ML20078C547
| ML20078C547 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 04/01/1994 |
| From: | Mcmeekin T DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9410310322 | |
| Download: ML20078C547 (100) | |
Text
.
Il I
DukeIbwer Company T.C hiotta a kicGuire Nuclear Generation Depanment Vice hesident 12700Hagers fenyRoad(htG01\\P)
(T&l)8754800 Huntenodle, NC280iM985 (Tr>f)8754809 Fax DLfKEPOWER October 20, 1994 U.
S. Nuclear Regulatory Commission Document Control Desk Washington, D.C.
20555
Subject:
McGuire Nuclear Station Docket Nos. 50-369 and 50-370 Pursuant to 10 CFR 50.59, please find attached a summary of Nuclear Station Modifications, Minor Modifications, Procedure Changes and miscellaneous changes made to the McGuire Nuclear Station for the period of April 1, 1993 to April 1, 1994.
Questions or problems should be directed to Kay Crane, Regulatory Compliance at (704) 875-4306.
Very truly yours, y
{.
T.
C. McMeekin, Vice President McGuire Nuclear Station Attachment cc:
Mr. Victor Nerses, Project Manager Office of Nuclear Reactor Regulation U.
S. Nuclear Regulatory Commission Washington, D.
C.
20555 1
Mr.
S.
D.
Ebneter, Regional Administrator l
U.
S.
Nuclear Regulat.ory Commission Region II 101 Marietta Street, NW - Suite 2900 Atlanta, Georgia 30323 Mr. George Maxwell Senior Resident Inspector McGuire Nuclear Stativ.a Q ".. f ' :
u: ua 9410310322 940401 PDR ADOCK 05000369 N
R e
PDR V
/,
,w w m v me r
l L
Duke Power Company McGuire Nuclear Station Summary of Nuclear Station Modifications Completed Under 10CFR50.59 i
NSM-12322
==
Description:==
This modification replaces the main steamline drain level switches with a Hydratect electronic control system. The main steamline drain level switches, SMLS5860, 5870, 5880, and 5890, are used to automatically open or close valves SM78, 84, 90 and 96.
The existing level switches do not reset properly.
This causes their associated train valve to remain open, allowing excessive steam bloudown to the condenser.
The Hydratect system consists of two vendor-supplied monitor panels and 4 electrode pairs, used to measure conductivity. Two electrode pairs will be wired to each monitor panel.
Each monitor panel will monitor two steamline drains, with one electrode pair (two electrodes) used on each drain line. Each electrode pair will be welded directly in a 2" pipe which will be installed in place of the existing level switch.
The existing root valves will be utilized, and their locations will remain the same.
When the water level in the drain line reaches the top electrode, the electrode senses a conductivity change.
The monitor panel will cause the respective steamline drain valve to open and an alarm to be activated on the Operator Aid Computer (OAC) in the Control Room.
The bottom electrode will be used for verification.
When the water in the drain line is removed (drained to the condenser) the system will automatically reset and the drain valves will close.
The existing low drain level alarm is not changed.
One electronics unit will be located in the interior doghouse, the other in the exterior doghouse. The Hydratect systen is non-safety related. The drain lines are non-safety related downstream of the Steam Line Drain Isolation Valves. The affected piping is Class G.
Safety Review and USQ Evaluation:
The potential exists, with the current design, for a Reactor or Turbine Trip due to simultaneous opening of 2 or more steamline drain valves. A failure within the added monitoring system could potentially cause 2 valves to open simultaneously, however, the main steamline drain valves are expected to operate (open and close when desired) more reliably with this modification.
The Hydratect monitoring system has proven reliable in a similar application at another Duke plant.
The modified Class G piping has not been degraded.
An Appendix R review was conducted with no concerns identified.
Therefore, the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
The main steamline drains are not accident mitigators. They do not perform plant safety functions, and the performance of safety functions are not degraded by this modification.
The steamline drains will continue to perform the same function. A potential common failure of 2 drain valve simultaneously to the OPEN position was reviewed and determined to be bounded by existing failures.
Therefore, the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR is not increased.
l No new failure modes are created. System faults or power failure may result in one or more drain valves failing OPEN, which is an existing potential failure mode. No accidents previously thought incredible are made credible by this NSM.
Thus the possibility of an accident or malfunction of equipment of a different 1
j
type than evaluated in the SAR will not be created.
There are no changes of safety limits, setpoints, or plant parameters because of this modification.
The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSM.
Therefore, the margin of safety as defined in the basis for any Technical Specification is not decreased.
No USQ exists.
NSM-52362
==
Description:==
The subject modification is for replacing the Computer Room Chillt'r "A" (tag # SB-C-1) due to corrosion-induced failure.
The replacement is an air-cooled York chiller.
The Computer Room HVAC system has redundant chillers, "A"
and "B"
(not Safety I
Trains A and B).
NSM MG-52312 was used to purchase and install a replacement for chiller "B"
(tag # SB-C-2) and condensers for both chillers.
Since later replacement of chiller A was anticipated, it was desirable to replace all condensers on the Service Building roof at the same time.
The new chiller, which includes the evaporator, compressors, expansion valves and control panel, will be mounted in the same location as the existing chiller, utilizing the same mounting pad. The new chiller is smaller than the old chiller in both size and weight.
The replacement HVAC equipment was sized on NSM MG-52312.
The subject modification (MG-52362) requires one new roof penetration for refrigerant lines.
The added lines will be class H, non-QA, non-seismic.
RN system piping, which provides cooling water f rom a non-essential RN header to the exinting chiller, will be cut and capped.
This modification, along with the previous replacement of chiller B, makes 264 gpm of RN water available for other plant uses.
Chilled water (YC) piping within the Computer Room Chiller system, from the evaporator to the condensers, will be modified to accommodate the new chiller.
A new transformer and cabling are required to furnish power for the new chiller.
A non-essential power source will be used.
Safety Review and USQ Evaluation:
The Computer Room HVAC system has no applicable QA condition. The Operator Aid J
Computer and the turbine control systems located in the computer room have no applicable QA condition.
Redundant chillers and condensers are employed for computer room cooling because a high degree of reliability is desirable for computer-type equipment, due to safety-related requirements for redundancy.
The Computer Room is located in the Service Building. The Service Building is a non-seismically qualified structure.
Therefore, no seismic qualification criteria applies to Computer Room equipment.
The function of the Computer Room HVAC system is to maintain room temperature within temperature specifications for computer-type equipment. Although this is not a nuclear safety-related function, malfunction of tha (non-safety related) turbine control equipment in the Computer Room, due to overheating, could cause a Turbine Trip. Turbine Trips are evaluated as ANS Condition 2 faults in FSAR Chapter 15.0.
This modification replaces the failed chiller, maintaining the j
design basis of the Computer Room HVAC system.
The replacement chiller is j
expected to be more reliable than the existing chiller.
Therefore, the probability of a Turbine Trip is not increased by the modification.
Chilled 2
l
Water (YC) system functions are not af fected by tM g modification. The modified portion is non-safety, internal to the Computer Room HVAC system, and not related to the Control Room Ventilation System (VC).
The Nuclear Service Water (RN) system supply that will be removed is from a non-essential RN header. Chillers A and B will be powered from separate load centers to maintain power supply diversity. This is desirable for computer room HVAC reliability.
However, no safety-related equipment is involved, therefore, it is not a nuclear safety requirement. Electrical separation criteria is not applicable. An Appendix R review was performed, with no concerns identified. Therefore,.the probability of an accident or malfunction of equipment important to safety previously evaluated in the SAR is not increased.
The Computer Room HVAC system performs no accident sitigation functions nor plant safety functions. The new chiller will perform the sa.ne function as the existing chiller. The new equipment will not interact with pitint equipment such that the performance of safety functions are degraded.
No common failure modes are created.
Therefore, the consequences of an accident of a malfunction of equipment important to safety evaluated in the SAR is not increased.
No new system functions are added or failure modes created by the chiller replacement. No accidents previously considered incredible are made credible by these NSMs.
Seismic and separation criteria are not applicable.
Thus the possibility of an accident or malfunction of equipment of a different type than any evaluated in the SAR will not be created.
There are no changes of safety limits, setpoints, or plant parameters because of the modifications.
The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSMs.
Therefore the margin of safety as defined in the basis for any Technical Specification is not decreased.
No USQ exists.
NSM - 12410, 22410
==
Description:==
This Urgent Modification is expected to improve the reliability of the Steam Generators Feedwater Control Valves (FCV), CFl7, 20, 23, and 32.
These valves, located in the Turbine Building, are modulated by a three-element feedwater control system using feedwater flow, steam generator water level, and steam flow as parameters.
Presently, two solenoid valves (Train A and Train B) in series close the respective FCV by a signal from either Train A or Train B.
These NSMs will add redundant solenoid valves of the same train in parallel.
The new solenoid valves will be Valcor Model V70900-65, under Minor Modification MM-5127 (for Unit 1) and under NSM MG-22410 for Unit 2.
Safety Review and USQ Evaluation:
The Feedwater System (CF) supplies feedwater to the four steam generators (per unit) at the temperature, pressure, and flowrate required to maintain proper vessel water levels with respect to reactor power output and turbine steam requirements.
In so doing, the CF System is also required to 1) provide feedwater isolation to prevent over-pressure of the containment in the event of a feedwater or main steam pipe rupture 2) preclude blowdown of all steam generators following a feedwater or main steam line break 3) provide containment isolation when required and 4) prevent overcooling on a steamline break. As part of this scheme the FCVs are closed on a safety injection signal.
The FCVs require air to open and they fail closed upon loss of air or loss of power to the. lenoid valve circuit.
In the original control scheme, a single circuit or so' 1oid coil failure causes the associated FCV to fail closed which 3
results in a unit trip due to low steam generator level.
In the new control scheme, if one solenoid valve fails the other solenoid valves keep the control pressure to the FCV.
Also, the solenoid valves associated with a single FCV and of the same train are placed on separate circuits to improve electrical reliability. Consideration has also been given to Main Feedwater flow isolation on high doghouse water level (FSAR Section 7.6.19).
Control valves CF20 and 23 are connected to the feedwater lines that extend through the interior doghouse while CF17 and 32 are associated with the lines that run through the exterior doghouse.
Doghouse high water level is provided as a trip for every FCV; i
therefore, the valves repaired electrically according to their piping connections.
The new solenoid valves are considered more reliable than the
{
existing solenoid valves.
The new tubing, although not safety related, is comparable to the present design. The power supply and breaker arrangement does not introduce any new or common failure modes. Based on this, the new solenoid arrangement is functionally as capable as before of closing the FCVs under all required conditions.
The FCVs are accident initiators as defined in FSAR 15.1.2.
This accident originates from a malfunction in the feedwater control system.
As discussed j
above, the FCVs will essentially perform functionally as before. There response time is not adversely affected.
Therefore, there is no increase in the probability of an accident evaluated in the SAR.
Appendix R review is complete with no concerns identified.
Train separation has ben considered.
No new or common failure modes have been identified.
Based on the above, there is no creation of the possibility for an accident of a different type than any evaluated in the SAR.
The FCVs also perform the accident mitigation function of isolating feedwater flow to the steam generators when called upon by a feedwater isolation signal, which is an Engineered Safety Feature.
The valves will respond within the Technical Specification limit of 12 seconds total valve response time. B: sed on this and the reasoning in the paragraph above, there is no increase in the consequences of an accident or of a malfunction of equipment important to safety l
evaluated in the SAR.
i Using the reasoning discussed previously, the system will perform as designed, respond within Technical Specification limits, and no new or common f ailure modes were discerned.
Therefore, there is no increase in the probability of=a malfunction of equipment important to safety evaluated in the SAR. Likewise, the possibility for a malfunction of a different type than any evaluated in the SAR is not created.
No safety limit, setpoint, or operating parameter will be changed by these NSMs; therefore, the margin of safety as defined in the basis of the Technical Specifications will not be reduced. No USQ exists.
NSM - 12269, 22269
==
Description:==
The Spent Fuel Cooling System (KF) removes decay heat from fuel stored in the i
unit Spent Fuel Pool and maintains water quality and clarity for pool operations.
The cooling capability which is considered safety related is provided by redundant full capacity cooling train for all active cooling components.
The system is designed to prevent draining of the pool below acceptable levels by either active or passive failure.
The modifications add manual throttling valves 1KF155 and 1KF165 (2KF155 and 2KF156 on Unit 2) to the 8 inch Spent Fuel Cooling System (KF) piping downstream of the KF. heat exchangers.
Four inch valve 1KF157 (2KF157) is added downstream of the KF demineralizer to provide throttling for this piping.
The throttling function was previously provided by valves IKF21, 1KF40, and 1KF77 and 4
i 1
i
corresponding Unit 2 valves which also serve as isolation valves for maintenance or shutting down one of the redundant trains of cooling. The use of the existing valves for throttling has led to excessive degradation of their ability to provide isolation.
This modification will allow the current valves to be dedicated to isolation while the added valves will provide throttling capability.
The 8 inch and 4 inch piping is being rerouted to accommodate the addition of the new valves and the pressure relief lines to bypass the current valves is rerouted to include bypass of the added valves.
Power to the motor operator for valve 1KF77 and 2KF77 is being disconnected. These valves will be operated manually as will all new valves. The 8 inch cooling lines are QA-1, Duke Class C while the 4 inch purification line is QA-2, Duke Class E.
All new piping and valves were specified to the appropriate quality level for the application.
Piping Analysis end Pipe Support Design was revised to qualify the revised piping layout.
Flow calculations have been revised to assure that the functional capability of the system is not degraded by this modification.
The valve additions and removal of remote operation for valves IKF77 and 2KF77 does not change the system function or design requirements for the system. No functional requirements are adversely impacted by this modification.
Safety Review and USQ Evaluation:
The addition of the throttling valves to the KF system will not affect any accident evaluated in the SAR.
Operation of this system is not required for mitigation of an accidents discussed in the FSAR.
Operation of this system is not required for mitigation of any accidents discussed in the SAR.
Failure of any of these valves would not prevent the KF system from performing its principle safety functi7n of providing cooling capability for the Spent Fuel Pool.
Therefore, these modifications will not increase the probability or consequences of any accident evaluated in the SAR.
The function and operation of the system is not changed so there is no new accident, not previously reviewed, created by this modification.
The addition of these valves increases the reliability of the KF System since 1KF21, 1KF40, and 1KF77 and the corresponding Unit 2 valves may be dedicated to providing isolation when required and will not be degraded by use as throttle l
valves.
Therefore, the probability of an equipment malfunction or the consequences of a malfunction is not increased.
The addition of these valves j
along with the valves currently in place, does not create the possibility of a malfunction different than any evaluated in the SAR.
Margin of safety is related to the confidence in the fission product barriers to f unction as designed. This NSM does not degrade any fission product barriers or
{
af fect any assumptions in the accident analyses. No safety limits, setpoints or j
limiting safety system settings are affected. Therefore, the margin of safety i
as defined in the basis to the Technical Specifications is not reduced. No USQ exists.
NSM - 12301, 22301
==
Description:==
These modifications will install additional pressure monitoring instruments, which will monitor wide range annulus pressure for the annulus ventilation system. These instruments will trip the exhausting fan should annulus pressure exceed a low-low setpoint
(-7.0" w.g.)
and the dampers are in the exhaust alignment. Without these new instruments, should the existing instruments fail l
in such a way that the system exhausts continuously, then annulus pressure could reach a vacuum as low as 22" w.g., which could cause deformation of the VP and VE ductwork.
Two computer alarms are also add (Low and Low-Low Annulus Pressure).
5
l Safety Review and USQ Evaluation:
The Annulus ventilation (VE) System is affected by this modifictition.
1 The annulus ventilation system provides a negative pressure zone between the steel containment structure and the concrete reactor building wall. The annulus ventilation system creates a negative pressure by exhausting air through the unit vent.
This negative pressure is monitored by transmitters that regulate the vacuum between -0.5 and -3.5" w.g.
Filtration of the air within the annulus is required to ensure that the thyroid and whole body doses at the exclusion area boundary and the low population zone are within the limits of 10 CFR 100 i
following a severe loss of coolant accident with substantial core melt.
The installed instrumentation will be QA condition 1 and seismically mounted.
The VE System is not an accident initiator.
Therefore, the probability of an accident evaluated in the FSAR is not increased.
The VE System is an accident mitigation system.
Installation of the new instrumentation will not introduce any common mode failures into the system. The VE System will be less likely to fail due to high vacuum (VE f an continuously on/ Exhaust damper continuously open).
The VE System continues to meet single f ailure criteria. Failure of one instrumentation loop will still allow operation of the other VE train. A 10CFR50 Appendix R review revealed no safety concerns.
Therefore, the probability of a malfunction of equipment important to safety is not increased.
No accident analysis assumptions are altered.
No new failure modes are introduced.
The radiological consequences of any Chapter 15 FSAR accident remain the same.
Therefore, the consequences of an accident or a malfunction of equipment important to safety are not increased.
No operating parameters are changed. The function of the VE System will remain the same.
No new failure modes are introduced.
No previously incredible accidents are made credible.
Therefore, the possibility of an accident or a malfunction of equipment important to safety dif ferent from any evaluated in the SAR is not increased.
No fission product barriers are affected adversely.
No plant safety limits, limiting safety setpoints, or design parameters are affected.
No Technical Specification changes are required. Therefore, the margin of safety as defined in the bases to the Technical Specifications will not be reduced. No USQ exists.
NSM-22413 Description 1
The charging and letdown functions of the Chemical and Volume Control System (NV) are employed to maintain a predetermined water level in the Reactor Coolant
)
System (NC) pressurizer, thus maintaining proper reactor coolant inventory during I
all phases of unit operation. This is achievad by means of a continuous feed and bleed process during this the feed rate is automatically controlled based on pressurizer water level.
The bleed rate can be chosen to suit various unit operational requirements by selecting the proper combinations of letdown orifices in the letdown flow path.
1 Two letdown orifices are arranged in parallel to reduce the pressure of the letdown steam to a valm, compatible with the letdown heat exchanger design. One of the orifices is sized to pass normal letdown flow; the other can pass less than the normal letdown flow.
A third letdown path is provided via a control valve.
Any combination of the orifices and control valve can be utilized in order to increase letdown flow such as during reactor heatup operations and maximum purification. This arrangement also provides a full capacity for control of letdown flow.
The two orifices are replaced in and taken out of service by G
remote manual operation of their respective isolation valves. The control valve is also controlled by remote manual operation. A low pressure letdown controller controls the pressure downstream of the letdown heat exchanger to prevent flashing of the letdown liquid.
Orifice 2NVFE6200 is sized to pass the normal flow of 75 gpm. Orifice 2NVFE6210 is sized to pass the normal flow of 45 gpm. This NSM will replace these orifices with new orifices that have been redesigned to pass the same flows at the same pressure drops, but with reduced cavitation.
In addition, piping upstream and downstream will be replaced, including a thermal well 92NVTW5110). This piping will be raised off the floor to allow adequate clearance for welding.
Where possible, all socket welds will be replaced with butt welds. Each of the three flowpaths has a thermal expansion loop. Two pieces of properly sized pipe will be welded together to make new thermal expansion loops and reduce the number of required welds.
All piping is austenitic stainless steel and all joints and connections will be welded. The function and operation of this part of the NV System is unchanged.
Safety Review and USQ Evaluation:
Since the components are designed in accordance with applicable codes with due consideration for the design and operating conditions, the materials are either austenitic stainless steel or a material compatible with reactor coolant, and joints and connections are welded, the structural integrity of the system is maintained. The piping and orifices are still Duke Class B.
Support / restraints have been evaluated. Since the structural integrity of the system is unaffected and no other piping is adversely affected, no loss of coolant accident is affected.
The system will function and be operated as before.
Based on the above, reactor coolant inventory and boron concentration are not affected.
Therefore, there is no increase in the probability or consequences of an accident evaluated in the SAR.
Since NV System parameters are not changed and the integrity of the system remains intact, there is no increase in the probability or consequences of a malfunction of equipment important to safety evaluated in the SAR.
No new failure modes have been identified.
Based on this and the above the possibility for an accident of a different type or a malfunction of a different type than evaluated in the SAR is not created.
No safety limit, setpoint, or operating parameter will be changed by these modifications. Therefore, the margin of safety as defined in the basis of the Technical Specifications will not be reduced.
No USQ exists.
NSM-42416
==
Description:==
The demineralized water (YM) system provides filter demineralized water to the upper surge tanks (CM system) for makeup and to other systems throughout the plant that require high quality water. For several years, a temporary filtering system has been used to ensure the quality of this makeup water.
The purpose of this modification is to prepare the location for the permanent YM filtering system which will replace the temporary one. A permanent YM Processing Building (pre-fabric 4ted metal building) will be located on the south side of the Auxiliary Electric Boiler Room with permanent electrical power for the building needs.
The electrical supply for the Reverse Osmosis (RO) skid and the recirculation pumps will remain as temporary power until the building is erected.
The piping will remain as temporary hoses or PVC piping located near the skid.
The RO skids and filter tanks will be placed in the building, thereby, eliminating the need for the temporary mobile processing trailer.
This safety evaluation considers the preparatory work that must be completed prior to 7
f l
actually erecting the building and considers the following 1.
Remove the polishing oven and delete the cable and cable trays to the oven.
2.
Relocate the security light on the outside of the Auxiliary Electric Boiler Room. The relocated lights and any new lights will be installed in corapliance with Security requirements.
3.
Provide temporary power to the RO Skip Unit to ensure no interruption in power to the unit.
The temporary cable that goes to it now is routed through the Poliching Oven and is being removed.
4.
Relocate the emergency shower and eyewash into the transmission yard along with its associated heat tracing. The relocated shower and eyewash will remain within 25 feet of any chemical tanks to comply with regulatory requirements.
5.
Cut and cap the YD line from the shipping dock that originally supplied YD to the relocated emergency shower and eyewash and provide a core drill and YD piping for another emergency - shower to be located in the new YM Processing Building.
6.
Provide a three foot by three foot painted concrete pad to be placed under the relocated emergency shower and eyewash.
Use safety yellow for the paint color. In addition, paint the curb on the Chemical Addition Pad in the Processing Building with black and yellow paint to show a tripping hazard.
7.
Remove the wing wall from the Auxiliary Electric Boiler Room wall.
8.
Relocate the damper from its current position on the Auxiliary Electric Boiler Room wall to a position higher up on the wall to clear the new YM Processing Building.
9.
Install pipe sleeves in the hole left by the relocated damper and fill in around the sleeves.
Safety Review and USQ Evaluations The YM system is not an accident initiator in any FSAR Chapter 15 accident analyses. The only safety related system that the YM system interfaces with is the control Room Area HVAC Chilled Water System. The YM system provides makeup water for this system. This modification will not change any tie-ins to other systems.
This modification is only preparing the area for the permanent YM Processing Building that is going to be built.
The modifications to the YD system, the electrical system, and the Auxiliary Electric Boiler Room do not involve any safety systems of IE power.
The changes to the security lighting will not result in a degradation to the Security Plan. The interferences listed above will not increase the probability of any accident or malfunction of equipment important to safety previously evaluated in the SAR.
The YM system performs the function of providing high quality water to various systems throughout the plant. This modification will not result in a degradation of the performance of the YM system or any system that it provides purified water to.
The modification is only preparing the area for the permanent YM Processing Building. Therefore, the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR are not increased.
The modification adds no new functions and no new failure modes are introduced.
No accidents previously considered incredible are made credible by this 8
modification.
Therefore, the possibility of an accident or malfunction of equipment of a different type than evaluated in the SAR will not be created Since the modification does not system that could affect a fission barrier, interact with any fission product barrier or a it will not result in a degradation to such barriers (RCS pressure boundary, containment, fuel pellets, and fuel cladding).
No assumptions made in any accident analysis are affected by the modification.
Technical Specification is not decreased.Therefore, the margin of safety as defined in the basis fo No USQ exists.
NSM - 12356, 22356
==
Description:==
These modifications provide additional lighting to improve visibility for personnel working in selected Reactor and Auxiliary Building areas.
will be added as follows:
New lighting Quantity of Fixtures Per Plant Unit Locations Reactor Building:
4 Pressurizer cavity 12 2 on the crane wall side of each Reactor Coolant Pump (NCP) at the lateral supports.
1 below each pump motor to illuminate the seal area.
8 2 at the top of each Accumulator Room.
6 3 at the top of each VL Fan Room.
8 2 under the steam generator platforms at the bottom of each.
Auxiliary Building:
3 Elevation 716' between columns EE-FF and 52-53, inside the
" midget hole".
Outboard Doghouse:
1 Elevation 767' in the outboard doghouse.
Safety Review and USQ Evaluation:
The lighting systems are not accident initiators in any accident analyses. The new light fixtures are added for personnel safety, but are not nuclear safety related.
The fixtures are mounted in accordance with QA Condition 4 (seismic) requirements.
Applicable electrical separation criteria was used for added wiring to supply power for the lights. An Appendix R review was conducted with no concerns identified. Therefore the probability of an accident or malfunction of equipment important to safety previously evaluated in the SAR is not increased.
The modifications add lighting to the normal lighting systems.
These systems perform no accident mitigation functions.
The modifications do not affect the 9
e modification.
Therefore, the possibility of an accident or malfunction of equipment of a different type than evaluated in the SAR will not be created.
Since the modification does not interact with any fission product barrier or a system chat could affect a fission barrier, it will not result in a degradation to suca barriers (RCS pressure boundary, containment, fuel pellets, and fuel cladding).
No assumptions made in any accident analysis are affected by the modification. Therefore, the margin of safety as defined in the basis for any Technical Specification is not decreased. No USQ exists.
i NSM - 12356, 22356
==
Description:==
These modifications provide additional lighting to improve visibility for personnel working in selected Reactor and Auxiliary Building areas. New lighting will be added as follows:
Quantity Of Fixtures Per Plant 1: nit Locations Reactor Building:
4 Pressurizer cavity
)
12 2 on the crane wall side of each Reactor Coolant Pump (NCP) at the lateral supports.
1 below each pump motor to illuminate the ceal area.
8 2 at the top of each Accumulator Room.
6 3 at the top of each VL Fan Room.
8 2 under the steam generator platforms at the bottom of each.
Auxiliary Buildings i
3 Elevation 716' between columns EE-FF and 52-53, inside the
" midget hole".
Outboard Doghouse:
1 Elevation 767' in the outboard doghouse.
1 Safety Review and USQ Evaluation:
The lighting systems are not accident initiators in any accident analyses. The new light fixtures are added for personnel safety, but are not nuclear safety related. The fixtures are mounted in accordance with QA Condition 4 (seismic) requirements.
Applicable electrical separation criteria was used for added wiring to supply power for the lights. An Appendix R review was conducted with no concerns identified. Therefore the probability of an accident or malfunction of equipment important to safety previously evaluated in the SAR is not increased.
The modifications add lighting to the normal lighting systems.
These systems perform no accident mitigation functions.
The modifications do not affect the 9
l
emergency lighting systems.
The lighting addition will not degrade any plant safety functions.
There are no common failure modes created.
Therefore, the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR are not increased.
There are no new functions added, nor any new failure modes created.
No accidents previously considered incredible are made credible by these NSMs. The additional lighting is equivalent in specifications to the existing lighting, and 1
is appropriate for use in the areas where added.
Thus the possibility of an accident or malfunction of equipment of a different type than evaluated in the i
SAR will not be created.
There are no changes of safety limits, setpoints, or plant parameters because of the modifications.
The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSMs.
Therefore, the margin of safety as defined in the basis for any Technical Specification is not decreased.
No USQ exists.
l NSM - 12293, 22293
==
Description:==
This modification changes out the in-containment signal cables for radiation monitors EMF 51A and EMF 51B to coaxial type, and routes the cables in sealed conduit.
The cables will be routed through new triax feed-throughs to be used with the existing Conax containment penetrations. The D.G. O'Brien containment penetrations, presently used for signal and power cables, will only be used for EMF 51A and EMF 51B high voltage connections.
This change is being made to resolve potential concerns over the degradation of signal cables for the Containment High Range Radiation Monitors during Design Basis Accident conditions.
The modifications will allow the post accident readings frcm the radiation monitors to remain within the accuracy recommended by Regulatory Guide 1.97, Revision 2.
Radiation monitors EMF 51A and EMF 51B, the new coaxial cables and conduit, and the containment penetrations are QA condition 1.
The new conduit will be installed according to approved QA Condition 4 (seismic) mounting requirement s.
An Appendix R review was conducted, with no concerns identified.
Safety Review and USQ Evaluation:
The radiation monitors and containment penetrations are not accident initiators in any FSAR Chapter 15 accident analyses.
There are no control functions performed by the radiation monitors or containment penetrations. The new conduit will be mounted per QA condition 4 requirements.
An Appendix R review was conducted.
The new cabling, conduit, and penetration feed-throughs are QA condition 1 qualified (which includes environmental qualification for the containment environment).
Cabling will be installed according to applicable train separation criteria.
Therefore, the probability of an accident or malfunction of equipment important to safety previously evaluated in the SAR is not increased.
The radiation monitors and containment penetrations will continue to perform the same functions.
The radiation monitors provide safety-related pre-and post-accident indication of radiation levels inside containment.
The containment penetrations allow cable penetration into containment while serving to maintain containment integrity (to prevent the release of radioactivity). The radiation monitors and containment penetrations will not interact with other equipment in 10
any way such that the performance of any plant safety functions are degraded.
Applicable safety train separation criteria will be followed for cable installation, so that no common failure modes are created.
Therefore, the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR are not increased.
No new failure modes are created.
The new Conax triax feed-throughs are QA Condition 1 qualified for in-containment use.
The post-accident radiation monitor readings are expected to be more accurate after the modifications, providing the plant operators with more reliable post-accident in-containment radiation readings.
The new conduit will be installed per QA Condition 4 requirements. No accidents previously considered incredible are made credible by these NSMs. Thus the possibility of an accident or malfunction of equipment of a different type than evaluated in the SAR will not be created.
No technical specifications or plant parameters are affected by this modification.
Existing Conax cable penetrations are utilized, with new QA condition 1 qualified Conax triax feed-throughs. There are no changed of safety limits, setpoints, or plant parameters because of the modifications. The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSMs.
Therefore, the margin of safety as defined in the basis for any Technical Specification is not decreased.
No USQ exists.
NSM-12398, 22398
==
Description:==
These NSMs revise the Containment Spray (NS) systems to add a 10 inch check valve in the vertical piping of each train of the NS system just above floor elevation 750' +0 and down stream of the NS pumps. The check valves, INS 83 through IND88 and 2NS110 through 2NS115, currently used as vacuum breakers, will be deleted by these modifications. Low point drains will be deleted by these modifications.
Also, low point drains will be added inside containment to drain any accumulated water between the inside containment isolation check valves and the spray nozzles.
Tees will be added to the piping just upstream of the added check valves to allow water in the portion of the spray header downstream of the inside containment isolation valve to drain out following a pumps top.
This will aid in preventing water hammer on pump restart.
The added check valves do not act as containment isolation valves. Containment isolation is provided by a closed chock valve inside containment and a closed system outside containment.
The check valves used to provide a vacuum breaker e
will be deleted, thus ensuring a closed system.
The NS system does not operate during normal plant operation.
This system is used only during accident conditions with containment pressure above.25 psig.
The operation of the NS system may be manually controlled or controlled by the Containment Pressure Control System as specified in Table 3.3-4 Item 6 of the plant Technical Specifications.
There are no changes to the CPCS system as a result of these modifications.
The added 10 inch check valves act to prevent backflow of water in the NS spray header during periods af ter initial operation of the NS system when the pumps are shut off.
Should backflow occur with the revised piping configuration, there is potential for waterhammer in the header pipe due to a vacuum forming between the check valve inside containment and the water head.
This was previously prevented by allowing air to enter the pipe through check valves INS 83 through INS 88 and 2NS110 through 2NS115. The water hammer will now be prevented by maintaining the water column above the added 10 inch check valves ar.d draining the water in the spray header inside containment.
The Containment Pressure Control System will not require modification as a result of the valve additions and will function as before to help maintain containment 11 j
i l
i
pressure within acceptable limits. The NS system will not change its function or operating parameters as described in 6.5.1 through 6.5.5 of the FSAR due to these modifications.
None of the changes to the piping or valve additions will significantly affect flow or pressure requirements for the system. There is no change in signals to start or stop the system or in any automatic valve controls as a result of these modifications. The system flow test may be modified and new inspection criteria may be added to ensure the operability of the system and each component, but these will not be a detrimental affect on the system.
These
+
changes will be made as needed to meet the current functional testing requirements in an efficient cost effective manner.
Safety Review and USQ Evaluation:
The NS system is designed to reduce Containment pressure following an accident which results in a large steam release inside containment and also serves to reduce airborne activity in containment by capturing fission products and containing them in the sump liquid.
The system does not operate in any normal modes except for testing as required by Technical Specification 3/4.6.2.
Therefore, the NS system is not an accident initiator.
These modifications do not change any operating modes for NS and thus there is no increase in the possibility of an accident evaluated in the SAR or creation of an accident different than those evaluated in the SAR as a result of these changes.
Since the operation of the system and components in an accident is unchanged, there is no increase in the consequence of an accident or the probability of a malfunction of equipment as evaluated in the SAR.
The modifications primarily change the method of preventing water hammer in the header piping on NS pump restart. The operating parameters are unchanged, therefore, there is no possibility for creation of a malfunction different than those evaluated in the SAR.
The NS system works in accident mitigation by reducing the pressure on the containment structure and reducing the airborne activity available for release from containment. The containment structure is a fission product barrier. Since the system operation is unchanged as a result of this modification, there is no reduction in the margin of safety as defined in the Technical Specifications for these modifications.
No USQ exists.
NSM-12266, 22266
==
Description:==
These modifications change load center (LC) and motor control center (MCC) breakers on the 4 KV power system, and upgrade several power cables.
The modifications are QA Condition 1.
These modifications are being made in response to an analytical model review of the 4KV power system, which was conducted to review the system because of changes made since its original design.
NSM MG-12266 modifies the Unit 1 4KV power system, and NSM MG-22266 modifies the Unit 2 4KV power system.
Summary of changes:
1.
Replace trip units on incoming breakers in MCCs lEMXH, lEMXG, 2EMXH, and 2EMXG, with units of higher amperage.
2.
Pull new cables to MCCs 1EMXH, IEMXG, 2EMXH, and 2EMXG to upgrade for higher amperage.
3.
Revise settings on breakers in LCs 1(2) ELXA, 1(2) ELXB, 1(2) ELXC, and 1(2)ELXD, that supply power to MCCs lEMXH, lEMXG, 2EMXH, and 2EMXG.
4.
Revise incoming breaker settings for LCs lELXA, lELXB, IELXC, lELXD, 2ELXA, 2ELXB, 2ELXC, 2ELXD.
5.
Revise the settings for the 32DGT and 59DGN relays on Unit 1 and 2.
12
6.
Pull new cables to the Unit 1 and 2 backup pressurizer heater power panels i
to upgrade for higher amperage.
Safety Review and USQ Evaluation:
I The 4 KV power system is QA Condition 1, however, the syntem is not an accident initiator.
The system is not addressed in plant accident analyses (McGuire Nuclear Station FSAR, Chapter 15). A seismic qualification review was determined to be unnecessary because breaker trip units in MCCs 1(2) EMXH and 192) EMXG were replaced with equivalent units having a higher amperage rating.
An Appendix R review was conducted, with no concerns identified.
Therefore, the probability of accidente previously addressed in the FSAR is not increased.
Cable and breaker sizing, as well as breaker coordination, for the af fected LCs, MCCs, and power panels have been considered.
Breaker trip units used for replacement in McCs 1(2) EMXH and 1(2) EMXG were ordered with 10CFR50, Part 21 dedication provided by the manuf acturer. Thus, the probability of a malfunction of equipment important to safety previously evaluated in the SAR is not increased, and no new malfunctiens are created.
