ML20070V110
| ML20070V110 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 04/01/1991 |
| From: | Tuckman M DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9104100311 | |
| Download: ML20070V110 (286) | |
Text
II I flukt herr 'ompany V ?!O VM hurleur Productium (Jvpt . Hce l' resident PO lha HIni Nudeat Operations Cruirlette, A C lA2011007 fiO4),1?3.iMl -
DUKEPOWER April 1,1991 U. S. Nuclear Regulatory Commission A'ITN: Document Control Desk -
Washington, D. C. 20555
Subject:
Catawba Nuclear S.tation, Units 1 and 2 Docket Nos. 50-413 and 50-414
~199010 CFR 50.59 Report Pursuant to 10 CFR 50.59, find attached a summary of Nuclear. Station-Modifications, Exempt Change Variation Notices, procedures changes, tests, and experiments which were completed under the provisions of 10 CFR 50.59 from November 1,1989 to September !
30,1990.
Very truly yours,
-i h)
M. S. Tuckman i l
CRL27 !
Attachments C
xc: - Mr. S. D. Ebneter -
Regional Administrator, Region II U. S. Nuclear Commission 101 Marietta St., NW, Suite 2900 -
Atlanta, GA 30323 i Mr. W. T. Orders NRC Resident Inspector Catawba Nuclear Station -
Mr. R. E. Martin Office of Nuclear Reactor-Regulations =
U. S. Nuclear Regulatory Commission OWFN, Mail Stop 9H3
-Washington, DC 20555
(
9104100311 910401 -
PDR ADOCK 05000413.
R PDR ,
~f-t;
Catawba Nuclear Station Summary of Nuclear Station Modincations Completed from 11/1/89 to 9/31/90 CN-10295
Description:
This modification replaced the hiagnetrol Model 80-4123-002 level switches used for lWLLS5760 and lWLLS6860 with a single FCL FR72LLMPS Multipoint Sensor Assembly with three sensing points and remotely mounted electronics. The designation for the MPS is lWLLS5760, Safety Evaluation: The FCI sensor will enhance the operational flexibility of the incore instrumentation sump pump and will reduce personnel exposure for calibration activities, This modification does not involve any safety related components or circuitry, with the exception of disconnecting and reconnecting a plug on containment penetration IPENT0116. No unreviewed safety question is judged to be created by this modincation, CN 10309
Description:
This modification provided a piping route for draining each diesels lube oil from the filters and urainers for routine maintenance or as required.
Safety Evaluation: No new failure roocs etc postulated so the possibility of new accidents or malfunctions of equipment important to safety are not created. Since the diesel lube oil pressure boundary piping will be seismically designed and supported, the probability or consequences of a Lliunction of equipment important to safety will not be increased. This NSM will not increase the likelihood or consequences of any accident previously evaluated in the FSAR since it doesn't reduce the n;ltigation capability of the diesel generators, and the initial conditions of the core and reactor coolant system will not be affected by this NSM. Also, no margins of safety as defined in the bases to any Technical Specifications are affected since no plant parameters or setpoints are altered by this NSM. There are no unreviewed safety questions associated with this NSM.
CN-10575
Description:
This modification provided control of BB taak pressure when venting to the "D" heaters. It removed the flow restricting orifice in the BB tank vent line to the atmosphere. It recalibrated and changed setpoints on the BB tank pressure monitoring instrumentation and changed setpoints on the BB demineralizer influent temperature monitoring instrumentation. It connected
! IBBP5250 to the computer.
Safety Evaluation: This modification will increase the reliability and-availability of the BB System. No unreviewed safety questions are judged to be created or involved as a result of this modification.
! i CN 10675
Description:
This modification revised the control circuitry to fail open valves ICA48 and ICA52 by de-energizing their respective solenoid valves at their respective SSF disconnect enclosure. Valve ICA50A had its control circuit modified to op;m valve upon transfer of plant controls to SSF. Valve INV101 A had its control circuit modified to fait close the valve by de-energizing the solenoid valve at the respective SSF disconnect enclosure. All electrical controls were deleted from valve ITE33A. ;
Safety Evaluation: This modification does not change the normal mode of operation for the CA and NV valves. The changes affect the valve operation only after control has been transferred to the SSF. Valve ITE33A, which is always open, will have power removed from it and will remain open.
Therefore, this will not increase the probability of an accident or equipment malfunction previously evaluated in the FSAR. This modiGeation will help ensure that emergency feedwater is available to the steam generators and also RCS isolation is attained once control has been transferred to the SSF.
Therefore, this will not increase the consequences of a previously evaluated accident or equipment malfunction. These valves will still fall in the fail safe position. Therefore, the possibility of a different accident or equipment malfunction will not be created. This modification will not affect any key safety parameters or design limits. Therefore, the margin of safety as defined in'the bases of the Technical Specifications will not be reduced.
CN-10695
Description:
This modification installed a 5 micron filter on l A'and IB Diesel ,
Generator sump tanks to remove particulate when making additions of tube oil from the clean lube oil storage tank.
Safety Evaluation: The function of the system will not be affected and maintenance-wise will be improved with better filtration. The reliability of the system will be improved. Normal Diesel Generator operation _(emergency situations, tests, etc.) will be unaffected by this modification. Functionally, the system will be identical to the existing system with better filtration.
Accordingly, this modification wi!! have no affect on the probability, consequences, or " possibility of new" accidents evaluated in the FSAR. Nor will it affect the probability, consequences or possibility of malfunctions of equipment important to safety evaluated in the FSAR. The margin of safety defined in the bases of the Technical Specifications is unaffected.
CN-10717
Description:
This modification changed the class break on piping downstream of valve ISB029 to 900.4 class G and installed a blind flange at the end of this piping section.
Safety Evaluation: No unreviewed safety question is judged to be created or involved as a result of this modification.
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l CN-10816
Description:
This modification installed WP piping for Low Pressure Feedwater Relief discharge drainage. It also installed CM piping with hangers and deleted and reset remaining relief valves on the Low Pressure Feedwater Heaters.
Safety Evaluation: All components affected are non-QA and are located in the Turbine Building, a non-QA structure, Equipment important to safety is not degraded by this modification, No unreviewed safety question is,iudged to be created or involved as e result of this modification, CN-10938
Description:
This modincation added "P" traps in the control air handling unit (ICRA-AHU 1) WL drain lines. >
k Safety Evaluation: The probability of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR will not be increased as a result of this modification Since no new failure modes have been created, the possibility of an accident or malfunction of equipment important to safety which is different than any already evaluated in the FSAR will not be created, Because the ability of the AHU's to fulfill their function is not degraded, the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR will not be increased.
No degradation of safety limits has occurred as a result of this modincation and the reactor vessel core parameters are unaffected by this modifcation.
The reliability of the equipment associated with this NSM is not deg ilded.
Therefore, the margin of safety as defined in the bases to any Techmcal Specification is not affected. No Unreviewed Safety Questions are judged to be involved with this modification.
CN-10971
Description:
This modification disconnects power from valves INV44A, INV55A, INV66A, and INV77A during normal operation while retaining the position indication for each valve.
Safety Evaluation: This modification will enhance the stations capability to shutdown post fire. The probability or consequences of an accident previously evaluated in the FSAR will not be increased. The operation of these valves is not an accident initiator. The possibility for an accident or malfunction of a different type than any evaluated in the FSAR will not be created since these valves have the same function and Uperating characteristics as before. Since no safety parameters or design limits have been adversely affected, no margin of safety as defined in the Technical Specifications is reduced. There are no unreviewed safety questions associated with this modification.
CN-11000
Description:
This modification revised the VI flow diagrams to show the lines downstream of the header isolation valves.
Safety Evaluation: This modification involved tracing the lines downstream of a
l
! the header valves, sketching the lines downstream of the identified header i valves, assigning tag numbers to the newly identified VI valves, and hanging new valve tags. There was no work performed on any system in the plant.
Therefore, this modification did not create any Unreviewed Safety Question or require a change to the Technical Specincations.
CN-11086
Description:
The purpose of this modification was to install vent valves upstream and downstream of NM022A, Safety Evaluation: The major safety function of the QA 1 portion of the NM system is to provide containment isolation on appropriate safety signal. The system also serves to maintain the RCpB integrity. Neither of these functions is impaired by the addition of a vent valve. A seismic analysis has been performed to assure the seismic qualification of the modined sample line.
The failure of an instrument line connected to the reactor coolant system is postulated as part of the FSAR accident analysis. The addition of a QA-1 vent valve line to the instrument line does not make this accident more probable.
This modification does not adversely affect and plant safety functions, so the consequences of this or any other accident is not increased.
l All potential instrument line failures are clearly bounded by the rupture of a 3" l CVCS line. In addition, no new failure modes or operating characteristics are created by this modification. Therefore, no new accident scenarios are created.
Since no new failure modes or operating characteristics are introduced, the probability of previously evaluated malfunctions of equipment important to safety are not increased, and the possibility of new malfunctions is not created.
