the level switches were removed and they reported any changes in level to the Shift Supervisor.
Therefore, this change did not' constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Jumper-Lif ted Lead-Bypass Change Number 2-87-035 Jumper Lifted Lead Bypass Change Number 2-87-035 entitled "Caustic Bulk Storage Tank Pump Lov Level Shutoff" is complete.
Description of Change The low level shutoff _svitches (71-2CND-LS318A and B) for the caustic bulk storage tank pumps-(P12A and P12B) were bypassed to permit the pumps to empty the caustic bulk storage tank (TK15). This jumper has been removed.
Reason for Change The caustic in the tank vas no good.
Safety Evaluation '
This jumper prevented the lov level switch from automatically stopping the pumps that were used to pump the bulk caustic tan' . The lov level -
alarm was functional as designed and the tank level indication remained operable to allow monitoring of the tank level during pump-dovn. The pump protection which vas lost with the jumper, vas pro-vided by the persons performing the tank pumpdown. The manual controls of the pump were not affected. The pump, controls, and the 1 tank did not support any safety systems.
Therefore, this change did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Jumper-Lifted Lead-Bypass Change Number 2-87-047 Jumper Lifted' Lead Bypass Change Number 2-87-047 entitled "Actuator Change-Out For Lov Pressure Safety Injection (LPSI) Boundary Valve" is complete.
Description of Change A mechanical block vas installed on the boundary valve for the Lov Pressure Safety Injection (LPSI) system (2-SI-657) which held the valve closed while the operator was changed. This jumper has been removed.
Reason for Change This mechanical block vas installed to prevent inadvertent cycling of the valve to the open position during the actuator replacement. The actuator replacement was required to meet Appendix R concerns.
Safety Evaluation The mechanical blocking device was designed to vithstand forces equal to or greater than the valve actuator stem that normally vould maintain the valve in the closed position and therefore has a lov probability for failure. However, the possible failure of this mechanical block to perform its required function meant that the valve could inadvertently open, thus no longer providing a boundary for the ,
LPSI system. The opening of this valve, or the effects of this valve remaining open did not affect flow to the Reactor Coolant System as l other manual locked closed valves vould provide back-up boundary isolation. j Therefore, this change did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Jumper-Lifted Lead-Bypass Change Number 2-87-048 Jumper Lifted Lead Bypass Change Number 2-87-048 entitled "'A' Diesel Generator Starting Air System" is complete.
Description of Change A temporary flex hose vas installed between the outlet of the existing air compressors and the relocated receiver tank inlet check valves.
This jumper has been removed. ;
Reason for Change This hose was installed to permit the replacement of the existing '
carbon steel piping and dryer between the compressors and the receiver tanks during the operating cycle.
Safety Evaluation The effect of the temporary loss of the air dryer system and the seismic configuration changes vere reviewed and vere found to be acceptable. The temporary loss of the air dryers was offset by the plant operators manually blowing down the receiver tanks once a shift to eliminate any water build-up. The relocation of the inlet check valve was seismically reviewed with a flex hose attached to it and was found to be acceptable. The system integrity was maintained. It was decided that the change did not have an adverse effect on the opera-tion of this safety system.
Therefore, this change did not constitute an unreviewed safety question per the criteria of 10CFR50.59. ,
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'I SETPOINT CHANGE INDEX Number Title 2-87-002 Steam Jet Air Ejector Radiation Monitor Sample Flov Rate 2-87-004 'C' Reactor Coolant Pump (RCP) Upper Seal Lov Pressure Alarm 2-87-005 'C' Reactor Coolant Pump (RCP) Middle Seal Lov Pressure Alarm 2-87-006 'D' Reactor Coolant Pump (RCP) Middle Seal Lov Pressure Alarm 2-87-007 Fire Pump Discharge Header Pressure 2-87-012 Steam Jet Air Ejector Radiation Monitor Sample Flow Rate 2-87-013 Lov Level Alarm Setpoint for the Diesel Fuel Oil Tank t 2-87-014 Auxiliary Building Ventilation Air Supply Fan ,
Flov Alarm 2-87-017 Overcurrent Trip for a Station Electrical Service Breaker l
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i Setpoint Change Numoer 2-87-002 Setpoint Change Number 2-87-002 entitled "Steam Jet Air Ejector Radiation Monitor Sample Flow Rate" is complete.
Description of Change The setpoint of the steam jet air ejector radiation monitor sample flow rate (FIS-5099) was temporarily changed from 3.5 g 1 CFM to 6.0 g
.25 cfM for the high alarm and from 2.5 g 1 CFM to 5.0 g .25 CFM for the lov alarm.
Reason for Change These setpoints vere increased in order to increase the sample flow rate so that it vould overcome the suction of the fan that moves the discharge of the steam jet air ejector to the Hillstone Unit I stack.
The old sample flow rate was not high enough to take a sample.
Safety Evaluation The vendor did not specify a flow rate for the radiation monitor. The sensitivity and accuracy of a gaseous radiation monitor was not affected by increasing and/or decreasing the sample flov. It responds to activity present and not on accumulation of activity as with ,
particulate filter type monitors. Therefore, adjusting the setpoints and their tolerances can be changed without affecting any safety systems or their operation.
Therefore, this change did not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 2-87-004 Setpoint Change Number 2-87-004 entitled "'C' Reactor Coolant Pump .!
(RCP) Upper Seal Low Pressure Alarm" is complete. !
' Description of Change l
The setpoint of the 'C' Reactor Coolant Pump (RCP) upper seal lov.
l pressure alarm (PAL-162) was temporarily-changed from 545 psig to 425 l psig. This change is still in effect. '
Reason for Change This change provided the operators v' S Indication that the pressure was dropping for the 'C' RCP upper 1.- . Prior to the change the annunciator was in and any further 6 sp in pressure vould not be alarmed.
Safety Evaluation l Changing this setpoint did not chunge ihn operation of any plant equipment since the alarm was used only for monitoring purposes.
Lovering the setpoint provided better control as a seal deterioration J vould be noticed quicker. '
Therefore, this change does not constitute an unreviewed safety l question per the criteria of 10CFR50.59.
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Setpoint Change Number 2-87-005 Setpoint Change Number 2-87-005 entitled "'C' Reactor Coolant Pump (RCP) Hiddle Seal Lov Pressure Alarm" is complete.
Description of Change The setpoint of the 'C' Reactor Coolant Pump (RCP) middle seal lov pressure alarm (PAL-161) vas temporarily changed trom 1290 psig to 800 psig. This change is still in effect.
Reason for Change This change provided the operators with indication that the pressure was dropping for the 'C' RCP middle seal. Prior to the change the annunciator was in and any further drop in pressure vould not be alarmed.
Safety Evaluation Changing this setpoint did not change the operation of any plant equipment since the ale.rm was used only for monitoring purposes.
Lovering the setpoint provided better control as a seal deterioration vould be noticed quicker.
Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setroint Change Number 2-87-006 Setraint Change Number 2-87-006 entitled "'D' Reactor Coolant Pump (RCP) Middle Seal Lov Pressure Alarm" is complete.
Description of Change The setpoint of the 'D' Reactor Coolant Pump (RCP) middle seal lov pressure alarm (PAL-181) was temporarily changed from 1290 psig to 1100 psig. This change is still in effect.
Reason for Change This change provided the operators with indication that the pressure was dropping for the 'D' RCP middle seal. Prior to the change the annunciator vas in and any further drop in pressure vould not be alarmed.
Safety Evaluation Changing this setpoint did not change the operation of any plant equipment since the alarm vas uscf only for monitoring purposes.
Lovering the setpoint provided better control as a seal deterioration vould be noticed quicker.
Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
Setpoint Change Number 2-87-007 Setpoint Change Number 2-87-007 entitled ire Pump Discharge Header Pressure" is complete.
Description of Change The setpoint of the. fire pump discharge header pressure switch (PS-7301) was changed from 95 psi, decreasing to 98 psi, decreasing.
Reason for Change The minimum Technical Specification limit for maintaining fire headcr pressure is 75 psi. The 95 psi setpoint turned on the fire pumps and picked up the pressure before it reached the Technical Specification ,
limit. Increasing the limit to 98 psi provided extra margin for i conducting testing. i Safety Evaluation An increase in the fire pump pressure switch setpoint did not affect the design bases since the fire systems ability to confine'or extin-guish a fire near any portion of safety-related equipment was not compromised, the system operability as defined by the Technical Specifications remains intact, and system performance assumptions concerning the sequential auto start capability of the site fire pumps was not altered.
1 Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 2-87-012 Setpoint Change Number 2-87-012 entitled "Steam Jet Air Ejector Radiation Monitor Sample Flov Rate" is complete.
Description of Change The setpoint of the Steam Jet A!. Ejector (SJAE) radiation monitor sample flow rate (FIS-5099) vas per';anently changed f rom 6.0 CFM .i .25 to 3.5 CFM i 1 for the high alarn .tnd from 5.0 CFM i .25 to 2.5 CFM i 1 for the lov alarm.
Reason for Change
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This setpoint change request returned the setpoints close to their original values and brought the alarm limits back to reasonable values. This was done based on the flow ratcr of the sample fan after it was rebuilt and the moisture separator was replaced.
Safety Evaluation
- The vendor did not specify a flow rate for the radiation monitor. The sensitivity and accuracy of a gaseous radiation monitor was not :
affected by increasing and decreasing the sample'flov. It responds to activity present and not on accumulation of activity as with particu- -
late filter type monitors. Therefore, adjusting the setpoints and their tolerances were changed without affecting any safety systems or their operation.
Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 2-87-013 Setpoint Change Number 2-87-013 entitled "Lov Level Alarm Setpoint for the Diesel Fuel Oil Tank" is complete.
Description of Change The setpoint of the diesel fuel oil storage tank lov level alarm (LS-7004) vas temporarily changed from 4'-8" (45 percent) to 2'-6" (20 percent).
Reason for Change This setpoint change restored a "dark board" (no alarm) condition at the annunciator vindow. The fuel level was below the alarm setpoint and oil was not going to be added to the tank until the 1988 refueling outage when the tank vas pumped down and inspected.
Safety Evaluation This change did not have any impact on the accidents evaluated in the design basis because this switch was used only to monitor tank level and did not affect operation of the Emergency Diesel Generators (EDG).
The EDGs are designed to operate for a week using the fuel oil in the day tanks, not this storage tank. Operating at this lover level vould require a tanker truck to fill the tank sooner if an accident vere to occur and the EDGs were running, but this can be accomplished without causing a safety concern. ,
Therefore, this change did not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 2-87-014 i Setpoint Change' Number 2-87-014 entitled "Auxiliary Building i Ventilation Air Supply Fan Flov Alarm" is complete. '
Description of Change-The;setpoint of the Auxiliary Building Ventilation air supply fan ,
- .(F16) lov flow alarm (FS-8609) was changed from 0.2 inch V.G. to 0.1 inch V.G.
Reason for Change 4
This sutpoint.vas decreased in order to clear-the lov flow alarm which-vas in alarm.
i Safety Evaluation i This alarm is used as a monit ting d vice only and does not affect any _;
safety system. Lovering this alarm still provided adequate fan pro-tection. !
i Therefore, this change does not constitute an unreviewed safety '
3 question per the criteria of 10CFK50.59.
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I Setpoint Change Number 2-87;0,17 Setpoint Change Number 2-87-017 ei.c.: lad "Overcurrent Trip for a Station Electrical Service Breaker" is s malete.
n Description of Change The time delay setpoint of the overcurrent trip setpoint for one of the two circuit breakers in series in the tie line between Hillstone Unit 2 and Millstone Unit 1 (A505) was permanently changed from 3.5 to 4.0 seconds and the trip value was changed from 960 to 1200 amps.
Reason for Change The changes provided more time for the downstream breakers to trip cn a fault before A505 trips. This change was analyzed during the 4160 volt bus coordination study performed under Appendix R vork and it was recommended that this change be made.
Safety Evaluation There is no change to the Final Safety Analysis Report (FSAR) description of the 4160 volt bus operation as a result of this change.
Bus protection details vere not provided in the FSAR. he change f.sd no effect on normal or emergency plant operations.
Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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TESTS INDEX Number Title T85-009 Electric Power Research Institute (EPRI)
Radvaste Ion-Exchanger Prototype Test T86-010 Emergency Operating Facility (EOF) Diesel Generator Load Test T86-016 Power Ascension Test - Cycle 8 T86-018 Main Exhaust Filter Test T86-020 Service Vater (SV) FJov Through the Reactor Building Cooling Vater (RRCCV) Heat Exchangers T86-021 Moisture Carryover /Feedvater Flov Rate Testing T86-035 Loose Parts Monitoring System Sensitivity Bump Test Verification T87-002 Steam Generator Leak Location Determination Satety Parameter Display System In-Service
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T87-004 Test T87-006 Condenser Tube Fouling Monitor Test T87-008 Reactor Building Closed Cooling Vater (RBCCV)
Temporary Demineralizer System T87-009 Condensate Polishing Facility (CPF) Discharge Rad Monitor Temporary Strainer Flov Test T87-014 6.9/4.16 kV Switchgear Room Cooling T87-015 Millstone 2 Plant Process Computer Memory Test T67-020 Process Computer Memory Test T87-021 Primary Makeup Vater (PMV) System Flush -
Turbine Building T87-022 Spent Fuel Pool (SFP) Clarity Test T87-023 Condensate Polishing Facility (CPF) Vaste Neutralization Discharge System Phase 1 Electrical Test
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TESTS INDEX (Continued)
Number Title T87-027 Pressure Drop Test on a Vestinghouse Fuel.
Assembly T87-031 Spent Fuel Pool (SFP) Cleanup System Operation Test Spent Fuel Pool (SFP)-Sludge Dispersion Test T87-034 T87-038 Safety Injection Valve Stroke Test T87-043 Motor Operated Valve Actuator Testing System (MOVATS) Testing of Main Steam Valves, +
2-MS-201, 2-MS-202, and SV-4188 T87-045 Motv- Operated '!:lyw Actuator Testing System (MOVATS) Testing of Auxiliary Feedvater Yalve !
2-FV-44 2
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NUMBER TITLE T85-009 Electric Power Research Institute (EPRI) ,
, Radvaste Ion-Exchanger Prototype Test Description of Test This testLinvolved the processing of aerated radvaste through a ._
prototype ion-exchange resin bed. This was an Electric Power Research .;
Institute (EPRI) sponsored evaluation test.
Reason for Test The purpose of this test was to determine =the effectiveness of a pilot g EPRI lon-exchange treatment system for processing aerated radvaste. !
Safety Evaluation .
The performance of this test did not affect any radvaste system
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functions and no safety systems were compromised.
1 Th'refore, this test did not constitute an unreviewed safety question -
per the criteria of 10CFR50.59.
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NUMBER TITLE T86-010 Emergency Operating Facility (E0F) Diesel Generator Load Test ;
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Description of Test <
The capability.of the Emergency Operating Facility (EOF) diesel' )
generator was verified. .
Reason for Test ;
To verify-that the_ diesel generator vas'not overloaded and that all-4- loads were being supplied sufficient power.
j Safety Evaluation ,
The testing of the E0F diesel generator did not affect any plant '
systems.
I Therefore, this-test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i NUMBER -TITLE -t T86-016 Power Ascension Test - Cycle 8 Description of Test This test brought Millstone Unit 2 from the completion of low power physics testing to full power operation. This-test also ensured that all reactivity related surveillances were completed.
Reason for Test This test verified that the Vestinghouse core physics design model was in agreement with actual plant response.
Safety Evaluation This test utili.ed normal operating procadures and controls. .
. Therefore, this test did not constitute an unreviewed. safety question
- per the criteria of 10CFR50.59. .
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I NUMBER TITLE 4
T86-018 Main Exhaust Filter Test Description of Test i
l- This test installed and determined the effectiveness of a nov !
polyester multi-ply roughing filter in the main-exhaust (ME) system. -
Reason for Test The new filters provided higher efficiencies and loadings than the existing Fiberglas filters.
Safety Evaluation 'f Filters L-26 and L-271 vere not safety related.- L-25 was an extension of the enclosure building ~ pressure boundary which is a safety-related j
-structure. Installation of Filter L-25 did not have any effect on the pressure retaining capacity and did not compromise any safety-related
] functions.
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! Therefore, this test did not constitute an unreviewed safety question l per the criteria of 10CFR50.59. >
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NUMBER TITLE T86-020 Service Vater (SV) Flov Through the Reactor Building Cooling Vater (RBCCV) Heat Exchangers Description of Test This test established a Safety Injection Actuation Signal (SIAS) valve line-up, failed the Reactor Building Cooling Vater (RBCCV) outlet temperature control valve open, and recorded flovs. Adjustments vere then made to the service water valves as required to meet flow requirements.
Reason for Test This test was performed to verify adequate Service Vater (SV) flow passes through the RBCCV heat exchangers when subjected to an accident (SIAS) valve line-up. This test also ensured that this line-up did not exceed SV pu9p capacity.
Safety Evaluation During the performance of this test all normal system functions were maintained.
Therefore, this test did not constitute an unrevieved safety question per the criteria of 10CFR50.59.
i NUMBER TITLE T86-021 Hoisture Carryover /Feedvater Flov Rate Testing Description of Test This test provided a procedure for injecting a lithium isotope in the main feedvater system and for sampling the main steam and blovdovn systems for the presence of the lithium isotope.
Reason for Test This test was performed to determine the actual moisture carry-over from the steam generators and mass flovrate to the steam generators.
The data from this test was used to determine the efficiency of the steam generators.
Safety Evaluation The sample injection equipment was designed for the conditions of use and did not have any deleterious effects on existing plant systems or personnel safety. Addition of the lithium was found to be acceptable.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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NUMBER TITLE T86-035 Loose Parts Monitoring System Sensitivity Bump Test Verification Description of Test This test involved striking the reactor vessel lover flange and steam generator bowls at known locations with a force hammer. The vibra-tions produced vere sensed by eight (8) installed accelerometers.
Each time the force hammer struck, the data was used to determine the a
sensitivity and satisfactory operation of each sensor channel. !
Reason for Test This test was developed to verify and document the Loose Parts Monitoring System sensitivity prior to start-up.
Safety Evaluation !
This test did not affect plant operation or any systems important to safety. The force delivered to create the reference signal was too small to cause any damage.
Therefore, this test did not constitute an unreviewed safety question '
per the criteria of 10CFR50.59.
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NUMBER TITLE '
T87-002 Steam Generator Leak Location Determination Description of Test This test started with a full steam generator, pressurized with nitrogen. The steam generator was then pumped down while monitoring leakage via cameras installed in the steam generator channel heads.
Reason for Test The purpose of this test was to determine the exact location of the tube leak.
Safety Evaluation This test utilized normal operating procedures and controls.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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' NUMBER ' TITLE T87-004 Safety Parameter. Display System In-Service Test 1
-Description of Test This test was written to provide a procedure to allow plant data to be
-collected and analyzed for_the Jafety Parameter Display System (SPDS) on the Plant Process Computer.
Reason for Test l The purpose of this test was to validate the plant process computer 7 data collection process for use in the SPDS system. '
Safety Evaluation This test was performed on the "test" host computer and did not affect the performance of the "master" host computer. Therefore, the per-formance of this test did not affect plant operations.
Therefore, this test did not constitute an unreviewed safety question [
per the criteria of 10CFR50.59.