No accidents previously thought incredible are made credible by this NSM, thus the possibility of an accident of a dif ferent type than evaluated in the SAR will not be created.
The 4KV power system supplies power for equipment used for accident mitigation.
The modification does not change the function or operation of the 4KV power system.
The NSM introduces no common failure modes.
The ability of the 4KV power system to perform its safety function is not degraded.
Therefore, the consequences of an accident or malfunction of equipment important to safety evaluated in the FSAR is not increased.
There are no changes of safety limits, setpoints, or plant parameters because of this modification. No assumptions made in any accident analysis are af fected by this NSM.
Therefore, the margin of safety as defined in the bases for any Technical Specification is not increased. No USQ exists.
NSM-12096, 22096
==
Description:==
A new Loose Parts Monitoring System (LPMS) will be installed for each plant unit by several modification packages over several refueling outages.
The new LPMS will replace existing LPMS which are outdated.
An installation in parts was chosen due to outage work scope and ALARA considerations.
The scope of this evaluation is limited to NSMs MG-12096 and MG-22096 at this time. The additional modifications that will complete the installation are included in this description as information to indicate the overall modification scope.
This calculation may be revised later to include the evaluation of one or more of the additional modifications.
n general, each new LPMS will have a final total of 22 sensors mounted at key locations on or near the reactor coolant (NC) system.
The sensors along with associated preamplifiers, and control room equipment will function uo detect loose parts by means of sound vibrations caused when loose parts impinge on piping walls and other interior surf aces within the NC system. The LPMS will be cazibrated to " tune out" normal NC system sounds so that abnormal sounds caused by loose parts may be detected. The scope specific to each modification package is described belows NSM MG-12096 7 LPMS sensors, 5 boxes containing 22 preamplifiers, and junction boxes for testing and calibration will be installed in Unit 1 containment. Cabling between the preamplifiers and 13
containment penetrations, and some cabling to the sensors, will be installed.
The new LPMS will not be operational at this time.
The old LPMS will continue in service.
NSM MG-12096 Install additional sensors and cabling in Unit 1 containment, Control Room equipment changes (includes removal of telemetering oscillator, boronometer, and old LPMS), and cabling-from control room to containment penetrations.
The new LPMS will be placed in service.
(Modification Installation and connection of sensors on the Unit 1 initiated later) steam generators.
NSM MG-22096 7 LPMS sensors, 5 boxes containing 22 preamplifiers, and junction boxes for testing and calibration will be installed in Unit 2 containment. Cabling between the preamplifiers and containment penetrations, and some cabling to the sensors, will be installed.
The new LPMS will not be operational at this time.
The old LPMS will continue in service.
NSM MG-22096 Install additional sensors and cabling in Unit 2 containment, Control Room equipment changes (includes removal of telemetering oscillator, boronometer, and old LPMS), and cabling from control room to containment penetrations.
The new LPMS will be placed in service.
(Modification Installation and connection of sensors on Unit 2 initiated later) steam generators.
Safety Review and USQ Evaluations 7
(For NSMs MG-12096 and MG-22096 only)
The LPMS is not an accident initiator in any accident analyses. Failure of the NC system is considered in various FSAR Chapter 15 design basis accidents. The NC system is both a fission product barrier and pressure boundary for the primary cooling system.
NC system failure is a Loss of Coolant Accident (LOCA).
The LPMS equipment and cabling is non-safety related.
The LPMS sensors do not penetrate the primary system pressure boundary. Three sensors will be clamped to the incore instrument tubing. Sensors near the reactor coolant pumps will be installed by drilling and tapping a support leg for each reactor coolant pump.
]
This installatior. method has been reviewed to ensure structural and functional 4
integrity of the support legs.
All LPMS equipment wil] be installed in acccrdance with QA Condition 4 (seismic mounting) requirements.
An Appendix R review was conducted, with no concerns identified.
The materials used in the LPMS have been reviewed for compatibility with the NC system material and to avoid chemical reaction with post-accident containment atmospheric environment.
The LPMS equipment and cabling is non-safety related.
Operation of the LPMS during accident scenarios is not required.
Therefore, the probability or consequences of an accident previously evaluated in the SAR is not increased.
The LPMS does not perform any accident mitigation functions.
The LPMS is required only to function during plant operation. The new LPMS will perform the same functions as the old LPMS.
The performance of plant safety functions will not be degraded.
Regulatory Guideline 1.133 sets forth a method acceptable to the Nuclear Regulatory Commission (NRC) staff for implementing a loose part detection program that meets regulatory requirements. Per the regulatory guide, "the presence of a loose part in the primary coolant system can be indicative of degraded reactor safety resulting from failure or weakening of a safety-related component."
The regulatory guide does not require an LPMS to be safety-classified (QA Condition 1), however, treatment of loose-part detection by the 14
regulatory guide and the presence of a technical specification on the LPMS indicate that the NRC considers an LPMS to be a system "important to safety".
A provision of Reg. Guide 1.133 (Section c.1.c.)
is that " instrumentation channels (e.g., cabling, amplifiers) associated with the two sensors recommended at each natural collection region should be physically separated from each other starting at the sensor locations to a point in the plant that is always accessible for maintenance during full power operation." The Duke Power Company Nuclear Guide 1.133 states that channel separation for the LPMS will be provided "as indicated by the station FSAR",
and that " channel separation will be determined by whether or not redundant sensors are installed at a particular location." This modification installs one sensor near each reactor coolant pump and three sensors in the collection region below the reactor vessel, along with associated cabling.
Each individual channel consists of a sensor, dedicated hardline and softline cables, and a preamplifier. Cabling from multiple sensors will be installed in common conduit as required to facilitate installation and preamplifiers for multiple channels will be located within the same electrical boxes. FSAR Section 7.7.1.12 will be revised to address channel separation for the operational, as per commitment in Duke Nuclear Guide 1.133.
The preamplifiers are located in containment at locations normally accessible during full power operation.
Therefore, the probability or consequences of a malfunction of equipment important to safety evaluated in the SAR are not increased.
There are no new failure modes created.
No accidents previously considered incredible are made credible by these NSMs. Thus, the possibility of an accident or malfunction of equipment of a different type than evaluated in t he SAR will not be created.
There are no changes of safety limits, setpoints, or plant parameters because of the modification.
Existing containment penetrations are used.
LPMS circuits passing through the containment penetrations are low energy, therefore, double fusing is not required.
The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSMs.
Therefore, the margin of safety as defined in the basis for any Technical Specification is not decreased.
No USQ exists.
NSM-22401
==
Description:==
Main steam line drain valves 2SM83, 2SM89, 2SM95, and 2SM101 were reviewed in accordance with Generic Letter 88-14.
The purpose of this review was to determine if air operated valves would perform their design function under a loss of air. These valves are currently normally open, fail open, and incident open air operated valves that are listed as Containment Isolation Valves in the McGuire FSAR, Table 6-113.
Concern was raised over the ability of these valves to isolate the steam generators following a seismic event.
Results of a subsequent operability evaluation revealed that the Chapter 15 accident analyses would not be adversely affected by this steam release path.
Dose consequences analysis considered this amount of steam release during the recovery period for accidents involving secondary side releases. Later safety analyses were revised to consider the phenomenon of tube bundle uncovery (TBU) which may occur during some accident transient scenarios. As a result of this change in assumed accident conditions, it is necessary to isolate the Main Steam line drain lines in some cases.
Accident analyses results require that the valves be closed in 30 minutes from the onset of the accident, assuming worst case conditions for accidents involving TBU.
General Design Criteria (GDC)-5 states that containment isolation valves for closed systems (outside containment) shall be either automatic, or locked closed, 15
or capable of remote manual operation. These valves fall under the category of GDC-57 valves and can be remotely operated. However, for accidents, credit can only be taken for local manual operation since the valves have air actuators with non-pafety related controls which are ascumed incapable of functioning during a Design Basis Accident.
To better meet the intent of GDC-57, these valves are being changed from all open to fail closed which is the more conservative safety related position.
This NSM will change the failure mode of the previously listed valves from fail open to f ail closed by replacing the operator on these valves.
The incident position will also be closed (i.e. loss of offsite power, Steam Generator Tube Rupture, accidents requiring Main Steam isolation, seismic events causing loss of downstream Class G piping, or loss of instrument air).
Safety Review and USQ Evaluation:
This change will help isolate the Main Steam System and containment during a design basis accident (DBA) with a loss of offsite power.
These valves are located in Duke Class B piping and perform a containment isolation function and therefore are considered QA1. The valve operators are also QA1 because the valve must be capable of being closed and remain closed to isolate containment (S/Gs) during DBAs.
This NSM will enable each valve to be closed by locally, manually removing the air from the operator allowing the spring to close.
All of the valve controls will be considered non-QA because the assured method of closing the valves does not involve the normal valve controls. It is also acceptable and possibly required for the valves to be closed for accident scenarios.
Since containment and Main Steam line isolation is enhanced and accident analyses are still acceptable, there is no increase in the probability or consequences of an accident evaluated in the SAR.
During accidents, the closure of these valves will not damage any safety related equipment.
Based on this and the above information, the probability or consequences of a malfunction of equipment important to safety evaluated in the SAR is not increased.
Normal valve operation will be essentially unchanged and there are no adverse affects on any system. Valve misalignments or failures will be determined as before and will allow corrective actions before system problems occur. Valve control is no less reliable than before.
Therefore, the possibility for an accident or for malfunction of a different type than any evaluated in the SAR is not created.
i No safety limit, setpoint, or operating parameter will be changed by this modification. No fission product barrier is diminished. Therefore, the margin of safety as defined in the basis of the Technical Specifications will not be reduced. No USQ exists.
NSM-42395 I
==
Description:==
Corporate Facilities is building a new office / shop facility on the east side of the plant across the road from the Unit 2 Turbine Building.
The facility will be located between the service road and the existing security fence.
This NSM will provide for the temporary relocation of the security fence in order for the new office / shop f acility to be constructed outside the protected area.
The facility is expected to take 10-12 months to build, then the security fence will j
be put back into it's original configuration by this modification.
once the temporary fencing is in place, a portion of the old protected area boundary will be opened up on the east side between existing microwave units for construction access.
The office / shop facility will be located a minimum of 20 feet from the bordering portions of the existing protected area fence.
No changes to the permanent fence location are expected.
16
-=
In order to provide as much construction clearance as possible on the west side of the new building, the temporary protected area will be far enough west to create violations of the 20 foot interior isolation zone with existing structures. The existing structures are the Waste Storage Facility, the steps to the Operations Procedure Group trailer, fire hydrant 1RY362, a sewage lift station, and stairs to the Environmental Management trailer.
Two existing CCTV cameras will be temporarily relocated in order to cover the temporary protected area boundary. One temporary CCTV camera will be added to cover the area.
A temporary Intrusion Detection System (IDS) will be provided to cover the temporary protected area boundary. Alarm / tamper monitoring will be provided for the temporary IDS and associated equipment in the alarm stations.
No additional security lighting or additional fence grounding is required.
Safety Review and USQ Evaluation:
Cable trenches and yard drains have been evaluated for barrier installations.
The existing structures within the temporary 20 foot interior isolating zone will be part of the Security Plan submittal to the NRC within the allowed 60 days from the date of implementation. The basis of the plan is that we are not degrading the effectiveness of the security program. The Fire Protection System (RY) has been reviewed and is not adversely affected. No accident initiators or safety equipment has been af fected by this modification; therefore, there is no increase in the probability of an accident or of a malfunction of equipment important to safety evaluated in the SAR.
Since no safety system is adversely affected and they are expected to respond as before, there is no increase in the consequences of an accident or of a malfunction of equipment important to safety evaluated in the SAR.
The level of security is not decreased. Based on this and the above, l
the possibility either for an accident or for a malfunction of a different type than any evaluated in the SAR is not created.
No safety limit, setpoint, or operating parameter will be changed by this modification.
Therefore, the margin of safety as defined in the basis of the Technical Specifications will not be reduced. No USQ exists.
NSM-12302, 12305
==
Description:==
These modifications replace the Main Steam Safety valves on the 1A, 1D, 2A, and 2D main steam headers.
Replacement valve bodies and internals, with flanged inlet and outlet connections, have been procured.
The inlet piping will be changed to accommodate flanged connection of the replacement valves rather than the butt-welded connections used for the existing valves. The outlet connections were flanged for the existing valves.
Additionally, this modification will install the replacement valves in the vertical position, rather than 15 degrees off vertical as for the existing valves, to prevent excessive seat leakage.
Safety Review and USQ Evaluation:
The Main Steam (SM) system is affected by this modification. The existing and replacement SM safety valves are QA Condition 1 (Nuclear Safety Related).
The valves are built to ASME Boiler and Pressure Vessel Code,Section III, Class 2.
The modifications have received, from the NRC, written relief from post installation hydrostatic testing as required by Section XI of the ASME Boiler and Pressure Vessel Code.
The safety valves are set for progressive relief at increasing pressure within the ASME Code allowed for range of pressure, to avoid more than one valve actuating simultaneously.
System shock may cause other valves to open.
The 17
safety valve discharge path is sized assuming all valves are discharging simultaneously. The safety valves have a combined capacity greater than maximum calculated heat balance steam flow conditions.
The relevant FSAR Chapter 15 accidents are Loss of External Load, Main Steam Line Break, Loss of Feedwater, Loss of Offsite Power, Loss of Condenser Vacuum, Inadvertent closure of MSIVs, Turbine Trip, Inadvertent Opening of a steam Generator Relief of Safety Valve, and Steam Generator Tube Failure.
In the event the non-safety class steam dump valves fail to open following a large loss of load, the steam generator safety valves may not lift, and the reactor may be tripped by a high pressure, high pressurizer level, or overtemperature signal.
The SM safety valves are accident initiators and accident mitigators.
The replacement valves are constructed to the same design, material, and construction standards as the existing valves. The replacement-valves will be tested prior to installation to ensure that SM system will not be operated outside of its design limits. The lift settings, orifice sizes, and flow capabilities for the replacement valves are the same as for the existing valves.
Blowdown and accumulation characteristics have been reviewed. The flanged joints meet all design requirements of. the SSC ( system, structure, or component). Therefore the probability of an accident or malfunction of equipment important to safety previously evaluated in the SAR is not increased.
The operational characteristics for the replacement valves are the same as for the existing valves.
No assumptions made in evaluating the radiological consequences of an accident have been altered. The function of the safety valves as a fission produce barrier has not changed. Therefore, the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR are not increased.
No new failure modes or operating characteristics are created.
No previously incredible accidents are made credible. The valves will be tested and have been seismically analyzed.
Thus the possibility of an accident or malfunction of equipment of a different type than evaluated in the SAR will not be created.
The replacement valves are construction to QA Condition 1 specifications. The fission product barrier function of the valves is not changed.
There are no changes of safety limits, setpoints, or plant parameters because of the modifications. The fission product barriers (RCS pressure boundary, containment, fuel pellets, aM cladding) are not degraded.
No assumptions made in any accident analysi: are affected by these NSMs.
Therefore, the margin of safety as defined in the basis for any technical specification is not decreased. No USQ exists.
NSM-12393 Descriptions This modification upgrades the 4" Main Feedwater (CF) system tempering lines, from the first block valves in the interior and exterior doghouses to the Turbine Building, as well as the 2" Main Steam (SM) drain lines from the exterior l
doghouse. The existing Class G piping will be analyzed and supported such that
{
-it meets Class F requirements.
For the 2" SM drain lines from the interior doghouse, flow restricting orifices j
will be added just downstream of Main Steam (SM) drain valves ISM 89 and 1SM95.
This will reduce steam flow in the 2" SM drain lines. Reduced steam flow along with the distance of the interior doghouse lines from the diesel and Control Room intakes, was judged to make it unnecessary to upgrade the interior intakes, was judged to make it unnecessary to upgrade the interior doghouse 2" SM drain lines 18 1
T
from Class G to Class F.
This NSM is in response to PIRs 0-M91-0088 and 0-M91-0100.
The PIRs were initiated to evaluate potential failure of the subject lines due to seismic loadings (PIR 0-M91-0088) and tornado wind / missile loads (PIR 0-M91-0100).
Safety Review and USQ Evaluation:
FSAR Section 15.1.5 addresses Steam System Piping Failure, and Section 15.2.8 addresses Feedwater System Pipe Break accidents, which involve the SM and CF systems. This NSM provides stress-analysis and design for upgrading the subject SM and CF system piping and supports from Class G to Class F.
This analysis and upgrade will satisfy the concerns of PIRs 0-M91-008 and 0-M91-0100, which were initiated to evaluate potential failure of the subject lines due to seismic loadings (PIR 0-M91-0088) and tornado wind / missile loads (PIR 0-M91-0100).
The class upgrade will increase the reliability of the subject lines for seismic and tornado events, precluding potential breaks of these lines in the vicinity of the Diesel Generator air intakes.
In the case of the 2" SM drain lines from the interior doghouses, engineering judgement determined that the distance plus reduced steam flow due to the added orifices eliminates the concern fer steam introduction into the diesel intakes. Therefore, the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
These parts of the Main Steam and Main Feedwater systems are not used for accident mitigation. The CF Tempering Flow is important to safety as it protects the CA nozzles against thermal shock. The performance of plant safety functions are not degraded by the modification. The functions performed by the upgraded piping remain unchanged.
There are no common failure modes created by this modification.
Therefore, the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR are not increased.
The upgraded or modified piping does not perform any new functions.
There are no new failure modes created. No accidents previously considered incredible are made credible by these NSMs.
The upgraded piping will be qualified for seismic and tornado wind loads. Thus the possibility of an accident normal function of equipment of a different type than evaluated in the SAR will not be created.
There are no changes of safety limits, setpoints, or plant parameters because of the modification.
The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSM.
Therefore, the margin of safety as defined in the basis for any Technical Specification is not decreased.
No USQ exists.
NSM-12398
==
Description:==
NSM MG-12398 revises the Containment Spray (NS) system to add a 10 inch check valve in the vertical piping of each train of the NS system just above floor elevation 750' + 0 and down stream of the NS pumps.
The check valves, INS 83 through 1NS88, currently used as vacuum breakers, will be deleted by this modi'ication. Also a low point drain will be added inside containment to drain any accumulated water between the inside containment isolation check valve and the spray nozzles. A 10X10X4 inch tee will be added to the piping just upstream of the added check valves to allow future rerouting of the flow test loop if desired. A drain hole will be added to the header inside containment to allow water in the portion of the spray header downstream of the inside containment isolation valve to drain out following a pump stop. This will aid in preventing water hammer on pump restart.
19
The added check valves do not act as containment isolation valves. Containment isolation is provided by a closed check valve inside containment and a closed system outside.
The check valves used to provide a vacuum breaker will be deleted, thus ensuring a closed system.
The NS system does not operate during normal plant operation.
This system is used only during accident conditions with containment pressure above.25 psig.
The operation of the NS system may be manually controlled or controlled by the containment Pressure Control System as specified in Table 3.3-4 Item 6 of the plant Technical Specifications.
There are no changes to the CPCS system as a -
result of this modification.
The added 10 inch check valves act to prevent backflow of water in the NS spray header during periods af ter initial operation of the NS system when the pumps are shut off.
Should backflow occur with the revised piping configuration there is potential for waterhammer in the header pipe due to a vacuum forming between the check valve inside containment and the water head.
This was previously prevented by allowing air to enter the pipe through check valves IND83 through 1ND88. The water hammer will now be prevented by maintaining the water column above the added 10 inch check valves and draining the water in the spray header inside containment.
The Containment Pressure control System will not require modification as a result of the valve additions and will function as before to help maintain containment pressure within acceptable limits.
The NS system will not change its function or operating parameters as described in 6.5.1 through 6.5.5 of the FSAR due to this modification.
None of the changes to the piping or valve additions will significantly affect flow or pressure requirements for the system. There is no change in signals to start or stop the system or in any automatic valve contorts i
as a result of thic modification. The system flow test may be modified and new inspection criteria may be added to ensure the operability of the system and each component, but these will not detrimentally affect the system.
These changes will be made as needed to meet the current functional testing requirements in an efficient cost effective manner.
Safety Review and USQ Evaluation:
The NS system is designed to reduce containment pressure following an accident which results in a large steam release inside containment and also serves to reduce airborne activity in containment by capturing fission products and containing them in a sump liquid.
The system does not operate in any normal modes except for testing as required by Technical Specification 3/4.6.2.
Therefore, the NS system is not an accident initiator.
This modification does not change any operating modes for NS and thus there is no increase in the possibility of an accident evaluated in the SAR or creation of an accident dif ferent than those evaluated in the SAR as a result of this change. Since the operation of the system and components in an accident is unchanged, there is no increase in the consequence of an accident or the probability of a malfunction of equipment as evaluated in the SAR.
The modification primarily changes the i
method of preventing water hammer in the header piping on NS pump restart. The operating parameters are unchanged, therefore, there is no possibility for creation of a malfunction different than those evaluated in the SAR.
The NS system works in accident mitigation by reducing the pressure on the containment structure and reducing the airborne activity available for release from containment. The containment structure is a fission product barrier. Since the system operation is unchanged as a result of this modification there is no reduction in the margin of safety as defined in the Technical Specifications for this modification. No USQ exists.
NSM-12131, 22131
==
Description:==
The existing Component Cooling Water (KC) System heat exchanger does not have 20
adequate venting and draining capabilities when tube cleaning is required. This modification will add a 2" vent line and a 2" drain line on the north end-cover of each KC heat exchanger.
The drain line on each Nuclear Service Water (RN) System strainer has experienced slow drainage during maintenance. The existing RN strainer 3" drain nozzles are presently reduced down to 3/5" but will be changed to 2" by removing the reducer and welding a 2" drain pipe onto the strainer. The 3/4" drain valves (RN285 and RN287) will be replaced by 2" valves. The valves will be Class C stainless steel ball valves. Both KC and RN drain lines added will be Class C stainless steel.
Safety Review and USQ Evaluation:
The function of the KC heat exchanger and the RN strainer is not af fected by this modification.
Both the RN and KC systems are QA1.
All valves added are Class C QAl stainless steel manual ball valves. The valves are oesigned to meet the system design conditions.
The piping is seismically qualified.
A stress analysis and support / restraint analysis was performed as required to ensure that the drain pipes and valves are seismically qualified and installed.
Although dissimilar metals (stainless steel to carbon steel). These welds will be tested as required. Upon completion of the modification the drain pipes and valves will be hydro tested for leakage.
These lines have similar design conditions to existing lines located in the area.
Thus, failure of this portion of the RN or KC Systems will not increase the probability of an accident or malfunction of equipment important to safety previously evaluated in the SAR.
The new vent and drain lines are Class C stainless steel.
The ball valves are manually operated and will be opened for maintenance only.
During power operation, these valves will be closed. The result of a violation of procedures to close valves after maintenance, would be no different from the failure to close existing valves of a similar service already existing. Any leakage from the RN strainers will be noticed by an increase flow to the Auxiliary Feedwater pump room sump pump as would flow from Jimilar existing tap lines located in the area.
Leakage from the KC heat exchanger would be detected by the KC/RN flow meter similar to existing tap lines in the area.
No new failure modes will be created by these changes so the possibility of an accident or malfunction of equipment important to safety different from any already evaluated in the SAR will not be created.
l The addition of vents and drains will not adversely impact the function of the KC heat exchanger and the RN strainer.
The ability to increase maintenance capability and reduce operation time under a limiting condition for operation due to inoperable heat exchange will aid in allowing the components to fulfill their accident mitigating function.
Therefore, the consequences of an accident or j
malfunction of equipment important to safety previously evaluated in the SAR will not be increased.
This modification will not adversely af fect the fission product barriers, safety limits, set points, or design parameters. Thus, the margin of safety as defined i
in the basis to any technical specification will not be reduced as a result of these changes. No USQ exists.
NSM-12269
==
Description:==
The Spent Fuel Cooling System (KF) r6 moves decay heat from fuel stored in the unit Spent Fuel Pool and maintains water quality and clarity for pool operations.
The cooling capability which is considered safety related is provided by redundant full capacity cooling trains for all active cooling components. The system is designed to prevent draining of the pool below acceptable levels by 21
either active or passive failure.
This modification adds manual throttling valves 1KF155 and 1KF156 to the 8 inch Spent Fuel Cooling System (KF) piping downstream of the KF heat exchangers. Four inch valve 1KF157 is added downstream of the KF demineralizer to provide throttling for this piping. The throttling function was previously provided by valves 1KF21, 1KF40,a nd 1KF77 which also serve as isolation valves for maintenance or shutting down one of the redundant trains of cooling. The use of the existing valves for throttling has led to excessive degradation of their ability to provide isolation. This modification will allow the current valves to be dedicated to isolation while the added valves will provide throttling capability. The 8 inch and 4 inch piping is being rerouted to accommodate the addition of the new valves and the pressure relief lines to bypass the current valves is rerouted to include bypass of the added valves.
Power to the motor operator for valve 1KF77 is being disconnected.
This valve will be operated manually. The 8 inch cooling lines are QA1, Duke Class C while the 4 inch purification line is QA2, Duke Class E.
All new piping and valves were specified to the appropriate quality level for the application. Piping analysis and Pipe Support Design was revised to qualify the revised piping layout.
Flow calculations have been revised to assure that the functional capability of the system is not degraded by this modification. The valve additions and removal of remote operation for valve 1KF77 does not change the system function or design requirements of the system. No functional requirements are adversely impacted by this modification.
Safety Review and USQ Evaluation:
The addition of the throttling valves to the KF system will not affect any accident evaluated in the SAR.
Operation of this system is not required for mitigation of any accidents discussed in the FSAR.
Failure of any of these valves would not prevent the KF system from performing its principle safety function of providing cooling capability for the Spent Fuel Pool.
Therefore, this modification will not increase the probability or consequences of any accident evaluated in the SAR.
The function and operation of the system is not changed so there is no new accident, not previously reviewed, created by this modification.
The addition of these valves increases the reliability of the KF System since 1KF21, 1Kl'40 and 1KF77 may be dedicated to providing isolation when required and will not
'>e degraded by use as throttle valves. Therefore, the probability of an equipoent malfunction or the consequances of a malfunction is not increased.
The addAtion of these valves along with the valves currently in place, does not create the possibility of a malfunction dif ferent than any evaluated in the SAR.
Margin of safety is related to the confidence in the fission product barriers to function as designed. This NSM does not degrade any fission product barriers or affect any assumptions in the accident analyses.
No safety limits, setpoints, or timiting safety system settings are af fected. Therefore, the margin of safety as defined in the basis to the Technical Specifications is not reduced. No USQ exitts.
NSM-1.1t3 58, 22358
==
Description:==
This modit4. cation will change the equipment and process used to measure Reactor Coolant Pump (NCP) oil reservoir level.
The existing float switches will be removed and replaced with Linear Variable Differential Transformer (LVDT) level sensors.
The existing float switches have contacts for only high and low oil level, 22 i
l
)
indicated by alarms on the Operator Aid Computer (OAC).
The LVDT level sensore will produce an analog signal that will be used to display actual oil level on an OAC graphics display for the NCPs, as well as generate the high and low oil level alarms.
LVDT leadwires will be terminated in a terminal box located on each NCP.
Existing cabling from the NCP terminal boxes to containment penetrations will be reused for the LVDTs.
New cabling will be routed from the containment penetrations to transmitters that will be mounted in the electrical penetration rooms.
For the Unit 2 pumps only, the new sensors will not physically fit in the existing lower reservoirs, which are different from the other reservoirs, therefore, each lower reservoir will be replaced with a new reservoir.
Safety Review and USQ Evaluation:
FSAR Section 15. 3.1, 15. 3.2, 15. 3. 3, and 15. 3.4 address potential events related to the NCPs that can cause a " Decrease in Reactor Coolant System Flow Rate".
Section 15.3.2 addresses " complete Loss of Reactor Coolant Flow" which is initiated by simultaneous loss of electrical supplies to all NCPs.
The subject modification does not affect the electrical supplies to the NCPs. Section 15.3.1 addresses " partial Loss of Forced Reactor Coolant Flow' which can be initiated by mechanical failure in a single reactor coolant pump.
Mechanical failure of an NCP can result from loss of oil to the pump motor bearings. Low oil level and high bearing temperatures are alarmed in the Control Room.
Receipt of these alarms require pump shutdown to avoid bearing failure. If the bearings fail, the babbitt metal surfaces on the bearing pads are designed to melt, preventing sudden bearing seizure to preclude either of the events evaluated in Sections 15.3.3 or 15.3.4,
" Reactor Coolant Pump Shaft Seizure (Locked Rotor)", or
" Reactor Coolant Pump Shaft Break".
In case of NCP pump failure, the resulting low reactor coolant flow in the affected loop will actuate a reactor trip to avoid a Departure From Nucleate Boiling (DNB) condition. The new instrumentation will provide continuous level indication and allow level trending in addition to high and low level alarms, whereas the existing instrumentation provides only high and low level alarms. Continuous level indication and trending will be an aid in diagnosing the need for pump maintenance before a low oil level condition occurs. QA Condition 1 qualification (including environmental qualification) is not required for the LVDT sensors, transmitters, cabling, and alarms, because NCP oil level measurement is not considered a safety function.
The new oil level instrumentation is expected to be as reliable as the existing non-safety oil level instrumentation.
Materials used for mounting the LVDT sensors are compatible with the sensors, the NCP oil, and reservoir material. QA Condition 4 mounting is not required for the LVDT sensors of the replacement Unit 2 lower reservoirs. The transmitters in the electrical penetration room will be mounted in accordance with QA Condition 4 (seismic) requirements.
New cables from the electrical penetrates to the transmitters were routed in accordance with cable separation criteria.
An Appendix R review was performed, with no concerns identified. For these reasons, the probability of an accident or malfunction of equipment important to safety previously evaluated in the SAR is not increased.
Continuous oil level indication is a new function added by the modifications as an operator and maintenance aid (not a required function).
Loss of oil level alarms due to instrumentation f ailure could occur with either the existing or new instrumentation.
The new instrumentation detects oil level using a method i
different from the existing level switches.
Possible failure of the new level
)
indication with a seemingly valid mid-scale indication is mitigated procedurally, in that the operators are required to stop a pump in case of high bearing temperature alarms. With the existing oil level instrumentation, low oil level alarm failure is also mitigated by the same procedural requirement.
There are 23
no accidents previously considered incredible that are made credible by this NSM.
Thus the possibility of an accident or malfunction of equipment of a different type than evaluated in the SAR will not be created.
There are no changes of safety limits, setpoints, or plant parameters because of these modifications.
The fission product barrAers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSMs.
Therefore, the margin of safety as defined in the basis for any Technical Specification is not decreased.
No USQ exists.
NSM-12293, 22293
==
Description:==
This modification changes out the in-containment signal cables for radiation i
monitors EMF 51A and EMF 51B to coaxial type, and routes the cables in sealed l
conduit. The cables will be routed through new triax feed-throughs to be sued with the existing Conax containment penetrations. The D.G. O'Brien containment penetrations, presently used for signal and power cables, will only be used for EMF 51A and EMF 51B high voltage connections.
This change is being made to resolve potential concerns over the degradation of signal cables for the Containment High Range Radiation Monitors during Design Basis Accident conditions.
The modifications will allow the post accident i
readings from the radiation monitors to remain within the accuracy recommended by Regulatcry Guide 1.97, Revision 2.
Radiation monitors EMF 51A and EMF 51B, the new coaxial cables and conduit, and the containment penetrations are QA Condition 1.
The new conduit will be installed according to approved QA Condition 4 (seismic) mounting requirements.
An Appendix R Review was conducted, with no concerns identified.
Safety Review and USQ Evaluation:
The radiation monitors and containment penetrations are not accident initiators in any FSAR Chapter 15 accident analyses.
There are no control functions performed by the radiation monitors or containment penetrations. The new conduit will be mounted per QA Condition 4 requirements.
An Appendix R Review was conducted.
The new cabling, conduit, and penetration feed-throughs are QA Condition 1 qualified (which included environmental qualification for the containment environment).
Cabling will be installed according to applicable train separation criteria.
Therefore, the probability of an accident or malfunction of equipment important to safety previously evaluated in the SAR is i
not increased.
The radiation monitors and containment penetrations will continue to perform the same functions.
The radiation monitors provide safety-related pre-and post-accident indication of radiation levels inside containment.
The containment penetrations allow cable penetration into containment while serving to maintain containment integrity (to prevent the release of radioactivity). The radiation monitors and containment penetrations will not interact with other equipment in any ways such that the performance of any plant safety functions are degraded.
Applicable safety train separation criteria will be followed for cable installation, so that no common f ailure modes are created.
Therefore, the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR are not increased.
i l
No new failure modes are created.
The new Conax triax feed-throughs are QA 24
)
Condition 1 qualified for in-containment use.
The post-accident radiation monitor readings are expected to be more accurate after the modifications, providing the plant operators with more reliable post-accident in-containment radiation readings.
The new conduit will be installed per QA Condition 4 requirements.
No accidents previously considered incredible are made credible by these NSMs.
Thus the possibility of an accident or malfunction of equipment
{
of a different type than evaluated in the SAR will not be created.
i No technical specifications or plant parameters are affected by this modification.
Existing Conax cable penetrations are utilized, with new QA Condition 1 qualified Conax triax feed-throughs. There are no changes of safety
)
limits, setpoints, or plant parameters because of the modifications. The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSMs.
Therefore the margin of safety as defined in the basis for any Technical Sp2cification is not decreased.
No USQ exists.
NSM-12421
==
Description:==
The Auxiliary Feedwater (CA) system has 10 pressure switches which detect low CA suction pressure in order to initiate automatic swapover to the assured Nuclear Service Water (RN) system supply. A problem with calibration drift, affecting the CA pressure switches, was investigated by PIP item 2-M92-0515.
A proposed resolution of the PIP item was to replace the CA pressure switches with a different vendor type.
NSM MG-12421/00 implements this proposed resolution.
The 10 CA system pressure switches listed below, currently United Electric (A) type, will be replaced with Static-O-Ring (SOR) type. SOR switches were selected due to more favorable setpoint drift characteristics, and a pressure process range closer to CA system design setpoints.
Instrument #
QA Condition Associated CA Purap ICAPS 5002 1
1A Motor-driven 1 CAPS 5350 1
1A motor-driven 1 CAPS 5042 1
Turbine-driven (Train A supply) 1 CAPS 5370 1
Turbine-driven (Train A supply) 1 CAPS 5012 1
IB Motor-driven 1 CAPS 5360 1
1B Motor-driven ICAPS 5381 1
Turbine-driven (Train B supply) 1 CAPS 5390 1
Turbine-driven (Train B supply) 1 CAPS 5044 N/A Turbine-driven (SSS supply)
ICAPS 5380 N/A Turbine-driven (SSS supply)
Minor Modifications MM-5200 and MM-5201 changed the setpoints for these switches to 3 and 4 psig from the original 2 psig, due to concerns with the seismic design of CA suction piping. Later, Minor Modifications MM-5230 and MM-5231 changed the setpoints to 5 psig, due to the concerns with setpoint drift.
It is desirable to have the setpoints as low as possible to prevent inadvertent swapover to the assured RN supply when not required. The new SOR pressure switches are expected to permit changing the setpoints back to 3 and 4 psig, however, additional data is required to qualify the SOR switches for operation during a seismic event for a 3 and 4 psig setpoint.
Switches 1 CAPS 5044 and ICAPS 5380 SSS supply to the Turbine-driven CA pumps are not required to operate during seismic events. The setpoints for ICAPS 5044 and ICAPS 5380 will be changed to 3 peig on NSM MG-12421.
Any setpoint changes for the remaining switches will be implemented by a future modification.
Pressure switch ICAPS 5381 will be relocated to a more accessible area on the south wall outside of the CA Turbine Pump Room at the 720' elevation. This will 25
allow deletion of the one hour fire blanket and simplify future calibrations.