The margin of safety as defined in the Technical Specifications is not reduced.
Therefore, no Unreviewed Safety Questions are invohed in this modification.
CN-11088
Description:
This modification rerouted piping to prevent water hammer in the drain lines to the Reactor Coolant Drain Tank (NCDT).
Safety Evaluation: Since the valve steam leakoff piping was routed with the proper design conditions and class and the required stress analysis was preformed, the probability and consequences of malfunctions of equipment important to safety previously evaluated in the FSAR will not be increased.
The function of the system is unchanged. The modification enhances operation; therefore, the probability and consequences of an accident previously evaluated in the FSAR is not increased. Since the tank system will operate as before, better design will decrease potential water hammer problems, and no new failure modes have been identified, the possibility of an accident or malfunction of equipment important to safety different than any
evaluated in the FSAR is not created. No safety / design limits are adversely affected so margins of safety as denned in the bases to the Technical Specifications are not reduced. There are no Unreviewed Safety Questions associated with this NSM.
CN 11103
Description:
This modification changes the indication of reactor cold leg temperature from loop A to loop C in the SSF.
]
Safety Evaluation: The RTD 5 cold leg C is being replaced with an identical model. The configuration of the cold leg C connection to the SSF resulting from this modification is the same as the existing cold leg connection to the SSF. Thus, no new failure modes exist as a result of this modiGeation.
Therefore, no possibility of an accident or equipment malfunction'different from any already evaluated in the FSAR is created. The components involved s in this modification are the same as or identical replacements for the components in existence. Therefore, neither the probability nor the consequences of an equipment malfunction previously evaluated in the FSAR is increased as a result of the modification. No pipe support or penetration is modified as a result of this NSM so neither the probability nor the consequences of an accident previously evaluated in the FSAR is increased.
The implementation of this NSM results in no change to any current operating parameters, setpoints, or safety limits. Thus, no margin of safety as defined in the basis of any Technical Specification is decreased as a result of this modiGcation.
CN-ill22
Description:
This modification installed a siphon in the auxiliary feedwater's turbine driven pump.
Safety Evaluation: Since the affected systems function is essentially the same as before and no FSAR accident initiators are altered, the probability of an accident os mdfunction of equipment important to safety previously evaluated in the FSAR is not increased. No equipment is adversely affected. Therefore, the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR will not be increased and the possibility of an accident or malfunction of equipment important to safety different than any evaluated in the FSAR is not created. Since no plant parameters or setpoints are altered by this NSM, the margin of safety as defined in the bases to the Technical Specifications is not reduced. There are no unreviewed safety questions associated with this modification.
CN-lll23
Description:
This modification replaced the stuffingbox on RN pump 1 A with a new stuffing box made of stainless steel and equipped with a flush connection.
Safety Evaluation: Since the new stuffingbox will perform all the functions of the old stuffingbox, the additional taps will only be in use during maintenance i
periods, and since the operation of the RN pumps will not be adversely ,,
affected; the consequences of previously analyzed accidents or malfunctions of equipment important to safety are not ncreased.
Neither the RN pump nor any other piece of equipment is degraded by the installation of this modincation. Therefore, the probability of a previously analyzed accident or malfunction of equipment important to safety is not increased.
The possibility of an accident or malfunction of equipment important to safety is not created by this modification because no failure modes have been -
introduced to the RN pump or any other piece of equipment, i
No safety limit, setpoint, or operating parameter is changed as a result of this modification. Therefore, the margin of safety as defined in the basis _ to any Technical Specincation is not reduced. There are no Unanswered Safety Questions associated with this NSM.
CN-11140
Description:
This modification replaced the Kerotest T-type globe valves currently installed on the Unit 1 Moisture Separator Rehcater Drain system (HS) with Lowe plug valves.
Safety Evaluation: This modification does not affect the function of the HS system in any way. This modification is not a QA condition item and has no impact on safety functions during design basis events. Therefore, an unreviewed safety question evaluation for this modification is not required.
CN-lll45
Description:
This modification adds control for valves INV032B and INV039A to the auxiliary shutdown panel.
Safety Evaluation: Valves INV39A and INV32B are not initiators of any FSAR accidents; therefore, the probability of any accident previously evaluated in the FSAR will not be increased. Normal operation will not be affected and ASP operation will initially automatically align charging flow; therefore, the probability of a malfunction of equipment important to safety, previously evaluated in the FSAR, will not be increased. Neither valve is part of an Engineered Safety Feature or used to mitigate the consequences of any accident; therefore, the consequences of an accident or malfunction of equipment important to safety previously analyzed in the FSAR will not be 3 increased. There are no new failure modes; therefore, the probability of an accident or malfunction of equipment important to safety, not previously analyzed in the FSAR, will not be created. No margins of safety, as defined in the Technical Specifications, are reduced since no plant parameters or setpoints are altered by this NSM. There are no Unreviewed Safety Questions-associated with this NSM.
l
i l
CN ill50
Description:
This modification replaced the blind flange connections with new l
flush / cleanout connections on the piping that connects the Nuclear Service Water System (RN) to the Spent Fuel Pool System (KF).
Safety Evaluation: Since the function of the assured make up line to the spent fuel will not be adversely affected,' the probability or consequences of an l
accident or malfunction of equipment important to safety, which was _.
i previously evaluated in tla FSAR, will not be increased. The possibility of an
- accident or malfunction of equipment important to safety which is different than any already evaluated in the FSAR will not be created because no new l l
failure modes are created, The safety limits, as defined in the bases to the Technical Specifications, are not reduced. No Unreviewed Safety Questions are judged to be involved with this modification.
t CN Ill56
Description:
This modification installed LVDT's on the main feedwater -
control and bypass valves providing full range analog indication of valve position.
1 t
Safety Evaluation: _ Since these modifications do nm affect operating mechanisms nor the manner in which they operate, there is no increase in the probability of accidents evaluated in the FSAR nor any increase in their consequences, l
These modifications do not adversely affect any safety _ equipment, nor any other equipment, so thee will not be any increase in the probability of a malfunction to safety equipment as evaluated in the FSAR nor:will the possibility of malfunctions not stated in the FSAR be created. Since these modifications do not adversely affect safety equipment, there is no increase in the consequences on any' malfunction of this equipment as evaluated in the
, FSAR. 'Since these modifications do not affect the feedwater control valve nor-the feedwater control bypass valve's operating mechanism nor operating modes, there is no increase in the possibility of an accident which is different-than any already evaluated. _Since no plant parameters or setpoints are altered -
. by this NSM, the margin of safety as defined in the bases to the Technical Specifications is not reduced. . There are no Unreviewed Safety. Questions associated with this modification.
CN-ill86
Description:
This modification replaced valves CA38,42,46,50,54,58,62, and 66 with a suitable replacement.
l Safety Evaluation: ' Since the purpose of these valves and their method of '
I 1 implementation remains unchanged, there is no increase in the probability of an accident as , valuated in the FSAR. Since the valves purpose, CA' System functions and operating. modes remai_n unchanged, there is no increase in the i
consequences of an accident evaluated .in the FSAR.-- Since no current purpose of these valves is being added to, deleted from or altered,_ there is no creation
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of the possibility for an accident of a different type than any evaluated in the FSAR. Since the new valves operate in the same manner as and are of the same class and construction material as the old _ valves, there is no increase in the probability nor any increase in the consequences of a malfunction of equipment important to safety as evaluated in the PSAR. Since no purpose of these valves is added, deleted or altered, nor is the method of implementation changed, there is no creation of the possibility of a mra mction of a different type than any evaluated la the FSAR. Since no plant parameters or setpoints are altered by this NSM, the margin of safety as described in the bases to any Technical Specification is not reduced. No Unreviewed Safety Question is judged to be created by this modification.
CN lll87
Description:
This modification replaces valves SA002 and SA00 with valves incorporating new design features and better tight shutoff capabilities.
Safety
Description:
Since no operating modes have been changed, added or deleted and the replacement of these valves does not adversely affect equipment important to safety, there is no increase in the probability of a malfunction of equipment important to safety as evaluated in the FSARc Since the functional requirements of the new valves have not changed (stroke time, flow capacity and shutoff capabilities), there is no increase in the consequences of a malfunction of equipment important to safety as evaluated in the FSAR.
Since the operating modes of these valves remain unchanged and no new failure modes were added, there is no increase in the probability of an accident as evaluated in the FSAR. Since the functional requirements of the new valves have not changed, there is no increase in the consequences of an accident as evaluated in the FSAR. Since no new operating modes _were added, and the new valves have similar characteristics and function as before, this modification does not create the possibility for an accident of a different type than any evaluated in the FSAR.
Since no plant parameters or setpoints are altered by this NSM, the margin of safety as described in the bases to any Technical Specification is not reduced.
There are no Unreviewed Safety Questions associated with this modi 0 cation.
Ld-ll217 Desciiption: This modification removed the Train A return air ring header -
from the Annulus Ventilation System (VE).