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.I NUMBER TITLE l l
T87-006 Condenser Tube Fouling Monitor Test Description of Test This test was developed to aonitor the fouling of condenser tubes by :
taking a slip streaa of circulating vater and routing it through the !
test equipaent. ,
Reason for Test This test was used to evaluate fouling rates and effectiveness of the l sodiua hypochlorite additions. .
i Safety Evaluation l l
This test incorporated appropriate precautions and controls to i circuavent all postulated failures.
Therefore, this test did not constitute an unreviewed safety question i per the criteria of 10CFR50.59.
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NUMBER TITLE ,
I l T87-008 Reactor Building Closed Cooling Vater (RBCCV)
Temporary Demineralizer System ,
Description of Test j t
This test installed a temporary demineralizer column on the discharge i of a Reactor Building Closed Cooling Vater (RBCCV) pump. Flov was ,
reintroduced to the system at a suction header low point drain. Flov l through this temporary demineralizer was limited to 25 gallons per t minute. I i
Reason for Test
- This inservice test was performed to determine the effectiveness of ;
the temporary demineralizers to remove chlorides in the RBCCV system. l Safety Evaluation i
Facility separation was maintained at all times during the performance ;
of this test.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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NUMBER TITLE T87-009 Condensate Polishing Facility (CFF) Discharge Rad Monitor Temporary Strainer Flow Test Description of Test This test was performed by installing various types and element size filters in the CPF discharge rad monitor piping.
Reason for Test The purpose of this test was to evaluate filter types to eliminate lov flov switch failures. This filter arrangement is temporary in nature.
r Safety Evaluation This test incorporated appropriate precautions for potentially contaminated systems and utilized approved Operations Department procedures.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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NUMBER TITLE T87-014 6.9/4.16 kV Svitchgear Room Cooling Description of Test This inservice test realigned service water system valves.to the 6.9/
4.16 kV svitchgear rooms and simulated a pipe break in the room. The test bypassed the control valve isolation function due to a leak and simulated a pipe break. The simulated break flov vas directed to a floor drain.
Reason for Test This test simulated a pipe break in the room to determine system hydraulic response. The results of this test vere used to support system improvements.
Safety Evaluation i
This test incorporated appropriate controls and precautions to prevent !
any compromise of electrical system integrity. ,
i Therefore, this test did not constitute an unrevieved safety question i per the criteria of 10CFR50.59. i i
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NUMBER TITLE 1
T87-015 Millstone 2 Plant Process Computer Memory :
Te.ct Description of Test 1
This test installed additional memory in the Millstone Unit 2 Plant l Process Computer (PPC). This test was vritten to demonstrate.the i stability and operability of PPC vhen configured with additional memory. As such additional memory and diagnostic software was installed in the "B" host computer.
Reason for Test i The objective of this test was to determine whether PPC system response can be improved by adding memory to the PPC.
Safety Evaluation
- This test incorporated a variety of precautions and diagnostics to i ensure no degradation of PPC functions occurred during the performance of this test. Upon completion of this test all temporary memory and
] software vas removed from the "B" host computer.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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NUMBER TITLE !
T87-020 Process Computer Memory Test Description of Test Each Host processor was powered down (one at a time) to install the ;
additional memory. The computer was restarted and diagnostic programs !
vere run to ensure the system operated properly. After the diag-nostics were completed, the computer was made fully operational and the general system performance vas. observed to determine if the system operated properly, i
Reason for Test I This test demonstrated the stability and operability of the plant !
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process computer host Central Processing Units (CPU) configured with 16 megabytes of memory. ..
Safety Evaluation l Since the plant process computer is not a safety-related component, this test did not affect the response of any safety systems.
t therefore, this test did not constitute an unreviewed safety question ;
per the criteria of 10CFR50.59. ;
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l NUMBER TITLE T87-021 Primary Makeup Vater (PNV) System Flush -
Turbine Building Description of Test This inservice test flushed soveral sections of the Primary Makeup Vater (PNV) lines in the Millstone Unit 2 Turbine Building. The flushed water was collected, analyzed for activity, and then disposed of via radvaste systems.
Reason for Test This test was performed to reduce contamination in the Turbine Building portions of the PMV system to less than minimum detectable activity levels.
Safety Evaluation This test utilized normal operating procedures and controls.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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NUMBER TITLE T87-022 Spent' Fuel Pool (SFP) Clarity Test ,
Description of Test i This test evaluated various combinations of temporary filters, demineralizers and. coagulants in the Spent Fuel Pool (SFP) purifi-cation sys:em.
Reason for Test The existing, installed purification system components were ineffective in removing the organic contaminants present in the SFP vater.
l~ Safety Evaluation t
System design and chemistry were evaluated with respect to their l effect on the SFP and SFP purification system and vere found to be l acceptable. Any abnormal operation vould be covered by existing analysis and operating procedures.
Therefore, this test did not constitute an unreviewed safety question :
per the criteria of 10CFR50.59.
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NUMBER TITLE E TS7-023 Condensate Polishing Facility (CPF) Vaste Neutralization Discharge System Phase 1 Electrical Test Description of Test This test performed three standard procedures for installation of nev !
breakers,' breaker overload heaters and logic changes. ,
Reason for Test This procedure vas used to provide a complete Phase 1 test for the installation of the new Condensate Polishing Facility (CPF) Tank 10 l and 11 sump discharge pumps.
l Safety Evaluation This procedure employed three (3) existing standard testing procedures !
into one procedure, i j i
Therefore, this test did not constitute an unreviewed safety question !
per the criteria of 10CFR50.59.
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NUMBER TITLE i
T87-027 Pressure Drop Test on a Vestinghouse Fuel Assembly Description of Test This test determined the hydraulic characteristics of a Vestinghouse fuel assembly.
Reason for Test This testing was required to support future fuel reload calculations by Advance Nuclear Fuels Corporation.
Safety Evaluation This test was evaluated for 1) ( iticality of the test 1oop,
- 2) failure of the fuel clad while the assembly is in the test loop.
- 3) leak in the test loop during the test, 4) potential adverse vater chemistry effects on the fuel during the test, 5) damage to the fuel assembly during installation and removal from the test loop,
- 6) hydraulic damage to the fuel assembly components during the test, and 7) potential damage to safety-related equipment during a seismic event. All evaluations vere found to be satisfactory.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
NUMBER TITLE T87-031 Spent Fuel Pool (SFP) Cleanup System Operation Test Description of Test This test operated and monitored the effectiveness of the temporary cleanup system on the Spent Fuel Pool (SFP).
Reason for Test The purpose of this test was to clean up the SFP vater.
Safety Evaluation The system used did not affect the chemistry of the SFP or the fu.
the pool. It was utilized in ronjunction with existing plant procedures.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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NUMBER TITLE 1
- T87-034 Spent Fuel Pool (SFP) Sludge Dispersion ;
Test ,
Description of Test
! i Various amounts of hydraulic fluid vere added to samples of Spent Fuel- .i j Pool (SFP) vater that vere known to contain sludge from the-bottom of l 1 the'SFF. The samples were agitated, maintained in a constant temper- l ature bath, and analyzed for turbidity and total organic carbon. !
1 Reason for Test l l
This test evaluated the ability of the Union Carbide VS 34 hydraulic j fluid to promote a stable suspension in the SFP vater. l I
4 Safety Evaluation ;
i The hydraulic fluid and SFP vater vere mixed and.placed in a container 1 in the SFP. The possibility existed that the container vould leak l
- allowing this aixture to enter the SFP. This was analyzed and found l 1
to be acceptable.
Therefore, this test did not constitute an unreviewed safety question j i
per the criteria of 10CFR50.59. ,
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NUMBER TITLE T87-038 Safety Injection Valve Stroke Test l
Description of Test i This test involved the stroking of twelve (12) High Pressure Safety Injection (HPSI) valves against a differential pressure 'reated by the !
"B" HPSI pump. The test was accomplished with the reactor vessel head removed while filling the Reactor Coolant System (RCS) for refueling operations.
Reason for Test The purpose of this test was to verify the operability of these valves under simulated design basis accident conditions (i.e., vorst caso differential pressure).
Safety Evaluation The performance of this test did not affect any accident analyses.
Additionally, no Technical Specification safety margins vere reduced during the performance of this test.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.50.
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NUMBER TITLE T87-043 Motor Operated Valve Actuator Testing System (M0 VATS) Testing of Main Steam Valves, 2-MS-201, 2-MS-202, and SV-4188 Description of Test This test involved the installation of Motor Operated Valve Actuator Testing System (MOVATS) test equipment and the subsequent stroking of three main steam system valves. The valves tested vere 2-MS-201 and 202 (Respective Main Steam Header Stop Valves to Steam Driven Auxiliary. Feed Pump) and SV-4188 (Main Steam Stop Valve to Steam Driven Auxiliary Feed Pump).
Reason for Test The purpose of this-test was to obtain an operating thrust and-current signature for each valve using MOVATS equipment. This testing was performed to verify proper torque svitch settings to ensure valve operability under simulated design basis accident conditions.
Safety Evaluation The performance of this test did not affect any accident analyses.
Additionally, no Technical Specification safety margins were reduced
} by the performance of this test.
Therefore, this test did not constitute an unreviewed safety question
, per the criteria of 10CFR50.59.
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1 NUMBER TITLE- !
I T87-045 Motor. Operated Valve Actuator Testing. System (H0 VATS) Testing of Auxiliary Feedvater Valve 2-FV-44 Description of Test '
This inservice test involved the stroking of Auxiliary Feedvater Valve, 2-FV-44. Vith a motor driven auxiliary feedvater pump ,
operating at minimum flow, valves 2-FV-44 and~2-FV-43B vere stroked open and closed.
Reason for Test The purpose of this test'vas to obtain an operating threst and current signature for 2-FV-44 using MOVATS equipment. This testing was performed to verify proper torque'svitch settings to-ensure valve
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operability under simulated des l.+1-basis accident conditions. ~l Safety Evaluation This test was performed with the plant operating in a r. eed mode to preclude any operability concerns with respect to the Auxiliary Feedvater System. The performance of this test did not affect any accident _ analyses. Additionally, no Technical-Specification safety margins were reduced by the performance of this test.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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EXPERIMENJT j i
~ Plant' Design Change 2-007-87,. Fuel Consolidation Demonstration-Program, was conducted under the provisions of 10CFR50.59 Special ;
Tests end Experiments. See the. Plant' Design Change Section, Plant j
Design Change ::;-ber 2-007-87, for more information. .
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CHALLENGES TO' RELIEF VALVES In accordance with the commitment made under Item II.K.3.3 of NICEG .;
0737 (THI Action Plan) in-the V. G. Council ~ letter.to D. G. Eisenhuc, 1
dated June 10, 1980,- the following is a report of challenges to relief / safety valves duririg 1987.
There vere no challenges to talief/ safety valves during 1987.
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C00 M IODINE spIgygg The specific activity of the primary coolant did not exceed the limits stated in the Technical Specifications.
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-I STEAM GENERATOR TULE IN-SERVICE-INSPECTION :
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There. vere no Steam' Generator Tube.In-Service Inspections-performed during 1987.
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e MILuSTONE UNIT 3 CONTENTS SECTION ~PAGE Changes Design Changes.. . . . . .. .- . . . 1 Procedure Changes . . . .. . . . . .. 91 Jumper-Lifted Lead-Bypass Changes . . . 93 Setpoint Changes . . .- . .- . . . . 99 Tests . . . . . . . . . . . . . . 114 Experiments . . . . . . . . . . . .- 134.
Occupational Radiation Exposure . .- . . . . .. 135 s Challenges to Relief Valves . . . .
. . . 136 s
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PLANT DESIGN CHANGE INDEX PDCR Number Title MP3-85-002 HVV Cable Pulls Through Sealed Racevays MP3-85-019 Add CO 2 Predischarge Alarms in Auxiliary Boiler Enclosure.
MP3-85-023 Addition of Cable Tray Bracing MP3-85-024 HVR Cable Splices, CES, HV0, HVR Cable Pulls Through Sealed Raceways MP3-85-025 Replace Diesel Fuel Oil Piping MP3-85-031 Replacement of 3CHS*V310 to Increase Dilution Flow MP3-85-032 Jacket Cooling Vater Lov Temperature Alarm MP3-85-035 HVV System-Cable Installation And Termination MP3-85-036 Fuel Handling Machine Control Circuit Modification MP3-85-044 Charging Pump and Component Cooling Vater Pump Area Supply Fan Modification MP3-85-046 Racevay Interference at 3CCP*A0V10B MP3-85-052 Containment Air Recirculation 3HVU-TC30A, B, C Setpoint Change MP3-85-054 3FVS-E2 Cooler Throttle Valve Addition MP3-85-055 Drain Isolation Valve Addition to ARC-RE21 MP3-86-002 Modify Tank Heater Circuits to Provide Failsafe Alarm MP3-86-003 Diesel Lube Oil Moisture Detector Alarm Setpoint MP3-86-008 Modify Auxiliary Turbine Oil Pump To Lover i Resonance Frequency l
MP3-86-013 Sample Valves To 3CCP-RE31 Not Cabled (UNS '
6993)
MP3-86-014 Sample Pump Heat Trace Circuit Interlock for Containment Hydrogen Monitoring MP3-86-030 Temporary Fill Connection for DVST
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l PLANT DESIGN CHANGE INDEX (Continued)
PDCR Number Title MP3-86-034 Provide Heat Trace To The Condensate Storage And Surge Tank Relief Valves MP3-86-039 Lead Shielding for Solid / Liquid Radvaste Disposal Lines MP3-86-042 keroute 3HVH-SCL1 Outlet Cooling Vater MP3-86-047 3BRS-RV95 Cap Leaks (DDR 960)
HP3-86-048 Main Turbine - Stop Loading Blocking Signals MP3-86-051 Revisions to the Boric Acid and Total Hakeup Flow Deviation Circuitry MP3-86-053 Containment Air Filtration Heater Flov Switch MP3-86-055 Cable Separation Barrier at 3TH413N MP3-86-057 Additional Overpressure Protection For 3CNS-TK1 and TK2 MP3-86-058 Boron Evaporator Feed Pump NPSH Modification MP3-86-060 Hot Vater Heating High Energy Line Break Isolation MP3-86-064 Addition of an Alarm for Vacuum Breaker Valves Auxiliary Circuit (DDR-491)
HP3-86-066 Pressure Loop 508 Modification for Vest'.nghouse Interface MP3-86-070 Rr.dioactive Liquid Vaste System Pumps' Circuit Modifications MP3-86-073 Revision of Reactor Containment Lighting Controls MP3-86-074 Deletion of HVAC Fire Dampers (3HVR*DHPF4,5; 3HVQ-DMPF1021)
HP3-86-075 Installation Details for Generator Leads Cooling Fans Circuit j MP3-86-076 Modifications To The Steam Generator Blovdown Sampling System MP3-86-079 Recircuiting of Emergency Battery Povered Lighting Packs in Containment l
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PLANT DESIGN CHANGE INDEX (Continued)
PDCR Number Title MP3-86-086 Installation of a Resistor in the Panel 3HVC .
PNL01A/B (DDR-890)
MP3-86-088 CCS Heat Exchanger Protective Coating MP3-86-105 Deletion of Pressure Indicators 3HVC-PI201A and 201B From the Control Room Envelope Pressur!Jation (Bottled Air) System MP3-86-114 Feedvater Pump Seal Vater Piping Modification MP3-86-122 Tube Union Addition (CP396724)
MP3-86-125 Separation Barrier on Tray Support C058C MP3-86-128 Installation of VAR Transducer MP3-86-138 Replacement of Cables in Chlorine Pit of Intake Structure MP3-86-153 3SVP*MOV102A-D Replacement MP3-86-155 Circuit Additions to the Chemistry and Health Physics Laboratories MP3-86-161 Circulating Vater Pump Control Circuit Time Delays MP3-86-202 Permanent Platform to Receive and Inspect New Fuel MP3-86-223 Elgar Inverter - Replacement of Solid Bus Bar MP3-86-261 Power Peed to Gate Operator and Control Station MP3-86-265 Main Steam Safety Relief Valves Drain Line Modification MP3-86-266 Main Steam Pressure Relieving Valves Open Belov Setpoint (UNS 7579)
HP3-86-272 Normal And Reserve Station Service Time Of Day Meter MP3-86-284 Replacement of Steam Generator Vater Level Recorders MP3-86-311 Reroute MP3 Pire Protection Yard Loop For MP3 Maintenance Shop Expansion PLANT DESIGN CHANGE INDEX (Continued)
PDCR Number Title MP3-86-312 Damaged Page/ Evacuation Cable 3COPCNX126 MP3-86-316 Screenvash Piping Replacement MP3-86-367 Added Insulation In Main Steam Valve Building For Heat Load And Personnel Protection MP3-86-379 Replacement of the Valve Positioners on Feedvater Regulating Bypass Valves MP3-86-386 Security Lighting HP3-86-387 Replace Terminal Lugs for Cable CCIAOC001 MP3-86-388 Installation of Separation Barriers on Elevation 24'-6" of Control Building MP3-86-408 Main Steam and Steam Generator Blovdown System Nitrogen Isolation Valves Removal /
Reorientation MP3-86-420 Installation of Cable Tray Cover / Conduit Vrap on Elevation 24'-6" of Auxiliary Building HP3-87-002 High Radiation Area Gate Alarms And Varning Lights MP3-87-008 Resolution of Instrument Cable Bend Radius Violations MP3-87-010 Installation of Cable / Conduit Vrap MP3-87-013 Damaged Cable (30SSNNX225)
MP3-87-014 First and Second Point Feedvater Heaters Operating Vent Piping Replacement MP3-87-018 Replacement of Post Accident Sampling System Flush Pump Motor MP3-87-020 Radiation Monitor System Computer Upgrade MP3-87-026 PORV Seat Material Change MP3-87-027 Feedvater Recirculation Valve Internals Changeout MP3-87-030 Penetration Protection for Non-Category 1 Circuits l
PLANT DESIGN CHANGE INDEX-(Continued) i PDCR Number Title . ,
MP3-87-034 Carbon Dioxide Fire Protection Keylock Switch Modification MP3-87-036 Feedvater Venturi Inspection Ports MP3-87-038 Safety Provisions for the Auxiliary Boiler Pressure Relief Line MP3-87-045 shielding of the Fuel Transfer Canal Traverse MP3-87-053 Transformer Cooling Controls MP3-87-061 Installation of Bracing to Cable Iray Supports to Auxiliary Building Elevation ;
43'-6" and MCC and Rod Control Area Elevation )
45'-6" l MP3-87-062 Refueling Cavity-Drain Line Modification f i
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l Plant Design Change Number MP3-85-002 Plant Design Change Number MP3-85-002 entitled "HVV Cable Pulls Through Sealed Raceways" is complete.
Description of Change Main Steam Valve Building HVAC (HVV) cables were rerouted to avoid '
construction interferences. This change provided details required-to reroute the cables.
Reason for Change I
The original routing of the cables required the cables to go through vall/ floor sleeves and blockouts that had already been sealed with :
grout. To remove the grout could cause damage to the existing cables l installed in the sleeves for blockouts. To avoid possible damage to the existing cables, the new cables were rerouted to spare sleeves or i new sleeves were installed.
Safety Evaluation This safety evaluation addressed the rerouting of the HVV cables as I described below:
The cables vere rerouted to spare sleeves or new blockouts via new conduit and junction boxes; the new installations being seismically supported. In some cases, core drilling a new sleeve parallel to the existing seal sleeve was required. Any new pene-trations through a fire, pressure or radiation barrier vere resealed to maintain the integrity of the barrier.
The rerouting of the cables meets the licensing requirements _for separation, Appendix R, hazards analysis.