In order to accept a right hand NPT port, pressure switch 1 CAPS 539 will be relocated to the existing 1 CAPS 5381 location inside of the CA Turbine Pump Room.
These relocations have been reviewed for Appendix R concerns.
Safety Review and USQ Evaluations The CA system is not an accident initiator in any FSAR Chapter 15 accident analyses.
10CFR50, Part 21 dedication is provided for the new switches (QA Condition 1 qualified). As noted in the preceding table, the switch applications are QA Condition 1 except for ICAPS 5044 and ICAPS 5380, which do not require QA
'ualification.
The new switches are expected to be more reliable than the ex(sting switches. Therefore, the probability of an accident or malfunction of equlpment important to safety previously evaluated in the SAR is not increased.
The CA system performs the function of providing decay heat or reactor coolant heat dissipation for any emergency or normal shutdown operation. The performance of this safety function is not degraded by the pressure switch changeout. The new CA pressure switches will perform the same functions as the existing switches. The relocation of ICAPS 5381 and 1 CAPS 5390 introduces no common f ailure modes between Safety Trains A and B.
Therefore, the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR are not increased.
This modification adds no new CA system functions.
No new failure modes are introduced. No accidents previously considered incredible are made credible by this NSM.
Thus, the possibility of an accident or malfunction of equipment of
]
a different type than evaluated in the SAR will not be created.
There are no changes of safety limits or plant parameters because of the modification. The lowered setpoint (3 psig) of '. CAPS 5044 and ICAPS 5380 remains above the calculated allowable minimum value of 1.87 psig and the Technical Specification allowable minimum value of 1 psig. The setpoints of the remaining switches will not be changed on this modification. The fission product barriers (RCS pressure boundary, containment fuel pellets, and cladding are not degraded.
No assumptions made in any accident analysis are affected by the NSM. Therefore, the margin of safety as defined in the basis for any Technical Specification is not decreased.
No USQ exists.
NSM-12326 Rev 1
==
Description:==
These modifications provide remote radio control for the 125/4 ton auxiliary fuel handling crane, to increase personnel safety and ease of operation. MG-12326 is for Unit 1 and MG-22326 is for Unit 2.
Existing control pendant is bulky and cumbersome to move, especially up the stairs from the truck bay to the fuel pool.
Also, there is no provision for securing the pendant when not in use. Several panels will be mounted on the crane which will interf ace the radio contorts with the existing controls. The added panels will be seismically mounted. The radio transmitter wattage is in the 50 to 250 miliwatt range.
Safety Review and USQ Evaluation:
FSAR Sections 15.7.4 and 15.7.5 address fuel handling accidents, and 15.2.3 address turbine trips.
The added radio control panels will be seismically mounted (QA Condition 4) on the fuel handling crane to prevent interaction with fuel in the pool in case of a seismic event.
Crane operating commands are transmitted by encoded radio signals to prevent crane misoperation (potentially leading to fuel damage) by radio signal interference. Crane operation will stop when radio transmitter power is lost or the control buttons are not continuously 26
t.
depressed, which is the failsafe operat ing mode. Station personnel have reviewed J
the use of radio frequency transmissions in the plant, and do not anticipate detrimental interaction with plant circuits that could cause a unit trip or are important for safety due to the low transmitter wattage involved. An Appendix R review was conducted with no concerns identified. The crane control equipment and wiring is not nuclear safety-related.
No equipment or cable separation criteria is applicable. Therefore, the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
The fuel handling crane and controls are not accident mitigators. They perform no plant safety functions, and the performance of safety functions is not degraded by the modifications. The modifications will not change the functions of the fuel handling crane. Use of radio frequency transmissions was reviewed by station personnel for effect on plant electronic circuits, with no concerne identified.
Therefore, the consequences of an accident or malfunctior. of equipment important to safety evaluated in the SAR is not increased.
Tha fuel handling crane has no new operational functions.
The use of rcdio i
control for the fuel handling crane was reviewed by station personnel, with no l
concerns identified.
No accidents previously considered incredible are made credible by these NSMs.
Thus the pc esibility of an accident or malfunction of l
equipment of a different type than evaluated in the SAR will not be created.
There are no changes of safety limits, setpoints, or plant parameters because of the modifications.
The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions made in any accident analysis are affected by the NSMs.
Therefore, the margin of safety as defined in the basis for any Technical Specification is not decreased.
No USQ exists.
NSM-52405
==
Description:==
This NSM removed the four existing Instrument Air (VI) system air dryers and associated af terfilters, and installed three new dryer and filter packages. The l
existing refrigerant-type dryers could not process air adequately to meet air quality standards defined by ISA-57.3-1975, and were addressed as a concern by the NRC in Generic Letter 88-14.
The new dryers are be desiccant-type. Due to problems with excessive moisture in the Station Air (VS) system, a new VI/VS connection was installed downstream of the new dryers. The new dryers are sized to supply air simultaneously to the VI and VS systems.
VI piping at the dryers was modified to accommodate the new dryer packages. The replacement was made for one dryer at a time, without interrupting the VI supply.
The inlet valves for each dryer was replaced with new isalation valves.
Test vents were installed for easier VI air quality testing. Valves were installed in the discharge piping from the dryers to provide isolation when needed.
The existing VI dryers are installed on concrete pads, located in the Service Building, Elev 739'+0".
These concrete pade were removed.
The new dryers are 3
mounted on steel skids.
The skids were mounted on steel base plates to be
]
attached to the concrete floor.
Four sleeved openings will be added in the i
Service Building wall, Column Line V, between Column Lines 32 and 34.
l The existing dryers were powered from non-safety shared 600V Motor Control Centers (MCC) SMXS and SMXT.
SMXS and SMXT are normally fed from Unit I and alternately fed from Unit 2.
The new dryers require 120 VAC.
They are supplied from three different 120 VAC power panels, fed from the MCCs.
New cables were installed from the panelboards to the new dryers.
Some cable trays in the 27
P Service Building were removed to accommodate dryer installation.
Existing pressure switches OVIPS5390 and OVIPS5380 monitor differential pressure across the existing afterfilters.
New afterfilters were provided with the new dryer packages. The existing afterfilters and OVIPS5390 were removed. OVIPS5380 was used to monitor discharge pressure for all three dryers and alarm a Control Room annunciator on low pressure. This signal will also cause automatic closure of the dryer exhaust valves IVIl838, IVIl839 and IVll840.
The dryer and filter packages are provided with differential pressure monitoring.
VN MG-52405B assigns valve numbers to vendor supplied valves on the dryer packages and adds valves to supply control air to valves on the dryer packages.
System drain valves were shown on flow diagrams.
VN MG-52405C revises an existing VI Control Room annunciator alarm, adds an additional alarm, and deletes an alarm. OVIPS5066, 5068, and 5069 was changed to measure VI system pressure downstream from the new dryers instead of measuring VI receiver system pressure (they will be retagged to OVIPS6440, 6441, and 6442 respectively).
VN's MG-52405D and MG-5245E are to accommodate installation and have no impact on design.
MG-52405F adds vents and drains as needed.
MG-52405G and MG-52405H concern piping reroutes to accommodate installation. MG-52405I and MG-52405J install air operated valves downstream of the dryers and control valves to close these valves on low discharge pressure.
MG-52405K adds valve numbers to instrumentation isolating valves for use with procedural references.
Safety Review and USQ Evaluation:
The VI system supplies oil-free, dried air for instrumentation, testing and control air requirements. The VS system supplies air for general station service air requirements. The VS system has no safety classification. The VI equipment and piping involved with the modification have no safety classification. Safety class equipment requiring compressed air is provided with an air reserve sufficient to perform its safety function should the VI system fail (Reference FSAR Section 9.3.1).
The portion of the Control Rcom annunciators (EMB system) affected by this modification are mounted in accordance with QA Condition 4 (seismic) requirements, but no other QA Conditions are applicable. The seismic mounting of the annunciators are not affected by the modification because no hardware, additions a e being made to the EMB system. Existing spare window tiles will be used for added VI system alarms.
While the VI and VS systems are not accident initiators in any FSAR Chapter 15 accidents or events, loss of VI can lead to a unit trip.
Cabling to the 1
annunciator system will be routed in accordance with applicable cable separation criteria.
No separation criteria applies to other VI system cabling involved with the modifications.
An Appendix R review was conducted with no concerns identified.
New dryer equipment is to be installed because the existing equipment cannot meet air quality standards (dewpoint).
Improved air quality should increase overall reliability of equipment that uses air supplied by VI and VS.
Therefore, the probability of an accident of malfunction of equipment important to safety previously evaluated in the SAR is not increased.
The VI dryers do not perform accident mitigation functions or other plant safety functions.
A sufficient air supply is stored in reserve for safety related equipment supplied by VI.
The modification does not degrade the performance of any plant safety functions. The annunciator window changes and additions have been reviewed per applicable human f actors guidelines. The modifications do not change any VI or VS system functions. There are no common failure modes created by the modification. Therefore, the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR is not increased.
There are no new system functions added by the modifications.
The new dryers will perform the same functions as the existing dryers, except using a dif ferent process. The overall reliability of equipment using air from VI and VS should j
l 28
be improved due to improved air quality.
No accidents previously considered incredible are made credible by the modifications. Thus, the possibility of an accident or malfunction of equipment of a different type than evaluated in the SAR will not be created.
There are no changes of safety limits, setpoints, or plant parameters because of the modification.
The fission product barriers (RCS pressure boundary, containment, fuel pellets, and cladding) are not degraded. No assumptions madt in any accident analysis are affected by these modifications.
Therefore the margin of safety as defined in the basis for any Technical Specification is not decreased.
No USQ exists.
NSM-52338
==
Description:==
This modification will add a permanent access platform for the condensate Boric Acid Batching Tank, providing a safe access for use in monitoring the tank's performance and maintenance. A 4" diameter stainless steel Borax delivery chute will be added to provide a safe and less time consuming means of adding Borax to the tank.
A water supply at the top of the Borax delivery chute will be added to clean the chute to prevent clogging.
Safety Review and USQ Evaluation:
The Condensate Boric Acid System (YA) and Condensate Feedwater System (CF) are affected by this modification. The condensate boric acid system allows addition of boric acid to the condensate water to arrest intergranular stress corrosion cracking in the steam generator tubes. The YA system is piping class G, stainless steel, 150 psig, non nuclear safety class, non QA condition, non seismic boundary. The relevant Design Basis Events are loss of condenaer vacuum and loss of coolant accident.
The YA system is an accident initiator for a loss of condenser vacuum / turbine trip event. The YA piping provides a flowpath that could cause a loss of vacuum in the hotwell, should the boric acid tank be empty, and the vent path isolation valves be open, or should the piping rupture.
Installation of the platform, chute, or water supply will not increase the probability of piping rupture, due to physical separation of these components and the piping.
Therefore, the probability of an accident evaluated in the FSAR is not increased.
The YA system is not an accident mitigation safety system; however, addition of boric acid to the steam generators will reduce the effects of intergranular stress corrosion cracking, making a primary to secondary LOCA less likely to occur.
This modification introduces no new failure modes, preserves the YA system function, and makes no changes to any accident analysis assumptions.
Therefore, the probability or consequences of a malfunction of equipment important to safety, or the consequences of any accident, evaluated in the SAR, is not increased.
No new f ailure modes are introduced. No previously incredible accidents are made credible. The operating characteristics and function of the YA system remain the same.
Therefore, the possibility of an accident or a malfunction of equipment important to safety different from those evaluated in the SAR is not created.
No fission product barriers are adversely affected.
No safety setpoints, limiting safety settings, or Technical Specifications are changed. Therefore the margin of safety as defined in the bases to the Technical Specifications is not decreased.
No USQ exists.
NSM-11913 29 1
l 1
i
==
Description:==
Wire gates presently in use throughout the auxiliary building to limit access to high radiation areas have recurring problems with the locking mechanisms.
NSM MG-11913 will replace 26 wire gates with hollow metal doors and will replace five wood framed partitions and wire gates with wire mesh partitions and hollow metal d.oors.
Safety Review and USQ Evaluation:
The pressure due to pipe rupture analysis performed for the McGuire Auxiliary Buildino assumed open flow paths to individual compartments. Hollow metal doors will significantly reduce the area of flow path available versus that provided by wire mesh gates.
Rooms identified as having wire mesh gates to be replaced by hollow metal doors either (a) have no safety related equipment or (b) have no postulated high or moderate energy line breaks. Therefore, the effects of pipe breaks on safety related equipment are not made more severe as a result of the new type doors. Excessiva pressure buildup in rooms with high or moderate energy line breaks would fail the door long before the structural integrity of the surrounding walls would be challenged. No door involved in this scenario has any safety related equipment as potential targets should the door become a missile.
It is judged that any damage from a door striking nearby walls would be negligible.
Therefore, the consequences of any accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
The doors and partitions will not be seismically mounted, nor are the existing wire mesh gates and wooden framed partitions seismically mounted. The dif ference in mass and rigidity between the existing doors and partitions and the replacements is not considered significant enough to require seismic mounting.
Therefore, there is no increase in the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.
Since no other interactions can be identified, no possibility of an accident or malfunction of equipment different from any previously evaluated in the FSAR is created.
No margin of safety as defined in the bases to any Technical Specification is reduced.
No USQ exists.
NSM-19410
==
Description:==
This modification is part of the Steam Generator Replacement Project (SGRP). The new B&W Steam Generators (S/Gs) will require feedwater to be supplied through an upper feedring nozzle.
Currently, feedwater is supplied through a lower preheater nozzle in the Westinghouse S/G.
The feedring design will resolve concerns with excessive wear and vibration associated with the preheater design of the existing S/Gs. Feedwater will enter the new S/G through the feedring and flow down through the downcomer to the U-tubes to provide a more uniform flow around the tubes.
This change in Feedwater nozzle location will require the Feedwater System (CF) piping to be rerouted when the S/Gs are replaced.
Interferences exist with the new CF pipe route.
The scope of this NSM is to remove interferences for the new CF pipe route and pipe supports during the outage prior to SG replacement.
This will reduce the amount of work and congestion in the S/G cavity during the replacement outage.
The Steam Generator Wet Layup Recirculation System (BW) instruments and instrument tubing, disabled under NSM MG-1221, will be deleted. The tap for this instrumentation on the Main Steam piping will be removed and capped in accordance with Class B pipe requirements.
The BW System isolation and root valves, reservoir, and piping associated with the SM connection will be removed. The BW System root and isolation valves, reservoir, and piping associated with the CF connection will be removed by NSM MG-19710 during the S/G replacement outage.
The BW instruments include transmitters, pressure gauges, pneumatic and electric receiver meter, and pneumatic to electric converters.
Associated cables will 30
also be removed involving panel wiring changes, electrical penetration work, and removal of field routed conduits /electray.
Neither the ATC panels nor the penetration will be degraded by this cable deletion.
All lights to be relocated are part of the normal lighting system and are not safety related. Areas will remain sufficiently lit.
The relocation of a sound powered phone jack also has no safety significance.
Abandoned electray, pipe supports, and BW tubing supports at the top of the S/Gs will be removed.
Additional abandoned steel attached to the crane wall and empty electray running along the crane wall will be removed. The removal of these components will not af fect any Structure, System, or Component (SSC) since they are not currently in use.
Some cable tray sections will be cut back.
The structural integrity of these cable trays is not affected by the change. The criteria for safety related or seismically hung cable trays and safety related cables is maintained.
Cables routed through cut out sections of tray will be supported by unistruts along the crane wall.
The structural integrity of the crane wall is not degraded by the unistrut addition.
Rerouted and repulled cables will meet present separation, independence, and Appendix R requirements.
Reactor Coolant Pump (RCP) power cables on S/G 1B and 1C will be replaced with new cables and rerouted. The new cables meet all present criteria for materials, class, separation, independence, and Appendix R requirements. New and existing RCP cable tray and supports meet all the present criteria.
Removal of some existing RCP cable trays that are not needed will not affect any SSCs.
Three new IE excore Instrumentation (ENB) cables and 1 new lE Reactor Coolant System (NC) cable will be pulled.
These new cables and cable routes meet all present criteria for materials, class, separation, independence, seismic, and Appendix R
requirements.
Pipe rupture analysis has been completed satisfactorily. Therefore, functional capabilities of the cables is not degraded from the present cable abilities.
The field routed tubing to be rerouted is safety related and serves CF instrument loops related to Engineered Safety Features (ESF), Reactor Trip System (RTS),
Process Control Cabinets, S/G level monitoring, and post Accident Monitoring.
The reroutes maintain channel independence and separation criteria, and are seismically qualified where required, thus meeting single f ailure requirements.
The response, accuracy, and functional capabilities of the instrumentation are not adversely af fected. Since this field routed tubing is rerouted below the S/G enclosure, pipe rupture analysis will be performed in conjunction with tubing installation and before startup. This calculation must be revised at that time to indicate satisfactory completion of this review.
Based on the above considerations, the tubing reroute will preserve the abilities of the instruments to perform their safety functions. Tubing insulation will meet design criteria and be installed securely.
The rerouted Makeup Demineralized Water System (YM) piping is the two inch Class H line that supplies the decontamination sinks in Containment.
The non-safety piping is less than four inches nominal diameter and is not subject to non-seismic interaction review. This reroute does not adversely affect any SSC. The modification does not change the function of any plant systems. Note that the BW System has been previously disabled. All new materials are suitable for use in post accident containment.
Safety Review and USQ Evaluation:
Rerouted components meet all the present criteria, as do all new components.
There are no new interaction consequences, or other adverse affects on and SSC.
The SM System tap removal and capping maintains the Class B pressure boundary.
31
All components will function as before.
Since the status quo is essentially maintained, there is no increase in the probability of an accident or of a malfunction of equipment important to safety evaluated in the SAR.
All separation, independence, and single failure requirements are maintained.
No adverse affects on any SSC were noted. No functions were added or deleted, and present function capabilities still exist.
Therefore, there is no increase in the consequences of an accident or of a malfunction of equipment important to safety evaluated in the SAR.
There is no degradation of any SSC nor adverse affect on any SSC.
No new f ailure modes were introduced. Component functioning capabilities are maintained and no new functional requirements were added. Based on this, the possibility for an accident or for a malfunction of a dif ferent type than any evaluated in the SAR is not created.
No safety limit, setpoint, or operating parameter will be changed by this NSM.
Therefore, the margin of safety as defined in the bass of the Technical Specifications will not be reduced. No USQ exists.
NSM-12050, 22050 Descriptions The operability of the diesel generators is of prime concern because of their required operation during safety-shutdown events. Because of the severe impact to plant operations a major engine failure would have, performance and trend analysis capability would be extremely valuable in analyzing problems, as well as identifying and correcting degrading situations before major catastrophic material and component failures occur. The existing diagnostic instrumentation is of limited scope and does not provide sufficient data for analysis and troubleshooting of problems. This modification will add a new diesel generator diagnostic monitoring system. Each diesel will have a dedicated CRT, keyboard, super-processor and termination cabinet located in the respective diesel generator room. One host computer will be provided to share between units. This will be located in the cable spreading room.
I i
Safety Review and USQ Evaluation:
These NSMs will add a computer based monitoring system to the diesel generators to be used during normal operation for diagnostic purposes.
The monitoring instrumentation and equipment are not required to be an QA condition because they are used for maintenance and testing only.
However, portions of the diesel generator diagnostic monitoring system are seismically qualified because of the interaction with QA1 equipment.
The system has no interlocks with the diesel generators. It is powered by the EPF System so that during blackout conditions, data will be available.
The modification does not change the diesel ganerator design bases or require new design bases or criteria. Therefore, the probability of occurrence of accidents previously described in the FSAR is not increased.
Neither the diesel generator operation during accident conditions nor accident mitigation systems are affected; therefore, the consequence of accidents previously evaluated in the FSAR is not increased.
The diesel generator diagnostic monitoring system has no interlocks with the diesel generator. Normal safety separation and isolation between A and B train and QA1 and non-QA equipment are maintained. Thus, there is no increased possibility of an accident not previously evaluated.
This modification will be located in the diesel generator room and the cable i
spreading room in the Auxiliary Building. There are no ALARA concerns associated with this NSM, and an Appendix R review indicated no auxiliary shutdown panel or standby shutdown systems design bases are impacted. The system was purchased to operate in the areas indicated and does not require equipment qualification. The HVAC requirements in the areas where the system will be located are not increased.
As part of the increased monitoring capabilities, additional 32
instrumentation will be added.
The instrumentation will utilize existing thermowells and existing RTD inputs. One exception, annubar flow elements to be installed inside a QAl pipe, will be purchased non-safety and installed using a QAl contoured weld fitting.
The orifice plates are QA1, Class C components.
Stress sand support / restraints were reviewed and appropriate calculations performed for instruments mounted on pipe or supports.
Standard separation criteria was observed for cable changen.
Seismic calculations of affected QAl equipment important to safety has been created nor are existing malfunctions previously evaluated more likely to occur. Since no accident mitigation systems are degraded, the consequences of a malfunction of equipment important to safety are not increased.
The modification will not af fect the plant initial conditions, key parameters or setpoints. Therefore, there is no reduction in the margin of safety as defined in the bases to any Technical Specification. No USQ exists.
NSM-12045, 22045
==
Description:==
The Valve Stem Leakoff Line to the Reactor Coolant Drain Tank will be rerouted such that it enters the tank through the recirculation line instead of through the normal tank inlet. Water hammer problems around the inlet nozzle have caused the nozzle welds to break on the Unit 1 tank. The problem has been analyzed and the determination made that this modification should alleviate the water hammer problem.
Safety Review and USQ Evaluation:
t The Reactor Coolant Drain Tank (NCDT) collects all deaerated recyclable liquids with entrained fission product gases that are released through piped up drains and leakoffs inside the containment.
The tank is part of the Reactor Coolant Drain Tank subsystem of the Liquid Waste Recycle System (WL).
During normal i
operation, liquids are sent to the recycle holdup tanks for reuse, with out further processing by the liquid waste recycle system.
One source of influent to the NCDT is piped up valve leakoffs located in the containment. This liquid is routed to the NCDT through 2" Class E pipe (passing through check valve WL296), to a 3"
(Class E) header to the NCDT.
Design temperature and pressure in both the 2" and 3" pipe is 200 degrees F.
and 100 j
peig. The modification will cut and cap the 2" valve stem leakof f lie at the 3" header and reroute the pipe to connect to a 3" header (Class E) return line to the NCDT from the Reactor Coolant Drain Tank Heat Exchanger. This inlet to the NCDT is normally flowing, and also has design temperature and pressure of 200 degrees F.
and 100 psig.
The modification is QA condition 2, and the NCDT and piping affected are not safety related nor seismically designed. Monitoring of fluid level within the NCDT will not be affected by the modification since check valve WL 296 will remain unaffected by the modification, and will prevent flow away from the NCDT through the valve stem leakoff piping. Measurement of fluid level and pressure in the tank is used in Reactor Coolant System Leakage calculations.
Piping changes for Unit 1 and 2 were reviewed for stress analysis.
Since the valve stem leakoff piping is routed to piping with the same design parameters and class, and the stress analysis was performed as required, the probability and consequences of malfunctions of equipment important to safety will not be increased. Also, since the NCDT is not involved in mitigation of any FSAR accidents, the consequences of FSAR accidents will not be increased. No new failure modes are created so no new accidents or malfunctions of equipment important to safety are created. No safety / design limits are adversely affected j
33
so margins of safety as defined in the bases to Technical Specifications are not reduced. No USQ exists.
?
i i
34
Duke Power Company McGuire Nuclear Station Summary of Minor Modifications Completed Under 10CFR50.59 Minor Mod i Valve 3969 1NB0260 3376 2NB0260 Descriptions NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for HOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
These Minor Podifications provide for the diagnostic testing and constitute part of the actions necessary for compliance to NRC Generic Letter 89-10.
The actual changene i.svolve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
The inaccuracies of the diagnostic test system used to f acilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluation:
The MOVs af fected are in the Borori Recycle system and serve as the containment isolation valve for the makeup water storage tank.
The safety functions of 1NB0260 and 2NB0260 are to close and their normal position is closed. Re-setting the open and close torque switches will not affect open and closure times. The existing stress analysis of the associated piping will not be affected by re-setting the open and close torque switches.
Since the mms will ensure these valves will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
Minor Mod i Valvo 3637 1NV0245
==
Description:==
NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power 1
stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
This Minor Modification provides for the diagnostic testing and constitutes part of the actions necessary for compliance to NRC Generic Letter 89-10.
The actual changes involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
The inaccuracies of the diagnostic test system used to f acilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the opurator or valve components.
Safety Review and USQ Evaluation:
The MOV affected by MM 3637 is in the Chemical and Volume control System.
The function of 1NV0245 is to provide containment isolation for the normal charging flow path.
INV0245 is an active valve that is normally open during Modes 1 through 4.
1NV0245 receives a safety injection signal to close to satisfy containment isolation requirements.
Re-setting the open and close torque switches will not af fect open and closure times of 1NV0245. The stress analysis of the piping associated with INV0245 will not be af fected by re-setting the open and close torque switches.
Since the MM will ensure that 1NV0245 will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
Minor Mod #
Valve 3413 IFWOO27 Descriptions NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions tk Duke Power Company nuclear stations will accomplish in order to comply v.a the Generic Letter.
This Minor Modification provides for the diagnostic testing and constitutes part of the actions necessary for compliance to NRC Generic Letter 89-10.
2
The actual changes involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present.
The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power specification DPS-1205.19-OO-0002 which establishes the parameters and c riteria used to determine the minimum required and maximum allowed thrust levels.
The inaccuracies of the diagnostic test system used to facilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluation The MOV affected by MM 3413 is in the Refueling Water System. The function of 1FWOO27 is to isolate the RWST from the ND pump suction after switchover to the recirculation phases of an accident. This valve also provides a flow path from the REST to the ND pump suction for injection phase of an accident. The safety position of 1FWOO27 is to open and close and it's normal position is open.
Re-setting the open and close torque switches will not af fect open and closure times of 1FWOO27. The existing stress analysis of the piping associated with 1FWOO27 will not be affected by re-setting the open and close torque switches.
Since this MM will ensure that 1FWOO27 will perform as required for design basis system conditions. the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
Minor Mod i Valve 3638 IVE0006
==
Description:==
NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
This Minor Modification provides for the diagnostic testing and constitutes part of the actions necessary for compliance to NRC Generic Letter 89-10.
The actual changes involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present.
The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
3
The inaccuracies of the diagnostic test system used to facilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be suf ficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluation:
The MOV af fected by HM 3638 is in the Annulus Ventilation System. This valve is used to isolate the Annulus from the containment structure and therefore the safety function of IVWOOD6 is to provide containment isolation.
The safety position of IVE0006 is closed and it's normal position is closed. Re-setting the open and close torque switches will not af fect open and closure times of IVEOOO6.
The existing stress analysis of the piping associated with IVE0006 will not be affected by re-setting the open and close torque switches.
Since this MM will ensure that IVWOOO6 will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
Minor Mod #
Valve 3452 1NC033 3418 1NC0031 3379 2NCO35 3378 2NCO33 3377 2NC0031 3453 1NC0035
==
Description:==
NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
These Minor Modifications provide for the diagnostic testing and constitutess part of the actions necessary for compliance to NRC Generic Letter 89-10.
The actual changes involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
The inaccuracies of the diagnostic test system used to facilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluation:
4 i
I The MOVs af fected by these minor modifications are in the Reactor Coolant System.
These valves serve as inlet isolation valves for the Power Operated Relief Valves (PORVs). The safety function is to close and the normal position is open.
Re-setting the open and close torque switches will not affect open and closure j
times.
The existing stress analysis of the piping will not be affected by re-setting the open and close torque switches. Since these mms will ensure that the MOVs will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety j
previously evaluated in the FSAR is not increased.
No USQ exists.
Minor Mod #
Valve 3428 1NIO430 3429 1NIO431 3782 2NIO178 3785 2NIO135 3427 1NIOl83 3455 1NIO136 3454 1NIO100 3771 2NIO150 3770 2NIO136
==
Description:==
NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
These Minor Modifications provide for the diagnostic testing and constitute part of the actions necessary for compliance to NRC Generic Letter 89-10.
The actual changes involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
The inaccuracies of the diagnostic test system used to f acilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluations The MOVs af fected by these modifications are in the Safety Injection System. The function of 2NIO150 is to open to provide a flowpath for safety injection into the RCS cold leg during the cold leg injection phase and cold leg recirculation phase of ECCS operation. 2NIO150 is normally open which is the safe position of this valve. Re-setting the open and close torque switches will not affect open and closure times of 2NIO150.
The existing stress analysis of the piping 5
associated with 2NIO150 will not be affected by re-setting the open and close torque switches. Since this MM will ensure that 2NIO150 will perform as required for design basis system conditions, the probability or consequences of an accident of malfunction or equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
The function of INIO430 and 1NIO431 is to open to provide nitrogen pressure to operate the RCS PORVs in the event that instrument air is lost. Low Temperature Overpressure Protection (LTOP) is also provided by having INIO430 and 1NIO431 opened. The safety position is closed and their normal position in closed.
Re-setting the open and close torque switches will not af fect open and closure times of 1NIO431.
The existing stress analysis of the piping associated with these valves will not be affected by re-setting the open and close torque switches.
Since these mms will ensure that INIO430 and 1NIO431 will perform as required for design basis system conditions, the probability or consequences of an accident of malfunction of equipment important to safety previously evaluated in the FSAR is not increased. NO USQ exists.
The function of INIO136 and 2NIO136 is to open during cold leg recirculation to allow RHR pump B to prime the safety injection suction header.
INIO136 and 2NIOl36 are normally closed which is the safe position of this valve. Re-setting the open and close torque switches will not affect open and closure times. The existing stress analysis of the piping associated with these valves will not be af fected by re-setting the open and close torque switches. Since these mms will ensure these valves will perform as required for design basis system conditions, the probability of consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
The function of 2NIO178 is to open during normal operation to allow the RHR pumps to inject borated water into the RCS via the cold legs.
This valve is closed during hot leg recirculation to isolate safety injection into the RCS cold legs.
Re-setting the open and close torque switches will not affect open and closure times. The existing stress analysis of the piping associated with 2NIO178 will not be affected by re-setting the open and close torque switches. Since this MM will ensure that 2NIOl78 will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated int he FSAR is not increased.
No USQ exists.
The function of 2NIOl35 is to open to allow flow to SIP and CCP suction header from the RHR pumps (s) during the recirculation phase of ECCS operation. 2NIOl35 is normally open which is the safe position of this valve. 'Re-setting the open and close torque switches will not affect open and closure times. The existing stress analysis of the piping associated with 2NIO135 will not be affected by re-setting the open and close torque switches.
Since this MM will ensure the 2NIO135 will perform as required for design basis system conditions, the probability or consequences of an accident of malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
The function of 1NIO100 is to open (normal position) during the cold leg injection phase of ECCS operation to allow flow from the REST to the NI pumps.
1NIO100 is also required to close upon operation action to isolate the REST from NI pumps and RHR during the sump recirculation phase of ECCS operation.
Power is normally removed from this valve to prevent inadvertent actuation. Re-setting the open and close torque switches will not affect open and closure times of INIO100. The existing stress analysis of the piping associated with INIO100 will not be af fected by re-setting the open and close torque switches. Since this MM will ensure that INI)100 will perform as required for design basis system conditions, the probability or consequences of an accident of malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
6
The function of 2NIO150 is to open to provide a flowpath for safety injection into the RCS cold leg during the cold leg injection phase and cold leg recirculation phase of ECCS operation.
2NIO150 is normally open which is the safe position of this valve. Re-setting the open and close torque switches will not affect open and closure times of 2NIO150.
The existing stress analysis of the piping associated with 2NIO150 will not be affected by re-setting the open and close torque switches. Since this MM will ensure that 2NIO150 will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
The function of 1NIO183 is to open only during hot leg recirculation to allow flow from the RHR pumps to RCS hot legs, loops 2 and 3.
Power is normally removed from this valve to prevent inadvertent actuation.
The safety position of INIO183 is closed and it's normal position is closed. Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the piping associated with 1NIOl83 will not be affected by re-setting the open and close torque switches.
Since this MM will ensure that 1NIO183 will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
Minor Mod #
Valve 3436 1NM0072 3437 1MN0075 3435 1MN0026 3433 1MNOO22 3397 2NM0025 3396 2NM0022 3395 2NM0006 3430 INM0003 3431 INM0006 3432 INM0007 3387 2NMOOO3 3434 1NM0025
==
Description:==
NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
These Minor Modifications provide for the diagnoctic testing and constitute part of the actions necessary for compliance to NRC Generic Letter 89-10.
The actual changes involve re-setting the open and close torque switches no that the motor operator will produce the necessary torque, that will be converted oy the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
7
-- -. ~ _ _
The inaccuracies of the diagnostic test system used to f acilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system ' pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluations f
The MOVs affected by these modifications are in the Nuclear Sampling System. The system function of 2NM0006, INM0003,1NM0006, INM0007 and 2NM0003 is to provide isolation for sampling the Pressurizer (water and steam).
These valves also serve a containment isolation function for penetration M235 and is -considered Active. ~ The safety function is to close and the normal position is closed.
Re-setting the open and close torque switches will not affect open and closure times. The existing stress analysis for the piping associated with these valves will not be affected by re-setting the open and close torque switches.
Since these mms will ensure the valves will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
The system function of INM0072 and INM0751s to provide isolation for sampling the NI Accumulator Tanks A, B,
C, and D.
These valves also serve a containment isolation function for penetration M280 and are' considered active. The safety function is to close and the normal position.is closed. Re-setting the open and close torque switches will not affect open and closure times.
The existing i
stress analysis of the piping associated with these valves will not be affected by re-setting the open and close torque switches. Since these MM will ensure the valves will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in'the FSAR is not increased. No USQ exists.
The system function of INM0022, INM0026, and 2NM0022 is to provide isolation for sampling the Reactor Coolant Loops 1 and 4.
This valve also serves a containment isolation function for penetration M309 and is considered active.
The safety function of 1NM0022, INM0026 and 2NM0022 is to close and it's normal position is open.
Re-setting the open and close torque switches will not affect open and closure times. The existing stress analysis of the piping associated with these valves will not be affected by re-setting the open and close torque switches.
Since this MM will ensure that the valves will perform as required for design basis system conditions, the probability or consequences of an accident of malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
The system functions of INM0025 and 2NM0025, is to provide isolation for sampling the Reactor Coolant Loops 1 and 4.
These valves also serve a containment isolation function for penetration M309 and are considered active.
The safety function of INM0025 and 2NM0025 is to close and their normal position is closed. P setting the open and close torque switches will not affect open and closure.imes.
The existing stress analysis of the piping associated with INM0025 av 2NM0025 will not be affected by re-setting the open and close torque switches. Since these MMr will ensure that these valves will perform as required for design basis system conditions, the probability or consequences of an accident of malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
Minor Mod #
Valve i
3398 2NS0001 3402 2NS0018 3446 INS 0038 3403 2NS0020 8
1 4
3447
'1NS0043 3399 1NS0003 3443 1NS0020 3440 INS 0012 3438 1NS0001 3439 1NS0003 3441 1NS0015 3444 1NS0029 3445 1NS0032 3407 2NS0043 3442 INS 0018 3406 2NS0038 3400 2NS0012 3404 2NS0029 3405 2NS0032 3401 2NS0015
==
Description:==
NRC Generic Letter 89-10, issued on June 28, 1985, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenanco of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
These Minor Modifications provide for the diagnostic testing and constitute part of the actions necessary for compliance to NRC Generic Letter 89-10.
The actual changes involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
The inaccuracies of the diagnostic test system used to facilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluation The MOVs affected by these modifications are in the Building Spray System. The function of 2NS0001 is to isolate the containment pump suction header for the containment spray pumps. This isolation occura during Modes 1 through 4.