Safety Evaluation: The VE system is a standby system and has no function during normal plant operations. No new failure modes were created by this modification. This modification does not result in any change to the interfaces of the VE system with other plant systems, Thus, this modification does not increase the probability of an accident evaluated in the FS AR nor does this modification create the possibility of an accident of a different type than any evaluated in the FSAR. Based on the safety review, this modification does not l
I increase the consequences of an accident evaluated in the FSAR. No new l equipment is added, no existing equipment is degraded. The environmental and seismic qualincation of the VE system remain unchanged. No new failure l modes were identified in this evaluation. No control equipment is modified. l Thus, the modification does not increase either the consequences or the l probability of a malfunction of equipment important to safety as evaluated in l the FSAR. This modification does not create the possibility of a malfunction )
of a different type than any evaluated in the PSAR. No setpoints, design limits, or Operating characteristics are changed as a result of this modification.
Therefore, this modification does not reduce the margin of safety as defined in the bases of the Technical Specifications. No Unreviewed Safety Questions are associated with this modification.
CN ll218
Description:
This modification installed a gate valve on the NV Let-down Heat Exchanger to stop leaking from other valves.
Safety Evaluation: Since the valve being added will be in the same operating position as the upstream valves and it performs a redundant isolation function, there is no increase in the probability of an accident evaluated in the FSAR.
Since the new valve will be operated in the same manner and for the ssme function as existing system valves, there is no increase in the consequemees of an accident evaluated in the FSAR.
Since the new valve is not used for normal system operation and perfocms a-redundant function to existing system valves, there is no creation of the possibility for an accident of a different type than any evaluated in the FSAR.
Since the new valve will not adversely affect equipment important to safety, there is no increase in the probability, nor the consequences of a malfunction of equipment important to safety evaluated in the FSAR. Since the new valve is not used for normal system operation, and its addition does not change the l system design basis nor provide a function that is not already performed by the-l system, there is no creation of the possibility for a malfunction of a different l
type than any evaluated in the FSAR.
Since no setpoints, safety limits, or design parameters are affected by this modification, the margin of safety as defined in the basis to any Technical Specification will not be reduced. No Unreviewed Safety Question are associated with this modification.
CN-20049
Description:
The purpose of this modification was to provide a piping route for draining each diesel's tube oil from the filters and strainers for routine maintenance or as required.
Safety Evaluation: No new failure modes are postulated so the possibility of
new accidents or malfunctions of equipment important to safety are not created. Since the diesel lube oil pressure boundary piping will be seismically designed and supported, the probability or conseqt: aces of a malfunction of equipment important to safety will not be increased. This NSM will not increase the likelihood or consequences of any accident previously evaluated in the FSAR, since it does not reduce the mitigative capability of the diesel generators, and the initial conditions of the core and reactor coolant system will not be affected by this NSM. Also, no margins of safety as denned in the bases of the Technical Specifications are affected since no plant parameters or setpoints are altered by this NSM. There are no Unreviewed Safety Questions associated with this NSM.
CN 20220
Description:
This modification installed a wet layup loop for the shell side of the Containment Spray Heat Exchanger.
Safety Evaluation: Nuclear Service Water (RN) system which flows through the Containment Spray Heat Exchangers has caused fouling problems on these heat exchangers on Unit 1 Catawba which has led to reduced heat transfer capability. No Unreviewed Safety Questions are judged to be created by this modification.
CN-20325
Description:
This modification added "P" traps in the control air handling unit (2CRA-AHU-1) WL drain lines.
Safety Evaluation: The probability of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR will not be increased as a result of this modification. Sir.cc no new failure modes have been created, the possibility of an accident or malfunction of equipment important to safety which is different than any already evaluated in the FSAR will not be created. Because the ability of the AHU's to fulfill their function is not degraded, the consequences of an accident or malfunction of equipment-important to safety previously evaluated in the FSAR will not be increased.
No degradation of safety limits has occurred as a result of this modification and the reactor vessel core parameters are unaffected by this modification.
The reliability of the equipment associated with this NSM is not degraded.
Therefore, the margin of safety as defined in the bases to any Technical Specificatic is not affected. No Unreviewed Safety Questions are judged to be involved with this modification.
CN-20400
Description:
This modification provides a 1 inch class G carbon steel drain line downstream of relief valves 2BB028 and 2BB161.
Safety Evaluation: All reference piping is non-safety related and is located in the non-QA Turbine Building. No Unreviewed Safety Question is judged to be created by this modification.
CN 20451
Description:
This modification placed a manual bypass on the P-14 safety signal ( S/G Hi Hi level) to prevent feedwater isolation during modes 4, 5, and 6.
Safety Evaluation: Since this modification does not affect the operating modes of the equipment affected nor does it introduce any new type of equipment, there is no increase in the probability of an accident as evaluated in the FSAR.
Since no operating modes, functions, or methods of accomplishing these functions for equipment affected by this modification are adversely affected, there is no increase in the consequences of an accident as evaluated in the FSAR, Sincc this modification does not add any new failure mechanism
- change any functions, or otherwise adversely affect equipment importa,a to safety, there is no increase in the probability nor any increase in the consequences of a malfunction of equipment important to safety as evaluated in the FSAR, Since no existing functions were altered nor added and the methods of accomplishing these functions were not changed, there is no creation of the possibility for an accident of a different type than any evaluated in the FSAR, It is for this same reason that there is no creation of the possibility for a malfunction of equipment important to safety of a different type than any evaluated in the FSAR, Since no plant parameters or setpoints are altered by this NSM, the margin of safety as defined in the bases to r.ny Technical Specification is not reduced. No Unreviewed Safety Question is -
judged to be associated with this modification,
- CN 20471
Description:
This modification rerouted piping to prevent water hammer in the drain lines to the Reactor Coolant Drain Tank (NCDT),
Safety Evaluation: Since the valve steam leakoff piping was routed with the proper design conditions and class and the required stress analysis was preformed, the probability and consequences of malfunctions of equipment important to safety previously evaluated in the FSAR will not be increased, The function of the system is unchanged. The modification enhances operation; therefore, the probability and consequences of an accident previously evaluated in the FSAR is not increased. Since the tank' system will operate as before, better design will decrease potential water hammer problems, and no new failure modes have been identified, the possibility of an accident or malfunction of equipment important to safety different than any evaluated in the FSAR is not created. No safety / design limits are adversely affected so margins of safety as defined in the bases to the Technical Specifications are not reduced. There are no Unreviewed Safety Questions associated with this NSM, CN-20485
Description:
This modification changes the indication of reactor cold leg temperature from loop A to loop C in the SSF, Safety Evaluation: The RTD in cold leg C is being replace'l with an identical
model. The configuration of the cold leg C connection to the SSF resulting from this modification is the same as the existing cold leg connection to the SSF Thus, no new failure modes exist as a result of this modification.
Therefore, no possibility of an accident or equipment malfunction different from any already evaluated in the FSAR is created. The components involved ,
in this modification are the same as or identical replacements for the components in existence. Therefore, neither the probability nor the ,
consequences of an equipment malfunction previously evaluated in the FSAR is increased as a result of the modification. No pipe support or penetrat' ore modified as a result of this NSM so neither the probability nor the consequences of an accident previously evaluated in the FSAR is increased.
The implementation of this NSM results in no change to any current operating parameters, setpoints, or safety limits. Thus, no margin of safety as defined
- in the basis of any Technical Specification is decreased as a result of this modification.
CN 20504
Description:
This modification replaced the stuffingbox on RN pump 2A with a new stuffing box made of stainless steel and equipped with a flush connection.
Safety Evaluation: Since the new stuffingbox will perform all the functions of the old stuffingbox, the additional taps will only be in use during maintenance periods, and since the operation of the RN pumps will not be adversely affected; the consequences of previously analyzed accidents or of malfunctions of equipment important to safety is not increased.
Neither the RN pump nor any other piece of equipment is degraded by the installation of this modification. Therefore, the probability of a previously analyzed accident or malfunction of equipment important to safety is not increased, i The possibility of an accident or malfunction of equipment important to safety is not created by this modification because no failure modes have been introduced to the RN pump or any other piece of equipment.
No safety limit, setpoint, or operating parameter is changed as a result of this modification. Therefore, the margin of safety as defined in the basis to any Technical Specification is not reduced. There are no Unanswered Safety Questions associated with this NSM.
CN-20529
Description:
This modification replaced the blind flange connections with new flush / cleanout connections on the piping that connects the Nuclear Service Water System (RN) to the Spent Fuel Pool System (KF).
Safety Evaluation: ace the function of the assured make-up liae to the spent l
l
fuel will not be adversely affected, the probabih!y or consequences of an ,
accident or malfunction of equipment important to safety, which was previously evaluated in the FSAR, will not be increased. The possibility of an accident or malfunction of equipment important to safety which is different than any already evaluated in the FSAR will not be created because no new failure modes are created. The safety limits, as defined in the bases to the Technical Specifications, are not reduced. No Unreviewed Safety Questions are judged to be irvolved with this modification.