This modification consists only of rerouting of cable to avoid construction interferences. This change had no impact on system operability. There is no voltage drop impact due to the rerouting.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
Plant Design Change Number MP3-85-019 1 Plant Design Change Number MP3-85-019 entitled."Add CO 2 Predischarge l Alarms in Auxiliary Boiler Enclosure" is complete. '
Description of Change This change added varning strobe lights to the areas adjacent'to the Auxiliary Boiler Enclosure Fuel Pit. The lights are actuated 60 seconds in advance of a discharge of the carbon dioxide fire sup-pression system and continue to flash throughout the discharge.
Reason for Change The purpose of this design change vas to enhance personnel safety.by ,
providing an additional level of varning in the areas adjacent to the carbon dioxide gas discharge. '
Safety Evaluation The addition of the alarms does not affect any safe shutdown system components or other safety-related systems or components.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. <
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l Plant Design Change Number MP3-85-023 l j
Plant Design Change Number MP3-85-023 entitled "Addition of Cable Tray Bracing" is complete.
Description of Change The change added bracing to five cable tray _ supports in the Control Building required to complete the design verification. ,
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Reason for Change- !
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The change was required based on a review of the cable tray supports l per structural design criteria for Category I cable tray supports. )
l Safety Evaluation
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The modification does-not alter the design function of the cable ;
trays. The supports vill not be subject to structural failure or I collapse based on the inherent designed conservatism combined with conservative analysis assumptions. {
1 Therefore, this change does not constitute an unreviewed safety j question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-85-024 Plant Design Change Number-MP3-85-024 , entitled."HVR Cable Splices, CES, HVQ, HVR Cable Pulls Through Sealed Raceways" is complete.
Description of' Change This change vas to splice some cables and reroute others to avoid construction interferences. Additional conduits were installed to reroute the cables.
Reason for Change The original design could not be completed because there was not enough slack in the cables to allov for the splicing. The additional conduits were required to facilitate the rerouting of the cables, due to original duct lines being sealed. In order to avoid damaging existing installed cables, it was decided that the cables vould be routed through new conduits.
Safety Evaluation This change provided the installation details for cables and conduits which vere already included in the design of the plant. There was no impact on the system operability of any of the affected systems due to the change. All the installations vere performed in accordance with the electrical installation specification and meet the licensing requirements for electrical separation, fire protection, and seismic consideration.
Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-85-025 Plant Design Change Number MP3-85-025 entitled "Replace Diesel Fuel Oil Piping" is complete.
Description of Change This change replaced a section of vendor supplied diesel fuel oil piping (copper) with stainless steel piping. This change was per-formed on both of the emergency diesel generators. _The piping'vas rerouted from the original to provide better support.
- ason for Change Vibration was causing the pipe to rub against other pipes, causing-severe fretting of the pipe. The change was made to have a material less vulnerable to wear and to change tne support system to minimize vibration induced pipe movement.
Safety Evaluation This change did not affect the design function of the diesel fuel oil system. The internal diameters of the previous and the new piping are essentially equal, thus there was no impact on the system function from either a fluid system or an engine operability standpoint.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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l Plant Design Change Number MP3-85-031 l
Plant Design Change Number MP3-85-031 entitled "Replacement of 3CHS*V310 to Increase Dilution Flow" is complete.
Description of Change Two-inch check valve 3CBS*V310 was replaced with a three-inch check valve. '
1 Reason for Change Startup testing identified that insufficient primary grade water was being supplied to the VCT when in the automatic makeup and dilute modes of operation. An engineering review of the system head losses revealed that a large differential pressure, 28 psi at 120 gpm, occurred across a two-inch check valve.. Vith a three-inch check valve, the differential pressure is less than 1 psi at 120 gpm.
Safety Evaluation The design change did not alter the function of the primary grade i vater system as the new check valve has the same capability to j restrict boric acid backflov into the chemical addition tank and l primary grade water system ss did the original check valve. j i
The change required the relocation of one pipe support by approxi- l mately three inches because of interference with a reducer. However, stress levels in the pipe are lov and are not substantially affected.
Failure of the new check valve is no different than the failure of the !
original check valve. l Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-85-032 Plant Design Change Number MP3-85-032 entitled "Jacket Cooling Vater Lov. Temperature Alarm" is complete.
Description of Change The emergency diesel generators' (EDG) lov jacket cooling vater tem-perature alarm circuitry was modified so the alarm was made inoperable during engine operation.
Reason for Change By design, the jacket cooling vater keep-varm pump stops whenever the engine runs. As a result, there is no flow in the piping that connects the pump with the main headers due to a. check valve on the pump discharge. Consequently, the water in the affected piping cools off. Since the alarm switch is located in this "dead" leg, a spurious alarm results.
Safety Evaluation The alarm is part of the jacket coolant standby system. This system operates only when an EDG is shut down. Since the alarm is bypassed only when_an engine operates, there is no affect on the design of the ,
lov temperature annunciation system. The additional viring meets the licensing requirements for separation, Appendix R, seismic and hazard analysis. The added viring is minor and has no impact on circuit voltage drop.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. 3 1
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Plant Design Change Number MP3-85-035 Plant Design Change Number MP3-85-035 entitled "EVV System-Cable Installation And Termination" is complete.
Description of Change This modification-provided revised cable routing, new conduit and.
terminal box installation details and revised cable termination details.
The_ existing cables were rerouted to spare or nev. sleeves. New raceway junction boxes were installed, seismically supported. The junction boxes were added to terminate the undamaged portion of the cables to new cables. These new added cables were routed in separate conduit to avoid the duct, which in turn vere wired to complete the circuits as intended in the original design.
Reason for Change 4
The following constructibility problems vere encountered during installation:
Cables were routed into conduit too full to permit installation of additional cables.
Cable had been routed through vall/ floor sleeves and blockouts which had been sealed.
Cables vere damaged while pulling ductline.
Additional termination and training space was required in junction boxes.
Safety Evaluation All electrical installation was performed in accordance with the applicable Electrical Installation Specification.
Any barriers in which new penetrations had been installed vere resealed to maintain barrier integrity.
Therefore, this modification does not constitute an unreviewed cafety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-85-036 Plant Design Change Number MP3-85-036 entitled "Fuel Handling Machine Control Circuit Modification" is complete.
Description of' Change This change added a 2000 ohm resistor to the fuel handling machine hoist up stop overload circuit.
Reason for Change The circuit component was added to correct a malfunction in the hoist up stop overload circuit.
Safety Evaluation Addition of the 2000 ohm resistor allowed correct operation of the '
hoist up stop overload circuit. Failure analysis shoved that open circuit failure vould return the circuit to the previous mode of operation. Short circuit failure vould stop hoist operation and allow '
no movement of hoist. dince the addition of the resistor could only improve circuit opere. ion, and could not produce an unsafe condition, the modification was considered safe.
Therefore, this modification does not constitute an unreviewed safety question per the critarls of 10CFR50.59.
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Plant Design Change Number MP3-85-044 Plant Design Change Number MP3-85-044 entitled "Charging Pump and Component Cooling Vater Pump Area Supply Fan Modification" is complete. ,
Description of Change This plant design change modified setpoints on two sets of lov flov sensing switches, increased the time delays on two relays, and removed instantaneous contacts from the control circuits of the charging pump and component cooling water pump area supply fans. The modification was made to allov the charging pump and component cooling vater pump area supply fans to function as they were originally designed.
Reason for Change This change was required to prevent the charging pump and component cooling water pump area supply fans from cycling following a loss of power on both emergency busses.
Safety Evaluation Previous testing had proven the modification vould function as desired. The change did not affect the design bases or safety evaluation of the Auxiliary Building ventilation system as described in Final Safety Analysis Report (FSAR) Section 9.4.3.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-85-046 Plant Design Change Number MP3-85-046 entitled "Racevay Interference at 3CCP*A0V10B" is complete.
Description of Change This change provided installation and support details for a conduit leading to the actuator for a component cooling vater air operated valve.
Reason for Change A previous design change required that the valve actuator on 3CCP*A0V10B be replaced with a different model. The conduit could not be installed on the existing seismic support, so a new support and installation details vere required.
Safety Evaluation This change did not affect any system operability, nor was the design function of the affected valve changed. The implementation of the change met all licensing requirements for electrical separation, fire protection, and seismic considerations.
, Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-85-052 Plant Design Change Number MP3-85-052 entitled "Containment Air Recirculation 3HVU-TC30A, B, C Setpoint Change" is complete.
Description of Change The setpoints on the temperature controllers for containment air l temperature were changed to 75 degrees Fahrenheit.
Reason for Change During the plant precore hot functional testing, the containment, temperatures were found to be unacceptably high. This setpoint change maintains the average containment temperature belov the design limit specified in the Final Safety Analysis Report (FSAR).
Safety Evaluation This change does not adversely affect the equipment environmental-qualifications lover temperature limitations. -The design basis of the system has not been changed by this modification, and neither has the operability of the system.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. j a
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4 Plant Design Change Number MP3-85-054 Plant Design Change Number MP3-85-054 entitled "3FVS-E2 Cooler Throttle Valve Addition" is complate.
Description of Change This change added a globe valve to the discharge of the Steam Generator motor driven feedvater pump air cooler (3FVS-E2).
Reason for Change This change ensured the design differential pressure of 6.5 psid across the cooler vould not be exceeded.
Safety Evaluation The' motor driven feedvater pump and its associated support systems is a nonseismic, nonsafety-related piece of equipment located in the nonseismic Turbine Building. It is not required to prevent or miti-gate the consequences of any accidents.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-85-055 Plant Design Change Number MP3-85-055 entitled "Drain Isolation Valve Addition to ARC-RE21" is complete.
Description of Change This change added a 3/8" needle valve to the drain line from an inline radiation monitor to provide isolation of the process fluid. This radiation monitor's purpose is to detect primary to secondary leakage by continuously monitoring the air discharged by the steam jet air '
ejectors.
t Reason for Change The radiation monitor was provided by the vendor without any means of isolating the drain system. The' moisture collected from the process is potentially radioactively contaminated and vould have drained onto the floor.
Safety Evaluation i
This change did not alter the performance of the radiation monitor and had no effect on the operation of any other systems. -
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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l PlanF Design Change Number MP3-86-002 ;
Plant Design Change Number MP3-86-002. entitled "Modify Tank Heater
, Circuits to Provide Failsafe Alarm" is complete.
I Description of Change This change modified the heater control circuits for various' outdoor tanks to provide positive annunciation upon heater circuit loss of ;
power.- The change was accomplished by simple reviring within-the '
heater controllers, with no.new equipment or components added. ~The heater control circuits on the following' tanks were modified by this
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changes 3LVS-TK3A/B Vaste Test Tanks-3PGS-TKlA/B Primary Grand Vater Storage Tanks t 3BRS-TK2A/B Boron Test Tanks 3BRS-TKlA/B Boron Recovery Tanks 3CNS-TKl- Condensate Storage Tank 3CNS-TK2 Condensate Surge Tank- .
3VTS-TK6 Vaste Treating Storage Tank 3FVA*TK1 Demineralized Vater Storage Tank Reason for Change ,
Initial design of the tank heater control circuits did not provide i annunciation for circuit loss of power. Heater loss of power could i 4 result in freezing of tank contents and serious tank damage. i l
Safety Evaluation This change increased dependability and availability of the affected )
tanks. The viring changes were all made within nonsafety-related '
j control cabinets (the heater control for 3FVA*TK1 is designated 9 nonsafety related).
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-003 Plant Design Change Number MP3-86-003 entitled "Diesel Lube Oil Moisture Detector Alarm Setpoint" is complete.
Description of Change The emergency diesel generators' (EDG) lube oil moisture detector alarm setpoint was increased from 0.2 percent to 1 percent of water content.
Reason for Change The change was initiated to prevent spurious lube oil moisture detector alarms from occurring. An analysis of the oil performed after the alarm was activated proved that the water content was well belov 0.2 percent. Due to the increase of suspended solids, but still vithin acceptable limits, the oil dielectric constant changes and thus adversely affects the sensitivity of the moisture detector.
Safety Evaluation The purpose of the alarm is to varn of a jacket vater system to oil system leak. The detection system is operated only when the e--ine is .
shut down, and then only during periodic surveillance checks. ether-vise, the detector oil pump is stopped and the detection system is isolated from the main oil header. The manufacturer (Colt Industries) is in agreement with the setpoint change. In addition to vater leaks ,
that are alarmed by the moisture detector, vater leaks vill also be detected from oil samples that are periodically sent to independent ,
laboratories for analysis, and from the observation of water in a lube t oil reservoir during engine operation. t Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-008 Plant Design Change Number MP3-86-008 entitled "Modify Auxiliary Turbine Oil Pump To Lover Resonance Frequency" is complete.
Description of Chan,ge The auxiliary turbine oil pump motor mount was modified to reduce the stiffness and lover the resonance frequency of the pump.
The lifting lugs located on the motor mount, which vere determined to be the cause of the frequency problem vere cut into with a hand held -
saw to reduce the stiffness of the pump casing.
Reason for Change The auxiliary turbine oil pump, 3TML-P2 vas experiencing a resonant vibration problem at 55 to 60 hz. This was causing a dynamic amplf-fication of the vi'uration levels in the direction perpendicular to the discharge flov (north-south direction). A "Bump Test" vas performed
'.n the north-south direction and was unacceptable in that the critical frequency was approximately equal to the rotational frequency.
Safety Evaluation The auxiliary oil pump is nonsafety related and nonseismic. Failure '
of the pump affects no safety equipment. dnder a "single failure,"
the auxiliary oil pump would cause the steam generator feed pump to shut down and the motor driven steam generator feed pump would start.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-013 Plant Design Change Number HP3-86-013 entitled "Sample Valves To 3CCP-RE31 Not Cabled (UNS 6993)" is complete.
Description of Change This change installed a svitch panel near radiation monitor 3CCP-RE31 to control 3CCP-A0V31A, B, C, and D. These valves determine which train of reactor plant component cooling (CCP) is to be sampled by the radiation monitor.
The switch contact arrangement precludes sampling from both CCP trains simultaneously.
Reason zor Change This change was implemented to allow operators to select which train of CCP vill be sampled by the radiation monitor.
Safety Evaluation This change provided a manual control station neer the radiation monitor skid which consists of a switch and two train indicator lights. This installation does not alter any information for either the CCP or radiation monitoring systems contained in the Final Safety Analysis Report Chapters 9.2 and 11.5, respectively.
A power interruption to the valve control circuitry vill cause the valves to fa.il closed, resulting in a lov flow condition in the radiation monitor. This vill be annunciated in the Control Room.
Since only the inlet valves are equipped with position indication switches, the switches have been interlocked with the opposite train indicating lights such that the light vill not energize until its respective inlet valve is fully open, and the opposite train inlet valve is fully closed. Both lights energized or de- energized is indicative of a failure.
As long as one sample valve is open, the ability to sample is not impaired. If the opposite outlet valve is open to the train being sampled, the sample in not af fected, but the water vill be directed to the opposite side of the CCP surge tank. However, this situation is of minor concern since th? surge tank is a baffled tank where the vater is permitted to mix.
Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
Plant Design Change Number MP3-86-014 Plant Design Change Number MP3-86-014 entitled "Sample Pump Heat Trace Circuit Interlock for Containment Hydrogen Monitoring" is complete.
Description of Change This change modified the control circuit of the sample pump heat trace in order to remove the power to the heat-trace circuit when the sample pump is not in service.
Reason for Change The supplier of the containment hydrogen monitoring system recommended the change, so that the heat trace circuit would not be required to operate continuously.
Safety Evaluation This change did not impact system operability for the containment hydrogen monitoring system as described in the FSAR, Section 9.3.2.6.
The failure of the heat trace circuit vill not cause the loss of function of the sample-pump, or any other safety-related equipment.
i Therefore, this modification does not constitute an unreviewed safety question'per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-030 Plant Design Change Number MP3-86-030 entitled "Temporary Fill Connection for DVST" is complete.
Description of Change The Demineralized Vater Storage Tank (DVST) loop seal was modified to accommodate the installation of a temporary fill connection for demin-eralized, deaerated water to the tank directly from an external source. The installation was temporary and was used only while damage to the condensate Storage Tank (CST) vas being repaired.
Reason for Change Due to damage to the CST during plant startup, an alternate source of makeup water to the DVST vas required in order to support continuation of plant startup.
!3 fety Evaluation The loop seal and all modifications to it vere nonsafety-related, class 4 piping changes. Also, as the design (CST) makeup source was not safety related, the alternate source was not safety related either. Since an overflow line was used as a fill point, an alternate overflow line was used by virtue of using another unused line to the DVST. After the CST vas repaired, all systems that were modified by this change vere returned to their normal design configuration.
Therefore, this modification did not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-034 ,
Plant Design Change Number MP3-86-034 entitled "Provide Heat Trace To The Condensate Storage And Surge Tank Relief Valves" is complete.
Description of Change This change provided heat tracing for both the breather relief valves,
- 3CNS-RV48, 3CNS-RV49, and the goose neck vents, located on top of the
, condensate storage (3CNS-TKl) and surge (3CNS-TK2) tanks.
Reason for Change Adding heat trace to the breather valves vill prevent the valves from becoming inoperable during cold weather, due to the formation of ice around the valve seat; thus preventing the tanks from exceeding their design pressure or vacuum ratings. Also, adding heat trace to the goose neck vents on both tanks vill ensure the operability of the rupture dises, installed under change number MP3-86-057.
Safety Evaluation The heat trace circuits provided are supplied from non-Class lE power sources and are mounted on nonsafety-related components. The circuits are run such that they do not pass through any Category 1 fire areas.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number HP3-86-039 Plant Design Change Number MP3-86-039 entitled "Lead Shielding for Solid / Liquid Radvaste Disposal Lines" is complete.
Description of Change Lead shielding was added to piping in both Solid and Liquid Radvaste Systems. Stainless steel banding was installed to hold the lead sections in place, and supports were added or modified as necessary to accommodate the additional veight.
Reason for Change Changes to the method of radvaste disposal requires the presence of orarating personnel in areas not accounted for in the original design.
Radwaste piping within these areas required the installation of lead in order to reduce radiation levels to as lov as reasonably achievable (ALARA).
Safety Evaluation The installed lead complied with the material / chemical specifications of ASTM B29. The lead vas installed in two half-sections contoured to fit over the outside diameter of the pipe with no gaps. Supports we';e added to the piping to accommodate a total veight of 190 pounds. The lead cas added to a nonsafety-related system installed in a nor. safety-related building.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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l Plant Dtsign Change Number MP3-86-042 4
Plant Design Change Number MP3-86-042 entitled "Reroute 3HVH-SCL1 Outlet Cooling Vater" is complete.
Description of Change This change rerouted the Turbine Plant Component Cooling Vater (CCS) outlet from 3HVH-SCL1 from the Turbine Building floor drains back into the CCS system. ,
Reason for Change The Turbine Building floor drains are ultimately pumped to the discharge canal and out into Long Island Sound.~ Since-the CCS system contains hydrazine, it cannot be discharged to the Sound.
Safety Evaluation The installation of the new piping was done in accordance with the original design and installation specifications. The installation was performed on c Class 4, nonsafety-related system located in a
, nonsafety-related area.
I Therefore, this modification does.not constitute an unreviewed safety question per the criteria of 10CFR50.59. i i
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J Plant Design Change Number MP3-86-047 Plant Design Change Number MP3-86-047 entitled "3BRS-RV95 Cap Leaks (DDR 960)" is complete.