This valve is also required to open to realign the spray pumps' suction from the REST to the containment sump. Therefore, the safety positions of 2NS0001 is to close and open and it's normal position is closed. This valve is considered active.
Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the piping associated with this valve will not be affected by re-setting the open and close torque switches.
Since
~
9
this MM will ensure that the valve will perform as required for design basis system conditions, the probability or consequences of an accident or nialfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
The function of 2NS0018 is to isolate the containment pump suction header for the containment spray pumps. This isolation occurs during Modes 1 through 4.
This valve is also required to open to realign the spray pumps' suction from the REST to the containment sump. Therefore, the safety position is to close and open and it's normal position is closed. This valve is considered active. Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the piping associated with this valve will not be affected by re-setting the open and close torque switches.
Since this MM will ensure that the valve will perform as required for design basis system conditions, the probability of consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
The function of 1NS0038 is to isolate the ar"iliary containment spray header and provide containment isolation. This isolats an occurs during Modes 1 through 4.
This valve is also required to open, by operator action, to provide a flowpath from the RHR pumps to the auxiliary containment spray headers.
Therefore, the safety position of 1NS0038 is to close and open and it's normal position is closed. This valve is considered active. Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the piping associated with this valve will not be affected by re-setting the open and close torque switches.
Since this MM will ensure this valve will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
The function of 2NS0020 is to provide a borated water flowpath from the REST to the containment spray pumps during Modes 1 through 4.
This valve is also required to close to realign the spray pumps' suction from the REST to the containment sump. Therefore, the safety position of 2NS0020 is to close and open and it's normal position is open. This valve is considered active. Re-setting the open and close torque switches will not affect open and closure times. The existing stress analysis of the piping associated with this valve will not be affected by re-setting the open and close torque switches.
Since this MM will ensure that this valve will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
The function of 1NS0043 is to isolate the auxiliary containment spray neader and provide containment isolation. This isolation occurs during Modes 1 through 4.
This valve is also require to open, by operator action, to provide a flowpath from the RHR pumps to the auxiliary containment spray headers.
Therefore, the safety position of INS 0043 is to close and open and it's norn al position is closed. This valve is considered alive. Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the piping associated with this valve will not be affected by re-setting the open and close torque switches.
Since this MM will ensure that 1NS0043 will perform as required for design basis system conditions, the probability or consequences of an accident of malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
The function of 2NS0003, INS 0003 and INS 0020 is to provide a borated water flowpath from the REST to the containment spray pumps during Modes 1 through 4.
These valves are also required to close to realign the spray pumps' suction from the REST to the containment sump.
Therefore, the safety position of 2NS0003, INS 0003 and INS 0020 is to close and open and the normal position is open. These 10
j i
valves are considered active. Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the piping will not be affected by re-setting the open and close torque switches.
Since these mms will ensure that the valves perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of 1
equipment important to safety previously evaluated in the FSAR is not increased.
)
No USQ exists.
The function of valves INSOO12, INS 0015, INSOO29, INS 0032, 2NS0012, 2NS0015, 2NS0029, and 2NS0032 is to isolate the containment spray header and provide containment isolation. This isolation occurs during Modes 1 through 4.
These valves are also required to open, by Sp signal, to provide a flowpath to the headers. Therefore, the safety positions of these valves is to close and open and their normal position is closed. These valves are considered active.
Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the piping will not be affected by re-setting the open and close torque switches. Since these mms will ensure that the valves will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
The function of 2NS0038 and 2NS0043 is to isolate the auxiliary containment spray header and provide containment isolation. This isolation occurs during Modes 1 through 4.
These valves are also required to open, by operator action, to provide flowpath from the RHR pumps to the auxiliary containment spray headers.
Therefore, the safety positions are to close and open and the normal position is closed.
These valves are considered active.
Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the piping will not be affected by re-setting the open and close torque switches.
Since these mms will ensure that the valves will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
The function of 1NS0001 and INS 0018 is to isolate the containment pump suction header for the containment spray pumps.
This isolation occurs during Modes 1 through 4.
These valves are also required to open to realign the spray pumps' suction from the REST to the containment sump.
Therefore, the safety position is to close and open and the normal position is closed.
These valves are considered active. Re-setting the open and close torque switches will not af fect open and closure times. The existing stress analysis of the piping will not be affected by re-setting the open and close torque switches. Since these mms will ensure that the valves will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
Minor Mod i Valve 3451 INSS5590 3450 1NSS5580 3408 2NSSTVL5560 3390 2NSSTVL5570 3389 2NSSTVL5590 3388 2NSSTVL5580 3449 1NSS5570 3448 1NSS5560
==
Description:==
NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and 11
maintenance of motor operated valves (Movs) so as to provide the necessary assurance that. they will function when subjected to design basis system i
conditions. The. level of testing,. inspection,_ and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
These Minor Modifications provide for the diagnostic testing and constitute part of the actions necessary for. compliance to NRC Generic Letter 89-10.
~
The actual changes involve re-setting the open and close torque switches so that-the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering. Calculation MCC-1205.19-00-0003 - and is provided by controlled document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19-00--
0002 which establishes the parameters and criterla used to determine the minimum required and maximum allowed thrust levels.
The inaccuracies of the diagnostic test system used to facilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Gafety Review and USQ. Evaluation:
The MOVs affected by these modifications are in the Containment Spray System.
Their function is to provide containment isolation and a pathway for wide range l
containment pressure sensing for the CPCS.
The safety position is close and 4
open.
Re-setting the open and close torque switches will not affect open and closure times. The existing stress analysis of the piping will not be affected by re-setting the open nd close torque switches. Since these mms will ensure the valves will perform as required for design - basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
Minor Mod #
Valve 3479 2VIO148 i
3458 lVIO150 3457 1VIO148 1
3480 2VIO150 3456 1VIO129 1
3478 2VIO129 3481 2VI0l60 3459 1 VIOL 60 i
==
Description:==
NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As l
12
1 required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
These Minor Modifications provide for the diagnostic testing and constitute part of the actions necessary for compliance to NRC Generic Letter 89-10.
The actual changes involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
The inaccuracies of the diagnostic test system used to f acilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluation:
The MOVs af fected by these modifications are in the Instrument Air System. The function of valves 2VIO148, IVIO150, IVIO148 and 2VIO150 is to isolate the instrument air supply to lower containment from the Duke Class G portion of the pipe.
The safety function is to provide containment isolation as part of the accident mitigation sequence.
The safety position is closed and the normal position is open. Re-setting the open and close torque switches will not affect open and closure times. The existing stress analysis of the piping will not be af fected by re-setting the open and close torque switches. Since these mms will ensure the valves will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
Valves lVIO129 and 2VIO129 isolate supply tank "A" from the Duke Class G portion of the pipe. The function of these valves is to provide containment isolation as part of the accident mitigation sequence. The safety position is closed and the normal position is open. Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the piping will not be affected by re-setting the open and close torque switches.
Since these mms will ensure these valves will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
Valves 2VI0l60 and IVIOl60 isolate supply tank "B" from the Duke Class G portion of the pipe.
The safety function of these valves is to provide containment isolation as part of the accident mitigation sequence. The safety position in closed and the normal position is open.
Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the piping will not be affected by re-setting the open and close torque switches. Since these mms will ensure these valves will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to 80fety previously evaluated in the FSAR is not increased.
No USQ exists.
i Minor Mod #
Valvo 13 l
i
3419 IND0001 3380 2ND0001 33B4 2ND0058 "424 1ND0058
'385 2ND0067 3786 2ND0068 3420 IND0004 3422 1ND0019 3381 2ND0004 3383 2NDOO19 3382 2ND0015 3423 1ND0030 3783 2ND0030 3421 IND0015
==
Description:==
NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
These Minor Modifications provide for the diagnostic testing and constitute part of the actions necessary for compliance to NRC Generic Letter 89-10.
The actual changes involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
The inaccuracies of the diagnostic test system used to f acilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be suf ficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluation:
The HOVs af fected by these modifications are in the Residual Heat Removal System.
The function of IND0001 and 2ND0001 is to provide ND system isolation from the NC system.
IND0001 and 2ND0001 are active valves that are required to close to satisfy their safety function. The normal position is closed. Re-setting the open and close torque switches will not af fect open and closure times.
The existing stress analysis of the piping will not be affected by re-setting the open and close torque switches. Since this MM will ensure the valve will perform as required for design basis conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
The function of 2ND0058 and 1ND0058 is to isolate the ND system from the NV system. These valves must open following the injection phase for the ND system 14
to supply the NV and NI system during the recirculation phases. These valves are active valves which are normally closed.
Re-setting the open and close torque switches will not affect open and closure times. The existing seismic analysia and the stress analysis of the piping will not be affected by re-setting the open and close torque switches. Since these mms will ensure the valves will perform as required for design basis system conditions, the probability or consequences of an accident of malfunction of equipment important to safety previously J
evaluated in the FSAR is not increased.
No USQ exists.
The function of 2ND0067 and 2ND0068 is to provide a recirculation flow path to protect the ND pumps from low flow conditions.
The safety position of these i
valves is open during low flow conditions to protect the ND pumps and close when the ND pumps are injecting to maximize the injection flow rate. These are active valves which are normally closed. Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the piping will not be affected by re-setting the open and close torque switches.
Since these mms will ensure the valves will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
)
The function of 1NDOOO4, 2ND0004, IND0019 and 2ND0019 is to provide ND pump suction isolation. These valves must be capable of closing for train separation during the recirculation phase of an ECCS actuation. These are active valve that are required to close and open to satisfy their safety functions.
The normal position is the open position.
Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the piping will not be affected by re-setting the open and close torque switches.
Since these mms will ensure that the valves will perform as required for design i
basis system conditions, the probability or consequences of an accident or i
malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
The function of valves 1ND0015, IND0030, 2ND0015 and 2ND0030 is to provide train separation during the recirculation phase of an ECCS actuation.
These valves must close when going from the injection phase to the cold leg recirculation i
phase and must open when going from the cold leg to the hot leg recirculation phase.
These are active valves that are require to close and open to satisfy their safety functions. The normal position is the open position.
Re-setting the open and close torque switches will not affect open and closure times. The existing stress analysis of the piping will not be affected by re-setting the open and close torque switches.
Since these mms will ensure the valves will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
Minor Mod #
Valve 3772 2KC0003 3774 2KC0228 3773 2KC0018
)
3414 1KC0345 3415 1KC0364 3775 2KC0230
==
Description:==
NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs 15 1
I meeting the selection criteria established by the Generic Letter is much greater than that ' previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
These Minor Modifications provide for the diagnostic testing and constitute part of the actions necessary for compliance to NRC Generic Letter 89-10.
The actual. changes involve re-setting the open and close torque switches so that-the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open ana/or fully close the valve disc when
-design basis systems conditions are present.
The minimum required and maximum allowed. thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled i
document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in j
accordance with the latest revision to Duke Power Specification DPS-1205.19 !
0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
1 The inaccuracies of the diagnostic test system used to facilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluation:
The MOVs affected by these Minor Modifications are located in the Component Cooling Water System. The function of 2KC0018, 2KCO228, 2KCO230 and 2KC0003 is to isolate the Reactor Building non-essential headers.and provide component cooling equipment train separation.
The safety position of these valves is closed and their normal position is open. Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of their associated piping will not be affected by re-setting the open and close e
torque switches. Since these mms will ensure the valves will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
The function of 1KCO345 and 1KCO364 is to isolate the thermal barrier cooling water of the Reactor coolant Pumps lA, 1B, 1C, and 1D respectively in case of a rupture of a thermal barrier cooling coil. The safety position of these valves is to close and their normal position is open.
Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of their associated piping will not be affected by re-setting the open and close torque switches. Since these mms will ensure the valves will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
Minor Mod i Valve 3461 1NV0221 3636 1NV0244 3635 1NV0222 Description NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system 16 L
conditions. The level of testing, inspection, and maintenance performed for MOVs j
meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company l
nuclear stations will accomplish in order to comply with the Generic Letter.
These Minor Modifications provide for the diagnostic testing and constitute part of the actions necessary for compliance to NRC Generic Letter 89-10.
The actual changes involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
The inaccuracies of the diagnostic test system used to facilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluation:
The MOVs affected by these mms are in the Chemical and Volume Control System.
The function of INV0244 is to provide containment isolation for the normal charging flow path.
INV0244 is an active valve that is normally open during Modes 1 through 4.
INV0244 receives a safety injection signal to close to satisfy containment isolation requirements. Re-setting the open and close torque switches will not affect open and closure times.
The stress analysis of the associated piping will not be affected by re-setting the open and close torque switches.
Since this MM will ensure the valve will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
The function of INV0221 and 1NV0222 is to open to provide a flow path from the REST to the NV pumps and to close to isolate the REST from the NV pumps. These are active valves that are normally closed during Medes 1 through 4.
These valves receive a safety injection signal to open and a manually initiated signal from the control room to close following the injection phase.
Re-setting the open and close torque switches will not affect open and closure times.
The stress analysis of the piping will not be affected by re-setting the open and close torque switches.
Since these mms will ensure the valves will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
MM-3223 The purpose of this modification is to remove pressure switches which monitor normal and emergency DG fans. The sole function of the pressure switches to be removed is to alarm upon sensing pressure below setpoint.
Each normal and emergency f an is monitored by one of these switches. Upon sensing low pressure, the " Diesel Building Ventilation Malfunction" annunciator in the DG control panel will locally alarm. Also, the " Diesel Panel Trouble" annunciator in the control room will alarm.
Given the nature of the control room alarm, an operator is forced to go to the DG room to see which local alarm has been activated. These 17
pressure switches have a history of high maintenance / replacement and frequent calibration and frequently alarm even though f ans are functioning normally. Fan on/off lights have been provided on the ventilation control panels in each DG room.
Safety Review and USQ Evaluation:
Neither the probability nor consequences of an accident previously evaluated in the FSAR will increase since these switches are non-safety related and are incapable of causing an accident or contributing to it's consequences.
The possibility of creating an accident other than those evaluated in the FSAR does not exist since the DG and its associated equipment are used in accident mitigation and are not normally in operation.
The function of these switches is not important to safety even though they monitor some fans which are.
Malfunctions of these switches have been a common occurrence, however, their failure is incapable of keeping the emergency fans from operating.
Therefore, the probability of a malfunction of equipment important to safety as evaluated in the FSAR is not increased by switch removal.
Since other means have been provided (on/off lights) locally and operators are already forced to go to the DG rooms during emergency situations, a malfunction of the emergency fans will be identified almost immediately.
In the event an operator is not present in the DG room during an emergency, OAC alarms sound at 115 degrees F and local high temperature thermostats trigger both the local and control room alarms which will bring an operator to the room. Therefore, removal of these switches will not prevent detection of an emergency f an malfunction any sooner than is currently possible with switches in place and consequences of the malfunction are unaffected. No USQ exists.
MM-3354 This modification is to expand selected tubes in the 1A, 1B and 2A Containment Spray (NS) Heat Exchanger (HX) to stabilize heat exchanger baffle plates.
The baffle plates support and stabilize heat exchanger tubes.
The expanded tubes will substitute for tie rods and tie rod spacer sleeves that normally hold the baffle plates in the correct positions.
The tie rod spacer sleeves have been weakened due to extensive corrosion. The weakened sleeves may permit the baffle plates to move, changing heat exchanger performance or allowing tube vibration leading to tube failures.
Six tubes in key locations within each heat exchanger will be selected for expansion.
The number of tubes required for expansion was based on load calculations.
Selection of the six tubes in each HX will be based on eddy current test results at the time of modification implementation.
The six selected tubes per HX will be hydraulically expanded above and below each baffle plate on both the inlet and outlet sides. The expanded tubes will be plugged to remove them from service.
Safety Review and USQ Evaluation:
The NS system is not an accident initiator in any accident analyses. There are no electrical changes involved with the modification.
Seismic loading was included in the calculation used as the basis for the number of tubes to be expanded. Therefore, the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
The NS system is used to mitigate high-high containment pressure resulting from accidents such as Loss of Primary coolant or Steam Line Break in Containment.
The ability of the NS system to perform its safety function is not reduced by the modification. The allowed tube plugging margin will not be exceeded because of 18
the modification. The performance of other plant safety functions will not be degraded because of the modification. The functions performed by the NS system or NS Heat Exchanger will not be changed.
No common failure modes are created between heat changers. Therefore the consequences of an accident or malfunction of equipment important to safety evaluated in the FSAR are not increased.
There are no NS system or heat exchanger functions added. No new failure modes are created. No accidents previously considered incredible are made credible by the modification.
Thus the possibility of an accident or malfunction of equipment of a different type than evaluated in the FSAR will not be created.
There are no changes to safety limits, setpoints, or plant parameters because of the modification. The fission product barriers are not degraded. No assumptions made in any accident analysis are affected by the modification. Therefore, the margin of safety as defined in the basis for any Technical Specification is not decreased. No USQ exists.
MM-3672 The subject modification will be used to replace the existing Standby Nuclear Service Water Intake trash racks with new racks using stainless steel materials and fasteners.
An inspection of the Standby Nuclear Service Water Intake structure revealed deterioration of the trash racks and f asteners. The existing racks and fasteners were galvanized steel.
Safety Review and USQ Evaluation:
The Standby Nuclear Service Water Pond provides the ultimate heat sink of shutdown of McGuire Nuclear Station in the event that normal cooling is not available from Lake Norman (due to an earthquake). The Standby Nuclear Service Water Intake structure is a Seismic Category I structure. The new trash racks and fasteners are seismically qualified. The trash racks perform the function of screening large debris (tree limbs, etc.) from the Standby Nuclear Service Water Intakes. The new trash racks will perform the same function as ef ficiently as the existing racks because of similarity in mesh size.
The new trash racks have been evaluated as a commercial grade item approved for use in this QA Condition 1 application. The new trash racks were reviewed to ensure structural integrity equal to or greater than the original racks.
The Standby Nuclear Service Water Pond, dam, intake structure, or trash racks are not accident initiators in any accident analyses.
The new trash racks are qualified to meet the seismic and QA requirements for use in the intended application.
Therefore, the probability of an accident or malfunction of equipment important to safety previously evaluated in the SAR is not increased.
The Standby Nuclear Service Water System serves as the ultimate heat sink for any accident scenario. The performance of this safety function will not be degraded by the modification. The new trash racks will function in the same manner as the existing racks. There are no common failure modes created by this modification.
Therofore, the consequences of an accident or malfunction of equipment important to safety evaluated in the SAR are not increased.
There are no new functions added by the modification. No new failure modes are created. No accidents previously considered incredible are made credible by the modification.
Thus the possibility of an accident or malfunction of equipment of a different type than evaluated in the SAR will not be created.
There are no changes of safety limits, setpoints, or plant parameters because of the modification. The fission product barriers are not degraded. No assumptions made in any accident analysis are affected by the modification. Therefore, the margin of safety as defined in the basis for any Technical Specification is not decreased. No USQ exists.
19
MM-2678 The purpose of this modification is to allow the replacement of a vendor identified potentially defective component.
The affected component is the
" emergency tappet" associated with the over-speed trip mechanism for the auxiliary feedwater turbine (Terry Turbine).
Safety Review and USQ Evaluation:
The emergency tappet is located within the trip mechanism housed under the turbine governor end bearing housing. Under turbine over-speed conditions, the tappet will receive an upward force f rom the governor fly' weight. The force, and resulting movement, will cause the trip mechanism to release the stop valve spring.
Spring action will close the turbine stop valve eliminating the over-speed condition.
The vendor has indicated the existing tappet may " bind" the mechanism preventing the closure of the stop valve.
The vendor supplied re-designed tappet will remove the potential for an inoperative trip mechanism.
Installation of this new tappet will improve the reliability of the Auxiliary Feedwater Turbine as well as providing improved personnel safety. The design and function of the turbine will not be affected in any manner by the installation of this new emergency tappet. The FSAR and Technical Specifications will not be affected in any manner due to the installation of this exempt variation notice.
The possibility, probability, and consequences of an accident not previously evaluated in the FSAR will not be increased due to the installation. The margin of safety as defined in the bases to any Technical Specification will not be decreased due to replacing this component. The system and component af fected are QA Condition 1 and nuclear safety related. No USQ exists.
MM-2938 The purpose of this minor modification is to install a remote leak rate test system for Containment Purge (VP) inside and outside containment isolation valves.
Safety Review and USQ Evaluation:
There are a total of nine containment penetrations for the VP system, five for purge air supply and four for purge air exhaust. These penetrations are in the upper compartment and lower compartment.
Installation of the remote leak rate test system is proposed for each of the VP containment penetrations.
The purpose of the Containment Purge System is to reduce radioactivity in containment by supplying fresh air which, in turn is exhausted through cleanup filters and unit vent to the atmosphere.
This system performs no safety function during a LOCA and is not nuclear safety related, however, the isolation valves and connection piece between the isolation valves are Duke class B, QA Condition 1.
This modification will be installed as QA Condition 1 since the instrument tubing for the remote leak rate test will be tied-in to the existing penetration pressure test isolation valves located on the piping between the containment isolation valves.
The existing penetration pressure test isolation valves will be placed in the open position and secured. Instrument turbing will be routed from the existing pressure test isolation valves to the new pressure test isolation valves.
The new pressure test isolation valves will be normally closed, and will only be used for containment penetration testing.
The instrument tubing will be 3/8" stainless steel and will be routed along the reactor building concrete wall.
Expansion loops will be installed on all tubing routes to minimize tubing stresses caused by expansion or contraction.
All tubing w311 be installed in accordance with existing installation specifications for class 6S2 instrumentation. The new pressure test isolation valves will be 1/2 inch Dragon 20
i i
valves. These valves are compatible for use in liquid or gaseous systems, and have demonstrated excellent characteristics against leakage through the valve seat as well as the valve stem.
Implementation of this minor modification will facilitate maintenance work on the
-VP system, and it will not degrade the safety of the system. The probability and consequences of an accident previously evaluated in the FSAR will not increase, and the possibility of an accident different than any already evaluated in the FSAR will not be created. The inargin of safety as defined in the bases to any l
Technical Specification will not be reduced. NO USQ exists.
j MM-3531, 3532 l
The purpose of this minor modification is to allow the installation of instrument impulse line " snubbers."
The snubbers are designed to attenuate the pressure spikes seen by the RN Strainer differential pressure switches. Elimination of.
c these process flow rate induced pressure spikes should prevent the RN Strainers j
from going into " backwash" unnecessarily.
Another purpose of this minor modification is to reset the automatic RN Strainer Backwash Timer from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
RN System testing has proven the automatic backwash function is not required under normal, high fouling conditions.
Reducing this frequency will reduce the corrective and periodic i
maintenance performed on the RN Strainer backwash control valves.
1 Safety Review and USQ Evaluations l
Neither change performed under this Minor Modification will have a detrimental l
impact on the design or function of the Nuclear Service Water System. In whole, both changes should lessen the maintenance requirements on four control valves.
The FSAR and Technical Specifications will not be affected in any significant manner due to the implementation of the minor modification.
The possibility, t
probability, or consequences of an accident not previously evaluated in the FSAR j
will not be increased due to these simple instrument and control circuit changes.
The margin of safety as defined in the bases to any Technical Specification will not be decreased in any significant manner due to this minor modification. The instrument impulse line change (snubber installation) is QA Condition 1 and nuclear safety related. The automatic strainer backwash timer change is non-QA l
condition.
i The automatic RN strainer backwash timer setpoint was reached by compromise. The Unit 1 timers have a span of 0-60 hours while the Unit 2 timers have a span of i
0-20 hours. For consistency, all timers will be set for 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> instead of the maximum available.
No USQ exists.
}
MM-3312, 3313, 3314 I
It has been documented in VIL-W 92-09 that there is a potential for inadvertent j
isolation of pressurizer power operated relief valves (PORV). This would be the i
result of a valve stem failure causing the PORV block valve to remain closed, thus inadvertently isolating the PORV. The valve steam failure is attributed to i
a combination of age-related embrittlement and Stress Corrosion Cracking (SCC).
I These minor modifications will replace the valve stems for PORV block valves 1NC31B, INC33A, and 1NC35B. The new stems will be made of SB 637 (Inconel 718).
This material was recommended by the valve manufacturer, Borg Warner Industrial-i Products, after reviewing the block valve design requirements including Reactor Coolant System chemistry and temperatures. The new stems are expected to improve I
block valve integrity and reliability.
Safety Review and USQ Evaluation:
f 21 l
{
r
The purpose of the NC System
- to transport heat from the reactor to the steam generators, where heat is tran ferred to the Feedwater and Main Steam Systems of the secondary side.
The NC system consists of four heat transfer loops connected in parallel to the reactor vessel.
Each loop contains a reactor coolant pump and a steam generator.
In addition, the system contains a pressurizer and a pressurizer relief tank. The pressurizer provides a point in the NC System where liquid and vapor can be maintained in equilibrium under saturated conditions for pressure control purposes.
The pressurizer is equipped with three power-operated relief valves. The PORVs and steam bubble function to relieve NC System pressure during all design transients up to and including the design step load decrease with steam dump.
Each PORV has a 7tely operated block valve to provide a positive shutoff elief valve become inoperable.
The PORVs and associated capability sho-block valve per.
?e following functions: 1) PORVs under manual control can control NC system ressure.
This is a function that is used for the steam generator tube rupture accident coincident with a loss of all of f site power and for plant shutdown. 2) Maintaining the integrity of the reactor coolant pressure boundary.
This function is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage. 3) Block valve manue'ly controlled to unblock an isolated PORV to allow it to be used for manual cont
-f NC system pressure and to isolate a PORV with excessive leakage.
- 4) POT -
- omatically controlled to control NC system
.c reduces challenges to the code safety valves pressure. This is a functit: _
- 5) Block valves manually controlled to isolate for overpressurization events.
a stuck open PORV.
The PORVs can be an accident initiator in the case where a failed open PORV results in a SBLOCA.
Since the block valves will be no less reliable and are available for isolation of the PORV, and otherwise will perform as before, there is no increase in the probability of an accident evaluated in the SAR.
From an accident and transient mitigation standpoint, the mms are expected to increase the likelihood that the POP"a and associated block valve will be available for performing the above safet.
lated functions. Therefore, there is no increase in the consequences of an a dent or of a malfunction of equipment important to safety evaluated in the SAR.
With the exception of the SBLOCA mentioned above, the PORVs and associated block valves cannot initiate accident sequences.
Based on this and since the mms do not affect the operation or function of the valves, there is no creation of the possibility for an accident of a different type than any evaluated in the SAR.
The subject valves and system will function as before. The new block valve stem are expected to make the valves and system more reliable. No new failure modes were identified.
Therefore, there is no increase in the probability of a malfunction of equipment important to safety evaluated in the SAR. Likewise, the possibility for a malfunction of a different type than any evaluated in the SAR is not created.
No safety limit, setpoint, or operating parameter will be changed by these mms.
Hence, the ma:qin of safety as defined in the basis of the Technical Specifications will not be reduced. No USQ exists.
MM 3970 Description NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs 22
i i
i 1
meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company i
nuclear stations will accomplish in order to comply with the Generic Letter.
)
This Minor Modification provides for the diagnostic testing and constitutes part
)
of the actions necessary for compliance to NRC Generic Letter 89-10.
j The actual changes involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering calculation MCC-1205.19-00-0003 and is provided by controlled document MCM-J 205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
The inaccuracies of the diagnostic test system used to facilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluation The MOV affected by Minor Modification 3970 is in the Reactor Coolant System.
INC0056 serves as a containment isolation valve on the nitrogen supply line to the PRT.
The safety function os INC0056 is to close and it's normal position is open.
Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the associated piping will not be af fected by re-setting the open and close torque switches. Since this MM will ensure that INC0053 will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. NO USQ exists.
MM 3974
==
Description:==
NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
This Minor Modification provides for the diagnostic testing and constitutes part of the actions necessary for compliance to NRC Generic Letter 89-10.
i The actual changes involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled
)
document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in l
l 23 i
l
accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine tne minimum required and maximum allowed thrust levels.
The inaccuracies of the diagnostic test system used to facilitate thrust testing l
have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test vill be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluation:
The MOV affected by Minor Modification 3974 is in the Reactor Coolant System.
1NC0056 serves ac a containment isolation valve for the reactor makeup water pumps and header to the PRT. The safety function of INC0056 is to close and it's normal position is open. Re-setting the open and close torque switches will not affect open and closure times. The existing stress analysis of the associated piping will not be affected by re-setting the open and close torque switches.
Since this Minor Modification will ensure that INC0056 will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment imporMant safety previously evaluated in the FSAR is not increased.
No USQ exists.
MM 3975
==
Description:==
NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
This Minor Modification provides for the diagnostic testing and constitutes part of the actions necessary for compliance to NRC Generic Letter 89-10.
The actual changes involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
The inaccuracies of the diagnostic test system used to facilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluation:
The MOV af fected by Minor Modification 3975 is in the Nuclear Service Water System.
1RN0069 isolates RN Train A supply to the CA pumps and will automatically open upon receipt of a low pressure signal from CA motor driven A 24
train pump of CA turbine driven pump. The safety function of 1RN0069 is to open and it's normal position is closed.
Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the associated piping will not be affected by re-setting the open and close torque switches.
Since this Minor Modification will ensure that 1RN0069 will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
MM 3972
==
Description:==
NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
This Minor Modification provides for the diagnostic testing and constitutes part of the actions necessary for compliance to NRC Generic Letter 89-10.
The actual changes involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
The inaccuracies of the diagnostic test system used to f acilitate thrust testing t
have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluation:
The MOV affected by Minor Modification 3972 is in the Chilled Water System. The system function of lYC0083 is to provide isolation for the YC System Train B from the demineralized water system.
This valve is considered Active.
The safety function of lYC0083 is to close and it's normal position is open. Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the associated piping will not be affected by re-setting the open and close torque switches. Since thia Minor Modification will ensure that lYC0083 will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction o.f equipment important to safety previously evaluated in the FSAR is not increased.
No USQ l
exists.
MM 3971 Descriptions NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power i
25
stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
This Minor Modification provides for the diagnostic testing and constitutes part of the actions necessary for compliance to NRC Generic Letter 89-10.
The actual changes involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maxinem allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
t The inaccuracies of the diagnostic test system used to f acilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluation:
The MOV af fected by Minor Modification 3971 is in the Chilled Water System. The system function of lYC0002 is to provide isolation for the YC System Train A from the demineralized water system.
this valve is considered Active.
The safety function is to close and it's normal position is open. Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the associated piping will not be affected by re-setting the open and close torque switches. Since this Minor Modification will ensure that 1YC0002 will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
MM 3973
==
Description:==
NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary assurance that they will function when subjected to design basis system conditions. The level of testing, inspection, and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
This Minor Modification provides for the diagnostic testing and constitutes part of the actions necessary for compliance to NRC Generic Letter 89-10.
The actual changes involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when 26
l design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering calculation MCC-1205.19-00-0003 and is provided by controlled document McM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
The inaccuracies of the diagnostic test system used to facilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be sufficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluation:
The MOV affected by Minor Modification 3973 is in the Make Up Demineralized Water System. 1YM0115 serves as a containment isolation valve for the reactor building demineralized water supply header.
The safety function is to close and it's normal position is open. Re-setting the open and close torque switches will not affect open and closure times. The existing stress analysis of the associated piping will not be affected by re-setting the open and close torque switches.
Since this Minor Modification will ensure that 1YC0115 will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
MM 3375
==
Description:==
NRC Generic Letter 89-10, issued on June 28, 1989, instructs nuclear power stations to develop a program to provide for the testing, inspection, and maintenance of motor operated valves (MOVs) so as to provide the necessary
~
assurance that they will function when subjected to design basis syscem conditions. The level of testing, inspection, and maintenance performed for MOVs meeting the selection criteria established by the Generic Letter is much greater than that previously performed by Duke Power Company nuclear stations.
As required by NRC Generic Letter 89-10, Duke Power Company has developed a comprehensive program plan that describes the actions that Duke Power Company nuclear stations will accomplish in order to comply with the Generic Letter.
This Minor Modification provides for the diagnostic testing and constitutes part of the actions necessary for compliance to NRC Generic Letter 89-10.
The actual changes involve re-setting the open and close torque switches so that the motor operator will produce the necessary torque, that will be converted by the stem nut to thrust, to fully open and/or fully close the valve disc when design basis systems conditions are present. The minimum required and maximum allowed thrust used as the test acceptance criteria has been determined by Engineering Calculation MCC-1205.19-00-0003 and is provided by controlled document MCM-1205.19-00-0039-001. This Engineering Calculation was performed in accordance with the latest revision to Duke Power Specification DPS-1205.19 0002 which establishes the parameters and criteria used to determine the minimum required and maximum allowed thrust levels.
1 The inaccuracles of the diagnostic test system used to f acilitate thrust testing have been included in the Engineering Calculation. The final output thrust level achieved during the diagnostic test will be suf ficient to allow valve operation at design differential pressure and system pressure without exceeding the limitations of the operator or valve components.
Safety Review and USQ Evaluation:
27
)
The MOV affected by Minor Modification 3375 is in the Diesel Generator Engine Lube Oil System. The function of 2LD0113 is to provide a bypass of the Full Flow Lube Oil Filter on high differential pressure. The safety position is open and it's normal position is closed.
Re-setting the open and close torque switches will not affect open and closure times.
The existing stress analysis of the associated piping will not be affected by re-setting the open and close torque switches. Since this Minor Modification will ensure the valve will perform as required for design basis system conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
Minor Mod #
Valve 3468 1NIOO65 3487 2SV0025 3490 2SV0028 3470 INIOO88 3471 INS 0038 3472 1NSOO43 3464 1CF0157 3469 INIC076 3474 ISV0026 3463 1CF0155 3467 1NIO354 3462 1CF0153 3475 ISVOO27 3473 ISV0025 3476 ISV0028 3488 2SV0026 3489 2SV0027 3482 2CA0006
==
Description:==
These modifications are being implemented as a result of the findings in NRC IEB-85-03 "MOV common Mode Failures During Plant Transients Due to Improper Switch Settings,"
and Duke Power's Response to NRC Generic Letter 89-10.
These valve operators are being modified to resolve concerns of not attaining suf ficient thrust to completely pull the valve disc out of its seat under normal or high DP conditions.
Presently on these valves, the torque switch bypass contact is located on the primary switch pack.
Due to the primary switch's characteristics, this bypassing action does not stay in the circuit but for a very short period of time.
This is normally on the range of 5% of total valve travel.
Due to system conditions with the present setup, after the bypass circuit opens, there still may be high resistance in the seating area that could cause the actuator motor to cut off due to insufficient torque. To assure that these valves will open fully, the modified torque switch bypass contact will be in the circuit for 50% of the valve's travel, +/- 25%. This will ensure that the maximum motor torque will be applied to the unseating action. After this bypass circuit drops out, the torque switch will be in the circuit to deenergize the actuator should a high resistance be present after complete unseating.
The torque switch bypass circuit will be moved to the AOP auxiliary switches, and the computer points will be moved to the primary switch pack. By moving the computer indication to the primary switches, this will also give the computer a more accurate stroke time.
Functionally, these valves will operate identically to their present operation.
With these modifications, these valves will be more reliable in obtaining the desired positions. Other indications and interlocks will not be affected.
Safety Review and USQ Evaluation:
)
28 l
The MOV affected by MM 3482 is in the Auxiliary Feedwater System. The function of 2CA0006 is to provide a flowpath from the CA CST for the condensate grade supply to any or all three CA pumps to avoid swapping over to the assured RN suction source.
This valve is normally positioned closed and it's safety position is open.
The existing seismic analysis and stress analysis of the associated piping will not be affected by the implementation of the 50% torque switch bypass modification.
Since this modification will ensure the valve will perform as required for design basis system and emergency conditions, the probability or consequences of an accident or malfunction of equipment important tot safety previously evaluated in the FSAR is not increased. No USQ exists.