CN-20567
Description:
This modification replaced valves CA38,42,46,50,54,58,62, and 66 with a suitable replacement.
Safety Evaluation: Since the purpose of these valves and their method of implementation remains unchanged, there is no increase in the probability of
, an accident as evaluated in the FSAR. Since the valves purpose, CA System functions and operating modes remain unchanged, there is no increase in the consequences of an accident evaluated in the FSAR. Since no current purpose of these valves is being added to, deleted from or altered, there is no creation of the possibility for an accident of a different type than any evaluated in the FSAR. Since the new valves operate in the same manner as and are of the
, same class and construction material as the old valves, there is no increase in the probability nor any increase in the consequences of a malfunction of equipment important to safety as evaluated in the FSAR. Since no purpose of these valves is added, deleted or altered, nor is the method of implementation j changed, there is no creation of the possibility of a malfunction of a different type than any evaluated in the FSAR. Since no plant parameters or setpoints are altered by this NSM, the margin of safety as described in the bases to any Technical Specification is not reduced. No Unteviewed Safety Question is judged to be created by this modification.
- CN-20568
Description:
This modification replaces valves SA002 and SA005 with valves incorporating new design features and better tight shutoff capabilities.
Safety
Description:
Since no operating modes have been changed, added or deleted and the replacement of these valves does not adversely affect equipment important to safety, there is no increase in the probability of a malfunction of equipment important to safety as evaluated in the FSAR. Since the functional requirements of the new valves have not changed (stroke time, flow capacity and shutoff capabilities), there is no increase in the consequences of a malfunction of equipment important to safety as evaluated in the FSAR.
Since the operating modes of these valves remain unchanged and no new -
failure modes were added, there is no increase in the probability of an accident as evaluated in the FSAR, Since the functional requirements of the new valves have not changed, there is no increase in the consequences of an accident as evaluated in the FSAR. Since no new operating modes were added, and the
+ r .
l i new valves have similar characteristics and . function as before, this modification does not create the possibility for an accident of a different type than any evaluated in the PSAR.
Since no plant parameters or setpoints are altered by this NSM, the margin of safety as described in the bases to any Technical Specincation is not reduced.
There are no Unreviewed Safety Questions associated with this modification. !
CN 20594
Description:
This modification removed the Train A return air ring header from the Annulus Ventilation System (VE).
Safety Evaluation: The VE system is a standby system and has no function during normal plant operations. No new failure modes were created by this modification. This modification does not result in any change to the interfaces of the VE system with other plant systems. Thus, this modification does not increase the probability of an accident evaluated in the FSAR no 'xs this modification create the possibility of r.a accident of a Gifferent type than any evaluated in the FSAR. Based on the safety review, this modification does not increase the consequences of an accident evaluated in the FSAR. No new equipment is added, no existing equipment is degraded. The environmental and seismic qualification of the VE system remain unchanged. No new failure modes were identiGed in this evaluation. Ne control equipment is modined.
Thus, the modification does not increase either the consequences or the '
probability of a malfunction of equipment important to safety as evaluated in the FSAR. This modification does not create the possibility of a malfunction of a different type than any evaluated in the FSAR. No setpoim design ',
limits, or operating characteristics are changed as a result of this modincation.
Therefore, this modification does not reduce the margin of safety as defined in 3 the bases of the Technical Specifications. 'No Unreviewed Safety Questions are associated with this modification. .
CN-20599
Description:
This modification modified the internals for valves 2CA174,
)
2CA175, and 2KC081B. The intent of this modification was to reduce seat leakage.
Safety Evaluation: The probability and consequences of any malfunction of equipment important to safety or any accident previously evaluated in the FSAR will not be increased, since the valves being modified will continue to function as they did prior to the modification and rio other equipment important to safety is adversely affected by this modification.
Based upon the fact that the components meet the necessary QA Condition and ervironmental requirements and since :hese components will not interact with other components different than they did prior to the modification, this modification will not create the possibility of a new accident or new malfunction of equipment important to safety.
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1 No setooints, operating parameters, or safety limits are affected by this modificatio.., therefore the margin of safety as defined in the Technical Specifications will not be reduced. There are no Unreviewed Safety Questions 6
associated with this modification. ,
l l CN-20606
Description:
This modification installed by3 sss lines to allow RN water to be flushed froni the upstream sides of valve RN250A and (N310B located at the ;
inlets to the Auxiliary Shutdown Panel Area dir Conditic ang Units (ASPSU) to the CA line between valves cal?4 and CAL 7%
Safety Evaluation: No syvem h adversely affected by this modification and does not impair the ability of any system to perform its safety function.
Therefore, the probability and consequences of accidents evaluated in the FSAR are not increased. Since neither the ASP, the RN backflush line to CA, nor the EMF are accident init:ators, the possibility for an accident of a different type than any evaluated in the FSAR is not created.
Since all safety systems are still able to perform thcir functions, there is no increase in the probability or consequences of a ralfunction of equipment important to safety evaluated in the FSAR. No n:w safety related equipment was introduced by this NSM and no other safety related equipment was affected. Furthermore, no new failure modes of existing safety related RN equipment were identified to be associated with this modificatio14. Therefore, there is no crc:aon of the possibility for a malfunction of a different type than l any evaluated in the FSAR.
l Since no setpoints, safety limits, or design parameters were changed, the margin of safety as defined in the bases to any Technical Specification will not be reduced. No Unreviewd Safety Questions are judged to be associated with this modification.
CN-20608
Description:
This modification provided an access hole in Steam Generator i 2A to be used for foreign object search and removal.
Safety Evaluation: The previously analyzed steam line break accident bounds any potential failure of the 2" S/G shell gasket closure; therefore, 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, the probability or consequences of any malfunction of equipment important to safety or accident previously evaluated will not be increased. No safety limit, setpoint or operating parameters will be changed by this modification; therefore, the margin of safety as derined in the basis of tlic Technical Specification will not be reduced. There are no Unreviewed Safety Questions associated with this modification.
_. . _ . _~ _ ._ _ .
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1 CN 20618
Description:
This modification installed a 3/4" line upsticam of NI95A which l connects just downstream of this same valve. Check valve NI471 was installed in the line.
Safety Evaluation: The ability of NI95A and NI96_A to function as designed is not adversely affected by this modification. Since the components affected or added are not accident initiators, the probability of any accident previously evaluated will not be increased. The probability or consequences of any malfunction of equipment important to safety will not be increased, since the modi 0 cation has no adverse affect on the containment isolation valves, the class B section of pipe, or any other piece of equipment. No new failure modes have been identined, therefore the possibility of any new accidents or new malfunctions of equipment important to safety have not been created. The consequences of any accident previously evaluated will not be increased, since no ace'ident mitigating systems have been adversely affected. No setpoints, operating parameters or saf; y limits are adversely affected by this modi 0 cation, therefore the margin of safety as defined in the Technical Specifications will not be reduced. No Unreviewed Safety Questions are associated with this modification.
CN-50124
Description:
This modification installed two instrument air line connections; one in the meter workshop and one in the relay workshop in the Service Building at elevation 609',
Safety Evaluation: Addition of instrumentation air line connections will provide an appropriate means for testing and maintenance of relaying and metering. No safety system will be degraded. No Unreviewed Safety Question is judged to be created or involved as a result of this modification.
CN 50133
Description:
This modification provides backflushing for IEMF52,2 EMF 52, and OEMF47.
Safety Evaluation: No plant parameter will be adversely affected by this modincation. The YM system will be isolated from each EMF by an isolation valve and a check valve to prevent contamination of the YM. Since this modification has no adverso interaction on safety and does not create a new release path, there is no increase in the probability of accidents or malfunctions of equipment impcrtant to safety previously evaluated in the FSAR. Since the performance of all components affected under t;.is NSM .is not degraded, there is no increase in the consequences of accidents in
- malfunction of equipment important to safety already in the FSAR. For the above reasons, no possibility of accident nor malfunctions of equipment important to safety different from any already evaluated in the FSAR is created. No plant parameters or setpoints are altered by this NSM so no L margin of safety as defined in the basis to any Technical Specification is l
l
reduced. There is no Unanswered Safety Questions associated with this NSM.
CN 50204
Description:
This modi 6 cation installed a permanent tank in the Aux Bld. for I dumping ice solution from the ice machines.
Safety Evaluation: No Unreviewed Safety Questions are judged to be c.ssoc!ated with this mod 10 cation.
CN 50325
Description:
'Ihis modification installed permanent connections from VS, WS (YM source), '.nd WL to the relief valve test bench in the Hot Machine Shop.
Safety Evaluation: Use of this VS line for the test bench is not significantly different from the line being used for temporary equipment connections.