Description of Change This change replaced the existing "plain lift" cap of 3BRS-RV95 vith a "screwed" cap.
Reason for Change The discharge of the subject relief valve is connected to a recircu-lation line back to the Boron Recovery Evaporator. The backpressure on this line was allowing contaminated fluid to leak out of the "plain lift" cap. This change eliminated this leakage.
Safety Evaluation The subject relief valve protects the Boron Evaporator Pottoms Cooler from inadvertent overpressurization. Neither the relief valve, nor ,
the bottoms cooler perform a safety-related function. This cap change !
vould not cause the relief valve to fail to operate.
Therefore, this modification does not constitute an unreviewed safety -
question per the criteria of 10CFR50.59.
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I Plant Design Change Number MP3-86-048 Plant Design Change Number MP3-86-048 entitled "Main Turbine - Stop Loading Blocking Signals" is complete.
l Description of Change The main turbine Electro Hydraulic Control (EHC) system turbine stop loading input signals were modified.
Reason for Change This modification was done to preclude main turbine EHC load hold system lockup. This lockup occurred whenever remote stop loading commands were received during the thirty second time delay between a turbine trip and a main generator trip.
l Safety Evaluation This. safety evaluation addressed the addition of inhibit signals to the main turbine stop loading P/C card located in the Turbine Supervi-sory Instrument (TSI) section of the EHC cabinet as described below:
The main turbine stop loading feature as installed vill lockup whenever external stop loading command to the EHC system occurs during a 30 second time delay between a turbine trip and a main ,
generator breaker trip.
l All electrical work was completed in accordance with the appro-priate electrical installation specif;ication.
A stop loading system lockup will not adversely affect the safe opera-tion of the turbine control system. It only creates an extra step for the operator, whereby the lockup condition must be cleared internally at the EHC cabinet before a turbine restart can be initiated. The i addition of the load hold inhibit vill block stop loading signals in the EHC system following a turbine trip. A failure of the load hold feature inhibit feature vould result only in a stop loading system lockup.
The proposed change does not impact the turbine control system operability as described in the Final Safety Analysis Report (FSAR) i Section 10.2.2.1. Additionally, it does not alter or change any j physical equipment other than the deletion and addition of a small ,
amount of vire inside the main turbine EHC cabinet. l l
Therefore, this modification does not constitute an unrevieved safety i question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-051 lj Plant Design Change Number MP3-86-051 entitled "Revisions to the Boric ;
Acid and Total Makeup _ Flow Deviation Circuitry" is complete.
Description of Change The. delays associated with boric acid flow deviation and total makeup flow deviation alarms were increased:from 15 seconds to 30 seconds.
In addition, the circuitry was modified to keep the boric acid makeup injection and boric acid dilution injection' valves closed af ter - -
receipt of a flow deviation alarm until the operator resets the circuitry. 4 Reason for Change ,
The boric acid injection valve cannot move quickly enough to adjust ,
for a borate demand signal without also causing a boric-acid high flov -6 deviation alarm that is set for 15 seconds. A 30 second delay time allows the valve adequate time to adjust flow deviations so the alarm !
von't occur and cause-inconvenient closure of makeup and dilution ,
valves. The circuitry modification provides additional stability by l preventing makeup and dilution valve oscillations that vould otherwise '
occur when a flow deviation alarm signal-is received and then cleared by closure of the boric acid makeup injection valve. i Safety Evaluation
- The increase in the time delay from 15 to 30 seconds results in an ;
insignificant additional error for boric acid concentration. The t
circuitry modification involved changes to the nonnuclear safety-related portion of the control circuits. Furthermore, the modifi-cation prevents unnecessary and inconvenient cycling of makeup valves.
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Therefore, this-modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. :
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Plant Design Change Number MP3-86-053 Plant Design Change Number MP3-86-053 entitled "Containment Air Filtration Heater Flow Svitch" is complete.
Description of Change This change deleted two flow stitches from the Containment Building filtration system heater control circuitry.
Reason for Change The filter bank heaters vere cycling on and off during filter train operation. Instead, they should remain continuously energized during filter train operation to enstire maintenance of desirable relative humidity conditions of the entering air st.:eam. The deletion of the switches allows the heaters to operate p operly.
Safety Evaluation The original system was equipped with overtemptrature protection for the heaters, along with lov flov switches. After the deletion of the flov svitches, the system still conforms to the National Electric Code Requirement for Duct Heaters because of the overtemperature protec-tion.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
Plant Design Change Number HP3-86-053 Plant Design Change Number HP3-86-Ca5 entitled "Cable Separation Barrier at 3TH413N" is complete.
Description of Change This design change installed a sheet metal separation barrier just belov cable tray 3TH413N to provide electrical separation between Class IE and Non-Class IE cables.
Reason for Change The change vas installed because significant cable congestion in the area rendered separation by way of cable vrapping unfeasible.
Safety Evaluation The installation of the cable separation barrier meets the require-ments of the Final Safety Analysis (FSAR) for vertical separation between Non-Class IE cables in tray and Class IE cables in air. The installation meets all seismic qualifications and has been reviewed for postulated seismic interactions imposed on ti.e existing seismic cable tray system.
Therefore, th:s modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i Plant Design Change Number MP3-86-057 Plant Design Change Number MP3-86-057 entitled "Additional i Overpressure Prote: tion For 3CNS-TK1 And TK2" is complete. I Description of Change This change modified the condensate storage tank (3CNS-TKl) and condensate surge tank (3CNS-TK2) to ensure greater overpressure protection. The change included the installation of rupture discs, pressure control valves, closed vent breather valves, and weather enclosures on the roof of the tanks.
Reason for Change Past experience has proven that the overpressure protection initially installed was inadequate. The tanks were provided with only a single means of tank venting, which was through a breather relief valve.
These valves have in the past frozen closed leading to severe tank damage. The modifications made under this design change vill ensure greater protection against over and underpressurization under all veather conditions.
Safety Evaluation All changes made vere to nonsafety-related systems, components, and structures.
Therefore, this modification does not constitute an unreviewed safety question per the critetla of 10CFR50.59.
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l Plant Design Change Number MP3-86-058 l Plant Design Change Number MP3 86-058 entitled "Boron Evaporator Feed ,
Pump NPSH Modification" is complete.
Description of Change A time delay relay was added to the trip circuit of each Boron Evaporator Feed Pump that blocked a lov suction pressure trip for five seconds after a start.
Reason for Change ,
Due to pump and piping system configuration, the pressure sensing switches opened on pump starts and caused unnecessary trips. The
- addition of a time delay relay to each pump trip circuit prevents the initial lov suction pressure spikes from tripping the pumps.
Safety Evaluation Should the pump experience a lov NPSH condition for greater than five seconds after initial starting, the pump vill be tripped automatically
- and must be restarted manually. A postulated failure of the timing relays added into the control circuits vould result in the pump not i starting, or tripping the pump if the pump was already started. The pump and the associated system are nonsafety related.
I Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. E e
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Plant Design Change Number MP3-86-060 Plant Design Change Number MP3-86-060 entitled "liot Vater Heating High Energy Line Break Isolhtion" is complete.
Description of Change This change removed high energy line break pressure switches from hot vater preheating lines in the Auxiliary Building, and replaced high energy line break pressure switches on hot water heating lines in the Auxiliary Building.
Reason for Change The Hot Vater Preheating System is not a high energy system. Pressure switches, therefore, are not needed for isolation in the event of a line break. Higher setpoints vere needed for hot water heating pressure switches to allov isolation of lines in the Auxiliary and Fuel Buildings. Pressure switches with an increased range were required to meet the higher setpoints.
Safety Evaluation Installation of replacement pressure switches in the Hot Vater Heating System does not alter the original design function of the pressure switches. The increased range of the replacement pressure svitches enables them to perform their intended high energy line break function. ,
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-064 Plant Design Change Number MP3-86-064 entitled "Addition of an Alarm for Vacuum Breaker Valves Auxiliary Circuit (DDR-491)" is complete.
Description of Change This design change installed an auxiliary relay in the Main Control Board Section 1 to monitor the loss of control pover of the water box vacuum breaker valves auxiliary circuit and provided an annunciator vindov at the Main Control Board.
Reason for Change The change enabled monitoring of the condenser vater box vacuum breaker valve control circuit for loss of control pover. Loss of control power to this circuit vould cause inoperability of the con-denser vater box vacuum breaker valves, which could contribute to circulating vater tunnel damage.
Safety Evaluation The change did not affect the operability of the condenser water box vacuum breaker valve auxiliary circuit. The change did not involve any safety-related system. The new cable installed in the Main Control Board satisfies design requirements for cable installation, including seismic support and physical separation.
Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-066 Plant Design Change Number HP3-86-066 entitled "Pressure Loop 508 Modification for Vestinghouse Interface" is complete.
Description of Change This change added a type NSA card to the Vcstinghouse 7300 system process control system cabinet for gain adjustment on the turbine driven feedvater pump speed control circuitry.
Reason for Change A previouc design change respanned a feedvater pressure indicator, but did not take into account that the transmitter feeding the . indicator ,
also provided input for the feedvater pump speed controller. The existing hardware could not condition the signal properly, so a new type NSA card was added to provide the required gain adjustment for the feedvater pump speed controller.
Safety Evaluation A postulated failure of the NSA card could cause the turbine driven pump speed to be driven high or lov, depending on the type and mode of a fsilure. A change, whether an increase or decrease, in the pump speed is bounded by transients that have already been analyzed for the Final Safety Analysis Report. This portion of the feedvater system is non- I safety related, and is not required for a safe shutdown. No other safety systems are affected by this change. The addition of this card i into the control circuitry improves the performance of the circuit and I does not affect the design basis of the feedvater system. ,
- Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-070 Plant Design Change Number HP3-86-070 entitled "Radioactive Liquid Vaste System Pumps' Circuit Hodifications" is complete.
Description of Change A time delay relay was added to the trip circuit of several Radio-active Liquid Vaste Pumps that blocked a lov suction pressure trip for five seconds after a start.
Reason for Change Due to pump and piping system configuration, the pressure sensing switches opened on pump starts and caused unnecessary trips. The add-ition of a time delay relay to each pump trip circuit prevents the initial lov suction pressure spikes from tripping the pumps.
Safety Evaluation Should a pump experience a lov NPSH condition for greater than five seconds after initial starting, the pump vill be tripped automatically and must be restarted manually. A postulated failure of the timing relays added into the control circuits vould result in the pump not starting, or tripping the pump if the pump was already started. The pump and the associated system are nonsafety related.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-073 Plant Design Cnange Number MP3-86-073 entitled "Revision of Reactor 1
Containment Lighting Controls" is complete.
Description of Change Prior to this change, the master control for all Reactor Containment lighting removed power from all lighting panels. This modification removed the vital lighting panels from the master control.
l Reason for Change Deenergizing all lighting power in the normally unoccupied Reactor Containment caused the battery povered emergency lights to energize and discharge their batteries. The emergency lights are charged from the vital lighting panels, which are no longer deenergized when the master lighting switch for the containment is operated.
Safety Evaluation The modified lighting system is not safety related and there are no design basis accidents affected. There are no manual actions required within the Reactor Containment to mitigate any design basis accidents.
This modification did not affect the function of the Emergency Lighting System. ,
Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFh50.59.
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Plant Design Change Number MP3-86-074 Plant Design Change Number HP3-86-074 entitled "Deletion of HVAC Fire Dampers (3HVR*DNPF4,5; 3HVQ-DHPF1021)" is complete.
Description of Change the change deleted two fire dampers located in ventilation duct vork within common fire areas.
Reason for Change Deletion of the two fire dampers was undertaken since they are installed in duct vork penetrating slabs within common fire areas.
The affected slab areas are not fire barriers but within common fire areas an documented by the Fire Protection Evaluation Report.
Safety Evaluation
. Deletion of the subject fire dampers vill not result in degradation of plant fire protection. Subject duct vork slab prctections vere out-fitted with fire dampers on the incorrect premise that these locations constituted fire barriers for separate fire areas. The Fire Protec-tion Evaluation Report assumed no credit for subject fire dampers.
Therefore, this modification does not constitute an unruviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number HP3-86-075 entitled "Installation Details for Generator Leads Cooling Fans Circuit" is complete.
Description of Change This change installed a time delay relay and associated viring into the control circuits for main generator leads cooling fans 3GHS-FNIA and 18.
Reason for Change Prior to implementing this change, the main generater leads cooling fans vould receive a ' start' signal at the same time its associated inlet damper received an 'open' signal. The fan vould come up to speed before the damper had a chance to fully open, causing a high differential pressure across the damper, preventing it from opening fully. The subject change added a time delay to allov the damper to go full open before starting its fan.
Safety Evaluation This change did not affect, alter, or modify any Category 1 system, component, or structure.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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1 Plant Design Change Number MP3-86-076 Plant Design Change Number MP3-86-076 entitled "Modifications To The Steam Generator Blovdown Sampling System" is. complete. .
-Description of Change Components of the steam generator blovdown sampling system vere replaced with equipment which have' greater flow capacities. Com-ponents replaced vere radiation monitor sample flov indicators and solenoid isolation valves, the blevdovn sample reclaim pump and the i pump suction line strainer and isolation valves. The inlet and outlet tubing of the reclaim pump was also replaced with 3/4 inch pipe.
Reason for Change l
The original steam generator blovdovn radiation monitor sample flow path had excessive head loss and did not provide a sufficient sample flow rate. The new flov indicators and solenoid valves substantially reduce the head loss and, therefore, allow adequate sample flov.
The original blovdown sample reclaim pump vas undersized causing the ,
sample reclaim tank to overflow. The new pump and associated ,
strainer, valves, and piping alleviated the overflow problem and accommodated the increased sample flov to the tank.
Safety Evaluation I None of the equipment involved in this change is safety related. No new types of equipment vere added by the change, just larger models of the original components. The operation and configuration of all of the compenents remained the same as that of the original installation.
Potential line breaks in the piping / tubing that was modified would be ,
mitigated by existing containment isolation valves. ;
L Therefore, this r.odification does not constitute an unreviewed safety l question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-079 Plant Design Change Number MP3-86-079 entitled "Recircuiting of Emergency Battery Povered Lighting Packs in Containment" is complete.
Description of Change This change allowed all lighting to be turned off in the normally unoccupied Reactor Containment during power operation.
Reason for Change The design of the battery backed emergency lighting system was such that whenever the lighting in the Reactor Containment was shut off for power operation, the emergency lighting vould energize and the battery packs discharge. This change provided power from independent circuits at the vital lighting panels such that the vital lights could be turned off, but the emergency lights vould remain on charge. The emergency lighting vill still energize upon the loss of the vital lighting panel.
Safety Evaluation The modified lighting system is not safety related and there are no design basis accidents affected. There are no manual actions required within the Reactor Containment to mitigate any design basis accidents.
This modification did not affect the function of the Emergency Lighting System.
Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-086 entitled "Installation of a I i
Resistor in the Panel 3HVC-PNLO1A/B (DDR-890)" is complete.
Description of Change ;
- j. This change installed a resistor in the Control Building amergency filter temperature control panels.
Reason for Change This change was installed to provide a continuous supervisory circuit and eliminate the unnecessary trouble alarm condition of the panels. j Safety Evaluation :
4 This change did not affect the operability of the panels. The modi-fication did not affect any safety-related systems, nor did it affect ,
j seismic or electrica7. separation requirements.
- Therefore, this modification does not constitute an unreviewed s'afety ?
question per the criteria of 10CFR50.59. l f
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Plant Design Change Number MP3-86-088 Plant Design Change Number K.'3-86-088 entitled "CCS Heat Exchanger Protective Coating" is complete.
Description of Change This design change involved the application of an epoxy coating on the Turbine Plant Component Cooling (CCS) heat exchanger inlet tube sheets and inlet tube ends.
Reason for Change This change was required to inhibit erosion of the inlet tube sheets and inlet tube ends of the CCS heat exchanger.
Safety Evaluation This change was a preventative maintenance activity which does not impact operability of the heat exchanger. If the epoxy coating should lose adherence, it vould pass through the service water system. The change has no impact on system performance.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. j l
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1 Plant Design Change Number MP3-86-105 Plant Design Change Number MP3-86-105 entitled "Deletion of Pressure Indicators 3HVC-PI201A and 201B From the Control Room Envelope Pressurization (Bottled Air) System" is complete.
Description of Change This design change deleted pressure indicators 3HVC-PI201A and 201B located in the pressure control valve manual bypass line.
Reason for Change The subject pressure indicators verc 2eing overranged during passive system conditions. The condition was attributed to minor leakage through the normally closed manua?. isolation valves to the system supply header, which is maintained at pressures greater than 2200 psig. The pressure indicators could not serve their purpose nor vere they necessary for system operability.-
Safety Evaluation The Final Safety Analysis Report did not take credit for the manual discharge capability of which these gauges are a part. Control Room Envelope Pressurization System acceptability is predicated on the existence of totally redundant systems with automatic discharge following a Control Building isolation signal. Removal of the gauges did not affect system operability nor did it compromise any safety system.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-114 Plant Design Change Number MP3-86-114 entitled "Feedvater Pump Seal Vater Piping Modification" is complete.
Description of Change This change modified the seal vater piping to the main feedvater pumps to allow the installation, at a later time and under a later plant design change, of two booster pumps to increase the pressure in the seal vater line.
Reason for Change The original design of the feedvater seal vater system assumed that there was sufficient differential pressure between the condensate pump discharge and the teedvater pump suction to provide sufficient flov to the seals. Operational experience indicated that sufficient differ- _
ential pressure existed only at power level above 70 percent, and the seals vere failing at a high rate due to lov seal vater flow.
Safety Evaluation This change had no effect on the operation of any system, since it was only a pipir.g modification to prepare for a later installation of equipment. All the valves that vere added vere left closed, and all pipe ends vere capped to prevent any inadvertent flow path changes.
The flow path of the feedvater seal vater did not change as a result of this change.
Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-122 Plant Design Change Number MP3-86-122 entitled "Tube Union Addition (CP396724)" is complete.
Description of Change A tube union was added to a Reactor Plant Sampling line.
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Reason for Change A process leak required repair on this Reactor Plant Sampling line.
The change was required to facilitate rework.
Safety Evaluation This change was installed in accordance with existing criteria for Category 1/ Group A tube installations. The change meets the original design criteria and stress requirements of the installation specifi- -
cation.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-125 Plant Design Change Number MP3-86-125 entitled "Separation Barrier on-Tray Support C058C" is complete.
Description of Change This design change installed a sheet steel electrical separation barrier between Class IE and non-Class IE cables in the Control :
Building. I Reason for Change l The change provided the necessary electrical separation barrier between Class IE and non-Class IE cables required to meet the separation criteria outlined in the Electrical Installation Specification.
Safety Evaluation The change meets the requirements of the separation criteria outlined in the Electrical Installation Specification. The change meets seismic specifications and maintains the structural integrity of the tray support system.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
Plant Design Change Number MP3-86-128 Plant Design Change Number MP3-86-128 entitled "Installation of VAR '
Transducer" is complete.
Description of Change This change entailed the installation of a MVAR transducer and asso-ciated viring/ cables to monitcr Hillstone Unit 3's gross megavar ;
output and automatically transmit gross MVAR data to Connecticut Valley Electric Exchange (CONVEX).
Reason for Change
?0NVEX required MVAR output data to determine the total VAR output of the Millstone Unit 3 generator. Initial plant design had no pro-visions to automatically transmit this data, so plant operators fre- '
quently phoned CONVEX personnel. This change utilizes the site svitchyard Supervisory Control and Data Acquisition (SCADA) system and
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telemetering circuits to transmit the data.