The MOVs affected by mms 3489, 3488, 3490, 3487, 3474, 3476, 3473 and 3475 are in the Main Steam Vent System. The function of these valves is to isolate main steam to the Power Operated Relief Valves (PORV).
These valves are normally i
positioned open during Mode 1 to allow steam to the PORVis which are normally
)
closed. The accident operating mode for the station is Mode 1, 100% power. The appropriate block valve will be closed by operator action should a main steam I
PORV fail to reseat after it lifts. To account for mispositioning, this valve should be capable to open across a differential pressure of 1205 psi. This valve does not receive safety power, and does not receive any ESF signal.
The existing seismic analysis and stress analysis of the associated piping will not be affected by the implementation of the 50% torque switch bypass modification.
Since this modification will ensure the valve will perform as required for design basis system and emergency conditions, the probability or i
consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
No USQ exists.
The MOVs affected by mms 3467, 3469, 3470 and 3468 are in the Nuclear Safety Injection System.
Their function is to allow cold leg accumulator safety injection into the reactor coolant system when RCS pressure drops below CLA pressure.
These valves are positioned open with power removed in Mode 4 and above. Their normal positions are open and safety and ESF positions are open.
These valves are considered Active.
The existing seismic analysis and stress analysis of the associated piping will be not affected by the implementation of the 50% torque switch bypass modifications. Since these modifications will ensure the valves will perform as required for design basis system and emergency conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
The MOVs af fected by mms 3463, 3462 and 3464 are in the Condensate System. Their function is to provide system isolation to each auxiliary feedwater nozzle. The eafety function is to close and their normal position is open.
The existing seismic analysis and stress analysis of the associated piping will not be affected by the implementation of the 50% torque switch bypass modification.
Since these modifications will ensure the valve will perform as required for design basis system and emergency conditions, the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. No USQ exists.
The MOVs affected by mms 3472 and 3471 are in the Containment Spray System.
Their function is to isolate the auxiliary containment spray header and provide containment isolation.
This isolation occurs during Modes 1 through 4.
These valves are also required to open, by operator action, to provide flow path from the RHR pumps to the auxiliary containment spray headers. Therefore, the safety position is to close and open and it's normal position is closed. These valves are considered Active. By installing the 50% bypass modification, the open and 29
closed stroke times will be affected, j
The existing seismic analysis and the stress analysis of the associated piping will not be affected by the implementation of the 50% torque switch bypass j
modifications. No USQ exists.
l MM-3677, 3678, 3715 and 3716 Description 1
Duke Power Company has initiated a progran that is intended to lead to a substantial reduction in the number of snubbers in use at McGuire Nuclear Station.
The reduction in the number of snubbers will be accomplished by replacing snubbers with Limit Stops manufactured by Grinnell.
The Limit Stops are passive devices.
There are not active mechanisms to lock up or otherwise fail and expose the piping to unanticipated loads.
Material problems are not anticipated because the Limit Stops are designed with large clearances to be tolerant of hostile environments and are made from austenitic stainless steel.
These characteristics will keep the devices and systems on which they are installed reliable.
Inspection and testing requirements will be reduced. The Limit Stops require no special inspection beyond normal nuclear safety related pipe support inspection.
The primary objective of the program is to provide evidence that snubbers at McGuire can be changed out on a one for one basis with Limit Stops with no unacceptable compromises in the licensing basis. The method to be employed is to perform an in-depth analysis of a comprehensive sample of the McGuire piping with both snubbers and Limit Stops. This comprehensive sample presently consists of four piping stress analysis problems.
Each MM addresses replacement of snubbers in a particular sample piping problem. The results of the analysis and the performance of the Limit Stops will show the licensing basis for the plant remains uncompromised. The samples chosen include diversity in ASME class, pipe size, pipe material, location, analytical problem size, flexibility, and design temperature.
Conclusions will be drawn from the engineering analyses, actual hardware performance and from full scale test results. The conclusions will be applied to the plant as a whole.
These mms replace snubbers with Limit Gap Stops supplied by Grinnell. MM-3677 will replace three snubbers in the Auxiliary Feedwater System with three limit stops. MM-3678 will replace four snubbers in the Main Steam Supply to Auxiliary Equipment System with four limit strops. An analysis was done for each problem comparing the existing analysis to the Limit Stop analysis. The original design analysis was performed with the SUPERPIPE program and was done using seismic j
spectra based upon Regulatory Guide 1.61 damping. The Limit Stop analysis uses j
the GAPPIPE computer program and ASME Code Case N-411 pipe damping. The N-411 damping is approved for use at McGuire.
The GAPPIPE program permits dynamic analyses of systems with all common types of nuclear plant dynamic loading.
GAPPIPE has been well correlated with full scale tests and approved by the NRC after in-depth review and independent verification by Brookhaven National Laberatory.
In both sample problems a seismic analysis was performed with all snubbers replaced with limit stops and the response spectra based on Code Case N-411 damping.
In each problem the conclusion was drawn from the results that there can be a one to one replacement of snubbers versus limit stops.
j l
Safety Review and USQ Evaluation:
The results of the analyses indicate that a one to one replacement of snubbers l
with limit stops is acceptable.
The calculated piping stresses remain within ASME Code allowables.
Computer programs used to analyze the problems are acceptable to the NRC. The characteristics of the limit stops make them at least as reliable as the snubbers they replaced.
Based on the above, the CA and SA systems are capable of performing their required functions during normal and all 30 l
1 other operating conditions. Therefore, the probability of an accident or of a malfunction of equipment important to safety evaluated in the SAR is not J
increased. Since the support and restraint of the systems has been adequately and properly analyzed and determined acceptable, there will be no adverse af fects on the CA or SA systems or on any other system. Based on this and the fact that all systems will respond during all events or accidents essentially as before, there is no increase in the consequences of an accident or of a malfunction of equipment important to safety evaluated in the SAR.
No new failure modes were identified.
The limit stops are at least as reliable as the snubbers.
The QA conditions of the limit stops are the same as the snubbers that they replaced.
There are no new line break locoLivas. Therefore, these mms do not create the possibility for an accident or for a malfunction of a different type than any evaluated in the SAR.
No safety limit, setpoint or operating parameter will be changed by these modifications.
Therefore the margin of safety as defined in the basis of the Technical Specifications will not be reduced.
No USQ exists.
MM-3331 and 3560
==
Description:==
MM-3331 will add 1 inch stainless steel turbine field routed from an existing RN vent closed by 1RNS35 to the Ground Water sump. Valve IRN1065 will be added in the turbine near the WZ sump and will be set to throttle flow from the vent line.
MM-3560 will similarly attach tubing to the vent closed by valve 2RN815 and be throttled by valve 2RN1065.
The 1 inch turbine will be field routed and supported using the instrument tube routir.g criteria for McGuire and thus will be seismically qualified.
The attachment to the RN vent line will be made according to the requirements of the instrument tube routing specification to ensure that the existing RN vent and process piping remain qualified. The added valves 1RN1065 and 2RN1065 will be seismically quelified to Duke class F criteria.
Valves 1RN835 and 2RN815 will normally be operated in the open position while new valves 1RN1065 and 2RN1065 will be operated in a throttled pocition to limit water flow from the RN system into the WZ sump.
The total flow into the sump from RN vents will be limited to ensure that not flooding will occur as a result of this venting during any anticipated accident or loss of sump pump.
Safety Review and USQ Evaluation:
The RN system is not an accident initiator for accidents evaluated in the SAR other than plant flooding. The added turbine is routed to seismic Duke class F criteria and the new piping supported to Duke Class F criteria to ensure that no pipe failure will occur and cause flooding. Therefore there is no increase in the probability of an accident aa evaluated in the FSAR. The venting of air and water through the vent during operation will not impair the RN system from supplying required flow at the necessary pressure to any equipment or system.
Therefore, the consequences of any accident evaluated in the SAR are not increased.
The added piping does not change the use or function of RN or other plant systems. The operating parameters of the af fected systems and components are not changed by this modification.
Therefore, no possibility of an accident of a different type than evaluated in the SAR is created by this modification.
The vent line is used to remove pockets of air from RN water which is available for use as supply to the Auxiliary Feedwater pumps during accidents or events requiring use of the Safe Shutdown Facility.
This modification decreases the probability of malfunction of the pumps should use of this suction path be needed. Therefore, there is no increase in the probability of a malfunction of equipment important to safety as evaluated in the SAR.
31
This modification does not af fect normal operation of safety related equipment, and helps ensure availability of a suction source for safety related CA pumps during accidents or other transients requiring CA pump operation.
Therefore, there is no increase in the consequences of a malfunction of equipment important to safety as evaluated in the SAR. Since RN system operation and other component operation is unaffected by this modification, there is no possibility for a malfunction of a different type than evaluated in the SAR.
The Margin of Safety defined in the basis to the Technical Specifications is related to the confidence in the fission product barriers.
The RN system provides assured cooling water to accident mitigation equipment.
This modification does not impair or degenerate the ability of the RN system to perform any safety function but serves to improve the quality of an available water supply for CA pump suction. This will help mitigate the consequences of an accident.
Therefore the modification does not reduce the margin of safety defined in the basis to the Technical Specifications. No USQ exists.
MM-3286
==
Description:==
The support skid for train A and B control room outside air pressure fan and filters is designated by the design drawings to be grouted once the skid is leveled. During the design SITA audit on the VC/YC system, the skids were found leveled with shim plates and not grouted.
PIR 0-M89-0318 was originated to address the operability of this equipment.
An operability evaluation was performed and the equipment was determined operable.
This equipment is QA condition 1, Safety Class 3.
Minor modification 3286 will revise the appropriate drawings to allow the skids to be shimmed in lieu of grout.
It is not feasible to grout the skids due to a varying gap from zero to 1/2 inch between the skid and floor, and the inability to properly prepare the substrate for grout. This modification will assure this equipment is properly mounted.
Safety Review and USQ Evaluation:
Structural operability of the Control Room outside air pressure fans has already been evaluated to meet all Category 1 loading conditions by Civil.
This minor modification will correct as-built drawings and work orders will be generated to verify existing shims and to install any additional shims needed to enhance the existing bearing area.
This modification will not adversely affect the safety function of the equipment, therefore there is no increase in the probability of an evaluated or possibility of an unevaluated accident or malfunction of equipment important to safety that has been or should have been included in the FSAR.
No setpoint, design limit, or operating parameter is affected by this modification, therefore the margin of safety as defined in the basis for any technical specification will not be reduced.
Consequences of an accident previously evaluated or dif ferent than any previously evaluated in the FSAR, are not increased by the modification. No USQ exists.
MM-3287 and HM-3288 These mms will replace Fuel Transfer Isolation Valve 1(2)KF122 with a new knife gate valve, and add / modify the support restraints for the new valve. The purpose i
of the modifications are to eliminate stem and drive system galling problems which caused the valve to fail. Valve replacement requires modification of two lateral supports on the transfer canal liner, and the addition of two spring hangers supported from the operating floor.
The fuel transfer isolation valve is non-QA, however the supports are QA condition 4 to ensure seismic interaction of the valve and the fuel transfer tube does not create excessive loads on the transfer tube.
Installation requires removal of the existing limit switches and connection of the vendor supplied Namco EA-170 limit switch for open indication.
The existing interlock with the fuel transfer system is not changed.
32
Safety Review and USQ Evaluations The Spent Fuel Cooling System (KF) is designed to remove heat from the spent fuel pool and maintain the purity and optical clarity of the pool water during fuel handling operations.
The purification loop is the primary means for removing impurities from either the refueling cavity, transfer canal water during i
refueling, or the refueling water storage tank water following refueling.
The fuel transfer tube connects the refueling canal (inside the reactor containment) and the spent fuel pool (outside the containment.)
The fuel transfer tube is closed on the refueling canal side by a blind flange at all times except during refueling operation.
The fuel transfer isolation valve located on the spent fuel pool side of the fuel transfer tube serves to isolate the fuel transfer tube during refueling to allow draining of the tube, and provide a low pressure (temporary) containment isolation during refueling modes.
The fuel transfer valve is normally open to provide a flow path of water from the spent fuel pool to the fuel transfer tube for suction to the standby makeup water j
pump for SSF.
The Fuel Transfer Isolation Valve is not considered a containment isolation valve, neither does it have a QAl function. Its function is to isolate the fuel pool. Valve leakage requirements have not changed. Reference to the valve being a containment isolation valve will be removed from the FSAR. The reason for this is that the McGuire FSAR currently reflects in part the original design of the KF system requiring valve KF122 to be locked closed. However, when the Standby Shutdown Facility was fully implemented and declared operable for compliance with 10CFR50 Appendix R, valve KF122 was then required to be locked open during plant operation in order to provide a suction source of borated water for the Standby Makeup Pump to be used as an alternate Reactor Coolant Pump seal injection system. Since the Standby Makeup Pump takes suction from the spent fuel pool via the fuel transfer tube, for Standby Shutdown System (SSS) operation, this valve must remain locked open in Modes 1 - 4.
Environmental conditions and material requirements have been considered.
The fuel transfer isolation valve is non-seismic. Failure of the valve will not cause fuel in the spent fuel pool to be uncovered due to the elevation and building structural arrangement.
The fuel transfer tube to which the isolation valve is attached is part of the containment boundary and is QA 1.
The deadweight and seismic loads transferred to the end of the fuel transfer tube are less than the original design; therefore, the original design basis and associated calculations are considered bounding.
The valve is not considered in any accident scenarios.
Therefore, the probability or consequences of an accident evaluated in the SAR is not increased.
Since the valve will function as before, will not adversely interact with the transfer tube, and is materially compatible with the system, there is no increase in the probability or consequences of a malfunction of equipment important to safety evaluated in the SAR.
No new failure modes were found and no new valve functions are required.
For these reasons and the ones discussed above, the possibility for an accident or for a malfunction of a different type than any evaluated in the SAR is not created.
No safety limit, setpoint, or operating parameter will be changed these mms. The margin of safety as defined in the basis of the Technical Specifications will not be reduced.
No USQ exists.
MM-3857 Evaluation:
This MM will add a piston rod adaptor to the snubber piston rods on pipe hangers 1-MCR-NC-681 and I-MCR-NC-781.
These adapters will allow the hydraulic snubbers associated with these hangers to operate through the necessary range of travel with apprcpriate margin in both expansion and contraction.
The adapters are 33
equivalent to the currently installed adapters except in length. These parts are QA-4.
The current adapters were installed as part of NSM MG-12333.
The total calculated movement of the Reactor Coolant System piping during operation at the hanger location causes one of the snubbers to travel to its fully compressed position.
The other snubber considered in this modification is near its fully compressed limit.
This condition is considered as a design error and is documented in PIP l-M93-1088.
The operability review associated with this PIP determined that both the hangers and piping components and valve would perform all their necessary safety functions and remain within acceptable design margins with the current snubber settings. However, since this condition is outside of the normal installation configuration for snubbers, the MM was generated to correct the snubber cold piston setting. Once implemented, the piping analysis i
will be as considered prior to the discovery of the dimensional error.
Safety Review and USQ Evaluation:
This MM changes safety related supports for the NC system piping in the area of Safety Relief Valves 1NCl, INC2 and 1NC3.
These valves and piping are part of the reactor coolant system boundary and are thus a fission product barrier.
These components are accident initiators in that their failure may result in a small break or a large break LOCA.
Operation of the safety relief valves is required in mitigation of several accidents evaluated in the Safety Analysis Report for McGuire.
Operability of these components is required by plant technical specifications. Proper operation of the pipe supports is required to ensure that these components will be able to perform their required function during all expected thermal and accident conditions.
The modification of the hangers will ensure that appropriate travel is available for all operating and accident movements required and that seismic support will be provided when necessary. Appropriate travel margin will be returned to the hangers such that the mechanical limit of motion for the snubbers is not reached.
The support will then perform as qualified in all analysis considered for this piping and support arrangement prior to PIP l-M93-1088.
Therefore, this modification does not increase the probability of an accident or malfunction of equipment as evaluated in the SAR.
For the same reasons there is no increase in the consequences of any accident or equipment malfunction evaluated in the SAR.
The function of all components and the manner in which they perform their function remains the same.
Therefore, the possibility for a new accident or malfunction of equipment is not created by this modification.
These items are unchanged from the values qualified prior to discovery of the dimensional errors associated with the pipe supports.
Therefore, there is no reduction of the margin of safety as defined in the basis for any technical specification. No USQ exists.
MM-2632 and MM-2633 Description These mms will replace existing pressure switches that automatically open and close the Main Steam PORVs with pressure transmitters feeding pressure switches.
Safety Review and USQ Evaluation:
The Main Steam PORVo are QA Condition components; however, the controls being modified under these mms are non-QA components except for QA-4 mountings.
The new control scheme will provide for more accurate opening and closing of the PORVs and should in no way degrade the valves' functions. However, in the event that the controls fail to open one or more PORV on high steam line pressure the Main Steam Code Safety Valves will open to provide overpressure protection.
In the event that the controls f ail to close one or more PROV as designed the safety related portion of the PORV controls will override the non-safety related portion to isolate the valves.
Also, each PORV has a block valve that can be manually 34
i isolated from the control room as needed.
The probability or consequences of accidents or equipment malfunctions discussed in the FSAR will not be affected by these modifications.
The probability or possibility of accidents or equipment malfunctions not-previously discussed in the FSAR will not be affected.
Also, the margin of safety as defined in the basis to any Technical Specification will not be reduced.
No USQ exists.
MM-3477
==
Description:==
The purpose of the MM is to redesign the instrument loop for Nuclear Service Water System instruments ORNPS6130 (train A) and ORNPS6150 (train b).
These pressure switches provide an " interlock" signal for the VC/YC chiller packages.
Without this signal, the chiller packagea will not start or continue in operation. The new instrument design will delete the "as-built" differential pressure switches used to provide the required signal. The "as-built" pressure switches indirectly monitor RN supply header pressure to the VC/YC chiller package condenser by the use of a 1" condenser bypass line.
The bypass line contains a throttling valve used to create a differential pressure which is sensed by the differential pressure switch.
Safety Review and USQ Evaluation:
This MM will significantly simplify the sensing circuit. A single input pressure switch will monitor the 8" RN supply header pressure and provide the necessary closed circuit when sufficient header pressure is present.
The new pressure taps, impulse lines, and pressure switch should be significantly more accurate and reliable due to the use of corrosion proof materials, location, and an improved pressure switch range.
The new instrument loops will not have any detrimental ef fect on the design or function of the Nuclear Service Water System
)
or this " equipment protection" instrument loop.
The FSAR and Technical i
Specifications will not be affected in any significant manner due to the implementation of the MM.
This mod should be an enhancement to VC/YC chiller package operation.
The possibility, probability, or consequences of an accident not previously evaluated in the FSAR will not be increased due the redesign of these two instrument loops. The entire instrument loop is QA condition 1.
The action of the instrument loop does not perform a safety related function -- it functions as a protection circuit for the VC/YC chiller package condenser. The margin of safety as defined in the bases to any Technical Specification will not be decreased in any manner due to the installation of these new components. No USQ exists.
MM-3676 and MM-3776
==
Description:==
Each unit of the McGuire Station has an independent fuel storage system which includes provisions for safe storage of spent fuel assemblies in a flooded pool, while maintaining a coolable geometry, preventing criticality, and protecting the fuel assemblies from excessive mechanical or thermal loadings.
In order to provide specified shielding and water volumes in the fuel pool during plant operation, system piping provides makeup capabilities.
Borated makeup water can be supplied to the spent fuel pool from the Refueling Water Storage Tank.
Demineralized water can be supplied to the pool by the Reactor Makeup Pumps, and emergency makeup water can be supplied to the pool from the Nuclear Service Water System.
All means of makeup are manually initiated and manually terminated.
35
1 1
i System piping is arranged so that f ailure of any pipeline cannot drain the spent i
fuel pool below the water level required for radiation shielding. A water level above the top of the stored spent fuel assemblies is maintained to limit direct gamma dose.
1RN113 (2RN214) is a check valve in the assured makeup line from the Train 1A (2B) essential header of the RN system to the Unit 1 (Unit 2) Spent Fuel Pool System (KF).
The only safety function of this valve is to allow flow in the l
forward direction.
However, this valve may prevent backflow of potentially contaminated water into the RN System.
It could also function to prevent siphoning KF inventory to the RN System.
The check valve, and thus the RN
{
essential header, is isolated from the KF system and a direct flow path into the Spent Fuel Pool by a manually operated RN valve and a manually operated KF valve.
These isolation valves are Class C normally closed valves and the KF valve is locked closed.
1KF126 (2KF136) is a normally closed vent valve used for flush and hydro.
These mms will remove the internals from check valve 1RN113 (2RN214) and allow them to be eliminated from the inservice testing program.
Additionally, the inte-sls from existing vent valves 1KF136 (2KF136) will be removed and Duke Cla.
elbows will be installed in place of the existing vent caps to direct any i
pot
.a1 flow down towards the pools.
Safety Revise and USQ Evaluation:
1 These modifications will not affect fuel handling. The amount of new material added in the fuel pool area is small enough that potential missile generation is j
bounded by previous tornado missile analysis. Therefore, there is not increase 4
in the probability of an accident evaluated in the SAR.
KF System makeup water is supplied by the same sources, essentially as before.
Makeup flows are not adversely affected, and no system is degraded by the changes.
Based on the l
above, there is no increase in the consequences of an accident or of a malfunction of equipment important to safety evaluated in the SAR.
- Likewise, j
these is no increase in the probability of a malfunction of equipment important i
to safety evaluated in the SAR.
1 RN System Train separation is maintained, as before, by procedure. The potential for siphoning the pools is unlikely since use of the RN System for makeup concurrent with an RN System Class C line break is improbable. Additionally, the now open vents to atmosphere through the valve bodies of 1KF136 (2KF136) will inhibit any siphoning ef fect. Potential backflow of contaminated water into the RN system has been reviewed and is considered acceptable. Based on this and the information in the paragraph above, there is no creation of the possibility for an accident or for a malfunction of a different type than any evaluated in the SAR.
No safety limit, setpoint, or operating parameter will be changed.
Therefore, the margin of safety as defined in the basis of the Technical Specifications will not be reduced. No USQ exists.
MM-3591 I
Descriptions I
I The Component Cooling System (KC) heat exchanger lA water boxes and tubesheets are corroding due to exposure to the RN system water.
The corrosion is potentially damaging to the pressure boundary of the heat exchanger RN side.
This modification will coat the interior surfaces of both the inlet and outlet water boxes and the tubesheets to prevent further corrosion of the carbon steel metal.
The coating material is a 100% solids non-solvent emitting epoxy resin made by Plastocor Inc (400U and 2000U.) Coatings applied will be approximately 200 mils thick on the tubesheets and approximately 40 mils thick on the water boxes. Before applying the coating to the tubesheets, the ends of the tubes will be flared to help lock the cladding (coating) material to the tubesheets. Duke 4
36
Power and vendor personal will jointly perform the installation of the coating material.
Safety Review and USQ Evaluation:
This modification will affect the Component Cooling System (KC), described in FSAR section 9.2.4, and the Nuclear Service Water System (RN), described in FSAR section 9.2.2, The only SSC directly affected is the 2B KC heat exchanger.
KC services as an intermediate system and second boundary between the Reactor Coolant System (NC) and RN and assures any leakage of radioactive fluid from the components being cooled is contained within the station.
During accident mitigation, the KC heat exchangers are used in the event of a LOCA to cool the ND heat exchanger and ND pump mechanical seal heat exchanger.
RN water (tube side) is used as the cooling source for the KC heat exchanger.
RN water is raw water from Lake Norman or the Standby Nuclear Service Water Pond (SNSWP). This water is untreated and therefore initiates corrosion on the carbon steel surfaces of the water boxes and tubesheets.
The shell side of the heat exchanger, on which the KC water flows, does not experience corrosion as a result of the water being chemically treated.
Since the coating will only be applied to the RN side of the water boxes and tubesheets, not to the tubes, heat transfer will not be affected. The entrance and exit of the tubes will be modified by flaring the ends of the tubes.
This j
flaring should enhance flow characteristics of the heat exchanger and help lock the cladding (coating) to the tubesheets.
KC heat exchanger 2B is ANS class 3 (Duke class C).
The coating material will be environmentally qualified as part of a commercial grade package. The material will be shown to be suitable for the application taking into account the ef fects 1
of temperature, pressure, radiation, seismic qualification, water chemistry, etc.
Apparent failure mechanisms of the coating (delamination) are blistering and erosion, with the most serious being blistering. If the coating was to come of f in a large enough sheet, the entrance to the tubes could be blocked, thus affecting the heat transfer of the heat exchanger and rendering it inoperable.
Also, chunks of the coating material could dislodge and interact with components downstream of the heat exchanger, affecting these components operation.
Blistering could occur from improper surface preparation or from cold wall effects.
Cold wall effects occur due to the side opposite the coating being colder than the coating side, causing blistering.
In this application, the KC side (side opposite coating) is warmer than the RN side (coating side) and this condition does not apply.
High quality surface preparation should prevent failure of the coating material. The surfaces of the KC heat exchanger will be prepared in accordance with Structural Steel Painting Counsel (SSPC) standard SP5, which will provide a high degree of quality.
The coating surface will be checked visually for degradation during heat exchanger maintenance during outages.
Inspection of surfaces prior to application of coating and regular scheduled inspections should prevent catastrophic failure of the coating and only allow for normal erosion of the material.
MM-5401
==
Description:==
Thic MM replaces the 525 kV Busline 2A underhung insulators associated with the Busline 2A Motor Operated Disconnect.
These insulators are being replaced as preventative maintenance based on failure of other insulators in a similar service environment. The insulators connect the horizontal structural support 37
steel to the stirrup assembly for the busline disconnect.
The replacement insulators meet all the requirements of ANSI TR No. 391 and are electrically and structurally appropriate for this application.
This modification is not a direct part replacement due to the long lead time for 1
direct replacement parts. All replacement parts are fully qualified for their application and no degradation of operation or reliability is expected as a result of these changes.
Safety Review and USQ Evaluation:
The busline insulator replacements potentially affect the Offsite Power System, which is described in FSAR Section 8.2.1.
The component directly affected is the Busline 2A MOD.
The failure of the Offsite Power System is considered in the FSAR as a Loss of External Load, Turbine Trip and Loss of non Emergency AC Power to Station Auxiliaries.
The changes implemented as a result of this minor modification have no adverse impact on the operation or reliability of any of the Offsite Power System components.
The replacement insulators meet all the requirements for the application for which they are installed. Therefore, there is no increase in the probability or consequence of any accident evaluated in the FSAR. The operation and configuration of the Of f site Power System is unaf fected by this modification.
Therefore, there is no increase in the probability or consequence of a malfunction of equipment important to safety as evaluated in the SAR.
Since the function of the replacement insulators is the same as the previously used components and the manner in which the function is performed is unchanged, no possibility of a malfunction of equipment of a dif ference type than j
evaluated in the SAR is created.
The Of f site Power System is not safety related and is assumed to be lost as part of the most safety evaluations.
The technical specifications require that independent offsite and onsite power sources are available in appropriate combination to ensure the availability of power for safety related components and systems required for safe shutdown and mitigation of accident conditions. This modification does not degrade the operability of the Offsite Power System and therefore does not reduce the margin of safety defined in the bases to the technical specifications. No USQ exists.
MM-5400 Descriptions i
This MM replaces the 525 kV Busline 2B underhung insulators associated with the i
Busline 2B Motor Operated Disconnect.
These insulators are being replaced due to f ailure of the Y phase underhung insulator and subsequent damage to switchyard components. The insulators connect the horizontal structural support steel to the stirrup assembly for the busline disconnect. The replacement insulators are l
manufactured by Locke Insulators.
The replacement insulators meet all the requirements of ANSI TR no. 391 and are electrically and structurally appropriate for these applications.
MM-5400 also replaces the rotating hinge insulator and jaw insulator on the Y phase of MOD 62R. These insulators were damaged when the Busline 2B MOD phase Y insulator f ailed. The replacement insulators are manuf actured by Lapp Insulator.
They meet ANSI TR No. 391 requirements and are electrically and structurally appropriate for this application.
These modifications are not direct replacement due to the long lead time for direct replacement parts.
All replacement parts are fully qualified for their applications and no degradation of operation or reliability is expected as a result of these changes.
The MODS are closed during normal operation.
The purpose of the MODS is to 38
provide isolation for maintenance.
These MODS provide no protective function when the bus is energized and are not designed to be operated under load.
The replacement of the insulators does not affect the operation of the MODS.
Safety Rev.lew and USQ Evaluation:
The busline insulator replacements potentially affect the Offsite Power System, which is described in FSAR Section 8.2.1.
The components directly affected are the Busline 2B MOD and MOD 62R.
The failure of the Offsite Power System is considered in the FSAR as Loss of External Load, Turbine Trip and Loss of Non Emergency AC Power to Station Auxiliaries. The changes implemented as a result of this MM have no adverse impact on the operation or reliability of any of the Offsite Power System components.
The replacement insulators meet all the requirements for the application for which they are installed. Therefore, there is no increase in the probability or consequences of any accident evaluated in the FSAR.
The operation and configuration of the Offsite Power System is unaffected by this modification.
Therefore, there is no increase in the possibility of an accident different from any evaluated in the SAR.
The replacement insulators meet all requirements for the application for which they are used.
The replacement insulators are no less reliable than the previously used insulators, and will perform the same function.
Therefore, there is no increase in the probability or consequence of a malfunction of equipment important to safety as evaluated in the SAR.
Since the function of the replacement insulators is the same as the previously used components and the manner in which the function is performed is unchanged, no possibility of a malfunction of equipment of a different type than evaluated in the SAR is created.
The Offsite Power System is not safety related and is assumed to be lost as a part of most safety evaluations.
The technical specifications require that independent offsite and onsite power sources are available in appropriate combinations to ensure the availability of power for safety related components and systems required for safe shutdown and mitigation of accident conditions.
This modification does not degrade the operability of the Offsite Power System and therefore does not reduce the margin of safety defined in the bases to the technical specifications.
No USQ exists.
MM-3594, HM-3604 and MM-3605 These mms are to install a portable waste processing unit consisting of two 50 cubic foot demineralizers with an associated booster pump and both a bag filter skid and a dual bag filter skid. This addition along with the two 25 cubic foot portable demineralizers currently being used will allow bypass of existing demineralizers and associated filters when desired.
'The additional waste processing equipment will be located in the waste processing building truck bay area and connected to the existing waste system through valves lWL-1054, 1055, 1058 and 1059. Since the waste processing equipment is portable, it is connected to the existing piping using flexible hose which is qualified for the service conditions by the vendor. All components associated with the waste processing units such as valves and instrumentation are provided as a part of the unit package by the vendor (Chem-Nuclear). Only MM-3605 is to install the portable unit. MM-3594 is written to modify building floor load analysis to qualify the building to carry the additional loads associated wit the unit and MM-3604 details the installation of a 600v power supply to the unit.
Safety Review and USQ Evaluation:
3 The waste processing system is not safety related and neither creates an accident analyzed in the licensing basis for the plant nor acts to mitigate the consequences of any analyzed accident.
This equipment neither acts to prevent an accident nor is relied on for accident mitigation. Failure of the equipment such that the process fluid is released may however result in release of 39 N
radioactive fluids. the equipment is located in a seismically qualified, curbed area with a capacity to contain a spill of 25000 gallons of liquid.
Additionally, a sump pump is located in the spill containment to transfer spilled fluid to a seismically qualified tank and thus contain any credible spill. The Auxiliary Floor Drain Tank and Auxiliary Waste Evaporator Feed Tanks which may be used to feed the demineralizers are limited to 25000 gallons capacity by administrative controls. The building structure is qualified to carry all loads identified with the addition of this equipment.
The mode of operation will therefore be similar to the current operation of the portable demineralizes in the area. The added power supply to the unit is not safety related and does not impact safety power to any equipment. No safety equipment is located in the area with the waste processing unit or the spill containment. Therefore, the failure of this equipment will not impact the ability of any safety equipment to perform its function.
The equipment is located and operated to meet the intent of Regulatory Guide 1.143 and ALARA principles for prevention of spills.
The waste processing equipment is not considered in any SAR evaluated accident.
The failure of this equipment will be contained by seismic structures and by sump pump systems in the drainage area of the equipment, such that no offsite release would be expected from this equipment. Therefore, there is no increase in the probability of an accident evaluated in the SAR.
The equipment is not located with safety related equipment and does not directly interact with safety systems.
Therefore, there is no impact on the consequences of any accident evaluated in the SAR as a result of these modifications.
The added portable demineralizers are similar in function and operation to current equipment located in the same area. Thus, no new accident or equipment malfunction, not previously considered, is created by this addition.
Since the equipment is not located near safety related equipment and does not interact with safety systems during operation, there is no increase in the probability of malfunction of safety equipment or the increase in the consequences of safety equipment malfunction as evaluated in the SAR.
The Margin of Safety defined in the basis to the Tech. Specs. is related to the confidence in the fission product barriers. The portable demineralizers do not perform any function as part of the fission product barriers. Therefore, this modification does not reduce the margin of safety defined in the basis to the Technical Specifications.
No USQ exists.
MM-3326 This MM has been generated to add a note to the pressurizer lower lateral support drawing to document concrete condition / damage found at the 91 -10'-25" wall connection at elevation 751+8 1/4".
The damage was identified during the unit 2 ISI in lower containment and reported per PIR 2-M92-0048.
The damage on the inboard side of the connection will be repaired during the next refueling outage (2EOC8) per station procedure MP/0/A/7700/18. The small gap (less than 1/8") on the outboard side of the connection, between the embedded plate and the concrete, will be left as-is to allow for thermal expansion at the connection.
Safety Review and USQ Evaluation:
Structural operability of the pressurizer lower lateral support has already been evaluated to meet all category 1 loading conditions by civil engineering per the es aluation for PIR 2-M92-0048. Therefore, the structural integrity has not been conpromised and the possibility of a new accident or new malfunction of equipment important to safety has not been created. Since no system or piece of equipment is adversely affected by this modification, as discussed above, the probability or consequence of any malfunction of equipment important to safety or accident previously evaluated will not be increased.
No safety limit, setpoint, or operating parameter will be changed by this MM, therefore the margin of safety as defined in the basis of the Technical Specifications will not be reduced. No USQ exists.
40
MM-3567
==
Description:==
The purpose of this MM is to allow the installation of two new ball type valves to serve as isolation between the Auxiliary Feedwater Condensate Storage Tank and 3
Flow Elements 1 CAFE 5330 and 1 CAFE 5340.
The "as-built" piping arrangement does i
not provide the necessary isolation for these instruments to allow periodic or 1
corrective maintenance.
With the existing arrangement, the AFWCST must be isolated and drained to support maintenance of these instruments or freeze seal employed. Since the AFWCST is a common component for both Unit 1 and 2 Auxiliary Feedwater Systems and draining the tank is neither desired or easily accomplished (42,500 gallons), the installation of the two ball type isolation valves is necessary. The design and function of the Auxiliary Feedwater System on Unit 1 or 2 will not be af fected or compromised in any detectable manner. Installation i
of these two new ball valves will have an insignificant ef fect on the ability of the Hotwell Pumps from either unit to supply the Auxiliary Feedwater Condensate Tank.
Safety Review and USQ Evaluation:
The FSAR and Technical Specifications will not be affected in any manner due to the installation of these two ball valves.
The possibility, probability, and consequences of an accident not previously evaluated in the FSAR will not be increased due to implementation of this MM.
This MM will eliminate the possibility of an " uncontrolled" loss of condensate should either of the affected flow instruments rupture. The margin of safety as defined in the bases to any Technical Specification will not be decreased by the implementation of this MM.
The affected piping, instrumentation, and valves are Duke Class G, non-nuclear safety related and Non-QA Condition. No USQ exists.