. Therefore, operation of the VS system will not be affecal by this modification. The mr.rgin of r.fdy as dcEned in the bases of the Technical Specifications will not be reduce occause no safety related portions of the YM, WS, WL, or VS systems are aff-cted. No Unreviewed Safety Questior, is judged to be associated with this modification.
CN 50383
Description:
This modification installed new level instrumentation on the YM (Deniineralizer Water) System Acid Day Tank in the Water Treatment Room.
Safety Evaluation: It has been determined that this modi 0 cation to provide safe level indication for the Aeld Day Tank will not create an Unreviewed Safety Function.
CN 50394
Description:
This modification downgc hd four WZ N .timy Building Sump Fumps to QA Condition 4.
1 Safety Evaluation: The subject pumps do not initiate or mitigate any accidents presently evaluated in the PSAR. Therefore, the probability or consequences of these accidents are not increased. Since two pumps remain safety rciated and the system can perform its function with a single failure and one pump l operating, the possibility for an accident of a different type than any evaluated in the FSAR is not ercated. Likewise, there is no increase in the probability or consequences of a malfunction of equipment important to safety evaluated ir the FSAR because any one pump van perform the system function and there are two safety related pumps and fout other QA 4 pumps. Since the single pump can handle maximur. : vater inflows and there is no adverse impact on other equipment, this modincation does not create the possibility for a malfunction of equipment of a different type than any evaluated in the FSAR, Since no safety parameters or design limits have been adversely affected, no margin of safety as defined in the bases to any Technical Specification is reduced. There are no Unteviewed Safety Questions associated with this modification.
i CN 50411
Description:
This modification deletes the RN pump lube injection crossover
- piping containing valve IRN26.
l Safety Evaluation: No Unreviewed Safety Question is judged to be associated with this modification.
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Catawba Nuclear Station Summary of Exempt Nariation Notices Completed from 11/1/89 to 9/31/90 CE 1723
Description:
This exempt change removed valve IRLib8 and removed the pusiibutton from IMCl3.
Safety Evaluation: Deleting this valve har no affect on the Technical Specifications. No Unrevicwed Safety Questions are judged to be created by this modification.
CE-1786
Description:
This exempt change removed the isolation and root valve piping configuration for test points ISMPX5860,5870, and 5880 and installed a pipe plug.
Safety Evaluation: Neither the function or operability on the SM system will be affected by removing the piping and valves and instalhng a pipe plug.
Installing the plug does not affect any analysis in the FSAR. This variation notice does not increase the probability nor consequences of an accident previously evaluated in the FSAR and it does not create the possibility of an accident not previously evaluated in the FSAR.
This portion of the SM system is not safety related and installing the pipe plug does not affect the function or operability of any safety related equipment.
Therefore, thia variation notice does not increase the probability nor consequences of safety related equipment malfunctions previously evaluated in the FSAR and it does not create the possibility of safety related equipment malfunctions not previously evaluated in the FSAR.
No setpoints, operating parameters or safety limits are affected. Therefore, the margin of safety as defined in the basis to any Technical Specification is not reduced. No Unreviewed Safety Question is judged to be created by this change.
CE 1797
Description:
This change revised flow dir. grams to remove locked designa%ns for valves INI239,281,240,444,445,449,450 / 2N1239,240, 281,444,445,449, and 450.
Safety Evaluation: Due to the UHI syst:m being removed from service, it is l no longer necessary to lock these valves. The UHI system has been isolated l from other plant systems and no longer performs any safety related function.
Therefore, the probability, possibility, or consequences of an accident or ;
l equipment malfunction will not be increased by this Variation Notice. No Technlol Specification will be affected by this change No Unreviewed i Safety Qacstion is involved with this Variation Notice since the UHI system is no longer operable.
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CE 2045
Description:
This exempt change installed a Five Star Series 86 double !
mechanical seal in "C" Heater Drain Tank Pumps ICl and IC2 alo.g with seal injection piping, it also installed vents between Pump ICI and valve 1HW294 and Pump 1C2 and valve 111W295. j Safety Evaluation: Installing the new mechanical seal and the high point vent will result in more reliabic operation of pumps ICI and IC2. The pumps will continue to function as evaluated in the FSAR. This variation notice does not increase the probability nor consequences of an accident previously evaluated in the FSAR and it does not create the possibility of an accident not previously evaluated in the FSAR.
Installing the new mechanical seal and the high point vent will not affect the function or operability of any safety related equipment. Therefore, this variation notice does not increase the probability nor consequences of safety-related equipment malfunction previously evaluated in the FSAR and it does not create the possibility of safety related equipment previously evaluated in the FSAR. For the same reasons, this variation notice will r.ot reduce the margin of safety as denned in any Technical Specincation basis. No Unreviewed Safety Question is judged to be created by this variation notice.
CE-2073
Description:
This exempt change replaced the carbon steel piping downstream of valve IShil38 and around Steam Trap T-06 with stainless steel piping, it also replaces the remaining schedule 40 carbon steel piping with schedule 80 carbon steel piping.
Safety Evaluation: The function or operability of the Shi system will not be affected by the piping replacement. The piping and Ottings installed will i
comply with the appropriate piping specifications for this pmtion of the Shi system. The ttainless steel and schedule 80 carbon steel pipmg will be less susceptible to erosion and corrosion and provide a more reliable piping system. This variation notice does not increase the probability nor consequences of an accident previously evaluated in the FSAR and it does not create the possibility of an accident not previously evaluated in the FSAR.
The piping replacement does not affect the tuntt!;.a or operability of any safety related equipment. The affected Shi system piping wi!! not be degraded in any way. Therefore, this variation notice does not increase the pwbability nor consequences of safety related equipment malfunctions previously evaluated in the FSAR and it does not create the possibility of safety reNed equipment malfunctions not previously evaluated m the FSAR.
No setpoints, operating parameters or safety limits are affected, so this variation notice will not reduce the margin of safety as denned in any Technical Specincation basis. No Unreviewed Safety Questions are judged to be created by this variation notice, l
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I CE 2121
Description:
This variation notice installed new 2" Olter racks on 1 VA-AH-1 0001 and 1 VA AH 0002.
Safety Evaluation: This modification will not require any changes to the Technical Specincations for the Auxiliary Ventilation System (VA). Because ,
the units will be out of service during the change and there is no change being made to the VA systems function, the chances of an accident will not he increased. Neither the probability ner the consequences of an accident previously evaluated in the FSAR will be increased since the VA system will continue to function as evaluated in the PSAR. No new accident will be created since the units will be taken out of service during implementation of this change. Also, the probability or consequences of malfunction of safety related equipment will not be increased because the function of the VA system will be unaffected by this modl6 cation. For the same reason, no new possibility of malfunction of safety related equipment is created and the modincation of these niter racks will not affect any margin of safety defined in the basis of any Technical Specification. No Unreviewed Safety Question is judged to be created by this variation notice.
CE-2156 Descriptio < This change replaced the expansion joint on RN Train I A with a spool piece. .
Safety Evaluation: The function and operation of the RN system will not be changed in any way by this modification. The RN system will still provide essential cooling during accident sequences, so the consequences of an accident or malfunction of equipment importar.' to safety previously evaluated in the
- FSAR will not be increased. - Since RN is a mitigating system and not an accident initiator, this mod 10 cation does not increase the probability of any accidents as previously evaluated in the FSAR. The spool piece meets ASME ;
Code requirements and De+!gned has verined seismic integrity.will bc l maintained. Therefore, the possibility of an accident or malfunction of -
eTiipment im;stant to safety different than any already evaluated in the FSAR : vill not be created. For the same reasons, this modification will not increase the probability of a malfunction of equipment important to safety-previously evaluated in the FSAR. Since the function of the RN system will not be impacted, this modification will not reduce the margin of safety as denned in the bases to any Technical SpeclGcation. Based on the discussion-presented, no Unreviewed Safety Questions are judged to be created or-involved in this modification.-
CE 2157
Description:
This char.ge replaced the expansion joint on RN TrainL1B with a spool piece.
Safety Eva'uation: The function and operation of the RN system will not be changed in any way by this modification. The RN system will still provide l.
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5 4
i essential cooling during accident sequences, so the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR will not be increased. Since RN is a mitigating system and not an .
accident initiator, this modl0 cation does not increase the probability of any '
, accidents as previously evaluated in the FSAR. The spool piece meets ASME
, Code requirements and Designed has verified seismic integrity will be maintained. Therefore, the possibility of an accident or malfunction of equipment important to safety different than any already evaluated in the -
FSAR will not be created. For the same reasons, this modincation will not increase the probability of a malfunction of equipment important to safety previously evaluated in the FSAR. Since the function of the RN system will
- not be impacted, this modification will not reduce the margin of safety as-defined in the bases to any Technical Specification. Based on the discussion presented, no Unreviewed Safety Questions are judged to be created or .
involved in this modification.
CE 2182
Description:
This modification removed the
- locked closed requirement" from
! valves 1 and 2 NV240.
Safety Evaluation: Valves 1 and 2 NV240 will continue to operate in the same manner as before, i.e., open and close according to established procedures.