Safety Evaluation This change did not affect the operability or availability of any Category 1 equipment or structures. Viring modifications were made to nonsafety-related circuits only. The transducer was mounted and viring trained in such a manner as to pose no threat to any safety-related equipment.
Therefore, this modification does not constitute an unrevieved safety :
question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-138 entitled "Replacement of Cables '{
in Chlorine Pit of Intake Structure" is complete. -
Description of Change j- l This change replaced all electrical ~ cables that terminate belov - !
j-elevation 14'-6" in the chlorine pit. In some cases, the damaged i cable was repaired by splicing a new cable in place of the damaged i part.' In other cases, the whole cable was replaced.
To accomplish this change, racevay tickets, cable pull tickets, and.
j viring diagrams were issued-to provide adequate information to {
4 respectively delete, spare,'and determ and reterm existing raceways j and cables.
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Any barriers in which new penetrations are installed were resealed'to _- '!
maintain the required integrity of the barrier.
Reason for Change i Due to c51orine pit flooding in the past, corrosion on wires and lugs I in limit switches left the integrity of lugs and termination indeter-minate. ,
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Safety Evaluation The rerouting of the cables does not jeopardize the intended function i
of the original design. The system operability is not impacted in any '
manner by this change. The new cables meet licensing requirements ;
concerning electrical separation, Appendix R, seismic, and hazards
, analysis.
j Therefore, this modification does not constitute an unreviewed safety
] question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-153 Plant Design Change Number MP3-86-153 entitled "3SVP*MOV102A ,0 Replacement" is complete.
Description of Change The Service Vater Pump discharge valves (3SVP*MOV102A-D) vere replaced with aluminum-bronze butterfly valves.
Reason for Change The previous design was subject to extreme corrosion damage when the rubber lining covering carbon steel separated. Corrosion from sea vater reduced the valve to near minimum vall thickness.
Safety Evaluation The flow rate through present valves is slightly lover than the
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previous valves. This, in turn, reduces service vater cooling flov to the containment recirculation spray coolers (6190 gpm to 6178 gpm).
This decrease does not affect original design basis of the system or any associated safety-related system.
Seismic design is not affected by the new valves. Valve corrosion resistance to the sea vater environment is superior to the replaced valves. The structural integrity of the valves remain the same.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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1 Plant Design Change Number MP3-86-155 Plant Design Change Number MP3-86-155 entitled "Circuit Additions to the Chemistry and Health Physics Laboratories" is complete.
Description of Change 120-Volt circuits and receptacles vere added to the Chemistry and
- Health Physics laboratories.
Reason for Change The new circuits and rec 9ptacles vere added to accommodate the Nuclear Data Acquisition and Analysis System Equipment. Excessive voltage drops vould have occurred if presently available outlets vere used.
Safety Evaluation The installation involved only nonsafety-related equipment and receptacles in a nonsafety-related building. Equipment and cable that vere installed in the safety-related Control Building met seismic requirements. Any new penetrations through a fire, pressure, radia-tion, C09, or tornado barrier vere resealed as required to maintain the barrier integrity.
Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Plant Design Change Number NP3-86-161 Plant Design Change Number MP3-86-161 entitled "Circulating Vater Pump control Circuit Time Delays" is couplete.
Description of Change This design change installed time delays for the circulating vater
! pump low lube water pressure and traveling screen high differential pressure trip signals.
Reason for Change This design change was implemented to prevent a trip of a circulating 3
vater pump (s) (and a potential subsequent turbine trip) due to momen-tary fluctuations in lube water pressure or screen differential ,
pressure.
Safety Evaluation This change to the circulating vater pump trip circuits is consistent with original design and has no impact on safety-related systems.
Therefore, this modification does not constitute an unreviewed safety l question per the criteria of 10CFR50.59,
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Plant Design Change Number MP3-86-202 Plant Design Change Number MP3-86-202 entitled "Permanent Platform to Receive and Inspect New Fuel" is complete.
Description of Change A permanent platform and ladder was constructed at the 37' elevation in the Fuel Building. The platform was attached to the north vall of the Spent Fuel Pool and the east vall of the Spent Fuel Pool filter cubicle. The ladder was attached to the north vall of the Spent Fuel Pool filter cubicle. In addition Service Air line, 3-SAS-750-187-4, was partially rerouted to facilitate placement of the platform.
Reason for Change During the initial fuel receipt and inspection process, temporary scaffolding was erected. The use of this temporary scaffolding was identified as a safety hazard by the Station Safety Administrator.
l Safety Evaluation The platform and its connections have been designed to resist seismic loading for Category 2 over 1 reasons. Peak resonant accelerations, from elevation 43 feet of the Fuel Building, have been applied to the structure. All members and connections have been designed to resist these seismic forces. The platform is designed for a live load of 125 pounds per square foot. All design stresses are within normal allow-able stress levels.
These loads are transferred to the north vall of the Spent Fuel Pool and the east vall of the spent Fuel Pool filter cubicles by surface mounted base plates. The forces transferred to these valls are quite l
small in relation to the capacity of the valls. Based on the size of l load being added compared to the capacity of the valls, no further analysis is required and the structures are not impacted by the addi- 1 tion of the platform.
The service air line, 3/4 inch diameter, Class 4, nonsafety related, i vas initially field run. A seismic qualification / hazards reviev was i performed, after the line vas relocated, which ensured that there is no negative impact on safety-related structures, systems, or com-ponents.
l Therefore, this modification does not constitute an unreviewed safety l
question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-223 Plant Design Change Number MP3-86-223 entitled "Elgar Inverter -
Replacement of Solid Bus Bar" is complete.
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Description of Change The solid copper bus bar between the inverter's internal fuse block and the shunt resistor on the filter panel was replaced with a 1
flexible braid. This connection carries direct current from the batteries to the inverter in the event that normal AC power is lost.
Reason for Change The vendor (Elgar) had determined that the previous design could have I
resulted in failure of the vital bus due to the bus bar being torqued against the plastic fuse block.
Safety Evaluation The modification did not affect the structural response of the cabinet or anchorage of the internal components, therefore, this modification did not affect the seismic qualification of the cabinet. This change
< did not change any of the operating parameters of the inverter, nor vere any new failure modes created.
Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-261 Plant Design Change Number MP3-86-261 entitled "Pover Feed to Gate Operator and Control Station" is complete.
Description of Change This change provided installation and viring details for cables that feed power to the gate operator in Varehouse 5.
Reason for Change This change was installed to bring the gate into conformance with its origina) design intent.
Safety Evaluation The change involved installation of nonsafety-related cables and viring them to provide power and control at the Varehouse 5 gate operator. This change met all licensing requirements for separation.
Appendix R, seismic, and hazard analysis. This change vas done within the confines of a nonsafety-related building Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i Plant Design Change Number MP3-86-265 Plant Design Change Number MP3-86-265 entitled "Hain Steam Safety Relief Valves Drain Line Modification" is complete.
Description of Change Unions were cdded to the folloving SVV drain lines to the main steam safety relief valves: 3 MSS *RV22A-D, 23A-D, 24A-D, 25A-D, and 26A-D.
Reason for Change This modification allowed for the servicing of the relief valves and their associated expansion joints for the main steam safety system.
This was not possible in the previous drain line piping arrangement.
By adding unions, it precludes the need for cutting and velding of the drain lines previously required each time maintenance on the SVV drain lines was to be performed.
Safety Evaluation The safety evaluation addressed the addition of unions to the SVV drain lines to the main steam safety relief valves, 3 MSS *RV22A-D, 23A-D, 24A-D, 25A-D, and 26A-D.
The SVV system is a Class 4, nonsafety-related system. This change vill not restrict the main steam safety valves from performing their intended function nor doea it compromise the functionability or reli-ability of the main steam system.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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h Plant Design Change Number MP3-86-266 Plant Design Change Number MP3-86-266 entitled "Main Steam Pressure l Relieving Valves Open Belov Setpoint (UNS 7579)" is complete.
Description of Change ,
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This modification revised the settings of 3 MSS-PIC 20A1, A2, B1, B2, C1, C2, D1, and D2 so that they will not cause the main-steam pressure
. relieving valves-to open during normal operating transients. Also, operating procedures vere revised to require the controllers to be svitched into the manual operating mode prior to any setpoint '
adjustments being made.
The main steam pressure relieving valve controller's reset and Hi/Lo i 1 limit settings vere changed to reduce the controller's sensitivity to :
pressure transients. ,
t Reason for Change 1 The potential existed for the main steam pressure relieving valves to open prior to reaching their relief setpoint. This was experienced t during start-up testing and is unacceptable during normal operating transients.
i Safety Evaluation i The main steam pressure relieving valves are safety-related so as to !
maintain the safety class 2 pressure boundary of the pipe they are ;
installed in (Reference Final Safety Analysis Report (FSAR) i Section 10.31.1). They perform no safety function and are not required for over pressure protection of the main steam system (provided by safety valves).
I Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. j I
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Plant Design Change Number MP3-86-272 ;
Plant Design Change Number MP3 86-272 entitled "Normal And Reserve >
l Station Service Time Of Day Heter" is complete.
Description of Change i This change installed equipment required to monitor and record Hillstone Unit 3's total electrical power usage during plant outages.
Circuitry including isolation relays, auxiliary relays, a totalizer, and time-of-use demand recording meter was installed and connected into the existing vatt-hour circuits for.the normal and reserve station transformers. The monitoring circuit is automatically put into operation when the main generator field circuit breaker is opened (generator off-line).
Reason for Change Initial design of the plant had no provisions for automatically and accurately recording plant "off-line" electrical power usage. Since the plant is not owned by Northeast Utilities alone, the cost for this power must be shared by all co-owners. This change provides a means of accurately monitoring and recording this data for billing purposes.
Safety Evaluation The metering equipment is nonsafety related. Cables have been run
- through Category 1 and fire protection OA boundaries, with any damaged ,
cable penetrations resealed in accordance with station procedures. i
! The-installation of cables and circuit components meet license '
requirements for separation, seismic, Appendix 'R', and hazard j analysis.
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Therefore, this modification does not constitute an unreviewed safety 4
question per the criteria of 10CFR50.59. l s
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Plant Design Change Number MP3-86-284 Plant Design Change Number MP3-86-284 entitled ' Replacement of Steam Generator Vater Level Recorders" is complete.
Description of Change The four Optimatic 100 recorders in the Control Room for recording of Narrov Range Steam Generator Vater Level vere replaced with Tracor Vestronics Model T4E3 recorders.
Reason for Change Both types of recorders were in use in the plant. While their tech-nical characteristics are virtually identical, the Tracor model had exhibited superior durability and performance.
Safety Evaluation The water level recorders for steam generator vater level do not provide any functions for control or protection. Normal indication is provided by meters on the Main Control Boards. The replacement recorders vere mounted iri seismically designed counts so that they vould not contribute to a seismic event. These meters have no impact on any safety systems or on the protective boundaries.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-311 Plant Design Change Number MP3-86-311 entitled "Reroute MP3 Fire Protection Yard Loop For MP3 Maintenance Shop Expansion" is complete.
Description of Change This change rerouted the existing HP3 Fire Protection Yard Loop, relocated several post indicator valves, and relocated one outside hose house.
Reason for Change This change vas implemented to allov for expansion of the MP3 Maintenance Shop.
Safety Evaluation Due to the extent of this modification, the potential existed during construction for isolating both the primary and secondary means of fire suppression in several areas (i.e., Containment and the Auxiliary Building) which contain redundant safety shutdovn equipment. It should be noted that the function of the Auxiliary Building and Containment vater curtain systems is to provide separation between redundant safety shutdown components during a fire in lieu of the separation requirements of 10CFR50 Appendix R, Section 3G. Detailed installation plans vere developed to minimize the amount of time these systems vere isolated during the construction phase of this modifi-cation. The installation and cross-connect plans vere based on hydraulic calculations and strategies approved by the NRC in Docket Bil760, dated October 1, 1985. The internal cross-connect made between system drain valves in the Auxiliary Building precluded the isolation of either the Containment or the Auxiliary Building stand-pipe and vater curtain systems during construction. This modification relocated existing components and piping only. The design and opera-tion of the system was not changed.
Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59. ;
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Plant Design Change Number MP3-86-312 4
Plant Design Change Number MP3-86-312 entitled "Damaged Page/
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Evacuation Cable 3COPCNX126" is complete. ;
Description of Change
- This design change was incorporated to facilitate repairs on inoperative speaker cable 3COPCNX126. The damaged portion of cable ,
2 vas cut and terminated in new termination box 3JB-3050A and additional, length of cablo vas routed from the new termination box to existing amplifier 3 COP-AMP 126 to complete the circuit. !
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Reason for Change '
l This design change (addition of termination box 3JB-3050A), avoided [
extensive rerouting of cable and reworking fire stops and seals while !
at the same time provided necessary circuit requirements for page/ [
, evacuation signal for amplifier 3 COP-AMP 126. Cable routing conformed i to all licensing requirements for separation and Appendix R.
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Safety Evaluation
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2 The new cable is rated the same as the old, and the cable and ter- I j mination box meet all requirements for separation, Apperdix R, and :
hazards analysis.
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- question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-316 Plant Design Change Number MP3-86-316 entitled "Screenvash Piping Replacement" is complete.
Description of Change ,
All of the original screenvash system piping was replaced with Fiberglas and stainless steel piping, and the six carbon steel j screenvash unit isolation valves were replaced with light-weight :
nonmetallic composite valves. The majority of the replaced piping is l' Fibercast, F-Chem type material with the section of piping at the strainer discharges being type 316L stainless steel. Further, ten inch bypass lines with restricting orifices at both pump discharges and six new pipe supports were installed and the control circuitry of t the pump discharge valves was modified to delay opening of the valves until approximately 75 seconds after pump start. i Reason for Change .
l The original screenvash system piping had extensive through vall failures caused by cyclic vater hammer loadings induced by screenvash ;
pump start-ups. The new piping materials, isolation valves and pipe j supports provide improved strength and rigidity and act to absorb loadings during system transients. The bypass lines and control i circuit modifications minimize cyclic water hammer loadings by allowing the system to fill slowly through the bypass lines before the l 16 inch pump discharge valves open. l l
Safety Evaluation-l The screenvash system is not a safety-related system and the modifica-tions accomplished by this design change do not affect any safety- l related system. The failure of the new piping could potentially l affect safety-related equipment in the service vater pump cubicles; j however, this failure is addressed in the Final Safety Analysis Report !
(FSAR). Specifically, failure of both trains of safety-related equipment is precluded by requiring that at least one of the two doors j to the service water cubicles is secured at all times. In addition, this design change decreases the probability of this occurrence because the modifications provide a more conservative design which is more reliable than that discussed in the FSAR.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-367 Plant Design Change Number MP3-86-367 entitled "Added Insulation In Main Steam Valve Building For Heat Load And Personnel Protection" is complete.
Description of Change The modification provided for the insulation of the steam drain, vent and valve irstrument root lines, associated valves and pipe supports in the 59 foot elevation of the Main Steam Valve Building.
Reason for Change Reduction of the heat load of the Main Steam Valve Building was necessary in an effort to maintain temperatures belov environmental qualification limits, and to provide personnel protection from hot obj ec ts.
Safety Evaluation The addition of the insulation in the Main Steam Valve Building does the following:
Reduces the heat load in the building - reducing the heat load ;
results in a lover ambient temperature, thus a change in the conservative direction with regards to Area Temperature Limitations in Technical Specifications. The only lov temperature constraint is pipe freezing. This problem is currently being addressed in another plant design change.
Causes a temperature increase on the items insulated - these temperature increases result in final temperatures that are still at or belov design temperatures.
Presents added veight to the insulated items - the additional veight to each item has been evaluated for the additional stress applied and determined not to be a concern. i i
Items that are required to remain uninsulated for proper operation ;
vere specifically left uninsulated and mentioned in the design i documents. Insulation of associated turbine plant drain lines causes a reduction in condensate supplied to certain turbine drain trapa which results in an increase in overall plant efficiency.
Therefore, this modification does not constitute an unrevieved safety
- question per the criteria of 10CFR50.59.
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Plant Design Change Number HP3-86-379 Plant Design Change Number MP3-86-379 entitled "Replacement of the Valve Positioners on Feedvater Regulating Bypass Valves" is complete.
Description of Change The existing Fisher valve positioners on the feedvater regulating
- bypass valves vere replaced with Hasonellan valve positioners, and volume boosters vere added to the air supply to the positioners.
Reason for Change The Fisher valve positioners vere subject to drift. The change in positioners and the volume boosters were installed to provide improved and more accurate valve response.
Safety Evaluation The seismic qualification of the valve is not affected, therefore, the change does not impact the consequences of seismic events. The new positioners are more accurate, have better response, and are more dependable than the originals. Testing associated with the imple-mentation ensured that the feedvater isolation function of the valve was not impacted.
Therefore, this modification does not constitute an unreviewed safety question per the etiteria of 10CFR50.59.
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Plant Design Change Number HP3-86-386 Plant Design Change Number MP3-86-386 entitled "Security Lighting" is complete.
Description of Change This change installed additional lighting fixtures in several areas of the yard.
Reason for Change Lighting levels in several yard areas vere less than that required by 10CFR73.55. This change increased lighting in these areas above the minimum required level.
Safety Evaluation This change enhanced existing lighting in several yard areas. The lighting system is nonsafety related, and has no connection to any unit specific safety related electrical distribution system. The backup power supply for this lighting was reviewed to ensure it could handle the additional load. The additional lighting does not jeop-ardize the existing lighting system. The installation of the new cables and conduit met all licensing requirements for electrical separation, Appendix R seismic, and hazard analysis.
Therefore, this modification does not constitute an unreviewed safety
, question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-86-387 !
l Plant Design Change Number MP3-86-387 entitled "Replace Terminal Lugs i for Cable CCIAOC001" is complete. l Description of Change A cable in the safety injection pump cooling system was disconnected and relugged with the proper vire lugs. The cable was then recon-nected to a spare portion of the terminal block and 6 jumper vires vere added to complete the circuit.
Reason for Change The original terminatiot. va= made with incorrect lugs. Due to the relugging process, the caule van not long enough to reach its original termination point, so it was necessary to reterminate the cable at a new termination point and add the internal jumper vires.
Safety Evaluation This change did not change the design function of the cable, nor did it impact system operability. All the installations in this modi-i fication meet the requirements of the Electrical Installation Specification.
Therefore, this modification does not constitute an unreviewed safety questiori per the criteria of 10CFR50.59.
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Plant Design Change Nigeber MP3-86-388 l Plant Design Change Number MP3-86-398 entitled "Installation of Separation Barriers on Elevation 24'-6" of Control Building" is
! complete. .
Description of Change ,
4 A series of cable vraps, one conduit vrap, and two cable tray covers were added to provide mandatory separation between the Class lE to- ,
Class lE and Class lE to Non-Class lE circuits. i i i i
Reason for Change t l
There were violations of separation criteria. .t l Safety Evaluation j These changes vere for the purpose of conforming to the installation ;
j requirements and provided no change in the design function of the '
l affected systems. There was no impact on the design basis of any of l i the systems. Therefore, this modification does not constitute an t
unreviewed safety question per the criteria of 10CFR50.59. l 1
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Plant Design Change Number MP3-86-408 Plant Design Change Number MP3-86-408 entitled "Main Steam and Steam Generator Blovdovn (BDG) System Nitrogen Isolation Valves Removal /
Reorientation" is complete.