MM-3829
==
Description:==
The purpose of this evaluation is to determine if an unreviewed safety question is involved with MM-3829 for plug repair of Unit 2 steam generator tubes. Steam generator tubes are periodically ECT inspected in accordance with station Technical Specifications to identify defective tubes. Defective tubes identified during the inspection are removed from service by installation of a.75" diameter rolled plug in the hot and cold leg side of the tube. This MM will document the installation of all Unit 2 steam generator tube plugs.
A detailed thermal hydraulic analysis and 50.59 evaluation has been performed by the general of fice Nuclear Engineering Safety Analysis Group which supports plugging of up to 15%
i of steam generator tubes in the McGuire units.
Reference is made to DPC l
calculation DPC 1552.0800-0118 and MCC 1552.08-00-0206.
Safoty Review and USQ Evaluation All field work associated with the installation of tube plugs during U2EOC9A will be performed under approved BWNS QA procedures and program. The maximum number of plugged tubes will be verified not to exceed 15% or 701 tubes per stemn generator therefore an unreviewed safety question does not exist with the installation process, procedures, and plug manufacture to ensure that the installation process for rolled tube plugs does not affect steam generator material strength to increase the likelihood of tube cracks or embrittlement.
The installation of tube plugs will not degrade primary boundary integrity. The i
probability or consequences of an accident previously evaluated in the FSAR will not be increased since the utilization of tube plugs in defective tubes serves to maintain the integrity of the RCPB while not af fecting the design function of 41 i
the steam generators or the reactor coolant system.
The probability or consequences of a malfunction of equipment important to safety previously evaluated or different than already evaluated in the FSAR will not be created.
No USQ exists.
j j
MM-3364-
==
Description:==
The purpose of this MM is to provide increased instrument loop process range for 2CALP5090,.2CALP5110, and 2CALP5120.
The existing span of indication for the safety related leg of the af fected. instrument loops is 0-400 GPM.
This range is insufficient for certain accident conditions. This MM will re-calibrate pressure transmitters 2CAFT5091', 2CAFT5101, 2CAFT5111 and 2CAFT5121 for an input range of 0-400 INWD.
This differential pressure span across the affected thin plate orifice corresponds to a flow rate indication of 0-600 GPM..This new loop range is sufficient for all expected accident conditions.
The revised span-for the affected instruments loops will not compromise the design or function of the I.
auxiliary feedwater system in any manner.
This instrument change'will provide-control room operators with significantly increased process monitoring range.
Safety Review and USQ Evaluations The FSAR and Technical Specifications will not be affected in any manner due to the implementation of this MM.
The possibility, probability, or consequences of an accident not previously evaluated in the FSAR will not be increased due to this instrument range revision. The af fected instruments are QA. condition 1 and nuclear safety related. Instrument loops 2CALP5090, 5100, 5110 and 5120 use one thin plate orifice to provide differential pressure signals to two pressure transmitters.
One leg is non-nuclear safety related and provides flow rate indication to the local pump control panels. The remaining leg is safety related and provides flow rate indication to the control room.
This MM affects the j
safety related leg of the instrument loop. The margin of safety as defined in the bases of any Technical Specification will not be decreased in any manner due to the implementation of this MM.
No new failure modes are created.
No USQ exists.
MM-3363
==
Description:==
The purpose of this MM is to provide increased instrument loop process range for ICALP5090, ICALP5110, and 1CALP5120.
The existing span of indication for the safety related leg of the affected instrument loops is 0-400 GPM.
This range is insuf ficient for certain accident conditions. This MM will re-calibrate pressure j
transmitters ICAFT5091, ICAFT5101, ICAFT5111 and ICAFT5121 for an input range of 0-400 INWD.
This differential pressure span across the affected thin plate i
orifice corresponds to a flow rate indication of 0-600 GPM.
This new loop range is sufficient for all expected accident conditions.
The revised span for the affected instruments loops will not compromise the design or function of the auxiliary feedwater system in any manner. This instrument change will provide control room operators with significantly increased process monitoring range.
Safety Review and USQ Evaluation:
l The FSAR and Technical Specifications will not be affected in any manner due to the implementation of this MM.
The possibility, probability, or consequences of an accident not previously evaluated in the FSAR will not be increased due to this instrument range revision. The affected instruments are QA condition 1 and nuclear safety related. Instrument loops 1CALP5090, 5100, 5110 and 5120 use one thin plate orifice to provide differential pressure signals to two pressure transmitters.
One leg is non-nuclear safety related and provides flow rate i
42
indication to the local pump control panels. The remaining leg is safety related and provides flow rate indication to the control room.
This MM affects the safety related leg of the instrument loop.
The margin of safety as defined in the bases of any Technical Specification will not be decreased in any manner due to the implementation of this MM.
No new failure modes are created.
No USQ exists.
MM-3840
==
Description:==
2NC14 is a 3 inch Walworth globe (item #4J-013) manually operated Duke Class A valve. This valve is the first valve of f of the C loop and is the isolation for valves 2NV1 and 2NV2 (normal letdown as shown on flow diagram MM-25553-1.0). The manufacturer's drawings for this valve are in the MCM 1205.00-0466 series of prints. Per Table 5-5, " Active and Inactive Valves in the Reactor Coolant System Boundary", this inactive valve is normally open and open in the Post-LOCA condition.
External leakage was found at the bonnet to yoke and yoke to body threaded connections during a field inspection to locate unidentified reactor coolant leakage.
This modification seeks to (1) reinject the bonnet clamp nut / bonnet / yoke / body interf aces as described in MM-3736, (2) seal weld the bonnet clamp nut to bonnet and yoke and (3) seal weld the yoke to the body.
This modification is covered by temporary procedure TM/2/A/9300/22.
Safety Review and USQ Evaluation:
This safety discussion will evaluate the following:
-Seating of the valve
-Removal of yoke material and welding the yoke to body interface
-Back seating of the valve
-Weld the bonnet clamp
-Leak sealant injection
-Welding Process Control
-Affects of welding on leak sealant
-Valve repair / replacement
-Discussions with Operations on the disposition of the valve due to this MM
-Unreviewed Safety Question (USQ) Evaluation Seating the valve Prior to welding the yoke to the body, the valve will be seated.
This action will apply a force to the yoke which will remove any thread clearances between the yoke and the body.
This event was discussed with Operations since letdown will be isolated by this action. Operations stated that since the Unit would be at mid-loop, letdown would not be inservice.
Removal of yoke material and welding the yoke to body interface In order to access this external leakage location, it may be necessary to remove yoke metal where the yoke turns from running perpendicular to the stem to approximately parallel with the stem near valve body. Removal of yoke material will not occur in the portion on the yoke which currently acts to retain pressure due to the leak sealant injection described in MM-3736. Note the Welding Process Control discussion for the seal welding of this leak path.
If yoke failure at the yoke material removal area occurs during system operation, system pressure (2485 psig design) will act to place the trim against the bonnet.
This action will keep the valve in the normally opened position. The bonnet as 43 l
\\
1 l
always will be pressed against the bonnet gasket and yoke (below the yoke to body threads) by system pressure.
l Beck seating of the valve The valve will be back seated prior to addressing the bonnet clamp. This seating of the machined surf ace on the disc locknut to the machined surf ace on the bonnet will be used to prevent or limit access of the NC system fluid to the packing area.
This is the typical function of the back seat.
Back seating will be accomplished by fully opening the valve with the handwheel.
Weld the bonnet clamp The bonnet clamp will be cleaned and tightened.
The bonnet clamp has be en modified to an oversized design such that it extends beyond its associated washer.
By mechanical means and controlled by welding process control, the bonnet clamp nut will be prepared for welding along with the adjacent bonnet and yoke material.
Welding will be controlled by welding process control as addressed below.
Leak sealant injection This injection will be through the injection ports and into the valve cavities as addressed in MM-3736. This injection will be as described in the appropriate maintenance procedure.
This procedure addresses the volume of the injection compound used, the type of compound used, injection pressures, and number of injections made.
The main purpose of this injection is to prevent any further water leakage in the seal weld area.
This valve has been uninjected/reinjected 4 times previously. During the field inspection that located this external leakage, the valve body temperature near the yoke was measured at a maximum of 225F.
Welding Process Control All welding associated with 2NC14 is installed Duke QA1. Welding process control will be generated per the Station Welding Program to control the welding on Duke class A pressure boundary.
The FWDS to be used are determined by the materials to be welded among other i
elements as per the Corporate Welding Program.
The bonnet, and body are fabricated using SA351 Grade CF8M (stainless steel.) The bonnet clamp nut and yoke are fabricated using SA216 Grade WCB (carbon steel. ) DPC FWDS L-365 and/or 1-264 will be used to seal weld carbon steel to stainless steel components.
L-
)
350 and/or L-255 will be used to seal weld the bonnet clamp nut to the yoke.
Affects of Welding on Leak Sealant A field test was performed by seal welding a carbon steel plate to a stainless steel plate with leak injection compound between the plates. This test found the j
seal weld was not acceptable to code standards but would be a suitable backing for injection compound.
Valve Repair / replacement This valve will be completely overhauled or replaced at the next refueling outage, 2EOC9.
MM-2309 I
The Robertshaw level switches presently installed in the containment and equipment sumps A and B provide Hi-Hi computer alarms when the water level 44 l
increases to 14 inches.
Separate level transmitters, lWLLT5250 and 5260, are also installed in the sumps to monitor the sumps' levels, and provide annunciator alarms. The alarms will provide the signal to manually start sump pump 1.
The computer alarm will provide the signal to manually start sump pump 2.
The level transmitters will automatically stop both pumps on the low level signal.
Currently, the Robertshaw level switches on both sumps are inoperable. The level switch cables are physically too long for the sumps, and water can not get inside the stand pipes due to sludge buildup at the bottom of the sumps.
This modification will modify the switch installation and the stand-pipe, so the level switch cable can be fit inside the stand pipe, and water can enter the stand-pipe regardless of the sludge buildup. Implementation of this VN will bring the level switches to operable status, and the operators will have the use of an alarm which they need in case the sump should fill to Hi-Hi level.
If the level switches are left out-of-service, the pipe chase area could be severely contaminated should the sumps overflow. Implementation of the modification will enhance the operator awareness of plant conditions, and will not degrade the safety of the WL system.
No USQ exists.
MM-3587
==
Description:==
This MM will replace the cartridge filter in the Spent Fuel Pool purification loop prefilter.
The filter presently identified for this application is one rated 0 3 microns per FSAR.
The replacement filter will be one rated @ up to 40 microns.
This purification loop uses a prefilter and a postfilter around the demineralizer. A third filter, also rated @ 3 microns, is used in the skimmer loop.
Initially 40 micron filters will be purchased from Pall Trinity, the original manufacturer of the filter housing, and the current supplier of the present filters. The replacement filter is a direct replacement for the present on in use in that the same part number is used, with a different type indicated for micron rating.
The filter housing is a QA Condition 2 component since it contains a contaminated fluid, however the cartridge is a Non QA Condition.
Safety Review and USQ Evaluation:
The increase in the rating of this cartridge will not have an adverse affect on the overall clarity of the pool since neither the post filter nor the skimmer loop filter will be affected by this modification. Presently, when the prefilter becomes clogged, the bypass around the filter is put in service and the resin trap upstream of the demineralizer collects debris.
When the trap requires cleaning, the demineralizer must be sluiced. This is a costly operation, and the installation of the 40 micron filter will extend the useable service of this prefilter, eliminating the need to operate it in the bypass mode. No USQ exists.
MM-5354
==
Description:==
The cover installed on the motor starter for EMF 52 is for safety reasons.
The starter electrical contacts were accidently contacted with a flexible hose.
Installation of an electrical box will eliminate the possibility of accidental contact with the electrical contacts.
Safety Review and USQ Evaluations Replacement of the FNA-5A fuse with FNQ-10A is to allow for proper operation of the sample pump motor. This motor draws approximately 27 amps starting current.
During testing, the FNA-5A fuse was replaced by FNQ-5A (FNA type fuses no longer used.) The FNQ-5A fuse repeatedly blew when starting the motor. The FNQ-10A is 45
i suitable for this application (hook-up wire used in this application is 16 AWG. )
These changes do not increase the probability or consequences of an accident evaluated in the SAR since they do not af fect the normal operation of the EMF-52 monitor.
EMF 52 is only used for monitoring purposes.
It is not an accident initiator. No USQ exists.
j MM-3570 i
Description The purpose of the MM is to allow the revision of the Test Acceptance Criteria for the Unit 1 Turbine-Driven Auxiliary Feedwater Pump.
The intent is to increase the operating margin for the TD CA Pump No.1 by widening the acceptable performance band from 90% - 104% to 90% - 105% of the baseline pump head curve.
The increase in band width is only 1%.
The TAC sheet change has been analyzed and the 1% increase was found to be acceptable. This change will have no impact on the design and/or function of the Auxiliary Feedwater System for Unit 1.
This change will improve the margin of acceptability for periodic Turbine-Driven Pump testing.
No fjeld work will be performed as a result of the implementation of this MM.
PT/1/A/4252/01 will be revised to reflect the new pump acceptance band.
Safety Review and USQ Evaluation:
The FSAR and Technical Specifications will not be affected in any manner due to the revision of the Unit 1 TD CA Pump No. 1 Test Acceptance Criteria.
The possibility, probability, or consequences of an accident not previously evaluated in the FSAR will not be increased due to this specification change. The affected pump is QA Condition 1 and nuclear safety related.
The margin of safety as defined in the bases to any Technical Specification will not be decreased due to the increase of 1% in operating margin for the Turbine-Driven Auxiliary Feedwater Pump No. 1.
This MM is not being issued to mask an underlying fault with the TD Pump, instead it is being issued to reflect actual pump performance in the upper region of acceptability.
MM-3593 & MM-3779 The purpose of these mms are to allow the revision of instrument control loops 1KCLP5160 and 1KCLP5210. These loops are designed to maintain control of valves 1KC149 and 1KC156 in a position as desired by a manual loader in the control Room.
The existing loop has several faults which tend to prevent satisfactory response to setpoint inputs. The "As-Built" pneumatic transmitter is exceedingly difficult to calibrate and does not hold its calibration satisfactorily.
The controller for this loop is physically several hundred feet remote from the valvo i
positioned. This situation contributes to poor setpoint response. The revision j
to the loop will include modifying the pneumatic transmitter for linear output
)
pressure, installation of a pneumatic square root extractor to provide a linear 1
output flow rate signal, and moving the "As-Built" controller to close proximity of the valve positioned.
This MM will not compromise the design or safety related function of the Component Cooling System.
It should improve the performance of this non-safety related KC load with the added result of significantly less corrective maintenance.
Safety Review and USQ Evaluation:
The FSAR and Technical Specifications will not be affected in any significant enner due to the implementation of this MM.
The possibility, probability, or
(
' sequences of an accident not previously evaluated in the FSAR will not be increased due to this instrument loop revision.
This instrument loop is non-nuclear safety related and non-QA Condition. The impulse lines from the KC Flow element to the input pneumatic transmitter are code 8 and form a safety related pressure boundary.
No instruments affected by this MM perform any safety 4G
function.
The margin of safety as defined in the bases to any Technical Specification will not be decreased due to the implementation of this MM.
MM-3644 The Fue. Handling system consists of the equipment needed for the refueling operation on the reactor core.
Basically, this equipment is compromised of cranes, handling equipment and a fuel transfer system.
The manipulator crane, is a rectilinear bridge and trolley crane spanning the refueling canal with vertical masts extending down into the refueling water. The masts support and guide the gripping and hoisting devices for handling, in the one mast, fuel assemblies, and in the other mast, rod control clusters.
the bridge and trolley motions are used to position either mast over the fuel assembly positions in the core.
The manipulator crane is rated QA-4.
The rod control cluster (RCC or control rod) mast contains a gripper for picking up the RCC element and a guide tube for containing and guiding the RCC element during removal or insertion. The guide tube is lowered out of the mast until it is seated on top of the fuel assembly containing the RCC element. The guide tube upper end is still contained by the mast when the lower end is engaged to the fuel assembly.
A pneumatic gripper is lowered down through the guide tube to grip the RCC element and withdraw it from the fuel into the guide tube. When the RCC element is completely withdrawn, the guide tube, with the RCC element inside, is withdrawn up into the mast.
RCC elements may be repositioned by the RCC mast on the manipulator crane; however, if this equipment is inoperative or if some non-standard operation is required, rod cluster control elements can be transferred from one fuel assembly to another by means of the rod cluster control changing fixture, or in the spent fuel pool.
A rod cluster control changing fixture, located on the refueling canal wall, can be used for transferring control elements from one fuel assembly to another as an alternative to the RCC handling device on the manipulator crane.
Safety Review and USQ Evaluation:
Use of the manipulator crane RCC mast is not part of the Operations procedure.
At McGuire, the entire core is removed from the Reactor Vessel and placed in the spent fuel pool. The appropriate core components are exchanged in the spent fuel pool verses using the RCC mast. As described above, an alternative is available to exchange an RCC in the Reactor Building if desired. Removal of the RCC mast will enhance maintenance activities and fuel handling.
The description of change outlines the procedure to remove the control rod mast.
The procedure secures all parts of the mast prior to being removed. The removal of the mast will be performed after fuel unload (no-mode) and after the reactor head is set.
Safety concerns have been addressed in the preparation of the mast prior to removal and the transport of it out of the building.
Deletion of the RCC mast will not adversely affect the function of the manipulator crane or affect nuclear safety in fuel handling activities. No USQ exists.
i MM-3997
==
Description:==
The purpose of this evaluation is to determine if an unreviewed safety question is involved with FM-3997 for plug repair of Unit 1 steam generator tubes. Steam 47
generator tubes are periodically ECT inspected in accordance with station Technical Specifications or as a result of a forced outage to identify defective tubes. Defective tubes identified during the inspection are removed from service by installation of a.75" diameter rolled plug in the hot and cold leg side of the tube. This MM will document the installation of all Unit 1 steam generator tube plugs. A detailed thermal hydraulic analysis and 50.59 evaluation has been performed by the general of fice Nuclear Engineering Safety Analysis Group which supports plugging of up to 15% of steam generator tubes in the McGuire units.
Reference is made to DPC calculation DPC 1552.0800-0118 and MCC 1552.08-00-0206.
Safety Review ar.d USQ Evaluation:
All field work associated with the installation of tube plugs during the 1/22/94 forced outage will be performed under approved BWNS QA procedures and program.
The maximum number of plugged tubes will be verified not to exceed 15% total for all 4 steam generators, therefore an unreviewed safety question does not exist and Technical Specification changes are not required. Adequate QA controls exist with the installation process, procedures, and plug manufacturer to ensure that the installation process for rolled tube plugs does not affect SG material strength to increase the likelihood of tube cracks or embrittlement.
The installation of tube plugs will not degrade primary boundary integrity.
The probability or consequences of an accident previously evaluated in the FSAR will not be increased since the utilization of tube plugs in defective tubes serves to maintain the integrity of the RCPB while not af fecting the design function of the SGs, or the reactor coolant system.
The probability or consequences of a malfunction of equipment important to safety previously evaluated or different than already evaluated in the FSAR will not be created.
No USQ exists.
MM-3997A
==
Description:==
The purpose of this evaluation is to review the process and affect of removing tube sleeve samples from the H/L side of the ID steam generator for an unreviewed safety question or concern.
Safety Review and USQ Evaluation:
All "yes" and "no" answers for this evaluation concerning the removal of the sleeve samples and the affect of leaving the remaining tube in the SG is substantiated by BWNT safety evaluation 51-1229198.00.
Potential concerns associated with drilling oversize hn'es in the tubesheet have been addressed.
The two oversized tubesheet holes are at least four tube pitches away from each other and will not exceed the minimum ligament dimension as identified in MCMO-1201.01-0352 - D2/D3 SG Misdrilled Hole Analysis.
All field work associated with the installation of the remote welded plugs in the H/L side of tubes R11-C75 and R9-C80 will be performed under approved BWNT QA procedures and program. Adequate QA controls exist with the installation process i
for the remote welded plugs to ensure that the installation will not degrade primary boundary integrity.
No USQ exists.
MM-3695
==
Description:==
This MM will document the modification on 1A KC heat exchanger.
One tube was plugged due to an unmovable obstruction in the tube.
There were 14 tubes previously plugged in the heat exchanger prior to implementation of this MM.
Safety Review and USQ Evaluation:
48 i
i There are two component cooling heat exchanger per unit. One heat exchanger is assigned to each train of KC equipment. One heat exchanger is required during normal plant operation, and is adequate to provide minimum engineered safeguards heat transfer requirements.
The heat exchangers are of horizontal, straight tube, single pass design with nuclear service water circulating through the tubes.
Each heat exchanger has 4100 tubes of which 10% or 410 tubes can be plugged without affecting the required heat transfer capability of the heat exchanger.
Plugging of the obstructed tube in the heat exchanger will prevent possible mixing of RN and KC waters, RN water may have high chloride content which is detrimental to the stainlesu steel components requiring KC water. Implementation of this MM will reduce possibility of tube break in the heat exchanger without affecting its required heat transfer capability.
The component cooling heat exchanger is a QA-1 safety related component.
No USQ exists.
MM-3688
==
Description:==
This MM will document the modification on IB KC heat exchanger. Nine tubes were plugged, four were plugged because of increasing pit depths in the tubes, and five were plugged because of obstructions in the tubes which could not be removed. There t ere seven tubes previously plugged in the heat exchanger prior to implementation of this MM.
Safety Review and USQ Evaluation:
F There are two component cooling heat exchanger per unit. One heat exchanger is assigned to each train of KC equipment. One heat exchanger is required during normal plant operation, and is adequate to provide minimum engineered safeguards heat transfer requirements. The heat exchanger are of horizontal, straight tube, single pass design with nuclear service water circulating through the tubes.
Each heat exchanger has 4100 tubes of which 10% or 410 tubes can be plugged without affecting the required heat transfer capability of the heat exchanger.
Plugging of the damaged / obstructed tubes in the heat exchanger will prevent possible mixing of RN and KC waters.
RN water may have high chloride content which is detrimental to the stainless steel components requiring KC water.
Implementation of the MM will reduce possibility of tube break in the heat exchanger without af fecting its required heat transfer capability. The component cooling heat exchanger is a QA-1 safety related component.
No USQ exists.
MM-3735
==
Description:==
The purpose of this modification is to add a hanger to support the operator of valve IRN21A. This valve serves as the 1A RN strainer backflush automatic supply isolation valve. The valve is horizontally mounted. The weight of the operator is binding the valve, preventing it from going fully closed.
The valve is normally closed, but opens automatically to initiate strainer back flush to protect RN pump A from cavitation due to low suction pressure.
The safety i
position of valve IRN21A is closed.
1 Safety Review and USQ Evaluation:
The hanger will support the valve operator so that the valve will perform its intended function. This modification will not add any additional loads on stress j
to the piping system.
No deEign parameters of this valve or the piping system will be changed.
This modification does not degrade the effectiveness of any
)
system, structure, or component important to safety.
The McGuire Technical 49
Specifications and the FSAR are not af fected and no change to either document is required.
MM-5132
==
Description:==
The purpose of this modification is to increase the torque switch settings for 1RN253 and 1RN276. As presently set-up, the torque switch acts as a back-up to the limit switch should a problem in the actuator or valve occur.
Safety Review and USQ Evaluation:
With a torque switch setting of 1 and flow and differential pressure acrosa the valve, the torque switch has prematurely opened causing the valve to de-energize prior to full valve travel. This revised setting will ensure proper operation under normal and emergency conditions. The valve operation will remain the same, opening and closing on limit.
The ability of the valve to perform its design function will be enhanced by the implementation of this change.
No valve or actuator structural concerns will be exceeded.
This modification will not adversely affect the operability of other components, structures, or systems.
No USQ exists.
MM-5361
==
Description:==
The thimble for location J-10 on Unit 2 has been bent and is therefore not accessible.
This is the calibration path for all 6 detectors.
Calibration of the detectors is still possible through the common mode in the
'C' ten path rotary device. Should the
'C' ten path experience some type of trouble, however, there would then be no acceptable means of calibrating the detectors.
This modification will simply change the tubing between the J-lO and G-9 core locations. Associated indications and OAC software will also change.
Safety Review and USQ Evaluation:
Core locations G-9 and J-10 are both located near the center of the reactor core.
Location G-9 is suitable for use for this application.
This change does not affect the incore system as described in FSAR sections 7.7.1.9.2, 7.7.1.9.3, 4.2, 4.4.6.1.
This modification will not lead to an increase in probability or consequences of an accident, create a possibility of another type of accident, increase the probability or consequences of equipment j
malfunction.
Tachnical Specifications require that at least 75% of all thimbles / core loc.ations be accessible, plus a minimum of two per quadrant. With the ben thimble at J-10, we will still be above these limits. No USQ exists.
l 50
I Duke Power Company McGuire Nuclear Station Procedure / Miscellaneous changes Completed Under 10CFR50.59 TT/2/A/9100/426 The purpose of this procedure is to setup and maintain Nuclear Service Water 1
System operation for sufficient duration to ensure satisfactory differential pressure testing of 2RN73A (KD HX 2A Outlet Throttle) as mandated by NRC Generic Letter 89-10.
Safety Review and USQ Evaluation:
Performance of this procedure does not place the RN System in any unanalyzed configuration.
It has no potential for degrading or compromising the design or function of the system. This procedure will dynamically VOTES test the af fected valve to ensure it is capable of its safety related function under design basis accident flow rate conditions.
Performance of this procedure has no impact on the operability or availability of the Nuclear Service Water System.
No USQ exists.
PT/2/A/4208/14 The purpose of this testing procedure is to measure the back leakage through check valves 2NS140 and 2NS141. These valves are located near the base of the risers that supply water to the Containment Spray (NS) rings in upper containment. These valves prevent drain down in the risers when the applicable NS pump is stopped thus minimizing water hammer upon restart of the pump.
Safety Review and USQ Evaluation:
The subject procedure performs a periodic leak check of the 10" NS check valves located near the bottom of the risers that supply the NS spray rings in upper containment. This test is performed in Modes 5, 6, or No Mode only. Since the NS system is not required to be operable during these modes, the performance of this test should have no impact on the possiollity or consequences of either an evaluated or unevaluated accident.
Also, there will be no reductions in any safety margins defined by plant Technical Specifications.
The test pressure is well below the design pressure of the NS piping / components.
In addition, the check valves to be tested are isolated from the rest of the NS system. There is the possibility that one or more of the test boundary valves could leak. If unnoticed, it might be possible to discharge water frcm the spray rings in upper containment. A caution has been placed in the. test stating this possibility and providing the test coordinator the option to abort testing if excessive leakage / makeup is observed.
As a result of the above, the implementation of this test should have no effect on the possibility or consequences of any evaluated or unevaluated equipment malfunction.
No USQ exists.
MP/0/A/7700/81 The purpose of this procedure is to allow tube plate hole sealing by tapered weld plug for Westinghouse Model D2, D3 and D5 Steam Generators.
This procedure is utilized when normal steam generator tube plugging methods are not possible, such as when tubes are removed from the tube plate region. Use of this procedure is controlled by Nuclear Station Modification or Exempt Variation Notice.
These documents provide information and detail on current licensing basis tube plugging limits.
Safety Review and USQ Evaluation:
The plugging of steam generator tubes has an impact on the Emergency Core Cooling System accident analysis as presented in chapter 15 of the FSAR.
Plugging also
has an effect on important plant performance characteristics.
Reactor Coolant System flowrate is significantly altered by tube plugging operations. As long as steam generator tube plugging is held within existing licensing basis limits then performance of this procedure will have no affect on the FSAR or Technical Specifications.
This maintenance procedure modifies QA Condition 1 and nuclear safety related equipment.
The possibility, probability, and consequences of an accident not previously evaluated in the FSAR will not be significantly increased due to the performance of this procedure. No Technical Specification changes are mandated by the implementation of this procedure. No USQ exists.
PT/0/A/4550/31 The purpose of this procedure is to control the reconstitution of fuel assemblies, including recage if necessary, by vendor personnel in the Spent Fuel Pool (SFP).
Safety Review and USQ Evaluations The majority of the changes in this reissue are editorial in nature and only serve to enhance the procedure. All applicable Tech Spec surveillance items in 3/4.9 will be followed and documented such that the fuel assembly reconstitution will be performed in a safe and orderly manner per this procedure. Procedure (s) used to do this work will be reviewed prior to work execution and then attached upon work completion.
No major changes were made in this reissue to this previously approved procedure.
However, this procedure does affect components (fuel) addressed in the FSAR in a significant manner since actual fuel rod (s) are removed / replaced during this work.
FSAR Section 4.2.1.2.2 notes that "Some assemblies feature a reconstitutable top nozzle design, which will allow fuel rod reconstitution of fuel assemblies." This statement eludes to the fact that the fuel reconstitution is a process that may be performed on the fuel.
In addition, TS 5.3.1 on Fuel Assemblies addresses that " Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods,.., may be used" which supports the acceptability of the fuel upon reconstitution. FSAR Section 15.7.4.2 and 15.7.4.3 describes fuel handling accidents in the Fuel Building.
The probability or consequences of these accidents are not increased by performing this procedure since only 1 rod is moved / handled at a time vs. an entire assembly drop or a weir gate drop onto assemblies. The procedure will be bounded by the limits established in the Technical Specifications and the analysis provided in the FSAR.
No USQ exists.
PT/2/A/4200/38 The purpose of this procedure change is to provide a prerequisite system condition for the performance of this venting procedure.
Safety Review and USQ Evaluation:
1 System Engineering conducted a review of the calculations with regards to the continuous venting flow rate acceptance criteria.
During this review, the original assumptions and experimental data was checked for accuracy and validity.
System Engineering believes the minimum of three (3) gym venting flow rate requirements is acceptable and contains prudent conservatism. System Engineering does not recommend that the procedure acceptance criteria concerning the
" continuous venting flow rate" be revised at this time.
This venting flow rate is based on the expected amount of " air" coming out of the solution under a design basis accident RN flow rate of approximately 12,000 gallons per minute. Under this condition, the 3 gpm continuous venting flow rate would ensure a water solid assured makeup source f or the Motor-Driven CA Pump 2A and Turbine-Driven CA Pump #2. Under certain system conditions such as low RN Pump 2A total flow rate and lower than normal lake level, it is possible for the
i 1
2A return header pressure to be at a value insufficient to drive the required 3 gpm flow rate.
This condition was observed on February 14, 1994.
To prevent future problems with the inability to meet the procedure acceptance criteria, System Engineering recommends the " continuous venting verifications be performed with the RN Pump 2A total flow rate at 8500 gpm or greater.
As long as this prerequisite system condition flow rate is less than the design basis accident flow rate, there is no impact on the system's ability to expel the " air" coming out of solution.
Essentially, the addition of a prerequisite system condition will attempt to standardize the test conditions and remove the unexpected variables that were encountered on February 14, 1994. This prerequisite system condition change will not alter the continuous venting system's ability to remove unwanted air pockets form the RN Train 2A header.
No USQ exists.
PT/0/A/4700/51 The purpose of this procedure is to evaluate an actual ARO, HZP no Xenon, Equilibrium Samarium critical boron concentration following a startup and compares it to the predicted.
Safety Review and USQ Evaluation:
The process is similar in nature to that explained in FSAR 14.3.2.1 and 14.3.3.4.
This procedure is used only as a tool for evaluation and not used as a tool for changing operating parameters of the plant. Since only numbers are generated by this procedure, the probability and consequences of any accident evaluated in Chapter 15 of the FSAR will not be increased.
There is no possibility of any accident different than evaluated in any section of the FSAR being created. This procedure does not require the operation of any equipment important to safety, therefore the possibility of any malfunction of equipment will not be created.
The margin of safety as defined in the bases to any Technical Specification will not be reduced. No USQ exists.
OP/0/B/6400/08 This procedure revision covers the operation of the YD system (Drinking Water),
and reflects the modifications that were made to delete portions of the system.
Safety Review and USQ Evaluation FSAR Section 9.2.6 (Treated Water Systems) describes the 5000 gallon drinking water storage and pressurizer tank, which equipment has now been removed from service / abandoned.
This system performs no safety function, therefore the probability or consequences of the malfunction of equipment important to safety does not apply.
An accident probability or consequence will not be increased due to this change.
No USQ exists.
TT/2/A/9100/419 The purpose of this procedure is to setup and maintain Residual Heat Removal System operation of sufficient duration to ensure satisf actory dif ferential pressure testing of Safety Injection System electric motor operated gate valve 2ND58A ( A NI and NV Suction from ND Isol. ) as mandated by NRC Generic Letter 89-10 and to verify 2NV-264 movement to close position system retest following leak repair of that valve.
Safety Review and USQ Evaluation:
Although this item does involve an infrequently run test, it does not significantly degrade the level of nuclear safety. The opposite train of decay i
~.
l r
P I
heat removal is available and the VOTES team will be located at valve 2ND-58A as it is being tested if it has to be operated for.any reason.
This' test performs differential pressure (dP) testing of motor operated valve 2ND-58A (A NI & NV Suction from ND 1801.).
It cycles the valve with ND Pump 2A aligned in recirculation to the Cold Legs with all other flowpaths on ND Pump 2A discharge header isolated except for miniflow and mini-mini flow.
The A Train of NI is aligned to the Loops 2 & 3 Hot Legs with the valve closed initially against ND full discharge pressure, which challenges operation of the valve under near design basis conditions by windmilling the NI Pump. This test also aligns ND Pump 2A discharge against check valve 2NV-264 with an otherwise open flowpath through the Boric Acid Blender to the Volume Control Tank.
The flow in this alignment is verified essentially zero to verify this check valve is closed.
These are normal alignments for the NI and ND systems during the NI check Valve Hovement Test PT/2/A/4206/09 (section 12.4) and do not affect operation of the opposite train of ND which will (as necessary) maintain NC system temperature control during the test.
Likewise this test will not affect operation of the opposite train of NI.
Should the valve fail to operate during the test, the ND and NI systems can be returned to normal alignments.
Proper oil levels are checked prior to the test to prevent equipment damage to the windmilling NI Pump and power is disabled for NI Pump 2A to prevent it from starting since the-suction from RWST is isolated.
Since this test will be performed within design conditions using normal system flowpaths, the probability or consequences of a previously evaluated accident are not increased and there is no possibility that any unevaluated accidents could occur. Likewise, the probability of or consequences of a previously evaluated malfunction or equipment should not be increased and there is no possibility of creating an equipment malfunction not previously evaluated. Existing Technical Specifications will be satisfied during the test and the margins of safety as described in the Technical Specification bases shall not be reduced. No USQ exists.
TN/2/A/9700/065 This procedure provides guidelines and requirements for the replacement of the 2B NS Heat Exchanger. This procedure outlines the necessary activities required to ensure that plant safety is maintained and makes implementation personnel i
aware of any limits and precautions necessary to avoid safety hazards and l
undesirable plant interactions.
This procedure will have enclosures which address specific activities required for implementation of'the heat exchanger replacement.
Safety Review and USQ Evaluations The administrative controls incorporated in implementation of this modification l
ensures that there will be no potential for a load drop to cause a spent fuel i
pool accident as discussed in section 15.7.5 of the FSAR.
Therefore, there is l
no increase in probability for this accident. No other analyzed accidents are affected by this modification. No systems or components which are required to be operable during the implementation period will be jeopardized by implementation tasks. Unit 2 will be in no-mode during heavy lifts to prevent the possibility of safety systems being required during periods that they could be damaged by a dropped load.
The accident analyses which assume operation of the containment spray system will not be detrimentally affected by this modification.
No NS system design or s
operating parameters are degraded which would increase the probability of an accident evaluated in the SAR.
The heat exchanger will be installed to meet all seismic and structural requirements to the same extent as the currently installed heat exchanger.
l The implementation of this modification requires that temporary Auxiliary Building boundaries be established to maintain operability of the VA system for
accident mitigation as required by Technical Specifications. The use of VA as an accident mitigation system is not considered in evaluating the consequences of accidents in the FSAR. The control of load paths will eliminate the potential
)
for dropping loads into the spent fuel pool which are different than those previously analyzed. Therefore the consequences of an accident evaluated in the FSAR is not increased.