For this reason, the probability or consequences of an accident previously evaluated in the FSAR will not be increased. Since the basic connguration and operating procedures are unchanged concerning these valves, the probability of a malfunction of equipment important to safety will not he -
increased. Since a malfunction of equipment will not be increased, there will not be an increase in the consequences of an equipment malfunction. The margin of safety as denned in the %ses to the Technical Specifications will not be reduced since the valves will continue to operate as before. No Unreviewed Safety Question is judged to be created by this change.
CE-2194
Description:
This variation notice extends the Condensate CE-2195 Booster Pump trip vme delay for low suction now from 5 seconds to 20 seconds, Safety Evaluation: The worst case accident that could result from these time delays being extended would be a loss of the pump. However, these pumps L are ".ot s%fety related and are not required for the safe shutdown of the reactor.
The casequences or probability of an accident or malfunction of equipment e.vehated in the FSAR will neither be increased nor created. The margin of mfety as defined in the basis to any Technical Specincation is not reduced. -
This modification does not create an Unreviewed Safety Question.
CE-2196
Description:
This variation notice extends the Mainfeedwater Pump trip time -
l delay for low suction flow and pressure from 5 seconds to 20 seconds.
Safety Evaluation: . The worst case accident that could result from these time -
, .a-.-,._-_.- -, - -- -- - . .-.----
delays being extended would be a loss of the pump. However, these pumps are not safety related and are not required for the safe shutdown of the reactor.
The consequences or probability of an accident or malfunct!on of equipment evaluated in the FSAR svill neither be increased nor created. The margin of safety as defined in the basis to any Technical Specification is not reduced.
This mc=dification does not create an Unreviewed Safety Question. ,
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9 CE-2479
Description:
This variation notice replaced valve 2SV066 with a l" Anchor-Darling double disc gate valve.
Safety Evaluation: Replacing valve 2SV966 will not affect the function of the SV system as evaluated in the FSAR. The draining function of the valve will not be affected. The design of the new valve meets the temperature and pressure conditions experienced in this portion of the SV system. The new valve will perform the same function as the existing one and also will be less likely to develop leakage past the seat. The additional weight of the valve does not create any scismic concerns or require any additional support. No Unreviewed Safety Question is judged to be created by this modification.
CE 2520
Description:
This modincation changed the VG Aftercooler piping drawings to reflect FS 300.4 Class O piping from the existing valves to the Goat traps rather than Class F piping.
Safety Evaluation: No Urreviewed Safety Question is judged to be created by this exempt change variation notice.
CE 2691
Description:
The purpose of this (xempt change is to update the Yard Drainage Catch Basin Schedule to indicate the present installed locations of the Type 1 Catch Basin Modification Cages.
Safety Evaluation: Since the system is now fully operable, it is capable of functioning at the capacity denned in the acceptance criteria and neither the probability nor consequences of an accident evaluated in the FSAR are increased. Since this system will operate in a manner that does not adversely affect safety related equipment, neither the probability nor consequences of a malfunction of equipment important to safety and evaluated in the FSAR are increased. The failure modes of this system have not been changed and no
. new failure modes are added. Therefore, no increase in the possibility of an equipment malfunction of a different type than previously evaluated in the PSAR exists. Since the system is_ fully capable of satisfying the acceptance criteria, it does not create the possibility of an accident of a different type than any evaluated in the FSAR. No plant parameters or setpoints are altered by this change, therefore, the margin of safety as described in the basis to any Technical Specincation is not reduced. No Unreviewed Safety Question is judged to be created by this variation notice.
CE-2741
Description:
This exempt change replaced valve ICF138 with a l" Anchor-Darling double disc gate valve.
Safety Evaluation: Replacing valve ICF138 will not affect the function of the CF system as evaluated in the FSAR. The draining function of the valve will not be affected. The design of the new valve meets the temperature and pressure conditions experienced in this portion of the CF system. The new-
valve will perform the same function as the existing one and also will be less likely to develop leakage past the seat. The additional weight of the valve does not create any scismic concerns or require any additional support. No Unreviewed Safety Question is judged to be created by this modification.
CE 2753
Description:
This exempt change deleted valves INM193,203,213, and 222.
Flanged in spool pieces were put back in place of these valves.
Safety Evaluation: None of the components affected by this exempt change are safety related. The replacement of these NM valves with spool pieces will not adversely affect system operation. No Unreviewed Safety Questions are judged to be created by this modification.
CE 2756
Description:
This exempt change relocated valve IRLO25 downstream of expansion joint IRLI. Also, a new expansion joint was installed.
Safety Evaluation: The RL system does not perform any safety function.
Relocatinp, valve IRLO25 downstream of the expansion joint will not affect the function or o%rability of the RL system. The design parameters associated with the system will not be affected. No Unreviewed Safety Question is judged to be associated with this exempt changc.
CE 2757
Description:
This exempt change relocated valve 1RLO28 downstream of its expansion joint. Also, a new expansion joint was installed.
Safety Evaluation: The RL system does not perform any safety function.
Relocating valve IRLO28 downstream of the expansion joint will not affect the function or operability of the RL system. The design parameters associated with the system will not be affected. No Unreviewed Safety Question is judged to be associated with this exempt change.
CE 2781
Description:
This exempt change replaced valve ICA204, item #6J 238, with a similar valve having item #6J 335.
Safety Evaluation: Based upon the similarities of the valves, this replacement
, is an acceptable substitute and will not degrade safety or performance factors in any way. No Unreviewed Safety Question is judged to be created by this exempt change.
CE-2809
Description:
This exempt change installed a four way connector on the discharge tubing for IEMF48 in place of the Tee currently installed.
Safety Evaluation: The components affected by this exempt change are QA Condition 2. The four way connector and cap will meet the QA requirements for this class E application and also meet the design requirements for system i
temperature and pressure. The Nuclear Sampling System (Nhi) and the IEhiF48 serve no emergency function and are not needed to bring the unit to a safe shutdown condition. No Unreviewed Safety Question is judged to be created by this modification.
CE 2814
Description:
This exempt change deleted the reach rod remote operators for valves IND72, IND79, hnd IND80.
Safety Evaluation: The reach rods for these valves are not a QA condition item and do not perform a safety related function. Removing the reach rods per this exempt change does not affect the functions of these valves in any way. No Unreviewed Safety Question is judged to be created by this exempt change.
CE-2821
Description:
Th!s exempt change designated dedicated injection points for the addition of Hot Well layup chemicals and secondary treatment chemicals.
Safety Evaluation: The designation of chemical addition points has no affect on system operation and does not degrade the function of system components.
This exempt change has no direct or indirect impact on safety functions during.
any design basis event. No Unreviewed Safety Question is judged to be created by this exempt change.
CE 2845
Description:
This exempt change will correct the error on drawing CN 1577-1.2. The Unit i VA Air Flow hionitor has the wrong unit I.D.
Safety Evaluation: No Unreviewed Safety Question is judged to be created by this modification.
CE-2857
Description:
This exempt change will correct the error on drawing CN-1565-1.9. Valve IWLE36 is shown normally closed and should be shown normally open.
Safety Evaluation: No Unreviewed Safety Question is judged to be created by this correction to a flow diagram.
CE-2862
Description:
This exempt change provided stainless steel chemical addition points on the Condensate system (Cht),
Safety Evaluation: The installation of chemical addition points has no affect on the function of the Chi system as described in the FSAR section 10.4.7.
The existing materials for injection piping are not compatible with high concentrations of boric acid. Therefore, new addition points made of stainless steel were necessary. The new valves and piping are in accordance with the design requirements for the system. No Unreviewed Safety Question is judged i
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'l to be created by this exempt change.
CE-2876
Description:
This exempt change makes permanent the installation of a 13 5/8" spool piece fabricated of 2" S.S. Sch. 4v ,.ipe,4 2* Sch.150# FF Bronze Flanges,2-1/8" Garlock Gaskets and 8 C.S. studs and nuts in the 2" aluminum Yhi make-up piping to the YV Chiller Building.
- sfety Evaluation: The spool piece is not installed in a safety related app!ication and has no direct or indirect impact on safety functions during desig:t basis events. No UnreviewM N - ^uestion is judged to be created by thh modification.
CE 2921 Desenption: This exempt chaage disables the hiain Steam Leak Detection CE 2922 Sys'.em alarm annunciator.
Safety Evaluation: The steam lines in the SG doghouse are in the break exclusion area. To ensure that a steam line break does not occur in any of the break exclusion aras, an augmented inservice inspection program has been instituted. The program meets the ash 1B Code Section XI, IWC 2000 and IWD-2000. These requirements do not specify a main steam led: detection system such as currently on line at Catawba.
Therefore, use of this system is not required for compliance ,vith the NRC acceptance criteria for an augmented inservice inspection program. No Unreviewed Safety Question is judged to be created by this exempt change.