Description of Change This change removed and scrapped valves 3 MSS *V075, 3 MSS *V076, 3 MSS *V085, 3 MSS *V086, 3 MSS *V099, 3 MSS *V100, 3 MSS *V109, and 3M u*V110.
l :vice l The existing lines that include these valves were removed from by capping off the existing piping connections.
This change removed existing valves 3BDG*V951, 3BDG*V952, 3BDG*V953, 3BDG*V954, 3BDG*V955, 3BDG*V956, 3BDG*V957, and 3BDG*V958. Identical new valves vere installed with valve flow direction opposite to those previously installed.
Reason for Change - r The above valves were installed assuming flov from the Nitrogen System to the Main Steam and Steam Generator Blovdovn System, thereby ,
exposing the valve packing to main steam pressure. A containment entry was required to tighten packing to minimize the resulting packing leaks that contributed to unidentified leakage inside con-tainment. Future steam generator blanketing vill be done via existing connections in the main steam lines, which are located in the main l
steam valve building.
j Safety Evaluation Severance of the in-containment nitrogen system / main steam system interface vere accounted for by using main steam piping connections l
located in the main steam valve building to provide a source of .
nitrogen for blanketing the steam generators. This blanketing operation vill also contribute to ALARA since a containment entry vill not be required in order to connect, via a spool piece, the nitrogen system to the main steam system.
All modifications complied with applicable codes and vere tested accordingly.
No degradation of either the main steam systen, steam generator
! blovdovn system or nitrogen system occurred as a result of this j modification.
l l Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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e Plant Design Chang,e_ Number MP3-86-420 .
-t Plant Design Change Number MP3-86-420 entitled "Installation of Cable :
Tray Cover / Conduit Vrap on Elevation 24'-6" of Auxiliary Building" is :
complete. f Description of Change-A ventilated cable tray cover, overlay plates, and siltemp vrapping l l
vere added to provide the necessary separation between Class 1E to non-Class 1E and Class 1E to Class 1E circuits.
I Reason for Change ;
There vere violations of separation criteria.
Safety Evaluation These changes vere for; the purpose of conforming to the installation --
requirements and provided no change in the design function of the '
affected system. There was no impact on the design basis of any of ;
l the systems. ;
c Therefore, this modification does not constitute an unrevieved safety l question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-87-002
, Plant Design Change Number MP3-87-002 entitled "High Radiation Area Gate Alarms And Varning Lights" is complete.
Description of Change This change provided reliable nonclosure alarms and varning lights at high radiation area access gates.
Reason for Change l These high radiation area access gates vere not equipped with nonclosure alarms and flashing varning lights. While these devices are not required where key locks are used to control access, it has been a convention to install them as a backup at such locations.
Safety Evaluation .
The power for these devices vas provided from 120 volt AC utility lighting panels which are not safety related. No safety-related equipment was affected by this change.
I Therefore, this modification does not constitute an unrevieved safety l
question per the criteria of 10CFR50.59.
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. Plant Design Change Number MP3-87-008 i Plant Design Change Number MP3-87-008 entitled "Resolution of l Instrument Cable Bend Radius Violations" is complete, j Description of Change i
i This change modified junction boxes.3RCS*J8495, 496, and 498 located !
- inside containment, to correct violations of minimum bending radius l J requirements of instrument cables that are terminated to the speed 2 sensing elements of the reactor coolant pumps. This rework involved d enlarging the splice / junction boxes and training the affected cables ;
to maintain an acceptable bending radius. '
- 1. ton for Change ,
1 i l Jes 3RCSIRX856, 3RCS2VX839, and 3RCS4YX813 vere in violation of the
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j nimum bending radius requirement within the junction boxes ~
1 .sRCS*JB495, 496, and 498. This change at: "hd the violation. !
Safety Evaluation l {t J This change did not aff A system operation. This change reduced the i j probability of cable fatl,re by installing an additional box to alle- i
- viste the minimum cable N ' is radius problem. The cables and the !
j junction boxes are safety related. This modification met the t
- licensing requirements for separation, Appendix R, seismic and hazard i
l analysis. No fire stops and seals vere affected by this rework. {
4 Therefore, this modification does not constitute an unreviewed saicty ;
l question per the criteria of 10CTR50.59. i I i i !
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Plant Design Change Number MP3-87-010 Plant Design Change Number MP3-87-010 entitled "Installation of Cable /
Conduit Wrap" is complete.
Description of Change This change installed a cable / conduit vrap to provide the required separation distance between redundant Class lE circuits in accordance with the Separation Criteria in the Auxiliary Building.
Reason for Change Cables training out of conduit on 3CX402BA2 were not adequately separated from cables in 3TX1160/3TX106R. A minimum distance of three feet horizontally and five feet vertically is required. A cable vrap is required for the cables / conduit of 3CX402BA2 to meet the require-ments of Regulatory Guide 1.75.
Safety Evaluation This modification met the licensing requirements for separation, Appendix R, seismic and hazard analysis. No fire stops or seals were affected by this change.
Therefore, this nodification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. ;
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Plant Design Change Number MP3-87-013 Plant Design Change Number MP3-87-013 entitled "Damaged Cable (30SSNNX225)" is complete.
Description of Change This change added a new terminal block into an existing-junction box and replaced the damaged portion cf a containment quench spray cable in order to reestablish electric".1 continuity of the circuit. The cable leads from the temperatur- detector in the refueling vater chemical addition tank to a poael in the Control Building.
Reason for Change A cable in the containment quench spray system was damaged by freezing vater in the conduit. In order to repair the cable, a new terminal block had to be added and the damaged portion of the cable was replaced with new cable. The conduit was sealed to the weather by -
using a gasketed cover on the conduit in order to prevent future damage to the cable.
Safety Evaluation This change did not change the design function of the cable or the temperature detector. Neither the cable nor the racevay is safety related. The modification met all the requirements for separation, fire protection, and seismic considerations of the electrical installation specification.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-87-014 Plant Design Change Number MP3-87-014 entitled "First and Second Point Feedvater Heaters Operating Vent Piping Replacement" is complete.
Description of Change The carbon steel vent. piping and valves on the first and second point feedvater heaters were replaced with 304L stainless steel piping and valves. Some of the piping was rerouted to provide a less severe flow path.
Reason for Change The vent piping had corroded / eroded away as carbon steel was inappr-opriate for this application. Some repairs had been made, but the repairs were also eroding / corroding.
Safety Evaluation- -
This change did not change the design function or operability of the vent piping, but rather increased the reliability of the system. The rerouting of the piping is similar, and had no impact on the feedvater and extraction steam systems operability.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-87-018 Plant Design Change Number MP3-87-018 entitled "Replacement of Post Accident Sampling System Flush' Pump Motor" is complete.
Description of Change This modification changed the motor in the sampling system flush pump from a 1/3 horsepower motor to a 1/2 horsepover motor.-
Reason for Change The 1/3 horsepower motor was tripping due to overload protection under certain valve alignment conditions. The motor was drawing excessive current to overcome pipe friction and short duration dead head situations.
Safety Evaluation _
This change did not affect any safety systems as it involved the replacement of an existing component in a nonsafety-related system.
The operation of the components involved did not change.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
Plant Design Change Number HP3-87-020 Plant Design Change Number MP3-87-020 entitled "Radiation Monitor System Computer Upgrade" is complete.
Description of Change The Central Processing Units in the Radiation Monitoring System computers were replaced to provide increased processing speed, memory, and disk storage capacity.
Reason for Change This change improved system response time and provided space for development of additional applications programs.
Safety Evaluation The Radiation Monitoring System is not safety-related, and has no impact on the operation of safety systems.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-87-026 Plant Design Change Number MP3-87-026 entitled "PORV Seat Material Change" is complete.
Description of Change The Pressurizer Pilot Operated Relief Valve (PORV) seats were replaced with seats manufactured in acccrdance with Crosby part number N96060.
These seats differed from the original seats only in material type.-
The replacement seats were fabricated of SB-637 No. 7718 (Inconel 718), whereas the originals were fabricated of Stellite 6B.
Reason for Change The seat material was changed to reduce seat leakage. Since Hillstone Unit 3 started up, the PORVs have experienced excessive leakage. The new seat material is better suited for steam inlet conditions and is superior to Stellite 6B in resisting steam cutting. -
Safety Evaluation The replacement material is solution treated and precipitation hardened, resulting in a Brinnell Hardness in excess of 331 and a yield strength in excess of 150 KSI. These seats are in all other aspects, other than material fabrication, in accordance vith the original PORV design specification. The Inconel 718 material has demonstrated superior resistance to steam cutting in use in pressurizer safety relief valves. Additionally, the hardness is equivalent to that of the Stellite 6B vhile the yield strength far exceeds that of the Stellite material. In the range of temperatures 1 that the seats vill be exposed to, the mean coefficients of thermal expansion of the materials are identical. This vill ensure geometric compatibility during periods of thermal strain.
Based on the increase in resistance to steam cutting, the equivalent !
mechanical properties, the quality program under which the seats vere manufactured and examined, and the testing of the reassembled valve, l l
it was concluded that the modifications did not affect the valve functionability whatsoever, and therefore did not affect the design l basis of the PORVs.
Therefore, this modification does not constitute an unreviewed safety )
I question per the criteria of 10CFR50.59.
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l Plant Design Change Number MP3-87-027 Plant Design Change Number MP3-87-027 entitled "Feedvater Recirculation Valve Internals.Changeout" is complete.
Description of Change This change replaced the existing seat rings.of the feedvater recirculation valves with seat ring / lover cage assemblies. The lower cage is to provide a sacrificial hard faced surface belov the valve seat to prevent valve body erosion.
Reason for Change The bottoms of the feedvater recirculation valve bodies were eroding due to flashing past the seat of the valves and subsequent cavitation on the lover surface of the valve bodies. The modification was recommended by the valve manufacturer as a way to reduce the erosion of the valve bodies and increase the life expectancy of the valves.
Safety Evaluation This change did not change the actual CV or reduce the maximum flow capability of the valve. Nor did the system operation change as a -
result of this design change.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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l Plant Design Change' Number MP3-87-030 l Plant Design Change Number MP3-87-030 entitled "Penetration Protection ,
for Non-Category 1 Circuits" is complete. !
f Description of Change This change provided viring details to complete the installation of secondary containment electrical penetration overcurrent protection into the following three'(3) non-Class 1E, 120VAC circuits: 3HVU-FS34A,B,C, 3HVU-PDIS20A, 21A, 22A, 23A, and 3HVU-PDIS20B,21B, 23B.
Reason for Change This change was made to provide secondary containment electrical penetration overcurrent protection devices into all nonsafety-related circuits entering containment where available fault current could exceed the penetration ratings.
Safety Evaluation The change added a greater degree of protection against an inadvertent breach of containment boundary due to an electrical fault inside containment.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-87-034 Plant Design Change Number MP3-87-034 entitled "Carbon Dioxide-Fire Protection Keylock Svitch Modification" is complete.
Description of Change This change replaced 39 toggle type actuation svitches with three-position keylock switches having the capability of:
- 1. Normal Operation (existing feature)
- 2. Discharge (existing feature)
- 3. Abort / Prevent Discharge (the new feature)
Reason for Change The actuation svitch was changed to a key locked design to preclude unauthorized individuals from initiating a carbon dioxide discharge. _
The "abort" feature was added to the switch to provide a quick means of terminating an undesired discharge or prevent an undesired discharge from initiating.
Safety Evaluation The automatic and manual initiation subsystem of carbon dioxide fire suppression is a 4 vire supervised Class A system as defined by the National Fire Protection Association. This means that each signaling line circuit and the devices connected to it are capable of operating
_ for their intended signaling services during a single break or a single ground fault condition of any signaling line circuit conductor.
The circuitry is supervised against open and short circuit.
The credible failures of the circuit are open and short circuit. The "abort" feature is energized to abort. The credible failures vould
, prevent the "abort" feature from operating, but the initiation of the ,
fire system would not be compromised, nor is it degraded from the l i
original design.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
Plant Design Change Number MP3-87-036 Plant Design Change Number MP3-87-036 entitled "Feedvater Venturi Inspection Ports" is complete.
Description of Change This change installed inspection ports upstream of the four main feedflow venturis.
Reason for Change This change allows periodic inspection and cleaning of the venturis as required by Technical Specifications.
. Safety Evaluation The inspection ports were analyzed, designed and installed in accordance with criteria that met or exceeded the original design criteria of the feedvater system. The modification allows better operation of the plant by ensuring accurate feedvater flow measure-ments.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-87-038 Plant Design Change Number MP3-87-038 entitled "Safety' Provisions for the Auxiliary Boiler Pressure Relief Line" is complete.
Description of Change The modification consisted of the construction of safety barriers and ladder to provide safe access around the auxiliary boiler pressure relief line on the roof of Varehouse #5 and the Auxiliary Boiler Building.
Reason for Change The auxiliary boiler pressure relief line had been identified as a potential safety hazard by the Station Safety Administrator.
Safety Evaluation Both the Auxiliary Boiler Building and Varehouse #5 are nonsafety-related conventional design structures; Varehouse #5 being also designed for seismic and tornado loads in the east-vest direction.
The attachments made to the existing structural steel vere fillet velds of 1/4 inch maximum size and the only new load introduced by this change is the dead load of the steel components. Based on the methods of attachment, and the insignificant additional load, cal-culations were not performed as impact on the design is insignificant.
Although not specifically designed for seismic and tornado loading, the steel members do not introduce any potential missiles that have not been considered in the plant design basis.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
Plant Design Change Number MP3-87-045 Plant Design Change Number MP3-87-045 entitled "Shielding of the Fuel Transfer Canal Traverse" is complete.
Description of Change This change installed lead block shielding and a locked vire mesh enclosure above the shake spaces in the vicinity of the fuel transfer canal traverse at the containment and the Fuel Building exterior valls.
Reason for Change It was determined through analysis that the passage of a spent fuel assembly through the fuel transfer canal could result in radiation dose rates as high as 2000 R/hr directly above the fuel transfer tube at the shake space areas between the Containment and Fuel Buildings.
The addition of 6 inches of lead shielding directly over the tube -
reduces this dose to 2 R/hr, and reduces the affected area so that a single locked cage can assure personnel safety.
Safety Evaluation This change did not affect the original design basis of the shake spaces, nor did it impair the structural integrity of the adjacent buildings. It did not af fect the operability of any of the plant systems.
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-87-053 Plant Design Change Number MP3-87-053 entitled "Transformer Cooling Controls" is complete.
Description of Change The intent of this change was to modify the cooling control circuitry for the Millstone Unit 3 main, normal, and reserve station trans-formers to be consistent with the following criteria:
- 1. Cooling fans / pumps vill not be shutdown upon initiation of the transformer fire deluge system.
- 2. Cooling fans / pumps will be automatically shutdown upon transformer fault detection.
- 3. Cooling fans / pumps which are started upon transformer ener-gization vill be shutdown automatically upon transformer de-energization.
The change required the installation of new cables, relaying, switches, control transformers and associated viring into the main board and transformer panels.
Reason for Change Millstone Unit 3 main transformers did not comply with 1, 2, or 3 above. The normal and reserve station transformers did not comply with 1 or 2 above.
Safety Evaluation All components added performed nonsafety-related functions and vere installed into nonsafety-related panels. Any cables pulled through Category 1 areas vere inspected for separation, and any dacaged Category 1 or fire protection 0A boundary cable penetrations vere )
sealed and reinspected in accordance with station procedures.
1 Therefore, this modification does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-87-061 Plant Design Change Number MP3-87-061 entitled "Installation of Bracing to Cable Tray Supports to Auxiliary' Building Elevation 43'-6" and MCC and Rod Control Area Elevation 45'-6"" is complete.
Description of Change This change added some bracing to cable tray supports to provide stability to protect nearby safety-related equipment.
Reason for Change The affected cable trays are nonseismic, but are in the vicinity of safety-related equipment. In order to ensure that the structural integrity of the nonseismic supports vould be maintained during and after a seismic event, some addition bracing was required on the supports.
Safety Evaluation This change did not affect the operability of any of the plant systems. None of the affected cable trays vere safety related and there was no impact on the design function of the systems involved.
Therefore, this modification does not constitute an unreviewed safety '
question per the criteria of 10CFR50.59.
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Plant Design Change Number MP3-87-062 Plant Design Change Number MP3-87-062 entitled "Refueling Cavity Drain Line Modification" is complete.
Description of Change This change added new supports and modified existing supports to seismically qualify the drain lines from the Refueling Cavity to beyond the first isolation valve.
Reason for Change This change was performed as a result of IE Bulletin 84-03. The Refuel Cavity drain lines vere not originally designed to be seismically qualified. This could have resulted in an unisolable drain path from the Refuel Cavity in the event of a break in one of the three drain lines, upstream of the first isolation valve. By seismically qualifying the lines to a point downstream of the first isolation valve, the probability of a break in the nonisolable portion of the line has been reduced.
Safety Evaluation Potential Refuel Cavity drainage paths from events other than the failure of the Refueling Cavity Seal have been identified to be the three drain lines from the bottom of the cavity. These lines provide a suction from the cavity to the nonsafety Spent Fuel Purification Pumps. These lines, originally designed as nonseismic, nonsafety related, have the potential to fail upstream of the first isolation valve, resulting in an unisolable drain path from the cavity. By installing this modification, the stiffness of the piping / restraint system has been increased to the point vere a seismic assessment has shown that the probability of failure has been minimized. Failure dovnstream of the first isolation valve is still possible and requires consideration. Since the isolation valve in one of the drain lines, 3SFC-V997 is locked closed when fuel is being moved in Containment, the maximum credible, though extremely improbable, leakage path occurs with the complete, double ended rupture of two drain lines (one 2" and one 3" line). This results in a maximum drainage flow rate of 1630 gpm. Previous calculations had shovn that the maximum leakage that vould result from the failure of the Refueling Cavity Seal Ring is 4750 gpm. Analysis of that event had shown that operators vould have 36 minutes to place any fuel in transit into a safe storage location.
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A change to Abnormal Operating Procedure A0P 3572 vas implemented to include the following actions to be taken upon symptoms of a failure of the cavity seal:
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Plant Design Change Number MP3-87-062 (Continued)-
Safety Evaluation (Continued)
- 1. Secure the Fuel Pool Purification Pumps.
- 2. Close drain valves 3SFC-V998 and V999.
Therefore, the failure of the Refuel Cavity drain lines was determined to be bounded by the failure of the Refuel Cavity Seal Ring. The analysis of the Refuel Cavity Seal Ring concluded that "Offsite doses are negligible and, therefore, well within the acceptable limits of 10CFR100."
Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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PROCEDURE CHANGE INDEX Procedure Number Title l OP 3204 At Power Operation ;
Rev. 1, Change 5 i i
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NUMBER TITLE OP 3204 At Power Operation Rev. 1, Change 5 Description of Procedure This change moved the normal-operating level of.the steam generators from 50 percent narrov range level to 58 percent level when reactor power vas above 70 percent.
Reason for Change This change was made to account for errors in the steam generator narrov range level sensing equipment. Due to a design problem with the steam generator narrow range level reference legs used for level measuremer.ts, as much as a 13.1 percent level error existed in the non conservative direction when applied to the lov-low steam generator i reactor trip setpoints. This prcblem was previously reported in LER 87-022-00. .