The consequences of analyzed accidents are not increased by the proposed modification. No system functions important to accident mitigation are degraded by this modification since the design parameters and operating characteristics of the replacement heat exchanger are equal to or superior to the currenuly analyzed equipment.
The potential for dropping a heavy load on the Auxiliary Building roof exists during implementation of this modification. The structures have been analyzed to ensure that any load drop along the prescribed load path will not penetrate the Auxiliary Building and damage equipment.
Therefore the implementation of this modification does not create the possibility of a new accident due to movement of equipment. The installation will replace most of all support steel and platforms around the heat exchanger.
The change in platforms and support steel will not create the possibility of a new accident.
The new equipment will operate in the same manner as the old equipment during accidents since design and operating characteristics are similar. Therefore, the replacement heat exchanger and support structure will not create the possibility of an accident of a different type than previously considered.
Implementation of this modification will occur with Unit 2 in mode 5,6, or no mode. The NS train being modified will not be required during implementation of this modification. No safety related equipment in the work or in areas which may be damaged by the mcdification work will be required to be operable during implementation.
Therefore, the probability of equipment malfunction is not increased during the implementation phase.
The replacement heat exchanger will be less likely.to malfunction due to the application of lessons learned in operation of the current equipment and the improved design of the replacement equipment. Therefore there is no increase in the probability of equipment malfunction during plant operation as a result of this modification.
The unit will be maintained in modes of operation which allow most unit equipment be inoperable during implementation. The equipment which must be operable during this modification will not be affected by this modification due to the use of temporary barriers or the separation of the safety related components in use from the work location.
Therefore the consequences of an equipment malfunction important to safety as evaluated in the SAR are not increased.
The consequences of a malfunction of equipment are not increased by this modification since the replacement heat exchanger design and operating parameters important to accident mitigation are equal to or better than the currently installed equipment.
Neither the installation or the operation of the replacement heat exchangers will create the possibility of a new malfunction of equipment different than that previously considered.
The operation of the replacement equipment is not different than the current equipment for accident purposes. Performance of the replacement equipment is expected to be better than the current equipment due to improved materials and maintenance.
1 The margin of safety as defined in Technical Specifications is related to the confidence in the fission product barriers. The NS system is a fission product barrier during the recirculation phase of a large break LOCA.
The use of NS to maintain containment pressure below design values in post accident conditions is important in the design margin for the steel containment vessel, which is also
___m
__m___
a' fission product barrier. The analysis of this revised system shows that the ability of the system to retain fission products during operation ' is not-degraded. The operation of the NS system to maintain low containment pressure and thus protect the Steel Containment Vessel is not diminished by this modification.
Therefore the modification does not reduce the margin of safety defined in the basis to the Technical ecifications.
NO USQ exists.
1 I
TT/1/A/9100/418
~
The purpose of this procedure is to setup and maintain Residual Heat Removal System operation of sufficient duration to ensure satisf actory dif ferential pressure testing of Safety Injection System electric motor operated gate valve 1 NI-136B (B NI and NV Suction from ND Isolation) as mandated by NRC Generic Letter 89-10.
Safety Review and USQ Evaluation:
Although this item does involve an infrequently run test, it does not significantly degrade the level of nuclear safety. The opposite train of decay heat removal is available and the VOTES team will be located a the valve as it is being tested if it has to be operated for any reason.
This test performs differential pressure (dP) testing of motor operated valve 1NI-136B (B NI and NV Suction from ND Isolation).
It cycles the valve with ND Pump 1B aligned in recirculation to the Cold Legs with all other flowpaths on ND Pump 1B discharge header isolated except for miniflow and mini-mini flow.
The B Train of NI is aligned to the Cold Legs with the valve closed initially against l
ND full discharge pressure, which challenges operation of the valve under near design basis conditions by windmilling the NI Pump. This is a normal alignment for the NI and ND system during the NI check Valve Movement Test PT/1/A/4206/09 (Section 12.4) and does not af fect operation of the opposite trains of ND which will (as necessary) ruaintain NC system temperature control during the test.
i Likewise this test will not affect operation of the opposite train of NI.
Should the valve fail to operate during the test, the ND and NI systems can be returned to normal alignments. Proper oil levels are checked prior to the test to prevent 1
equipment damage to the windmilling NI Pump and power is disabled from NI Pump
]
1B to prevent it from starting since the suction from RWST is isolated.
I l
Since this test will be performed within design conditions using normal system flowpaths, the probability or consequences of a previously evaluated accident are not increased and there is no possibility that any unevaluated accidents could
)
occur.
Likewise, the probability of or consequences of a previously evaluated 1
malfunction of equipment should not be increased and there is no possibility of creating an equipment malfunction not previously evaluated. Existing Technical j
Specifications will be satisfied during the test and the margins of safety as described in the technical specification bases shall not be reduced.
No USQ exists.
PT/1/A/4255/03C PT/2/A/4255/03C The purpose of this procedure is to measure Main Steam Isolation valve travel distance from OPEN to CLOSE at valve normal operating temperature. Valve stroke time is also measured. These two valve operating parameters if within acceptance criteria will ensure valve capability to perform its designed function.
Safety Review and USQ Evaluation:
The following plant safety issues are evaluated when plant systems are set up to
test these valves:
1.
The risk of steam line pressure transient when Main Steam Isolation valve (MSIV) is closed is evaluated.
MSIV is tested one at a time.
All four steam lines are interconnected by an equalization header. With one valve closed and the other three open, the equalization header would maintain the same header pressure downstream of the closed MSIV.
Therefore the pressure transient, if any, should be insignificant.
The Main Steam isolation Bypass Control is kept OPEN during the MSIV test to minimize any pressure transient.
2.
The risk of inadvertent initiating Main Steam Isolation on steam rate signal is also evaluated.
P-ll (1955 psig) the main steam isolation signal is not armed on rate. Au mentioned above, the valve is testud one at a time with the other 3 MSIVs open.
The steam equalization header should minimize any pressure differential across the closed MSIVs.
Steps are in the procedure to open the main steam isolation bypass control to further eliminate any DP across the valve prior to the valve opening.
- Also, the pressure on both sides of the closed MSIVs are monitored prior to valve opening.
3.
Each valve shall be stroked from open to close position.
Close position is the valve itil-safe position.
Therefore, testing the valve does not create an/ now failure mode.
4.
Control room operators shall be walked through the procedure prior to its performance.
5.
The MSIV(s) would isolate on a main steam isolation signal should an actuation occur during the testing.
MP/0/A/7650/78 The purpose of this procedure is to define the method for applying epoxy compounds to containment bellows assemblies.
Safety Review and USQ Evaluation:
This procedure is for application of an epoxy compound to seal existing bellows leaks.
Use of this procedure will have no adverse affect on the containment structure as designed.
The materials used for repair are compatible with the bellows materials during normal operation and accident conditions. The materials used for repair have been analyzed for use on stainless alloys below 20 degree F. Accident analysis during a LOCA event projects maximum containment atmospheric temperatures of 250 degrees F.
Due to the construction and location of the outer ply bellows, these temperatures are not expected.
During normal operation, process fluid temperatures are approximately 567 degrees F.
At these conditions, the surface temperatures of the damaged bellows assemblies have been measured at 130 degrees F maximum, with the actual repair areas measured at approximately 125 degrees F.
This f alls within the materials compatibility range of the epoxy compounds.
This procedure is for repair of equipment.
The consequences of failure / malfunction of this equipment is not affected. The surface _ preparation requirements and use of the prescribed epoxy compounds will create no new equipment failures. The proposed repair is to increase the margin of safety as defined in the Technical Specifications. No USQ exists.
OP/1/A/6100/lOJ OP/2/A/6100/10J The purpose of these procedure changes is to give the operator guidance to make a four hour notification per RP/0/A/5700/10 as an ESF actuation.
This change also gives guidance to perform a channel check within four hours of receiving an
1
" Ice Condenser Lower Inlet Doors Open" alarm.
Safety Review and USQ Evaluation:
These two changes clearly do not meet any of the IOCFR50.59 screening criteria.
The third part of this change gives the shif t guidance to contact the operations Superintendent, or designee, to consider the option to hold the lower inlet doors closed under the referenced conditions.
The Superintendent of Operations or designee must ensure that an unreviewed safety question is not introduced by considering current plant parameters as listed. The initial parameters are that Containment pressure hrs returned to normal and stabilized and the reactor is subcritical. The additional parameters that the OPS Superintendent must consider are the core's decay heat load, primary system temperature and pressure, secondary system temperature and pressure, and containment temperature and pressure. Therefore, this change in and of itself does not create an unreviewed safety question and get the proper level or management involved to ensure safe plant operation. No USQ exists.
MP/1/A/7650/60 MP/2/A/7650/ll6 The purpose of these procedures is to provide acceptable load paths in upper containment for use in lifting and moving heavy loads using the Polar Crane. The procedures provide guidance on inspection, training, and planning prior to making a heavy load lif t and define acceptable load paths for heavy lif ts commonly made during outages. The procedures cover items particular to movement of heavy loads and do not detail all required conditions and steps in moving any particular load.
Operator training and job pre planning are used to further ensure that loads are moved safety and proper precautions are in place for each lift.
For heavy load movements using the Polar Crane which are not directly addressed by these procedures, or which deviate from the procedures, the appropriate forms for designating the load path and approving it in accordance with Duke commitments is provided.
The Polar Crane is not safety related. It is designated QA-4 since its failure has the potential te damage safety related equipment.
Use of the crane is controlled by procedure to minimize the potential for radiological release, from damage to safe-shutdown equipment and decay heat removal systems from any accidental load drop while moving loads in upper containment. The consequence of any load drop accident must be within the limits and assumptions used in analysis of a spent fuel aseembly drop or other accidents as analyzed in the SAR.
The use of the crane during periods when fuel or other systems and equipment may be damaged by a drop is thus of particular concern in meeting commitments to NUREG-0612. Load paths are defined such that loads are not unnecessarily moved over areas which contain systems er equipment required operable for nuclear safety.
Movement over these areas is minimized at all times to lessen the potential for damage which could result in loss of cooling water. The procedures provide sufficient guidance to ensure that movement of heavy loads will be controlled in a safe manner.
The main hook of the Polar Crane meets the definition of a heavy load even when it is not being used.
The procedure provides steps to be taken to ensure that the main hook is restrained from use or from being dropped when auxiliary hoist use makes it necessary to move the main hoist over restricted areas. These measures have been reviewed and are in compliance with Duke's commitment to NUREG-0612.
Safety Review and USQ Evaluations l
The procedures are written to ensure that the chance of a heavy load drop that could result in consequences beyond those considered in the safety evaluation for a spent fuel assembly drop inside containment is minimized. This is accomplished by procedurally controlling the initial conditions to be met prior to any heavy load lift and procedurally controlling the path along which the heavy loads are l
moved. The procedures minimize as much as practical the opportunity for damage to systems or equipment required for safety by a heavy load drop or for cooling i
l l
i water from the spent fuel to be lost.
Therefore, there is no increase in the probability of an accident analyzed in the SAR or the consequences of any accident analyzed in the SAR.
NUREG-0612 is the report on industry practices on use of cranes which handle heavy loads in the vicinity of safety systems and components and the potential consequence of heavy load drops.
Duke's response 4
to this NUREG and the technical evaluation of potential accidents resulting from I
heavy load drops, bound the potential accidents from credible drops within the load paths and other restrictions of these procedures.
Therefore, these procedures do not create the potential for a different type of accident than those previously considered and the consequences of any accident credible using these load paths is not beyond those considered previously. The load paths are consistent with the Duke response to NUREG-0612, therefore, there are not j
accidents made credible by these procedures which were not considered in the Duke response.
The procedures outline the inspections and training required for equipment operation, but do no chanq<; any equipment used in moving heavy loads.
The main hook is considered a heavy load because of its weight, the potential for inadvertent or uncontrollea lowering of this component during crane operation, and because there are not single failure proof mechanisms to prevent this load from falling.
The procedures define this component as passive when power is disconnected from the main hoist such that the main hook is then not considered a load.
This is consistent with the treatment of active components throughout the station.
Therefore, the procedures do not create the possibility for a malfunction of equipment different than any evaluated in the SAR.
The procedures administratively ensure that the fuel and safety systems required to protect the fuel are protected during crane operation. Crane operation is not directly addressed in Technical Specifications. Based on the discussion above, the margin of safety as defined in the basis for technical specifications is not reduced. No USQ exists.
IP/0/A/3090/30 The purpose of this temporary modification is to test the ability of Calgon CL-4000 dispersant to prevent the buildup of sedimentation and fouling of RN system heat exchangers and piping. The actual temporary modification will be to modify the RN pump 1B suction and discharge pressure gauge impulse lines to provide a system tie-in point for the dispersant injection pump. The affected instruments do not serve a safety related function but the impulse lines are a safety related pressure boundary.
In its pure and undiluted form, the dispersant will aggressively attack carbon steel material.
For this TM, the injection pump supply and discharge line material will be stainless steel. The injection pump
" wetted" components are stainless steel. As an added precaution, the dispersant will be heavily diluted prior to injection into the suction side of RN pump 18.
i This is accomplished by creating a pump bypass loop by joining the pump suction and discharge pressure gauge impulse lines. The dispersant will then be injected into this bypass flow rate.
Safety Review and USQ Evaluation The design and function of the RN system will not be compromised in any manner due to the installation of this Temporary Modification. The physical parameters of the affected RN pump or the RN system will not be affected by the use of the dispersant. The dispersant injection pump will only be inservice while RN pump 1B is in operation.
Dispersant pump operation will be controlled by " pump-motor / pressure switch interlock." All wetted components in the RN system have been reviewed and analyzed for interaction with the Calgon CL-4000 dispersant.
At the dilution levels expected, approximately 1-5 ppm, the dispersant will have no significant or noticeable effect on the materials.
No USQ exists.
TT/1/A/9700/103 TT/1/A/9700/lO4 The subject TTs are infrequently run tests.
However, its performance does not significantly degrade the level of nuclear safety since the test utilizes normal
ND flowpaths and the opposite ND train is available to supply the required cooling.
In addition, the tested train will remain operational throughout the performance of the test.
If needed, it could be aligned as necessary to supply any needed cooling.
Safety Review and USQ Evaluation:
The subject TT utilizes normal ND flowpaths to test the operation of 1ND158 and 1ND30A. The "A" and "B" train of ND will be aligned to provide the required test flow and D/P. The "A" and "B" train of ND will be isolated from the "A" and "B" train and in service maintaining NC temperature at acceptable levels.
The "A"
and "B" will remain operational throughout the performance of the TT.
If needed it could be realigned to provide any additional cooling to the NC system.
Consequently, neither the probability or consequences of any unevaluated or previously evaluated accident or equipment malfunction will be created or increased. The requirements of the applicable Technical Specifications will be complied with during the performsnce of the TT.
Therefore, the margin of safety associated with these Technical Specification (s) will not be reduced.
No USQ exists.
TT/2/A/2293/00/AEl TT/2/A/2293/00/AE2 These procedures provide guidance and direction for the "A"
and "B"
train electrical implementation of SM MG-2-2293.
This NSM will upgrade the in containment signal cable for radiation monitors 2 EMF-51A and 2 EMF 51B.
Safety Review and USQ Evaluation:
All work required to implement NSM MG-2-2293 is currently scheduled for modes 4, 5,
6 and no-mode.
Equipment being modified by this NSM is not required during modes 4, 5,6 and no-mode.
2 EMF 51A and 2 EMF 51B have no control functions.
No unreviewed safety questions are created by the implementation of this NSM.
No fission product barriers are affected.
MP/0/A/7150/59 This maintenance procedure provides for use of the Man-Can to act as a radiation shield for a worker and pressure boundary for the refueling cavity during maintenance of a Reactor Coolant Pump.
The Man-Can is installed in the RAP housing after removal of the pump motor and pump impeller. The can is designed to provide shielding to workers working on the pump flange and to provide a seal on the Reactor Coolant System to prevent contamination of the work area and to allow filling the refueling cavity during RAP maintenance. The Man-Can fits into the pump casing, resting on the bottom chelf of the casing.
A double sealing inflatable seal with redundant air supply around the top of the can seals the RCS from the containment atmosphere just below the main RAP flange.
This sealing arrangement allows the flange surface to be un-obstructed for maintenance work.
The Man-Can is held in place by gravity.
The total wet weight of the can is given as 33,458 lbs and is capable of maintaining its position with an uplift head of up to 31.1 feet under static conditions.
This resistance is based on gravity resistance to static head without consideration of seal. friction or dynamic load effects. This resistance is sufficient to ensure that the can will remain in place during maximum flood-up of the reactor cavity with static conditions.
The seal is located at or below elevation 742-11 3/4, while the maximum possible elevation of the pool is 771-10. This represents a maximum head against the seal of 28.85 feet.
The can thus has a margin of 2.25 feet.
In order to ensure that the Man-Can will remain in place as designed during use, the equipment weight will be verified as acceptable prior to flooding the refueling cavity and the seals will be hydro-tested.
I i
1 1
The Man-Can 19 not qualified to maintain the seal during a seismic event.
This j
is not required during no mode operation when there is no fuel in the reactor or the refueling cavity and thus there is no risk of fuel damage due to f ailure of this or other Reactor Building components. The fuel will be stored in the Spent Fuel Pool during use of the Man-Can and will be isolated from the refueling cavity by valve KF-122.
Should the Man-Can fail during movement or storage of the reactor core barrel, there is a potential to drain the refueling cavity to a level which would expose the high activated portions of the core barrel and create very high dose rates in upper containment. The draining of the refueling cavity through the RAP casing would result in the flooding of lower containment with refueling water which is mildly contaminated.
These consequences of seal failure represent a significant commercial risk and ALARA dose concern to the company but do not create a hazard to the public.
The redundant seal design, previous experience with use of the Man-Can at other stations along with extensive hydro-testing of the equipment prior to each use will provide suf ficient assurance that the seal will maintain integrity during the period when the refueling cavity of filled.
Safety Review and USQ Evaluation:
The use of the man-can as a pressure boundary during no mode operation does not increase the probability of consequences of an accident evaluated in the SAR.
No fuel is in the Reactor Building during this mode of operation. There is no significant source of radioactive effluent in the building and thus no accidents are analyzed for this mode of operation. The loss of refueling water during the period that the core barrel is removed is not analyzed in the SAR. This accident is possible due to failure of the man-can seal. This potential accident is not considered since the core barrel is not a significant source of radiation outside the upper containment area except for streaming paths such as the equipment hatch which do not pose a hazard to the general public.
The lower containment would be flooded by the loss of refueling water but this is within the LOCA design of the building and components located in the potentially flooded area. Water in lower containment can be removed by the decay heat removal pumps and containment sump pumps to the Refueling Water Storage Tank or other tanks to allow the man-can to be repaired. The core barrel if uncovered, does pot create 3
an unacceptably high radiation field in lower containment where the man-can is located.
Offsite dose from the core barrel is insignificant.
Therefore, the failure of the man-can to maintain a pressure boundary for the refueling cavity can be considered an accident that does not jeopardize the public and can be readily recovered from, using the equipment available. This is not considered to be a nuclear safety significant accident and is therefore not considered as a different type of accident than those evaluated in the SAR.
There is no increase in the probability of a malfunction of equipment important to safety as evaluated in the SAR.
Since there are not significant sources of dose to the public and no high energy systems in operation, the safety systems and components in the reactor building are not required during no mode operation when the Man-Can will be in place.
For this same reason, there is no increase in the consequence of a malfunction of equipment important to safety.
As discussed above, the loss of water from the refueling cavity, with the core barrel exposed, is not considered as an accident in the SAR.
This is due to the lack of significant safety risk to the public from this accident.
Therefore, this is not considered a nuclear safety significant malfunction which must be addressed in the SAR.
The margin of safety is related to the ability of the fission product barriers to prevent release of ef fluent to the public as defined by the Technical Specifications. There are no Technical Specification applicable to the systems and components involved during no mode operation and thus there is no reduction in the margin of safety. No USQ exists.
T0/1/B/9600/068 This temporary operating procedure provides instructions for placing in service the temporary air supply hoses installed by a TM Addendum to WO 93019487. These
I I
l four hoses cross-connect the "A"
and "D"
trains of the Unit'1 VI blackout
. headers.
They are to be used to supply' air from one blackout header train to components served by the opposite train while the opposite train's main supply header is depressurized for valve ' maintenance.
The 10CFR50.59 evaluation
)
associated with the referenced TM evaluated the installation of the hoses, their ability to pass the required flows, and the probability and consequences of.a hose ruptuce/ leak.
This 10CFR50.59 will evaluate the operation of the hose as well as reiterate the portions of the TM 10CFR50.59 evaluation pertaining to the' probability and consequences of a hose rupture / leak.
Safety Review and USQ Evaluation:
i This procedure provides instructions for depressurizing a blackout header's main supply piping while ensuring a continuous supply of operating air to components-served by the header. It also specifies specific plant conditions during which the hoses can be placed into service. Only one blackout header train at a time shall be depressurized.
The opposite train will remain operational and will supply loads served by the depressurized header via the temporary hoses.
Note that this procedure temporarily crossconnects opposite trains of the VI blackout headers.
Consequently, measures have been taken to help ensure that a rupture / leak in one train does not reduce pressure in both trains to unacceptable levels.
Flow restrictors installed at both ends of the temporary air supply i
hoses will limit maximum flow from a header to a value well within the excess capacity of the VI compressors.
As an added precaution against loss of VI l
pressure, this procedure provides instructions for ensuring that the applicable train of the Diesel Generator Starting Air system (VG) is available to provide a backup source of air, if needed.
In the unlikely event that, despite the above, hose leakage, inappropriate valve manipulations, or loss of VI pressure secure air flow to components served by the blackout headers, they will fail to their safe positions as designed. The consequences of the f ailure of these loads would be further mitigated by the fact that this procedure requires that the Unit be in No Mode while the hoses are in service. With the exception of TS 3/4.9.4, the Technical Specification associated with systems serviced by a depressurized blackout header are not applicable during No Mode.
TS 3/4.9.4 requires an operable air lock door during core alterations or the movement of irradiated fuel within containment.
Consequently, this procedure also contains a requirement that the hoses not be in service during these evolutions.
Some concerns have been expressed about the consequences of a hose failure concurrent with the f ailure of a flow restrictor to limit flow from the blackout headers to within the capacity of the VI compressors.
Such a double failure could possibly depressurize the VI system before operator action could be taken.
This could possibly deflate the reactor cavity seal. If the lower internals were i
removed, the resultant loss of water in the pool could have severe radiological consequences. Given these severe consequences, this procedure also contains a requirement that the hoses not be in service when the lower internals are removed.
As a result of the above, neither the probability or consequences of a previously evaluated or unevaluated accident / equipment malfunction shall be created or increased by the implementation of this procedure. No USQ exists.
TT/2/A/9100/416 TT/2/A/9100/417 Although this item does not involve an infremently run test, it does not significantly degrade the level of nuclear safety.
This test will not affect either train of decay heat removal and the VOM'S team will be located at the valve as it is being tested if it has to be operated for any reason.
Safety Review and USQ Evaluation:
This test performs differential pressure testing of motor operated valves 2NI-118A and 2NI-150B which are the injection valves between the train A and train i
4
B side of the NI Pump discharge header and the cold leg safety injection lines.
It cycles the valve with NI Pump A and B operating aligned with suction from the FWST, miniflow returned to the FWST, and an open flowpath through 2NI-121A to loops 2 & 3 and 2NI-152B to loops 1 & 4 hot legs.
Although not a normal alignment for the NI system, this alignment does not exceed design limits of the system, and does not affect operation of the ND system which will maintain NC system temperature control during the test if necessary.
Sufficient attention is given to monitor NC system inventory during the test and initial conditions allow for accepting inventory from the FWST during the test.
A vent path is ensured to avoid overpressurizing the NC system at the low temperature which will exist during the test.
Should the valve fail to operate properly during the test, the system can be returned to a normal' alignment.
Since this testing will be performed within design conditions using normal system flowpaths, the probability or consequences of a previously evaluated accident are not increased and there is no possibility that any unevaluated accidents could occur.
Likewise, the probability or consequences of a previously evaluated malfunction of equipment should not be increased and there is not possibility of creating an equipment malfunction not previously evaluated. Existing Technical Specifications will be satisfied during the test and the margins of safety as described in the technical specification bases shall not be reduced.
No USQ exists.
TT/1/A/9100/419 Although this item does not involve an infrequently run test, it does not significantly degrade the level of nuclear safety.
this test will not affect either train of decay heat removal and the VOTES team will be located at the valve as it is being tested if it has to be operated for any reason.
Safety Review and USQ Evaluations This test performs differential pressure testing of motor operated valve 1NI-58A (A NI and NV suction from ND Isol.) It cycles the valve with ND Pump 1A aligned in recirculation to the cold legs with all other flowpaths on ND Pump 1A discharge header isolated except for miniflow and mini-mini flow.
The A Train of NI is aligned to the cold legs with the valve closed initially against ND full discharge pressure, which challenges operation of the valve under near design basis conditions by windmilling the NI Pump. This is a normal alignment for the NI and ND systems during the NI check valve movement test PT/1/A/4206/09 and does not affect operation of the opposite train of ND which w'.ll (as necessary) maintain NC system temperature control during the test. Likewise this test will not affect operation of the opposite train of NI.
Should the valve fail to operate during the test, the ND and NI systems can be returned to normal alignments. Proper oil levels are checked prior to the test to prevent equipment damage to the windmilling NI Pump and power is disabled from NI Pump 1A to prevent it from starting since the suction from RWST is isolated.
Since this testing will be performed within design conditions using normal system flowpaths, the probability or consequences of a previously evaluated accident are not increased and there is no possibility that any unevaluated accidents could occur.
Likewise, the probability or consequences of a previously evaluated malfunction of equipment should not be increased and there is not possibility of creating an equipment malfunction not previously evaluated. Existing Technical Specifications will be satisfied during the test and the margins of safety as described in the Technical Specification bases shall not be reduced.
No USQ exists.
TT/2/A/9100/422
Although this item does not involve an infrequently run test, it does not significantly degrade the level of nuclear safety.
This test will not affect either train of decay heat removal and the VOTES team will be located at the valve as it is being tested if it has to be operated for any reason.
Safety Review and USQ Evaluation:
This test performs differential pressure testing of motor operated valve 2NI-162A which is the isolation valve between the NI Pump discharge header and the cold leg safety injection lines.
It cycles the valve with NI Pump 2V operating aligned with suction from the FWST, miniflow returned to the FWST, and an open flowpath through 2NI-150B to loops 1-4 cold legs. This is a normal alignment for the NI system. Although this test will challenge the valve under dP considerably higher than he design basis dP, this test will not exceed design limits of the system, valve or operator. Also, this test does not affect operation of the ND system which will maintain NC system temperature control during the test if necessary. Sufficient attention is given to monitor NC system inventory during the test, and initial conditions allow for accepting inventory from the FWST during the test. A vent path is ensured to avoid overpressurizing the NC system at the low temperature which will exist during the test.
Should the valve fail to operate properly during the test, the system can be returned to a normal alignment.
Since this testing will be performed within design conditions using normal system flowpaths, the probability or consequences of a previously evaluated accident are not increased and there is no possibility that any unevaluated accidents could occur.
Likewise, the probability or consequences of a previously evaluated malfunction of equipment should not be increased and there is not possibility of creating an equipment malfunction not previously evaluated. Existing Technical Specifications will be satisfied during the test and the margins of safety as described in the technical specification bases shall not be reduced.
No USQ exists.
TT/2/A/9100/418 Although this item does not involve an infrequently run test, it does not significantly degrade the level of nuclear safety.
This test will not affect either train of decay heat removal and the VOTES team will be located at the valve as it is being tested if it has to be operated for any reason.
Safety Review and USQ Evaluation:
This test performe dif ferential pressure testing of motor operated valve 2NI-136B (B NI and NV Suction from ND Isol.) It cycles the valve with ND Pump 2B aligned in recirculation to the cold legs with all other flowpaths on ND Pump 2B discharge header isolated except for miniflow and mini-mini flow.
The B train of NI is aligned to the cold legs with the valve closed initially against ND full discharge pressure, which challenges operation of the valve under near design basis conditions by windmilling the NI Pump. This is a normal alignment for the NI and ND system during the NI Check Valve Movement Test PT/2/A/4206/09 and does not affect operation of the opposite trains of ND which will (as necessary) maintain NC system temperature control during the test. Likewise this test will not affect operation of the opposite train of NI.
Should the valve fail to operate during the test, the ND and NI systems can be returned to normal alignments. Proper oil levels are checked prior to the test to prevent equipment damage to the windmilling NI Pump and power is disabled from NI Pump 2B to prevent it from starting since the suction is isolated.
Since this testing will be performed within design conditions using normal system flowpaths, the probability or consequences of a previously evaluated accident are not increased and there is no possibility that any unevaluated accidents could occur.
Likewise, the probability or consequences of a previously evaluated malfunction of equipment should not be increased and there is not possibility of creating an equipment malfunction not previously evaluated. Existing Technical
Specifications will be satisfied during the test and the margins of safety as described in the Technical Specification bases shall not be reduced.
No USQ exists.
TT/2/A/9100/421 Although this item does not involve an infrequently run test, it does not significantly degrade the level of nuclear safety.
This test will not affect either train of decay heat removal and the VOTES team will be located at the valve as it is being tested if it has to be operated for any reason.
Safety Review and USQ Evaluation:
This test performs dif ferential pressure testing of motor operated valve 2NI-135B (B NI and NV Suction from ND Isol.) It cycles the valve with ND Pump 2B aligned in recirculation to the cold legs with all other flowpaths on ND Pump 2B discharge header isolated except for miniflow and mini-mini flow.
The A train of NI is aligned to loops 2 and 3 hot legs with the valve closed initially against ND f ull discharge pressure, which challenges operation of the valve under near design basis conditions by windmilling the NI Pump.
This is a normal alignment for the NI and ND systems during the NI check valve movement test PT/2/A/4206/09 and does not affect operation of the opposite train of ND which will (as necessary) maintain NC system temperature control during the test.
Should the valve fail to operate during the test, the ND and NI systems can be returned to normal alignments. Proper oil levels are checked prior to the test to prevent equipment damage to the windmilling NI Pump.
In order to isolate NC flow to the RWST, the common miniflow line for both NI pumps is isolated as well as the NI suction from RWST. Also both NI Pumps are racked out to prevent damage resulting from inadvertent pump starts with miniflow and the suction from RWST isolated.
Since this testing will be performed within design conditions using normal system flowpaths, the probability or consequences of a previously evaluated accident are not-increased and there is no possibility that any unevaluated accidents could occur.
Likewise, the probability or consegmences of a previously evaluated malfunction of equipment should not be increasec and there is no possibility of creating an equipment malfunction not previously evaluated. Existing Technical Specifications will be satisfied during the test and the margins of safety as described in the Technical Specification bases shall not be reduced.
No USQ exists.
TT/1/A/9700/101 Although this item does not involve an infrequently run test, it does not significantly degrade the level of nuclear safety.
This test will not affect either train of decay heat removal and the VOTES team will be located at the valve as it is being tested if it has to be operated for any reason.
Safety Review and USQ Evaluation:
This test performs dif ferential pressure testing of motor operated valves 1ND-15B (train B ND to hot leg recirc. isol.) and 1NI-183 (ND to B and C hot leg isol.)
It cycles each valve individually with ND Pump 1B aligned in recirculation with ND crossover, ND heat exchanger bypass and all other flowpaths on ND pump 1B discharge header isolated except for miniflow and mini-mini flow.
Then ND is aligned to the hog legs with each valve closed initially against ND full discharge pressure, which challenges operation of the valve under design basis conditions. This is a normal alignment for the NI system during the decay heat removal injection test PT/1/A/4206/10 and does not affect operation of the opposite train ND which will maintain NC system temperature control during the test.
Should the valve fail to operate during the test, the system can be returned to a normal alignment.
Since this testing will be performed within design conditions using normal system
flowpaths, the probability or consequences of a previously evaluated accident are not increased and there is no possibility that any unevaluated accidents could occur.
Likewise, the probability or consequences of a previously evaluated malfunction of equipment should not be increased and there is no possibility of 1
creating an equipment malfunction not previously evaluated. Existing Technical I
Specifications will be satisfied during the test and the margins of safety as described in the technical specification bases shall not be reduced.
No USQ exists.
TT/1/A/9700/102 Although this item does not involve an infrequently run test, it does not significantly degrade the level of nuclear safety.
this test will not affect either train of decay heat removal and the VOTES team will be located at the valve as it is being tested if it has to be operated for any reason.
Safety Review and USQ Evaluation:
This test performs differential pressure testing of motor operated valve IND-30A (train A ND to hog leg recirc. isol.) It cycles the valve with ND Pump 1A aligned in recirculation with ND crossover, ND heat exchanger bypass and all other flowpaths on ND Pump 1A discharge header isolated except for miniflow and mini-mini flow.
Then ND is aligned to the hot legs with the valve closed initially against ND full discharge pressure, which challenges operation of the valve under design basis conditions. This is a normal alignment for the NI system during the decay heat removal injection test PT/1/A/4206/10 and does not affect operation of the opposite train ND which will (as necessary) maintain NC system temperature control during the test.
Should the valve fail to operate during the test, the system can be returned to a normal alignment.
Since this testing will be performed within design conditions using normal system flowpaths, the probability or consequences of a previously evaluated accident are not increased and there is no possibility that any unevaluated accidents could occur.
Likewise, the probability or consequences.of a previously evaluated malfunction of equipment should not be increased and there is not possibility of creating an equipment malfunction not previously evaluated. Existing Technical Specifications will be satisfied during the test and the margins of safety as described in the technical specification bases shall not be reduced.
No USQ exists.
TN/2/A/9700/064 This procedure provides guidelines and requirements for the VA test which is required prior to the 2B NS heat exchanger replacement. The VA test will ensure that the VA system remains operable while the concrete hatch cover is removed during the 2B NS heat exchanger replacement.
This procedure outlines the necessary activities required to ensure that plant and personnel safety is maintained and makes implementation personnel aware of any limits and precautions necessary to avoid safety hazards and undesirable plant interactions.
The Auxiliary Building Filtered Exhaust System (VA) is required to be operable when either Unit 1 or Unit 2 of the station is in modes 1, 2,
3 or 4.
The VA test is being conducted to ensure that the system remains operable when the Nuclear Service Water system (NS) heat exchanger hatch plug is removed.
The implementation procedure is directed toward removal of the 2B NS HX hatch plug to allow the VA testing to be done.
The actual VA test is conducted under a separate procedure.
The implementation procedure provides for removing the sealing material from around the 2B NS HX hatch plug, and lifting the hatch using 4 hydraulic cylinders. Each cylinder has a rated capacity of 10 tons. The combined lifting capacity is 80,000 lbs compared to the hatch weight of 71000 lbs. Even though the jacks are capable of lif ting the entire weight of the plug, they may not be able to break the seal created by any sealing material remaining around the plug at the beginning of the lift. The Manitowoc 4100W Ringer Crane
i will be used to partially unload the jacks as necessary during the initial lift to aid in breaking the seals. Once the seals are broken and the hatch is free, it will be lifted using the four jacks only.
The pumps for the hydraulic cylinders have hydraulic relief valves provided to ensure that the jacks will not l
be overloaded during any part of the lifting procedure.
j The Manitowoc crane will be inspected and tested prior to use to minimize the chance of crane failure during use.
The ringer crane is only used during the initial lifting phase to break the plug seal.
The crane loads are monitored using a load cell. The maximum load and lif t of the crane are limited such that the crane will not be overloaded should the hydraulic jacks fail.
The maximum drop the crane can experience during its use to unload the jacks is 1/4 inch.