CE-2928
Description:
This exempt change revised drawings to renect that valve INB397, item #61235, has been replaced with a similar valve having item ChiV-489.
Safety Evaluation: Based upon similarities between the two valves, this replacement is an acceptable substitute and will not degrade safety or performance factors in any way. No Unreviewed Safety Question is judged to be created by this modification.
CE-3021
Description:
This exempt change replaced valve 2NM190A, item #9J 544, with a similar valve having item #9J 554 Safety Evaluation: Based upon similarities between the two valves, this replacement is an acceptable substitute and will not degrade safety or performance factors in any way. No Unreviewed Safety Question is judged to be created by this modification.
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Catawba Nuclear Station Summary of Procedure Changes, Tests, and Experiments l Completed From 11/1/89 to 9/30/90 -- Volume 1 IP/0/A/3817/12 Chango #11 This change adds "763 422" to the transmitter inspection sheet.
Exempt Chango CE-2223 adds Barton Transmitter Model 763 Sorial Number 422 to Note 4 on Index Manual Pago: R10.2 of the EQRI Manual. Note 4
. was added by Exempt Change CE-2030. It requires that epocific trans-l mitters have their internal pin connectors soldered por CNM-13999.60-0031 prior to being installed in a plant system. The 7
location of this Transmitter is unknown, but all model 763 Barton f Transmitters installed in the plant have boon inspected and veriflod
- not to be serial number 422. There is no unroviewed safety quest l7n associated with this chango.
4 MP/0/A/7600/37 Change #11 This chango is to incorporato a caution concerning the removal and reinstallation of the bonnet retainer ring. This caution is the result of Nuclear Regulatory Commission (NRC) Information Notico 89-62. As described in the Information Notico, gross back leakage cun occur due to improper location of tho bonnot retainer. This change adds NRC Information Notico 89-62 as a reference. Stop 6.4 was added to require the mechanic to contact Maintenance Engineering Sorvices (MES) prior to disassembly of the valve for special instructions concerning the removal /roinstallation of the bonnet retainer ring.
Also, a sign-off step was added to Data Shoot pago 1 to record the information about u r; acting MES in stop 6.4. This is a precautionary measure to ensure that the same problems described in the Information Notico are avoided. This change will act to reduce the probability or equipment malfunction. Consequently, no Unreviewod Safe.ty Question (USQ) is created,
~
PT/1/A/4200/17 Change #10 This change involves the method of recording stroke timo for valvo 1BB-56A. This valve has a Technical Specification (Toch. Spec.)
required stroke time of 10 seconds per Tech. Spoc. 3.6.3. That Spoc.
i also says timing will be pursuant to Tech. Spoc. 4.0.5 which referenc-es American Society of Hochanical Engincors (ASME)Section XI Code.
The code (IWV-3413(b)) allows stroke timing to the nearest second.
After discussion with station personnel, General Office Complianco and Design personnel, NRC Resident Inspectors, and NRC Region II person-nel, timing to the nearest second would have no effect on the proba-bility of an accident or malfunction of equipment.
1BB-56A has shown no degradation (steady increase in stroke timos) over the past two years of testing por IWV. Should degradation occur
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l and the stroke timo increano te greator than or equal to 10.5 or, condo, i the valve would be declared inoperable and corrective action 1:iltiat- ;
od. No new possibilities of accident or equipment malfuncticn are created. The margin of safoty is not reduced.
PT/1/A/4450/05B Change #23 A change was mada to this proceduro to allow using Digisnap model DSA-1000 clamp-on ammotor for the measuroment of motor currents. The acceptance criteria for motor currents were adjusted to encompass the additional error associated with the now ammotor. A change was mado to allow a Containment Air Rolceso and Addition System release during the performance of section 12.4. During the performance of section 12.4, the auction valvo (1VX2B) is closed when Hydrogen Skimmer Fan (HSF) IB is running, and no air is moved. Thoroforo, airborne contam-ination in containment will not be increased *y section 12.4.
A change was made to opon, then close, the L uaker for HSF-1B if the fan is operated by the procedure. Problems were encountered with the breaker on HSF-2A tripping, and a recommendation was made to cycle the breaker associated with HSF-1B after running the fan to unsure that the breaker trip latch lover is fully roset. After the breaker is re-closed, the indication on 1Mc4 is checked to verify that power has been returned to HSF-18. The procedure requires that HSF-1B is declared inoperable during this test, and HSF-1A will remain operable during this test as required by Tech. Specs, cycling of the breaker will decrease the possibility of an equipment malfunction. Opening and closing the breaker, and verifying that power is returned after re-closing the breaker, is indopondently verified within section 12.5 of the proceduro, other changes were made to improvo clarity, dolote confusion within the procedure, and correct typographical errors. The order of steps and alignment of systems was not changod. Therefore, an unroviewed safety question is not created by this procedure chango.
PT/2/A/4450/05B Change #12 l A change was made to this procedure to allow using Digisnap model )
DSA-1000 clamp-on ammeter for the measuromont of motor currents. The acceptance criteria for motor currents were adjusted to encompass the i additional error associated with the new ammeter. A change was made I to allow a VQ release during the performance of section 12.4. During <
the performance of section 12.4, the suction valve (2VX2B) is closed I when HSF-2B is running and no air is moved. Therefore, airborne l contamination in containment will not be increased by section 12.4.
l A change was made to open, then closo, the breaker for HSF-2B if the fan is operated by the procedure. Problems were encountered with the breaker on HSF-2A tripping, and a recommendation was made to cycle the l breaker associated with HSF-2B after running the fan to ensere that I the breaker trip latch lever is fully reset. After the breakar is 2
s re-closed, the indication on 2MC4 is checked to verify that power has been returned to HSF-28. The procedure requires that HSF-1B is declared inoperable during this test, and HSF-2A will remain operable during this test as required by Tech. Specs. Cycling of the breaker ;
will decrease the possibility of an equiptent malfunction. Opening -
I and closing the breaker, and verifying that power is returned after re-closing the breaker, is independently vexified within section 12.5 of the procedure.
Other changes were made to improve clarity, delete confusion within the procedure, and correct typographical errors. The order of steps and alignment of systems was not changed. Therefore, an unreviewed safety question is not created by this procedure change.
PT/0/A/4450/08C Change #1 The purpose of this restricted procedure change is to ensure operabil-ity of Train B Control Area Ventilation (VC) while closing / opening IPFT-HVD-2. This is being done to allow work on 1CR-D-9. This change requires continuous monitoring-of the critical parameters assumed in the Safety Analysis for VC to ensure that they stay within safety limits while closing / opening 1PFT-MVD-2. System Manual Volume Dampers (MVDs) may be adjusted to maintain these parameters. 1PFT-MVD-2 will be closed / opened in increments, and critical parameters will be verified each time the MVD is moved. Final position of 1 PPT-MVD-2 will be the same as:the "As Found" position to ensure that A train balan,ce will not be.affected. Therefore, the margin of safety will not be reduced. Installation / removal of test instrumentation, secur-ing duct access doors, and securing all adjusted MVDs are independent-ly verified within section 12.0 or 13.0 of this change. Also, the .
final position of all adjusted MVDs will be marked, and the method of marking will be documented on Enclosuro 13.11. There is no USQ created by this change.
PT/0/A/4450/08C Change #2 The purpose of this restricted procedure change is to ensure operabil-ity of Train A VC while closing / opening 2PFT-MVD-2. This is being done to allow work on 2CR-D-4, 2CR-D-9, and 2CR-D -10. This change requires continuous monitoring of the critical parameters assumed in l the safety Analysis for VC to ensure that they stay within safety limits while closing / opening 2 PPT-MVD-2. System MVDs may be adjusted -
to maintain these parameters.- 2PFT-MVD-2 will be closed / opened in increments, and critical parameters will be verified each time the MVD is moved. Final position of 2PFT-MVD-2 will be the same as the "As Found" position to ensure that B train balance will not be.affected.
Therefore, the margin of safety will not_be reduced. Installa-tion / removal of test instrumentation, securing the cap at the dis-charge of 1CRA-PFT-1, and securing all adjusted MVDs are independently verified within section 12.0 or 13.0 of this-change. Also, the final position of all adjusted MVDs will be marked and the meth,od of marking-3
I will be documented on Enclosuro 13.11. Th2re is no USQ created by this chango.
MP/1/A/7150/42 Re-type, Changes 0 to 12 Incorporated This procedure performs removal / replacement activities required on the Reactor Vessel Head during refueling outages.
This ovaluation is for changes mado during the procedure review following Catawba Unit 1 End of Cycle 3 outage. These changes were required to clarify steps for better understanding and add manufactur-ers information on new equipment. See proceduro changes listed below.
- 2.1.6 - Added CNM 1144.28.0037-001, Instruction manual for Reactor Vessel (R.V.) Nozzle Inspection Hatch Covers.
- 8.2 - Added 0-300 torque wrench.
- 8.3.12 & 8.3.13 - Reduced tygon tubing and rope length.