To account for the errors in narrov range level, the steam generator low-lov level reactor trip setpoint was raised 13.1 percent. This reduced the margin between the normal operating level and the trip setpoint to 13.4 percent. The margin of 13.4 percent vould have made operation and water control difficult. To increase this margin and prevent unnecessary reactor trips, th e normal operating level of the ;
steam generators was increased by 8 percent.
Safety Evaluation By operating the steam generators at 58 percent level, the margin to the reactor trip setpoint was increased and the potential for unplanned reactor trips reduced. The new operating level of 58 percent above 70 percent power did not change'any of the assump-tions stated in the accident analysis.. ,
i Therefore, this modification does not constitute an unreviewed safety question per the criteria of 10CFR50.59. :
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JUMPER-LIFTED LEAD-BYPASS (J-LL-B) CHANGES INDEX J-LL-B Number J-LL-B Title 3-86-100 Vater Treatment Filter Installation 3-87-060 3RCS*PIB Frame Vibration Indication 3-87-068 Turbine Emergency Trip Lov Oil Pressure Alarm and Trip Override 3-87-083 Supply Temporary Power to the Nuclear Instrument Rack During the Purple Train Outage 3-87-140 Reversal of Steam Generator High-High Level
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Jumper-Lifted Lead-Bypass Change Number 3-86-100 '
Jumper Lifted Lead Bypass Change Number 3-86-100 entitled ~"Vater. i Treatment Filter Installation" is complete.
Description of Change A jumper was installed to bypass the Vater Treating Demineralizers to
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supply the Ecolochem Reverse Osmosis unit directly from the Ultra-filter System. This is still installed.
Reason for Change The Reverse Osmosis unit requires a final polishing demineralizer to remove hydrazine injected for oxygen removal. This final polishing demineralizer performs the same function as the permanent demin-eralizer system. Therefore, it is unnecessary to run the permanent demineralizing package. _
Safety Evaluation The only credible failure modes associated with this bypass jumper are degraded water quality and the rupture of temporary hose. All tem-porary hose is contained within the Turbine Building southeast corner.
There is no safety-related equipment in this area. Failure of tem-porary hose vill have no impact on any design basis accident.
Degraded water quality is precluded by normal chemistry sampling ,
procedures.
Therefore, this bypass jumper does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Jumper-Lifted Lead-Bypass Change Number 3-87-060 Jumper Lifted Lead Bypass Change Number 3-87-060 entitled "3RCS*P1B Frame Vibration Indication" vas installed and restored.
Description of Change This bypass jumper involved temporarily removing vibration indication and alarm, vertical component, for 3RCS*PlB from service.
The vertical frame vibration channel of 3RCS*PIB vas alarming spuri-ously due to a fr.ulty voltage to displacement converter mounted on the pump. The spurious alarms were eliminated by lifting the field cable from the channel to the monitoring panel until the channel can be repaired during the next cold shutdown. The horizontal frame vibra-tion channel was still available for monitoring frame vibration, and the two shaft vibration channels were unaffected.
Reason for Change .
Spurious alarms and erratic indication vere constantly received.
Plant operation prevented access to perform repairs.
Safety Evaluation The Reactor Coolant Pump (RCP) vibration monitoring system provides indication and alarms; it does not provide any equipment protective features. The RCP vibration monitoring system is not a safety system, and does not interact with any safety system. The RCP vibration monitoring system does not provide indications of the initiation of accident conditions, and is not required for safe operation or a shutdown of the plant.
Therefore, this bypass jumper did not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Jumper-Lifted Lead-Bypass Change Number 3-87-068 Jumper Lifted Lead Bypass Change Number 3-87-068 entitled "Turbine Emergency Trip I)v Oil Pressure Alarm and Trip Override" vas installed and restored.
Description of Change While performing surveillance testing of the turbine control valves, one channel of "emergency trip fluid pressure" vas jumpered. The condition of the plant for this change was that of "the number of operable channels was one less than the total number of channels but '
within the minimum channels operable requirement."
Reason for Change ,
The purpose of this change was to minimize the possibility of a spurious reactor trip during turbine control. valve testing due to oil pressure fluctuations while exercising valves. ~
Safety Evaluation This is described in Technical Specifications as an acceptable configuration for turbine control valve testing.
Therefore, this bypass jumper did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Jumper-Lifted Lead-Bypass Change Number 3-87-083 Jumper Lifted Lead Bypass Change Number 3-87-083 entitled "Supply Temporary Power to the Nuclear Instrument Rack During the Train B ;
Outage" vas installed and restored. '
Description of Change I During the Train B electrical outage, the power for the Train B Nuclear Instrument Rack vas supplied from the installed non-Category 1 supply. ,
Reason for Change The requirements for operational nuclear instruments during fuel movement could not be met with the Train B Nuclear Instrument inoperable. The instrument was povered from nonsafety-related power to provide redundancy with the other nuclear instrument povered from the opposite electrical train.
Safety Evaluation To allow fuel movement during the Train B outage, Technical Speci-fications require two operable source range nuclear instruments. This change meets that goal by povering the Train B nuclear instruments from nonsafety-related power during the train outage. The Train B Nuclear Instrument System had met all Technical Specification tests for operability while povered from the alternate source. The alter-nate power source vas of identical voltage rating to the original source and had adequate capacity to supply the additional load.
The use of the alternate source did not subject the safety-related equipment to any condition that vould affect its environmental qualifications.
There vas no effect on separation as the jumper was internal to the nuclear instrument. This change vas for use in refueling mode only.
Therefore, this bypass jumper did not constitute an unreviewed safety l question per the criteria of 10CFR50.59.
Jumper-Lifted Lead-Bypass Change Number 3-87-140 Jumper Lifted Lead Bypass Change Number 3-87-140 entitled "Reversal of Steam Generator High-High Level Bistable Action" vas installed and restored.
Description of Change This change reversed the bistable action of the steam generator high-high level bistables such that no alarm signal vould be present with an indicated high-high level in the steam generators.
I Reason for Change The Feedvater Isolation Signal had to be removed to perform Loss of Power and Engineered Safety Features testing. The Feedvater Isolation Signal present was due to an indicated high level in the steam generators due to the reference legs being drained while the steam _
generators were empty for maintenance, and later, due to the steam generators being full of water in vet layup. The modification removed 4 the feedvater isolation by making the bistable output the same as the -
output vould be for normal vater level.
Safety Evaluation This change evaluated the impact on the reversal of the bistable outputs of all 16 steam generator high-high level bistables going to Solid State Protection for feedvater isolation and turbine trip. This change has been evaluated for refueling mode, where there is no -
requirement to keep steam generator level below the high-high setpoint and the action of this trip, that of mitigating an overcooling situ-ation, does not exist. Based on this set of conditions, the change is !
safe. This change was restored prior to commencing a heatup. !
i Therefore, this bypass jumper did not constitute an unrevieved safety i
- question per the criteria of 10CFR50.59.
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SETPOINT CHANGE INDEX Number Title 3-87-002 70 Percent and 90 Percent 4160 Volt Undervoltage Relays 3-87-003 Increase in Reset Point of Instrument Air System Pressure Switch 3IAS-PS14 3-87-004 Engineered Safeguards Features Building Ventilation Radiation Monitor Alert Setpoint 3-87-005 Alert and Alarm Setpoints for 3RMS-RE19 3-87-006 30SS-LS60 - Refueling Vater Storage Tank Level Alarm Setpoint Hodification 3-87-007 Turbine Building Floor Drains Sump Radiation Monitor Lov Pressure Alarm Setpoint 3-87-008 Supplementary Leakage Collection and Release System Radiation Monitor Alert Setpoint 3-87-009 "A" Fuel Building Filtered Exhaust Fan Lov inlet Plessure Svitch 3-87-010 "B" Fuel Building Filtered Exhaust Fan Lov Inlet Pressure Svitch 3-87-011 3CHS-PDS1012 (Thermal Regeneration Chiller Compressor Inlet Vane Closed Svitch) 3-87-012 Total Thrust Limit for High Pressure Safety Injection Valves 3SIH*HV8923A and 3SIH*MV8923B l 3-87-015 Total Thrust Limit for High Pressure Safety Injection System Valves 3SIH*MV8802A and
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3SIH*MV8802B 3-87-016 Total Thrust Limit for High Pressure Safety l Injection System Valves 3SIH*HV8801A and 3SIH*MV8801B 3-87-017 Total Thrust Limit for High Pressure Safety Injection Valves 3SIH*MV8802A and 3SIH*MV8802B ,
Setpoint Change Number 3-87-002 Setpoint Change Number 3-87-002 entitled "70 Percent and 90 Percent 4160 Volt Undervoltage Relays" is complete.
Description of Change Setpoints for both trains of the Class lE 4.16 kV undervoltage relays were raised slightly. The relays affected were the 90 percent and 70 percent ur<dervoltage relays. The 90 percent undervoltage relays vere changed from 3710 volts to 3745 volts. The 70 percent under-voltage relays were changed from 2800 volts to 2835 volts.
Reason for Change Prior to implementation of this change, setpoint documents required that the relays be set'to a value which is the minimum permitted by plant Technical Specification Table 3.3-4. The original setting did not allov for any relay setpoint drift in the dovnvard direction. At the first scheduled recalibration surveillance, some of the relays ,
vere found with settings below the Technical Specification limit.
Safety Evaluation This setpoint change has been reviewed for impact on safety-related !
systems and equipment and determined to be safe to implement since the l change makes the setpoints more conservative.
Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 3-87-003 Setpoint Change Number 3-87-003 entitled "Increase in Reset Point of Instrument Air System Pressure Switch 2IAS-PS14" is complete.
Description of Change The reset point of.3IAS-PS14 has been increased from 95 psig to 103-psig.
Reason for Change Testing during start-up of the air system demonstrated that a 10 psi deadband is the best that could be achieved between the pressure switch actuation point and its reset point. The new reset point is still 7 psi belov the 110 psig air compressor cutoff.
Safety Evaluation -
This change did not affect the operation of any safety-related equipment. Therefore, this modification did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 3-87-004' f Setpoint Change Number 3-87-004 entitled "Engineered Safeguards Features Building Ventilation Radiation Monitor Alert Setpoint" is complete. i Description of Change 7
Changed Radiation Alert Setpoint from 2 E-7 microcuries/cc to 1.03 E-6 i microCuries/cc. ,
Reason for Change The Radiation Alert Setpoint is based on twice the calculated normal radiation level. Measurements indicated the normal value had changed- !
as operation proceeded, dictating a change to the Radiation Alert Setpoint. ,
1 Safety Evaluation -.
l The alert setpoint does not provide protective actions: the alarm :
setpoint (1.5 E-5) provides isolation of Engineered Safeguards i Features normal ventilation. The Engineered Safeguards Features '
Building is then ventilated by Secondary Leakage Collection and
- Release System.
i' Therefore, this change does not involve an unreviewed safety question per the criteria of 10CFR50.59. 1 a ,
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Setpoint Change Number 3-87-005
'Setpoint Change Number 3-87-005 entitled "Alert and Alarm Setpoints ;
for 3RMS-RE19" is complete. I Description of Chsnge This setpoint change revised the alarm and alert setpoints for 3RMS-RE19 (Vaste Disposal Building area radiation monitor). The alert and alarm setpoints were changed from 0.4 mR/hr, 0.8 mR/hr to 2 mR/hr, 4 mR/hr respectively.
Reason for Change The Radiation Monitor Manual requires that the alert and alarm setpoints be set at 2 times and 4 times the normal background levels.
The actual background levels in this area vere 1.mR/hr. The initial anticipated background level was 0.2 mR/hr. . _. l Safety Evaluation This change was in accordance with design intent and procedures.
Therefore, this change does not constitute an unreviewed safety l
question per the criteria of 10CFR50.59.
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Setpoint Change Number 3-87-006 Setpoint Change Number 3-87-006 entitled "30SS-LS60 - Refueling Vater Storage Tank Level Alarm Setpoint Modification" is complete.
De,stription of Change The Refueling Vater Storage Tank (RVST), 30SS*TK1, high level alarm s9tpoint was changed from 57.7 feet to 58.15 feet. This increased the band between the RVST's high level alarm and the low level alarm from 5.4 inches to-10.8 inches.
Reason for Change !
The original setpoint band between the RVST-high level alarm and lov i level alarm was too narrow. Once the lov level alarm cleared, the high level alarm cycled in and out due to vave action in the tank. In order to eliminate the number of nuisance alarms and to prevent con-tinued operation in an alarmed condition, the band between the high level and lov level alarms was adjusted. l 1
Safety Evaluation This setpoint change to a nonsafety-related instrument has been reviewed for impact on safety-related systems and equipment and determined to be safe to implement.
Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 3-87-007 ,
Setpoint Change Number 3-87-007 entitled "Turbine Building Floor '
Drains Sump Radiation Monitor Lov Pressure Alarm Setpoint"-1 s complete. 1 Description of Change ,
i Changed low pressure alarm setpoint from 0.00 PSI to 1.00 PSI.
- Reason for Change Turbine Building Floor Drains Sump Radiation Monitor has flow through ~'
the monitor only when the sump pumps are running to reduce sump level.
The lov pressure' alarm setpoint at 0.00 PSI was causing continuous low pressure alarms. Raising the setpoint to 1.00 PSI eliminated the continuous alarm condition, while providing valid low pressure alarms when the sump ~ contents were being removed.
Safety Evaluation 7
This change does not affect any safety-related component, system, or structure.
Therefore, this change does not involve an unreviewed safety question per the criteria of 10CFR50.59.
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t Setpoint Change Number-3-87-008 r Setpoint Change Number 3-87-008 entitled "Supplementary Leakage ;
Collection and Release System Radiation Monitor Alert Setpoint" is complete. ,
' Description of Change ;
Changed radiation alert setpoint from 2.0' E-6 microcuries/cc to 8.0 E-5 microcuries/ce.
Reason for Change ,
The radiation alert setpoint was based on twice the calculated normal i radiation level. Measurements required to calculate theinormal value vere difficult to obtain, since the ventilation. paths to the monitor are_ frequently changed, and the normal radiation level calculated did not provide as good a basis.for the radiation alert setpoint_vhen com- ,
pared to monitors with continuous ventilation. The radiation alert .
setpoint was changed to 25 percent of the radiation alarm setpoint, which is based on release limits. This radiation alert setpoint is-high enough to avoid numerous alarms, yet lov enough to alert operators to releases. The corresponding release rate (400 .
i microcuries/sec) is well below plant Technical Specification limits (260,000 microcuries/sec). ;
i Safety Evaluation ;
j This change is in keeping with design intent - alarm on twice the '
normal radiation level. It provides an alarm far belov Technical ,
Specification limits. No safety-related system operation is affected. ;
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Therefore, this change does not involve an unreviewed safety question i per the criteria of 10CFR50.59.
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Setpoint Change Number 3-87-009 Setpoint Change Number 3-87-009 entitled ""A" Fuel Building Filtered Exhaust Fan Lov Inlet Pressure Svitch" is complete.
Description of Change The pressure svitch setpoint was changed from -6.0 inches vater column to -3.0 inches vater column.
Reason for Change Negative pressure at the inlet of the fan vas not great enough to reset the pressure switch. This resulted in closing of the fan inlet and outlet dampers and then tripping of the fan.
Safety Evaluatig The lover negative pressure setpoint allows the fan to be operated while still providing the safety function of closing the fan inlet and outlet dampers if the fan is idle. The change does not affect the operation of the safety-related Fuel Building Ventilation System and is necessary to allow the system to accomplish its safety function.
Operability of the "A" Fuel Building Filtered Exhaust Fan is verified on a monthly basis per a Technical Specification surveillance requirement which ensures that setpoint drift vill not prevent fan operation.
Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 3-87-010 Setpoint change Number 3-87-010 entitled ""B" Fuel Building Filtered Exhaust Fan Lov Inlet Pressure Svitch" is complete.
Description of Change The pressure svitch setpoint was changed from -6.0 inches vater column to -3.0 inches vater column.
Reason for Change Negative pressure at the inlet of the fan was not great enough to reset the pressure svitch. This resulted in closing of the fan inlet and outlet dampers and then tripping of the fan.
Safety Evaluation The lover negative pressure setpoint allows the fan to be operated while still providing the safety function of closing the fan inlet and outlet dampers if the fan is idle. The change does not affect the operation of the safety-related Fuel Building Ventilation System and is necessary to allow the system to accomplish its safety function.
Operability of the "B" Fuel Building Filtered Exhaust Fan is verified on a monthly basis per a Technical Specification surveillance requirement which ensures that setpoint drift vill not prevent fan operation.
1 Therefore, this change does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 3-87-011 Setpoint Change Number 3-87-011 entitled "3CHS-PDS1012 (Thermal Regeneration Chiller Compressor Inlet Vane Closed Svitch)" is complete.
Description of Change The setpoint of 3CHS-PDS1012 (Thermal Regeneration Chiller Compressor Inlet Vane Closed Svitch) was changed from 40 psid to 30 psid.
Reason for Change The setpoint was changed based on vendor recommendation during startup of the unit.
Safety Evaluation The setpoint change was specific to the Boron Thermal Regeneration Chiller skid. Since the switch functions as a permissive in the chiller start circuitry, the change had no effect on other systems and did not contribute to any safety failu...
Therefore, this change does not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 3-87-012 Setpoint Change Number 3-87-012 entitled "Total Thrust Limit for High Pressure Safety Injection Valves 3SIH*MV8923A and 3SIH*MV8923B" is complete.
Description of Change, Increase the target thrust at total thrust (open and close) from 5,986 pounds to 8,849 pounds.
Reason for Change The torque switch could not be set lov enough to meet the maximum torque switch setpoint as indi;ated in SP87-3-3, Procedure For Testing Limitorque MOVS Using H0 VATS.
Safety Evaluation The maximum inertial stem force allowed by the torque switch trip and the contactor drop out time does not stress the valve beyond levels acceptable for normal operation and seismic loads and including Final Safety Analysis Report design basis seismic event. A maximum total thrust of 8,849 pounds is also within the operator manufacturer allovable of 13,900 pounds. There are no additional, credible failure modes, since these changes are to ensure that the valve meets its intended safety function.
Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Changu Number 3-87-015 Setpoint Change Number 3-87-015 entitled "Total Thrust Limit for High Pressure Safety Injection System Valves 3SIH*MV8802A and 3SIH*MV8802B" is complete.
Description of Change Increase the target maximum thrust at torque switch trip (open and close) from 16,377 pounds to 17,200 pounds and increase the target thrust at total thrust from 17,839 pounds to 21,700 pounds (open only).
Reason for Change The torque switch could not be set lov enough to meet the maximum torque switch setpoint as indicated in SP87-3-3, Procedure For Testing Limitorque MOVS Using H0 VATS.
Safety Evaluation f
An evaluation has determined that a maximum thrust at torque switch trip of 17,200 pounds vill not violate the maximum allovable thrust for the operator and is below the thrust value which the operator can deliver at 80 percent voltage.
These valves open on limit only. An open thrust vill not be seen by these operators during normal or accident conditions and is outside the design basis.
Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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Setpoint Change Number 3-87-016 Setpoint Change Number 3-87-016 entitled "Total Thrust Limit for High Pressure Safety Injection System Valves 3SIH*HV8801A and 3SIH*HV8801B" is complete.
Description of Change Increase the target maximum thrust at torque svitch trip (open and close) from 16,377 pounds to 18,060 pounds and increase the target thrust at total thrust from 18,439 pounds to 21,000 pounds (open) and 18,439 pounds to 19,340 pounds (close).
Reason for Change The torque switch could not be set lov enough to meet the maximum torque svitch setpoint as indicated in SP87-3-3, Procedure For Testing Limitorque MOVS Using H0 VATS.