The potential damage should the crane fail and the boom collapse onto the roof has been evaluated. The boom may damage the Auxiliary building roof but would not damage any safety related equipment. The roof damage will be of the nature that is can be repaired sufficiently to ensure operability of the VA system in a short period of time as allowed in VA technical specifications. Once the hatch seal is broken and the jacks have begun to lif t the plug, the ringer crane will be unloaded and returned to its safe storage position.
During the hatch lift the weight of the hatch is transferred through the hydraulic cylinders to the Auxiliary Building roof.
The roof is qualified to support this load. Wood blocks are placed under the hatch plug during the lift to limit the distance the plug can f all should the jacks f ail. The maximum load drop is limited to 1.5 inches by use of the blocks. The auxiliary building roof can withstand this drop without f ailure. This also ensures that the plug cannot turn suf ficiently during any drop and f all through the hatch. Therefore, there is no safety consequence to equipment inside the Auxiliary building should the hatch be dropped during lifting.
The weather forecast will be checked prior to beginning the procedure.
Any forecast conditions which would cause the procedure to be unsafe will result in cancellation of the test until safe conditions exist.
The hatch plug will be lifted above the Auxiliary Building roof to the maximum height of the hydraulic cylinders and will provide a gap of at least 8 3/4 inches to allow VA testing.
Support blocking will be provided over the full lift distance to prevent a drop of more than 1.5 inches. VA testing will be conducted and ventilation parameters adjusted to ensure that VA will remain operable with the hatch removed.
The hatch will be lowered back into place in 1.5 inch increments with blocks in place to prevent a drop of greater than 1.5 inches.
The procedure will be implemented under the control of operators in the control room.
If necessary, the procedure can be stopped at any time and the ringer crane unloaded to allow full access to the station for accident response.
The hatch can be closed within one hour if necessary.
Safety Review and USQ Evaluation:
The load path of the ringer crane has been evaluated to ensure that failure of the crane will not initiate an accident as evaluated in the SAR or other accidente not evaluated in the SAR. The lif ting procedure for the 2B NS HX hatch plug ensures that any equipment failure during implementation of the procedure will not initiate an accident or damage equipment important in responding to or mitigating the consequences of an accident. Therefore, the use of this procedure does not create the possibility of an accident of a dif ferent type than evaluated in the SAR or increase the probability of an accident evaluated in the SAR.
The VA system is the only system important to safety that is challenged by this test.
l Duke analysis does not consider the operation of this system in evaluating accident consequences of a malfunction of equipment evaluated in the SAR.
The VA test is being performed too ensure the operability of the VA system the 2B NS
HX hatch open.
Sufficient sealing of Auxiliary building penetrations is done
-prior to the test to provide confidence than the VA test will be successful. The procedure provides for the test to be suspended if necessary and the hatch replaced if necessary to return VA to service if it is determined to be inoperable. Therefore, there is not an increase in probability of a malfunction of equipment important to safety evaluated in the SAR or creation of a malfunction of a different type than evaluated in the SAR.
Margin of safety is related to the confidence in the fission product barriers to function as designed. Implementation of this procedure and test do not degrade j
any fission product barrier or affect any assumptions in the accident analysis.
No safety limits, setpoints or limiting safety system settings are affected.
l Therefore, the margin of safety as defined in the basis to the technical specifications is not reduced.
No USQ exists.
1 TT/2/A/9100/420 TT/1/A/9100/426 TT/1/A/9100/428 l
The purpose of these procedures is to setup and maintain Nuclear Service Water System Operation for sufficient duration to ensure satisfactory differential pressure testing of 1RN174B (KD HX 1B Outlet Throttle), 2RN174B (KD HX 2B outlet throttle) and 1RN73A (KD HX 1A outlet throttle) as mandated by NRC Generic Letter 89-10.
Safety Review and USQ Evaluation:
Performance of these procedures does not place the RN System in any unanalyzed configuration. There is no potential for degrading or compromising the design or function of the system. These procedures will dynamically VOTES test the af fected valves to ensure the capability of their safety related functions under design basis accident flow rate conditions.
The tests have no impact on the
" unavailability" for the diesel generators or the Nuclear Service Water System.
No USQ exists.
e PIP 0-M94-0130 The subject PIR identified a deviation from a commitment made with regard to NUREG-0612 (Control of Heavy Loads); specifically, the polar crane was not to be operated while the af fected unit was in Mode 1 or 2.
Contrary to that, the crane has been used to move the pressurizer enclosure hatch while in those modes.
An evaluation was performed and it was determined that limited use of the polar crane was acceptable during modes 1 and 2, provided the crane's activities are limited to conducting periodic maintenance on the crane, and lifting the pressurizer enclosure hatch.
This is based on the ability of the CRDM missile shields to withstand a drop of the crane main block and cable, and on the ability of the operating deck to withstand a drop of the pressurizer enclosure hatch.
Station procedures that require operation of the polar crane have been revised to reflect appropriate modes for which use of the cranes is approved.
NO USQ exists.
i
\\
\\
i l i !
(
DPC-1552.084X)-Ol l 8 10 CFR 50.59 Evaluation For 15% Steam Generator Tube Flugging - Thennal-Hydraulic Impact I. OBJECTIVE The objective of this calculation file is to document per the requirements of 10 CFR 50.59 that plugging and sleeving of up to the equivalent of plugging 15% of the steam generator tubes in the McGuire and Catawba Nuclear Stations does not constitiute an unreviewed safety question and that no Technical Specification changes am required. This evaluation covers the thennal-hydraulic and dose consequence impacts of plugging tubes and does not cover the actual design or installation of plugs and sleeves including mechanical and material considerations.1he cunent licensing basis is valid for 10% steam generator tube plugging, which may be exceeded due to continued degradation of the tubes. This is a QA Condition I calculation file.
b II. DESCRIFTION OF ANALYSIS The analysis begins with an overview of the thermal-hydraulic impacts resulting fmm steam generator tute plugging on the operation and performance of the unit and on the FSAR licensing
(
basis analyses.1he potentially affected analyses are those in FSAR Chapters 6.2.1 and 15. Each of the analyses in these chapters is individually reviewed to determine if the results of the analysis are affected. Many of the analyses are insignificantly affected, and a conclusion is reached that the additional plugging has no impact. For several of the analyses them clearly is a potential for the results of the analyses to be impacted, and a reanalysis has been perfonned. The results of these reanalyses are summarized in this calculation ille, Then, the questions associated with 10 CFR 50.59 am addressed to determine if an unreviewed safety question exists, or if a Technical Specification change is required.
111. DETAILS OF ANALYSIS Backeround and Methodology 10 CFR 50.59 allows modifications to nuclear stations without prior approval of the NRC pmvided that an evaluation is documented to show that no unreviewed safety questions exist.
The intent of this review process is to detennine the potential impact on nuclear safety and license requirements. The guidance in Reference 1 is utilized in this evaluation. The existing t
FS AR analyses are based on tube plugging levels up to 10%. The 50.59 process is used in this calculation file to justify an increuse in the allowable steam generator tube plugging level imm 10% to 15% for the McGuire and Catawba Nuclear Stations, with the current Westinghouse Model D steam generators. The discussion in this calculation file includes the use of sleeves as well as plugs. Depending on the sleeve design, there exists a number of sleeves equivalent to one plug,in tenus of the increased hydraulic loss of the tube bundle. Although not discussed further in this calculation, it is acceptable to combine the number and types of s!ceves into an k
equivalent number of plugs when comparing the overall effective plugging level of a steam genemtor. Also, for steam generator plugging levels up to 11%, the cutTent analyses remain
4 y
valid provided that the assumptions in those analyses remain valid. The 11% limit is based on the analyzed 10% plus an additional 1% which can be justified by evaluation (Note that this j
evaluation has not been documented). This approach has typically been allowed by the NRC.
'Ihc analyses and justification in this calculation file are specifically valid for a mlatively symmetric plugging level between steam generators. liighly asymmetric plugging would require additional evaluation, although it is not expected to significantly impact the results of this analysis.
'Ihc thennal-hydraulic impacts of plugging steam generator tubes fall mainly into the following three amas:
Decrease In NC Flow: NC flow will decrease following any modification that incmases the loop hydraulic losses. Since plugging or sleeving a steam generator tube decmases the total flow area in the tube bundle, the velocity in the unplugged tubes will increase as will the pressure drop. His increase in loop pressum dmp will cause the NC pump to shift to a lower flow operating point. Although this impact is typically thought of in terms of water flow, it is also applicable to steam flow around the loop following a LOCA.
Decmase in Heat Transfer Area: As steam generator tubes are removed fmm service by plugging, there is less active heat transfer area. For transients that are characterized by an increase in NC heat removal, the rate of heat transfer to the steam generator secondary will be mduced. For transients that are characterized by a decrease in NC heat removal, 3
a reduction in heat transfer area will funher degrade heat transfer. A decrease in heat transfer area also results in lower initial steam generator pressure. This impact also
('
affects some transients.
Decrease In Primary Volume: Although not as significant as the two above impacts resulting from tube plugging, there is some potential for a smaller primary volume to contribute to a more adverse transient response.
in addition to the main contributors listed above, there are other smaller potential consequences resulting from tube plugging. These would include the addition of a small mass of passive metal due to the plug itself and the previously active tube metal now being inactive for heat transfer.
There might also be some small localized heat transfer and flow effects near plugged tubes.
There might also be some impact on steam generator intemal recirculation flow or other bulk processes. All of these minor effects am judged to be insignificant and are not modeled in the analyses or discussed funher in this calculation file.
Of the potential impacts of tube plugging discussed above, only the decrease in NC flow has a Technical Specification impact. NC flow is specifically given with a numerical value in Technical Specifications 2.1,2.2, and 3.2.5. The current value is 385,000 gpm for McGuim and for Catawba Unit 1, and 387.600 gpm for Catawba Unit 2. In the near future Catawba Unit 2 will implement a Technical Specification change to also be at 385,000 gpm. Since NC flow is a Technical Specification number, and since this evaluation is only valid if the allowable plugging limit is increased without changing Technical Specifications, the situation where increased tube plugging causes a reduction in NC flow to below the current Technical Specification values is not addressed by this calculation file. Should tube plugging or other cause result in a decrease in NC flow to below the Technical Specification value, then a separate Technical Specification
' change package will le required along with NRC approval, or the reduction in tractor power
{
permitted by existing Tecimical Specifications (when NC flow does not meet the full power value) will be required.
%c reason that this appmach was taken is that reactor power may have to be decreased as tube plugging increases due to other causes (other than low NC flow). The decrease in steam pressure that results from tube plugging will eventually cause the turbine contml valves to reach the wide-open position. Beyond this point the valves cannot admit the steam generation associated with full power, and the reactor will have to operate below full power. Current projections indicate that this might occur at between 12-15% plugging. If it occurs before the NC flow decreases to below the Technical Specification value, then the reactor power will have to be reduced anyway, and there will be no reason to pursue lowering the NC flow Technical Specification value. There is also an activity in pmgress which may raise the indicated NC flow and restore additional flow margin. If this margin gain can be licensed before the flow dmps to below the current Technical
- Specification limit, then there may not be a need to have the limit lowered.
Le overall impact of the above approach to addressing the decrease in NC water flow as it relates to additional steam generator tube plugging is that it will be ignored in this calculation file. This greatly mduces the scope of the reanalysis effort, since only the impact of increased NC loop hydraulic resistance to steam now during LOCA, less heat transfer area, and less primary volume need be addressed. Therefore, the potential reduction in NC water Dow will not be discussed further.
He NRC-approved Duke Power topical report DPC-NE-3002-A, FSAR Chapter 15 System Transient Analysis Methodology (Reference 2), pmvides the conservative initial conditions and
(
analysis assumptions employed in the analysis of most FSAR Chapter 15 transients. Three other Chapter 15 events are discussed in the NRC-appmved Duke Power topical report DPC-NE-3001-PA, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology (Reference 3). These reports identify for each Chapter 15 transient whether SG tube plugging should be assumed to be at the maximum or minimum (usually zero) plugged percentage, or if the event is insensitive to tube plugging (Note that NC flow is considered separately, not as a part of tube plugging). These reports are used to screen out those events which are insensitive to tube plugging, and to identify which must be reanalyzed or evaluated for an increase in tube plugging. The LOCA analyses documented in FSAR Section 15.6.5 are performed by the fuel vendors, and must be evaluated or reanalyzed for an increase in tube plugging. In addition, the contaimnent mass and energy release analyses in FSAR Section 6.2.1 need to be evaluated for an increase in tube pluggmg.
l l
Lmpact Of Increased Tube Plugging On FS AR Non-LOCA Events The FSAR Chapter 15 non-LOCA transients are currently a cembination of Westinghouse analyses and Duke Power analyses that superceded some of the previous Westinghouse analyses.
For the Westinghouse analyses, the impact of tube plugging on the Westinghouse methodology can be detennined by an earlier Westinghouse safety evaluation that was the basis for increasing the allowable tube plugging to 10% (Reference 4). Westinghouse detennined that the 15.2.6, 15.2.7, and 15.2.8 transients required reanalysis. In addition, if the tube plugging increase resulted in a decrease in NC flow, there is a long list of other transients that require reanalysis.
DPC-NE-3002-A describes the analytical methods and assumptions used to analyze most of the FSAR Chapter 15 events. Based on this report, the following events are not affected by an increase in steam generator tube plugging from 10% to 15% (other than the potential for NC flow I
to decrease). Le basis for this detemiination is discussed in the topical report and is not repeated here, since it is an NRC approved methodology. It is noted that some of these transients
4
{
are identified as being affected by tube plugging, but that they are bounded by other transients.
For those situations the bounding transient is evaluated or reanaly7.ed.
15.1.1 Feedwater Temperature Reduction (Bounded by 15.1.2 or 15.1.3) 15.1.2 Feedwater Flow Increase 15.1.3 Increase In Secondary Steam Flow 15.1.4 Inadvertent Opening Of A SG Relief Or Safety Valve (Bounded by 15.1.5) -
15.2.3 Turbine Tiip (Peak Secondary Pmssure) 13.2.4 Inadvertent MSIV Closure (Bounded by 15.2.3) 15.2.5 Loss of Condenser Vacuum (Bounded by 15.2.3) 15.2.6 Loss of AC Power (Core Response - Bounded by 15.3.2)
(Peak Secondary Pressure - Bounded by 15.2.3)
(Peak Primary Pressure - Bounded by 15.3.2) 15.2.7 Loss of Main Feedwater (Bounded by 15.2.3) 15.2.8 Feedwater Line Break (Shon-Term Core Cooling) 15.3.1 Partial Loss of Flow 15.3.2 Complete Loss of Flow 15.3.3 Locked Rotor 15.4.1 Uncontmiled Bank Withdrawal From Zero Power (Core Response) 15.4.3c Statically Misaligned Rod 15.4.4 Stanup of an Inactive Reactor Coolant Pump 15.4.6 Boron Dilution Transients 15.4.7 Inadve'rtent Loading of a Fuel Assembly in an Improper Posi* ion 15.5.1 Inadvertent ECCS Actuation
(
15.6.1 Inadvenent Opening of a Pressurizer Relief of Safety Valve 15.6.3 Steam GeneratorTube Rupture (Core Cooling)
From the above list it can be seen that the 15.2.6 and 15.2.7 events detemiined by Westinghouse to be affected by tube plugging are identified as bounded by other events in the Duke methodology. The 15.'2.8 event (shon-tenn core cooling)is identified as insensitive to tube plugging, and the 15.2.8 (long-term core cooling)is reanalyzed (see below). Rettfore, the events which are not reanalyzed with Duke methods att evaluated as being bounded or insensitive with Duke methods.
The DPC-NE-3001-PA topical report discusses analytical methods for the following FSAR Chapter 15 non-LOCA events:
15.1.5 Steam Line Break Core Response: On p. 5-10 of Reference 3, justification is presented for assuming no steam generator tube plugging is conservative. Since the steam line break is an overcooling transient, maximizing the cooldown rate would be achieved by maximizing the heat tmnsfer area.
15.4.3a & b Control Rod Drop: On p. 6-5 of Reference 3, justification is presented for assuming no steam generator tube plugging is conservative, ahhough it is also noted that this parameter is not significant for this transient.
15.4.8 Rod Ejection: On p. 4-25 of Reference 3,it is pointed out that most parameters have little impact on the pressure response due to the very shon duration of the
(
analysis. De peak pressure is teached in 1.9 seconds (Note: Revision 2 of Reference 5 indicates a time of 1.85 seconds). There is no time for the
/
(
secondary side to have any affect on the primary side peak pressure. The slightly smaller primary volume caused by the increase in tube plugging would also have little impact. In fact, the analysis does not model any tube plugging (Reference 5, p. 4) The peak pressure is 2758 psia (Reference 5). Based on these arguments and the large margin to the 3000 psia acceptance criterion, there is no need for reanalysis.
Based on the DPC-NE-3001-PA analytical methods and results, it can be concluded that additional steam generator tube plugging up to 15% or higher has no impact on the FSAR steam line break, control rod drop, or rod ejection transients.
Certain FSAR Chapter 15 events are not applicable to McGuire / Catawba (e.g. BWR transients),
or are addressed in the FSAR as being bounded by other analyzed transients. These events, which follow, are unaffected by any change to the plant including tube plugging.
15.2.I N/A forMcGuire/ Catawba 15.2.2 1 Ass of Load (Bounded by 15.2.3) 15.3.4 Shaft Break (Bounded by 15.3.3) 15.4.5 N/A for McGuire / Catawba 15.5.2 CVCS Malfunction That increases inventory (Bounded by 15.5.1) 15.5.3 N/A for McGuire / Catawba 15.6.4 N/A for McGuire / Catawba 15.6.6 N/A for McGuire / Catawba
(
The impact of additional tube plugging on the following FSAR Chapter 15 events is as follows:
15.6.2 Instrument Line Break: This analysis is a dose consequence analysis that is unaffected by steam generator tube plugging.
15.7 Radioactive Release From a Subsystem or Component: These events have nothing to do with steam generator tubes and are unaffected.
15.8 Anticipated Transients Without Scram: These events are analyzed generically as referenced in the FSAR, however, the current regulatory treatment of ATWS is based on risk management. The generic analyses have not been updated to reflect changes in inputs. The key mitigating equipment such as pressurizer code safety relief valves, pressurizer PORVs, the AMS AC circuitry, and the CA System, merit review if modifications to them are contemplated. For other analysis assumptions, such as the level of steam generator tube plugging, there is no risk-significance. A review of the generic FSAR analyses (Reference 6) delennined that many sensitivity studies'were perfomied, but tube plugging is not mentioned. This further confinns that ATWS is insensitive to this-parameter.
Reanalysis of Affected Non-LOCA Transients A revicw of FSAR Chapter 15. DPC-NE-3002-A, and DPC-NE-3001-PA analyses enables climination of the above non-LOCA transients as being insensitive orconservatively impacted by
(
the af fccts ofincreased steam generator tube plugging. Consequently, the remaining non-LOCA l
/
\\
~
(
transients require reanalysis. Each reanalysis has been documented in an approved calculation file. A brief discussion of each follows:
15.2.3 Turbine Trip (Peak Primary Pressure): This event was reanalyzed to detemiine the impact of increasing the analyzed tube plugging level from 10% to 15%.
De results, documented in Reference 7, show that the peak pressure increases fmm 2674 psig to 2677 psig. Since the peak pressure remains below the ecceptance criterion of 2750 psia, the increase in tube plugging is acceptable.
15.2.8 Feedwater Line Break (Long-Term Core Cooling): This event was reanalyzed to determine the impact ofincreasing the analyzed tube plugging level from 10% to 15%. The results, documented in Reference 8. show that the hot leg temperature does not exceed saturation. Therefore, the increase in tube plugging is acceptable.
15.4.1 Uncontmiled Bank Withdrawal From Zero Power (Peak Primary Pressure): This event was reanalyzed to determine the impact ofincreasing the analyzed tube plugging level from 10% to 15%. He results, documented in Reference 9, show that the peak pressure in 7.ases fmm 2698.88 psia to 2700.44 psia. Including an additional 1.63 psi for convening the RETRAN Volume 2 pressure for elevation head to the bottom of the reactor vessel, there remains a large margin to the acceptance criterion of 2750 psia. Herefore, the increase in tube plugging is acceptable.
15.4.2 Uncontrolled Bank Withdrawal At Power: This event was reanalyzed to detennine the impact of increasing the analyzed tube plugging level from 10% to 15%. De results are documented in References 10 & 11. The results of the com response reanalysis show that the minimum DNBR^does not change due to an increase in tube plugging. Since the same minimum DNBR has the same allowable power peaking, the result is acceptable. The results of the peak primary pressure analysis show that the peak pressure increases from 2737.63 psig to 2742.68 psig. There remains 7.3 psi margin to the acceptance criterion of 2750 psia. Therefore, the increase in tube plugging is acceptable.
15.4.3d Single Rod Withdrawal: This event was reanalyzed to determine the impact of increasing the analyzed tube plugging level from 10% to 15%. The results are documented in Reference 12. The minimum DNBR result is the same for both 10% and 15% tube plugging. Since the same minimum DNBR will result in the same number of fuel pin failures, and since the previous failed fuel percentage resulted in acceptable offsite doses,15% tube plugging is acceptable.
15.6.3 Steam Generator Tube Rupture (Offsite Dose inputs): This event was reanalyzed to detemline the impact of increasing the tube plugging level from 10% to 15%
The results are documented in Reference 13. Reference 13 concluded that there was no significant incmase or decrease in the primary-to-secondary leakage available for offsite release and attributable to an increase in tube plugging.
%cicture an increase of up to 15% tube plugging is acceptable i
l
s l
ESAR Chapter 6 Mass andfnergy Releuq 6.2.1.1.3.1 Peak Reverse Differential Pressure Mass and Energy Release: A LOCA mass and energy release is used in the analysis of the peak reverse differential pressure across the containment divider deck. The key parameters for this analysis are all related to the modeling of the containment response. Any effect related to the slight difference in the mass and energy release caused by an increase in tube plugging would be insignificant. A discussion of the impact of tube plugging on LOCA mass and energy release is given in the next section.
6.2.1.3 Peak Containment Pressure Mass and Energy Release: The intent of these analyses is to maximize the energy released into containment nree effects will result from an incmase in tube plugging. For a LOCA, reverse heat transfer from the secondary to the primary occurs as the pdmary temperature drops to below the steam generator temperature. Clearly this revesse heat transfer will be maximized with a maximum heat transfer area (no plugged tubes). Also, with minimum tube plugging the secondary temperature is higher, which means more stored energy available for reverse heat transfer. He third effect is reduced pdmary mass due to less primary volume, which translates into a smaller mass and energy release. Since all of these effects will result in a greater mass and energy release, minimum tube plugging is conservative. Therefore, these is no adverse impact due to an increase in tube plugging.
6.2.1.4 Peak Containment Temperature Mass and Energy Release: A spectmm of
(
steam line breaks are analyzed to ensure that the limiting case is found. Due to the interraction of processes going on, there is no obvious worst case without running a spectrum of cases. Of greatest importance is the timing of tube bundle uncovery, which starts the phase of the release involving superheated steam.
This is dependent rgainly on the size of the break and the CA flow assumed. We primary coolant temperature during the uncovery period strongly influences the temperature of the steam exiting the break, and consequently the peak containment temperature. The effect of tube plugging on these analyses is insignificant. Although there would be some shift in the timing of events and the results for an individual case, the spectrum approach used in the analysis adequately covers the variation in the analyses due to a change in any individual parameter, such as tube plugging. He lower initial steam generator temperatures associated with increased tube plugging would result in less initial stored energy being available for release. The reduction in heat transfer area also would decrease the rate of heat transfer from the primary to the secondary.
These two items are the same arguments used for the steam line break core response analysis. It is concluded that the effect,if any on the steam line break mass and energy release would be that an increase in steam generator tube plugging would make the results less severe.
6.2.1.5 LOCA Mass and Energy Release Used in the Minimum Containment Pressure Analysis: "This mass and energy release calculation was explicitly reanalyzed for the increased tube plugging analysis. Refer to the following section for a discussion of the results of the reanalysis.
i
'{
LOCA Reanalyziji e
%c currem FSAR LOCA analyses and evaluations address both B&W and H fuel assembly
. designs. The FSAR analyses and evaluations are valid for 10% plugging, so it was necessary to determine the scope of reanalysis for demonstrating acceptable ECCS performance for at least 15% tube plugging. Including current peak clad temperature (PCF) penahics being carried, the following are the PCTs of reconi:
. E LDLOCA MNS PCT = 2139 + 18 + 3.4 -25 = 2135.4 F (Note: 3.4 from Reference 14) i (Note:-25 from Reference 15)
CNS-1 PCT = 1703.7 + 238 -25 + 2 = 1918.7 F (Note:-25 and 2 from Reference 15)
CNS-2 PCT = 1703.7 + 250 -25 +2= 1930.7 F (Note:-25 and 2 fmm Reference 15)
. ESBLOCA MNS PCT = 1488 + 102 + 72 = 1662 F (Note: 72 from Reference 14)
. B&W LBLOCA PCT = 1945 + 40 = 1985 F (Note 40 fmm Reference 16)
. B&W SBLOCA (Uses E analysis)
Based on discussions with both vendors, it was expected that an increase in tube plugging above 10% would result in a small penalty for LBLOCA, and that the SBLOCA penalty would be insignificant for the tube plugging percentages of interest.. In spite of the expected small k
penalties, reanalysis was required due to historical NRC expectations to explicitly reanalyze increases in tube plugging greater than 1% above the previous analysis. Rather than explicitly reanalyze both fuel types, it was decided to analyze one and justify the penalty as valid for both fuel types. Due to the W analysis PC r of record being much higher, E was contracted to reanalyze the LBLOCA and SBLOCA. The results of this project are given in Reference 17.
I which is included as Attachment 1.
Numerous assumptions in the E reanalyses were revised, along with the correction of evaluation model and input errors for which penalties were being carried. One evaluation model error correction related to the grid rewet model was recognized as having the potential to significantly
(
improve the PCT result. The current F-Q and k(z) limits were maintained. None of the assumptions in the analyses required Technical Specification changes, although several changes are justified by the reanalyses and can be pursued in the future.
e The H reanalyses resulted in a LBLOCA PCT of 1945* F for all units, with a assumed steam generator plugging level of 18%. The local Zr/II20 reaction rate is 4.55%, and the whole core Zr/ll2O reaction rate is <l%. The resulting SDLOCA PCT is 1264 F for 18% plugging. The impruvement in the SBLOCA PCT is due to higher assumed ECCS flow and NOTRUMP code revisions. The ECCS perfomiance acceptance criteria given in 10 CFR 50.46 were met for all of the analyses. It is noted that these PCT results are less than the previous analysis results for 10%
plugging. This resulted from the net effect of the numerous changes incorporated into the reanalyses. An increase in tube plugging by itself will result in a higher PCT. Based on previous analyses E provided a PCT penalty factor of 4 F per 1% tube plugging in Reference 17. Based
~.
(
on this factor, a siand-alone increase in tube plugging from 10% to 18% would result in a 32 F PCT penaity.
For the B&W fuel an explicit analysis was not perfomied. Rather, the above PCT penalty factor based on the W Evaluation Model will be conservatively applied to the B&W PCT by doubling the penalty factor. The PCT penalty applied to B&W fuel will then be 64 F. which results in a PCT of 2049 F. Ris approach is justified based on the conservative doubling of the E penalty factor, the similarity of the PCTs for the two evaluatian models, and on the remaining margin to the regulatory limit of 2200 F PCT. For the other ECCS performance acceptance criteria given in 10 CFR 50.46, it is concluded that there should similarly be little difference in the results due to the minor differences between fuel assembly designs. For example,in Reference 16, B&W states that an increase in PCT of 18 F results in a maximum increase in the local Zr/H2O reaction rate of 6% (+6% of the previous result). De previous result was 4.9% local and 0.55%
whole core. An increase of 64 Fin PCT would translate into an increase of(64 /18) x 6% =
22%. De 4.9% local Zr/H2O result would then be 6.0%.which leaves a large margin to the 17%
acceptance criterion of 10 CFR 50.46. A similarincrease in the whole core Zr/H2O reaction would also maintain a large margin to the 1% acceptance criterion.
Based on the results of these reanalyses and evaluations, it is concluded that the LOCA response for steam generator tube plugging levels up to 18% meets the 10 CFR 50.46 PCTlimit of 2200 F. The other criteria of 50.46 are also met. His conclusion is valid for both E and B&W fuel assembly designs, and with the current Technical Specification limits.
(
IV. RESULTS Based on the results of the evaluations and reanalyses presented in Section III,it can be concluded that the technical requirements for the FSAR Chapter 6.2.1 and 15 transients which are potentially impacted by an increase in steam generator tube plugging are met. De events which were impacted were reanalyzed, and the results met the acceptance criteria. The next step in the process in to address each of the questions comprising the 10 CFR 50.59 review. He responses to these questions will determine if an unreviewed safety question exists, and if a Tecimical Specification change is required. If the answer to each question is "NO", then prior NRC appmval of additional plugging up to 15% is not required with respect to the thermal-hydraulic issues. His calculation file does not address the design of the plugs or sleeves, the installation procedure. or the stress and compatibility aspects of the plugs and sleeves within the tubes.
The questions required for a 50.59 review per Reference 1, and the associated responses are as follows:
May the modification increase the probability of an accident evaluated in the SAR7 No. Steam generator tube plugging does not increase the probability of an accident evaluated in the SAR. Steam generator tube plugs are passive components which are not assumed to fail based on the design of the plugs meeting all acceptance criteria (Note that demonstrating that j
tube plugs meet all design criteria is beyond the scope of this review). Ecrefore, there are no i
(
accident initiators associated with an increase in the amount of tube plugging. Furthemiore, the j
reanalyses of FS AR events did not identify any thennal-hydraulic consequences ofincreased tube i
l i
t
{
plugging which increased the probability of any event. Consequently there is no impact on the probability of occurrence of any accident.
May the modification increase the consequences of an accident evaluated in the SAR?
No.1he results of evaluations and reanalyses to address the increase in tube plugging documented in or referenced by this calculation file demonstrate that all acceptance criteria for i
cach accident continue to be met. This was shown by confirming no increase in die source terms for offsite dose consequences (i.e. the DNBR results remained acceptable and there was no increase in fuel failures for all events). This was also shown by demonstrating that there was no breach of the pressure boundary, and therefore there was no increase in the release of radioactivity. For the steam generator tube rupture event it has been shown that there is no increase in the primary-to-secondary leakage and there are no other adverse impacts attributable to the increase in steam generator tube plugging. Therefore. there are no increases in offsite dose consequences.
May the modification create the possibility for an accident of a different type than any evaluated in the Sand i
No. The installation of additional steam generator tube plugs and sleeves does not create the possibility for an accident of a different type. Since there already exists steam generator tube plugs and sleeves in the steam generator, nothing new is introduced by additional plugging.
May the modincation increase the probability of a malfunction of equipment important to
(
safety evaluated in the SAR?
No. The equipment that is important to safety that is most impacted by a tube plug is the tube.
The review of that intemction is beyond the scope of this calculation file. The reanalyses of the affected FSAR events resulted in no component or system design limits being exceeded. In particular, the acceptance criteria related to primary design pressure were not exceeded.
Therefore, since no congsnent or system design limits were exceeded in the reanalysis, them is no increase in the probability of a malfunction of equipment important to safety.
May the modification create the possibility for a malfunction of a different type than any evaluated in the SAR?
No. Additional tube plugging does not intmduce any new types of malfunctions since tube plugs already exist. The macmscopic impact of an increase in the number of tube plugs on the steam generator in temis of stress and material considerations is beyond the scope of this review.
l Therefore, there is no potential for the thermal-hydraulic impact of additional tube plugging to result in a different type of malfunction.
May the modification increase the consequences of a malfunction of equipment important to safety evaluated in the SAR9 i
f No. Since tube plugging already exists, and since additional tube plugging neither introduces
[
any new malfunctions of equipment important to safety nor makes previously evaluated malfunctions more likely, there are no increased consequences.
f
61 f
f Will the modification reduce the margin of safety as defined in the basis for any technical specification?
No. 'lhe analyses and evaluations of FSAR Chapter 6.2.1 and 15 events have confirmed that the impact of additional tube plugging does not result in exceeding the acceptance criteria for each accident. The minimum DNBRs, the failed fuel percentages, the peak primary pressures, and the offsite doses all are met. Therefore, the margin of safety is not reduced as defined in the Technical Specifications. Consequently, no Technical Specification changes are required.
Based on the answers to the above questions all being "NO", it can be concluded that the increase in steam generator tube plugging fmm 10% to 15% for the McGuire and Catawba Model D steam generators does not constitute an unreviewed safety question, and that no Technical Specification changes are required. Therefore, prior NRC approval is not required pmvided that the tube plugging level does not exceed ISE This calculation file addresses the thermal-hydraulic impacts and is not applicable to the design of the tube plugs and sleeves, installation, or mechanical or material considerations.
V. ASSUMPTIONS / COMMENTS
- 1. The design of the plugs and sleeves, installation, and mechanical and material considerations are beyond the scope of this calculation file.
- 2. The reduction in NC flow that resnits from tube plugging is not addressed in this 50.59
(
review. A reduction in flow below the Technical Specification value requires a Technical Specification change, so the 50.59 process would not be used.
- 3. The impact of tube plugging on B&W fuel was not explicitly analyzed. A factor relating the peak clad temperature penalty to the percentage increase in tube bundle plugging was conservatively applied instead. A phone discussion with the NRC (Frank Orr) on 7/16/92 resulted in an understanding that it would be acceptable to explicitly reanalyze one fuel type and make suitable justification of the impact on the other fuel type.
- 4. This evaluation assumes symmetric tube plugging between the foursteam generators.
Although some amount of asymmetry is not significant, extreme asymmetry would require 2
additional evaluation.
I
- 5. The existing LOCA analyses and evaluations which are the basis of the current allowable 10%
tube plugging remain valid unless the plugging level exceeds 11%
i i
w
VI. REFERENCES
{
- 1. Nuclear System Directive 209,10 CFR 50.59 Evaluation of Nuclear Facility Modifications, Rev.1, Nuclear Policy Manual
- 3. DPC-NE-3001-PA, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology
- 4. Letter, S. S. Kilbom, Westinghouse, to D. W. Murdock, Duke Power, DAP-89-591, 8/2/89
- 5. DPC-1552.08-00-0(M1, Revision 2, FSAR Section 15.4.8 - Rod Ejection System Thermal-Hydraulic Analysis
- 6. WCAP-8330. Westinghouse Anticipated Transients Without Trip Analysis
- 7. DPC-1552.08-00-0081, Revision 1, FS AR Section 15.2.3 - Turbine Trip
- 9. DPC-1552.08-00-0065, Revision 2, FSAR Section 15.4.1 - Bank Withdrawal At Zero Power
- 10. DPC-1552.08-00-0059, Revision 2, FSAR Section 15.4.2 - Rod Withdrawal At Power
- 12. DPC-1552.08-00-0077. Revision 1, FSAR Section 14.4.3d - Single Rod Withdrawal
- 13. DPC-1552.08-00-0119. FSAR Section 15.6.3 - Steam Generator Tube Rupture Sensitivity To Tube Plugging
- 14. MCC-1553.06-00-0059, Safety Evaluation for the ReactorInternals Upflow Modification
- 15. Letter, J. M. Itall, Westinghouse, to W. M. Sample, Duke Power, DPC-93-201,1/25/93
- 16. Letter, L. L Losh (B&W) to IL J. Lee, Jr. (Duke), MC-92-128, 8/31/92
- 17. Duke Power Steam Generator Tube Plug;ug Level increase Loss of Coolant Accident Analysis / Evaluations, Westinghouse, March 1993. DPC-93-039, March 10,1993 4
4
.