Safety Evaluation An evaluation has determined that a maximum thrust at torque switch trip of 18,060 pounds vill not violate the maximum allovable thrust for the operator and is belov the thrust value which the operator can deliver at 90 percent voltage.
The maximum target thrust at total thrust of 19,340 pounds meets all requirements as detailed in the original valve specification. In addition, this load vill not exceed the Limitorque operator allovable of 24,900 pounds. These valves open on limit only. An open thrust vill not be seen by these operators during normal or accident condi-tions and is outside the design basis.
Therefore, this change does not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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f Setpoint change Number 3-87-017 r
r Setpoint Change Number 3-87-017 entitled "Total Thrust Limit for High !
Pressure Safety Injection Valves 3SIH*HV8802A and 3SIH*MV8802B" is complete. }
Description of Change !
I Increase the target thrust at total thrust from 21,700 pounds to ;
24,000 pounds (open) and 17,839 pounds to 19,340 pounds (close). i Reason for Change The torque switch could not be set lov enough to meet the maximum torque svitch setpoint as indicated in SP87-3-3, Procedure For Testing Limitorque MOVS Using NOVATS.
Safety Evaluation The maximum target thrust at total thrust of 19,340 pounds meets all :
j requirements as detailed in the original valve specification. In i addition, this load vill not exceed the Limitorque operator allovable i of 24,900 pounds. These valves open on limit only. An open thrust !
vill not be seen by these operators during normal or accident con- !
- ditions and is outside the design basis. i 1
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- adequate to close and open the valve against a delta pressure of 1,750 i i pounds per square inch as specified in the Impell report for IEB ;
j 85-03.
1 Therefore, this change does not constitute an unrevieved safety
] question per the criteria of 10CFR50.59.
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'I 1 TESTS INDEX i
a Number Title 1
- IST 3-86-015 Carbon Dioxide System Modifications IST 3-86-022 Boron Thermal Regeneration System - Pre-Resin l Loading-O IST 3-86-040 Emergency Seal Oil Pump Test IST 3-87-001 Core Flow Data Monitoring IST 3-87-004 Motor Driven Auxiliary Feedvater Pumps (MDAFV) Suction and Discharge Piping IST 3-87-006 Motor-Driven Feedvater Pump 50 Percent Trip '
' ~
IST 3-87-011 Containment Air Recirculation (CAR) Fan Effects On Containment Temperatures 4
IST 3-87-013 In-Service Test Plan For Spent Fuel Pool J
Purification IST 3-87-015 Security Electrical Distribution IST 3-87-016 Turbine Oil Pump Testing 1
IST 3-87-018 Shock Chlorination of Circulating Vater System j IST 3-87-019 Euergency Diesel Generators Air Dryers Flush IST 3-87-024 Pressurizer Henters 86E Relay Test j IST 3-87-025 Refueling Cavity Seal Failure Action j Verification j IST 3-87-027 Vet Layup Skid Functional Test - Skid 1A
- IST 3-87-031 Refueling Cavity Pool Seal Leak Test IST 3-87-032 Emergency Diesel Generators Air Start Test q IST 3-87-035 Vet Layup Skid Functional Test - Skid 1B IST 3-87-036 Vet Layup Skid Functional Test - Skid IC 1
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NUMBER TITLE IST 3-86-015 Carbon Dioxide System Hodifications Description of Test .-
Alarms and indicators added to the carbon dioxide fire suppression system since initial start-up testing vere verified to operate.
Key operated "interrupt" switches vere tested to ensure that they had the capability, as designed, to prevent the fire detection system from sending an actuation signal to the fire suppression system and that a "trouble" alarm vould annunciate whenever this feature was used.
Retransmittal of alarms due to manual actuation of the carbon dioxide system was verified at the Main Fire Detection console in the control Room.
Modifications to ventilation system response to the fire suppression system vere verified. ,
Audible alarms vere verified to be sufficiently loud to reach all protected areas. t I
Reason for Test This test constituted a retest for functions which failed initial acceptance testing and as an initial acceptance test for modifications to ensure that the system operated to its design capacity.
Safety Evaluation All fire signals were sirulated electrically. Pntumatic signals were actual with the exception of the discharge valves seing disconnected from their actuating lines so that carbon dioxide ciuld fill the system as designed, but not discharge to occupied spaces. The cycling of the fire suppression system caused some ventilation systems to temporarily realign, but this response is as specified in design documents. No safety systems vere operated in a manner not consistent with their expected manner and no malfunctions of safety equipment occurred.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i NUMBER TITLE IST 3-86-022' Boron Thermal Regeneration System - Pre-Resin ;
- Loading !
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Description of Test
'l This test encompassed testing of the Boron Thermal Regeneration System .;
' . chiller, setting limits on the chiller and several control valves, and ,
testing system logic. !
t j Reason for Test !
l This test was performed following component testing during initial ;
start-up testing of the Boron Thermal Regeneration System.
i Safety Evaluation ;
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IST 3-86-022 tested previously established design parameters utilizing i normal operating procedures and controls. !+
J Therefore, this test did not constitute an unreviewed safety question i
per the requirements of 10CFR50.59. l i i t
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TITLE IST 3-86-040 Emergency Seal Oil Pump Test Description of Test This test consisted of auto starting the emergency seal oil pump by turning off the main seal oil pump vith 75 pounds of air in the generator.
Reason for Test This test was undertaken to collect pressure, current, and rpm data on the main generator D.C. emergency seal oil pump when it is auto started on loss of the main seal oil pump. This testing vas required by the main generator vendor, General Electric.
Safety Evaluation This test was run utilizing normal operating procedures where possible. The test deviated from normal operating procedures in that the generator was pressurized with 75 pounds of air and the main seal oil pump vas manually turned off to auto start the emergency seal oil pump. The credible failure mode for system operation in this manner vould be failure of the emergency seal oil pump to auto start and failure of the main seal oil pump to be manually restarted before the air in the generator is released through the generator seals. Any credible fire generated from this failure vould be mitigated by the existing bearing fire protection deluge system.
No safety systems or equipment vould be affected by this failure mode.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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NUMBER TITLE l IST 3-87-001 Core Flow Data Monitoring i Description of Test This test was performed by connecting pen recorders with high gain DC isolation amplifiers on the inputs to the four power range detectors, four reactor coolant loop flov detectors (one from each loop), and eight incore thermocouple detectors (tvo per core quadrant). Data was collected for a four hour period.
Reason for Test This test was performed to determine if the core vas demonstrating the type of core flov anomaly that was discovered at other Vestinghouse designed cores (Callavay and Volf Creek).
Safety Evaluation This test was for data collection only. There vere no vires or leads disconnected from any plant parameter. All temporary recorders had input isolation amplifiers to ensure facility separation. In addi-tion, the flov and power range connections were done at isolated i
outputs in the protection racks.
Therefore, this test did not constitute an unrevieved safety question per the criteria of 10CFR50.59.
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[ NUMBER TITLE IST 3-87-004 Motor Driven Auxiliary Feedvater Pumps l (HDAFV) Suction and Discharge Piping
(
j Description of Test L -
2 Both Motor Driven Auxiliary Feedvater (MDAFV) pumps were operated in- !
, recirculation mode and system pressures, suction pressure trip switch r status, and pump vibration data vere monitored or recorded. ;
Reason for Test 3 The test was performed in order to establish why the Auxiliary ;
, Feedvater pumps tripped on lov suction pressure. Licensee Event
, Report 87-004-00 describes.in detail the conditions under which pump '
, trips occurred, and the investigations conducted to determine the root ;
cause.
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i The MDAFV pumps vere operated in accordanco with approved operating ;
! procedures. Test equipment was connected to selected test connections
! and pressure switches, and cooling vater flov to the pumps' lube oil t coolers was temporarily adjusted during pump runs. However, the equipment installations and flov adjustments did not affect the l l
- capability of the MDAVF pumps to perform their safety function. ,
l l Therefore, this test did not constitute an unreviewed safety question !
per the criteria of 10CFR50.59.
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i NUMBER TITLE IST 3-87-006 Motor-Driven Feedvater Pump 50 Percent Trip l
Description of Test This test was performed to measure the vibration of the feedvater piping, from the discharge of the motor-driven feedvater pump up to !
the Main Steam Valve Building, during a trip, from 50 percent power, of the motor-driven feedvater pump. During implementation of the t test, the reactor was tripped manually; this reactor trip vas the beginning of a scheduled plant maintenance outage. ,
Reason for Test 1 The test was performed to complete the required testing which had not i been completed during the plant start-up. It was for data collection l purposes only, and there was no acceptance criteria.
i Safety Evaluation I- The transient caused by this test is within the bounds of a previously l analyzed accident - loss of normal feedvater. The design basis for I this accident assumes a coincident loss of offsite Alternate Current (AC) power. Performance of the test required the Engineered Safety Features Actuation System to be operable, because it caused a plant f trip, but it did not impact the operability of the system. The transient imposed upon the plant by this test was one for which the ;
plant is designed. ,.
Therefore, this test did not constitute an unrevieved safety question per the criteria of 10CFR50.59. [
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NUMBER TITLE IST 3-87-011 Containment Air Recirculation (CAR) Effects On Containment Temperatures Description of Test This test was for data acquisition only. The test systematically obtained Containment Building temperature data using various com-binations of the Containment Air Recirculation (CAR) fans.
Reason for Test This test was undertaken in an attempt to determine what affects various fan alignments had on containment area temperatures. The primary reason for the test was part of an investigation into the causes and corrective actions for temperature excursions experienced in the pressurizer cubicle.
Safety Evaluation The test utilized normal operating proceduros and controls. No abnormal fan alignments vere used.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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NUMBER TITLE IST 3-87-013 In-Service Test Plan For Spent Fuel Pool i Purification Description of Test This test completed preoperational testing for the Spent Fuel Pool Purification System. Flushing, pump testing, and fuel pool boration j vere performed. System logic vas demonstrated to be acceptable per design.
Reason for Test This test assured that during the refueling of the reactor, the Spent Fuel Pool Purification System vould function as designed. ,
1 Safety Evaluation This test utilized normal operating procedures and controls. i Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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NUMBER TITLE IST 3-87-015 Security Electrical Distribution Description of Test i
This test started the security diesel and loaded it with all security related loads. The diesel vas then run for one hour to verify that the engine could operate reliably under loaded conditions.
Reason for Test This test was performed to verify that the diesel could handle additional loads added to the security power distribution system.
Safety Evaluation This test verified that the security diesel could operate as designed in accordance with normal operating procedures. It had no effect on plant equipment.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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NUMBER TITLE Turbine 011 Pump Testing IST 3-87-016
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I j Description of Test This test consisted of performing the turbine lube oil pump veekly :
auto start testing with lov lube oil pressure turbine trip pressure ,
switch pressure, bearing header pressure and Alternating Current (AC) i- turning gear oil pump discharge pressure monitored by pressure trans-j ducer and recorded on a strip chart recorder. In addition, the l' turbine lov lube oil pressure 2/3 turbine trip vas disabled during the ,
(' test.
Reason for Test i This test was undertaken to obtain data in order to determine the
? cause of the lov lube oil pressure turbine trip which occurred on June i 14, 1987.
Safety Evaluation j
! The credible accidents associated with this test vere an inadvertent i y turbine trip while lifting and installing the lov lube oil pressure trip and turbine damage due to an actual low pressure condition during testing.
E An inadvertent turbine trip when lifting and installing the low lube :
! oil pressure trip was minimized by independent verification prior to lifting and installing the trip. l 1
Turbine damage due to lov lube oil pressure was precluded by two independent lov oil pressure trips on the hydraulic oil breaker and on the bearing oil header at the front standard. Further, Control j
- Operators monitored lube oil pressure during the test evolution with :
) instructions to trip the turbine at the lov oil pressure trip set- l i point. ,
i Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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NUMBER TITLE IST 3-87-018 Shock Chlorination Of Circulating Vater System Description of Test This test added sodium hypochlorite to the circulating vater systeu .
and monitored unit electric output for an output gain.
Reason for Test This test was performed to determine the effectiveness and safe 3 chlorination level of the main condenser. The main condenser had undergone biofouling and needed to be cleaned.
Safety Evaluation Sodium hypochlorite vas added to the circulating vater in quantities that did not exceed environmental discharge limits. The balance of the test collected data only. Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59. ,
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NUMBER TITLE i
IST 3-87-019 Emergency Diesel Generatots Air Dryers Flush ,
l Description of Test
! Equipment and piping associated with the recently installed air dryers j vere inspected, and cleaned or flushed prior to placing into operation ;
l for the first time. i I !
! Reason for Test r The test was performel in order to assure all equipment and piping was [
t adequately cleaned and prepared for initial operation. ;
Safety Evaluation j I
All equipment and piping cleaned and flushed was non-Category 1. The test was performed so it did not affect the operability of any emer-gency diesel generator. l Therefore, this test did not constitute an unreviewed safety question p,r the criteria of 10CFR50.59. :
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NUMBER TITLE IST 3-87-024 Pressurizer Heaters 86E Relay Test Description of Test Relays installed in the control circuits for Pressurizer Backup Group Heaters 'A' and 'B' vere confirmed to function properly. Heater group breaker closure and Loss of Pover (LOP) signals vere simulated along with manual control signals from Main Board 4 in order to assure the lockout relays functioned properly under all expected conditions.
Reason for Test The test was performed in order to assure relays that vere installed during the recent outage vould function as designed to prevent the Backup Heater Groups 'A' and 'B' from automatically reenergizing after a Manual Start Block (MSB) signal cleared subsequent to an LOP. The test was performed under simulated conditions to assure operability prior to the heaters actually being energized during start-up of the plant following an outage.
Safety Evaluation The test was performed on nonoperational equipment to verify relays installed in accordance with design documents functioned satisfac-4 torily prior to release for operation.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i NUMBER TITLE i l . I
- . IST 3-87-025 Refueling Cavity Seal Failure Action ;
- Verification !
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- Description of Test l
l j This test measured the time required to take'a fuel assembly from the j reactor vessel to a safe storage location, j 1
Reason for Test !
i This test was performed to ensure that a fuel assembly latched onto i the SIGMA Refueling Machine could be placed in a safe storage loca- !
! tion, in the event of a' failure of the refueling cavity seal ring, !
I before the refueling cavity would drain down to a level where the fuel i j assembly vould be exposed to air. This test was required to be (
) performed as a result of our response to NRC Bulletin 84-03.
Safety Evaluation l This test used the normal fuel handling equipment in accordance with }
normal fuel handling operating procedures. Safety interlocks vere root !
j allowed to be bypassed for this test without an approved bypass l i jumper. i 5
l Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59. ,
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NUMBER TITLE l IST 3-87-027 Vet Layup Skid Functional Test - Skid 1A Description of Test This test verified correct operation of Steam Generator Vet Layup Skid [
3BDG-SKID 1 A. This test also verified that all valves functioned }
correctly, there vas no excessive vibration or bearing temperature, 4
and that there was no leakage on the skid piping and hoses. j i
l Reason for Test '
L The vet layup skid was a new installation. i i
Safety Evaluation j
IST 3-87-027 operated the vet layup skid within its design capabili-
- ties. At its vorst this system could have been detrimental to the environment in the event that the hose connecting to the steam {
generator broke or came loose. It was determined that the boundary !
l imposed by the vet layup system was sufficient to maintain the con- ;
tainment boundary criteria while the plant was in cold shutdown. j
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This test utilized normal operating procedures and controls. [
- Therefore, this test did not constitute an unreviewed safety question !'
per the criteria of 10CFR50.59.
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NUMBER TITLE 1
IST 3-87-031 Refueling Cavity Pool Seal Leak Test i Description of Test This test determined if the refueling cavity pool seal vould ade-quately seal once a head of vater was established above the seal.
Tubing was run from the inner set of gaskets on the seal to the refueling floor. While the refueling cavity water level was raised, air leakage from the seal was monitored and a TV camera under the reactor vessel vatched for gross vater leakage. Vhen the water level reached approximately one foot above the seal, all air leakage stopped
! and a leak test per the original installation procedure was performed.
The leak test was successful and the In-Service Test terminated.
Reason for Test The inner set of gaskets on the cavity seal vould not pass the leak test required by MP 3790AN, the installation procedure for the refueling cavity pool seal.
Safety Evaluation While equipment not part of the normal operation of the pool seal vas utilized, the safety evaluation determined that failure of any of the additional equipment vould not result in an unsafe plant condition. .
Damage to other plant equipment caused by water leakage past the seal
- vas addressed and closely monitored during the test. .
Therefore, this test did not constitute an unreviewed safety question f per the criteria of 10CFR50.59. ,
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NUMBER TITLE IST 3-87-032 Emergency Diesel Generators Air Start Test Description of Test Normally, each of the two air receivers is lined up to provide air to its own independent air header. Each air header provides starting air to half of the diesel engine cylinders. In this test, one of the receivers for each diesel engine was removed from service with-the other air receiver left on line with a normally closed cross-connect valve open, thus making one receiver available to provide air to all the engine cylinders. The diesel engines were then started in order to determine starting times.
Reason for Test The test was performed in order to determine start times with just one receiver providing air to all cylinders rather than just half of the cylinders (cross-connect valve closed). The "B" diesel engine started in less than 10 seconds; the "A" diesel engine started in 10.32 seconds.
Safety Evaluation Testing was performed on engines that vere nonoperational at the time of the test. Upon completion of the test, the air systems vere returned to a standby condition per the operating procedure.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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l NUMBER TITLE IST 3-87-035 Vet Layup Skid Functional Test - Skid 1B Description of Test This test verified correct operation of_ Steam Generator Vet Layup Skid 3BDG-SKID 1B. This test also verified that all valves functioned correctly, there was no excessive vibration or bearing temperature, and that there was no leakage on the skid piping and hoses.
Reason for Test The vet layup skid was a new installation.
Safety Evaluation IST 3-87-035 operated the vet layup skid within its design capabili-ties. At its vorst this system could have been detrimental to the environment in the event that the hose connecting to the steam generator broke or came loose. It was determined that the boundary imposed by the vet layup system was sufficient to maintain the containment boundary criteria while the plant was in cold shutdown.
This test utilized normal operating procedures and controls.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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i NUMBER TITLE IST 3-87-036 Vet Layup Skid Functional Test - Skid 1C Description of Test i This test verified correct operation of Steam Generator Vet Layup Skid '
3BDG-SKID 10. This test also verified that all valves functioned correctly, there was no excessive vibration or bearing temperature, and that there was no leakage on the skid piping and hoses.
Reason for Test The vet layup skid was a new installation.
Safety Evaluation IST 3-87-036 operated the vet layup skid within its design capabili-ties. At its vorst this system could have been detrimental to the environment in the event that the hose connecting to the steam generator broke or came loose. It was determined that the boundary imposed by the vet layup system was sufficient to maintain the con-tainment boundary criteria while the plant was in cold shutdown.
This test utilized normal operating procedures and controls.
Therefore, this test did not constitute an unreviewed safety question per the criteria of 10CFR50.59.
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.J EXPERIMENTS 2
There' vere no. experiments performed under.the provisions of Title 10, Code of Federal-Regulations, Section 50.59, during 1987. q T -
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CHALLENGES TO RELIEF VALVES In accordance with the commitment made under Item II.K.3.3 of NUREG 0737 (TMI Action Plan) in.the W. G. Council letter to D. G. Eisenhut, dated June 10, 1980, the following is.a report of challenges to relief / safety valves during 1987.
There were no challenges to relief / safety valves during 1987.
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