- 10.2 - Added note on hrsd shielding blanket installation - moved from step 10.6.1.
- 11.3.2 - Added 2744 SWR to stop.
- 11.3.8 thru 11.3.8.2 - Deleted these steps due to new mechanical seals.
- 11.4.24.1 - Increased torque on stud hole plugs to 60 ft. Ibs.
This recommendation from plug manufacturer to reduce leakage.
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- 11.4.33 through 11.4.37.1 - Revised to add installation instruc-tions for mechanical sealing nozzle covers por manufacturer's manual.
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- 11.5.14 & 11.5.15 - Changed wording to require inspection on head l lifting rig.
- 11.5.23 - Added caution stop for stud holo cleaning tool brush rotation.
- 11.5.33 - Changed wording for operations sign-off to allow l
l tensioning to begin and provide a time for expected Mode 5 requirements.
- 11.5.69 thru 11.5.72 - Revised to add removal instructions for nozzle covers.
- Enclosure 13.5 - Added nozzle port cover locking sequence.
- Enclosure 13.1, Page 14 of 23 - Relocated stop 11.5.39 for clarity.
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l This procedure will be used to maintain the Reactor Vessel Head in its original design requirements and specifications. The Catawba Final Safety Analysis Report (FSAR) and Technical Specifications are not affected by this procedure change. The probability of an accident or a malfunction previously addressed will not be increased nor will any l unreviewed safety question be involved.
l MP/0/A/7600/23A Initial Issue This procedure provides a method for disassembly, inspection, reassem-bly, and corrective maintenance for BIF Butterfly Valves (Handwheel ONLY). Technical information was obtained from manual CNM 1205.02-111 and drawings CNM 1205.02-32, 33, 38, 39, 41, 42, 49, 51, 53, 60, 64, 174, and 223. This procedure also contains information specific to Catawba.
The purpose of this ovaluation is to describe the changes made to MP/0/A/7600/23A as part of the procedure upgrado process and the identified technical enhancements. While many of these changes are not significant in content and are editorial in nature, some technical information was added.
The following is a summary of the changes made to this procedure:
- Provided more detail on the seat preparation and installation
- Addition of scribe marks to positively indicate the disc pocition inside the valvn
- Addition of specific instructions on the method of replacing the disc on valves with a retaining ring
- Addition of steps to pressure test the valvo after repairs and/or before return to service
- Provided more detail on the setting of the operator mechanical stops
- Addition of steps to install an upgraded position indicator and I
to verify that it is accurately set
- Addition of steps to record specific valve information to maintain a complete history for the valves
- Addition of enclosure denoting torque values retrieved from the outline drawings The purpose of this procedure is to correct and improve the perfor-mance of these valves within their original design requirements and specifications.
The FSAR and Technical-Specifications will not be affected by the-changes described above. The probability of an accident or 5
malfunction previously addressed will not be increased. No unreviewed safety questions are involved.
MP/0/A/7600/23B Initial Issue This procedure provides a method for disassembly, inspection, reassom-bly, and corrective maintenance for BIF Butterfly Valves (Electric Motor Operated ONLY). Technical information was obtained from manual CNM 1205.02-113 and drawings CNH 1205.02-33, 35, 39, 42, 56, 58,. 62, 66, 68, 148, 149, 176, 223, 225, 243, and 296. This procedure also contains information specific to Catawba.
The purpose of this evaluation is to describe the changes made to MP/0/A/7600/23B as part of the procedure upgrade process and the identified technical enhancements. While many of these changes are not significant in content and are editorial in nature, some tect.nical information was added.
The following is a summary of the changes made to this procedure:
- Provided more detail on the seat preparation and installation
- Addition of scribe marks to positively indicate the disc position inside the valve
- Addition of specific instructions on the method of replacing the disc on valves with a retaining ring
- Addition of steps to pressure test the valve after repairs and/or before return to service
- Provided more detail on the setting of the operator mechanical stops
- Addition of steps to install an upgraded position indicator and to verify that it is accurately set
- Addition of steps to record specific valve information to maintain a complete hietory for the valves
- Addition of enclosure denoting torque values retrieved from the outline drawings The purpose of this procedure is to correct and improve the perfor-mance of these valves within their original design requirements and specifications.
The FSAR and Technical Specifications will not be affected by the changes described above. The probability of an accident or malfunc-tion previously addressed will not be increased. No unreviewed safety questions are involved.
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PT/1/A/4150/12B Re-type, Changes 0 to 3 Incorporated CHANGES:
- 1. Original Enclosures 13.3 and 13.5 dela'ted since they are no
, longer necessary for determination of Doppler Tamparature Coeffi-cient and Differential Boron Worth. This data is now provided by z Duke Power Startup and Operational Report (SOR) (instead of Westinghouse Nuclear Design Report (NDR)) in a straightforward manner.
- 2. Reference to Westinghouse NDR deleted.in Step 2.2 and "new" Enclosure 13.4, and Duke Power SOR. added as data source for this procedure to adopt use of this design data in lieu of West ng-house data.
- 3. Reference to Boronometer deleted from steps 8.6.3 And 12.3 and Enclosure 13.1 to reflect deletion of this plant component. Step 12.3 deleted entirely as result of this change.
- 4. Steps added to "new" Enclosure 13.4 (formerly Enclosure 13.6) to determine Differential Boron Worth from Duke Power SOR.
- 5. Steps added to "new" Enclosure 13.5 (formerly Enclosure 13.7) tv determine Doppler Temperature coefficient from Duke Power SOR.
- 6. Multiple Sign-off Form added to procedure as Enclosure 13.7.
EVALUATION:
The only substantive change in the reissue of this procedure is-the replacement of the Westinghouse Nuclear Design Report (NDR) with the Duke Power Design Engineering Startup and Operational Report (SOR) as .
the core design data resource for completing the calculations involved in the test. The Duke Power SOR is as rigorously reviewed and quali-fled a document as the Westinghouse NDR, making this replacement permissible. Calculation of the Moderator Temperature coefficient at End of Life (<300 ppmB critical boron concentration) for-the purpose of ensuring compliance with Tech.-Spec. 3.1.1.3 is performed in exactly the same way; the data to do-so is simply obtained from a different reference source. Due to thorough safety. analysis of the Duke Power SOR methodology and results, neither the probability nor the consequences of any accident or safety related equipment malfunc-tion (either analyzed or unanalyzed) will be increased by this change.
The margin of safety as defined in Tech. Spec. bases will be main-tained through verification of compliance with Tech. Spec.-3.1.1.3 via acceptable methodology.
PT/2/A/4200/13K Re-type, Changes 0 to 8 Incorporated The purpose of this procedure is to perform IWV valve stroke tests of Auxiliary Feedwater (CA) valves 2CA149, 2CA150, 2CA151, and 2CA152.
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These valves are normally tested during Cold Shutdown (Modes 5, 6, or No Mode). However, upon review of past procedure changes, it was found that temporary changes were writton when those valves had maintenance performed on them and were ratosted in other modos, usually at the end of an outage while Unit 2 was attempting startup (Modes 3 and 4). Testing these CA valves involves stroking each valve from open to closod, thus 1:elhting flow to the affected steam genera-tor upper nozzle. However, as long as the lower nozzle is receiving 85% or less of total rated Main Feodwater (CF) flow to the steam gennrator (total CF flow to the affected steam generator at 100% rated power), the CA valve may bo isolated. Although theso CA valves could be tested in all modes, this procedure limits testing to the follow-
! ing: Modos 2, 3, 4, 5, 6, and No Mode. Additionally, when isolation occurs during testing, if steam is being produced in the steam genera-tors, and Auxiliary Feedwater (CA) is the sole supplier of feedwater, the water level of the isolated steam generator could become low enough to exceed the Low-Low level trip point, causing a CA autostart signal. Procautionary steps and measures are being inserted in thir, re-type to warn against low steam generator levels which could result, should those CA valves be left closed for a prolonged period.
Since the flow path of 2CA149, 2CA150, 2CA151, and 2CA152 may be isolated during Modos 2, 3, 4, 5, 6, or No Mode, no CA valve is being made inoperable as a result of this test. As a result, neither the probability of nor consequences of any accident previously evaluated in the FSAR is increased. No accident different than those already evaluated in the FSAR is created. The probability of, and consequenc~
es of, a malfunction of equipment important to safety different from thoso already evaluated in the FSAR is not created. The margin of safety as defined in the Technical Specification Bases is not reduced.
PT/0/A/4450/08C Initial Issue The purpose of this procedure is to optimize the flow of both Trains of the Control Area Ventilation (VC) system to the Control Room and the control Room Area. In an accident condition, the primary purposo l of the VC System is to provide uncontaminated, filtered air to the Control Room (CR) and to pressurize the CR to ensure no inleakage of contaminated air. Outside air, taken normally from two intakes, is used to pressurizo the CR and the CR area. During an accident, one (or both) of these intakes may isolato due to high r6diation, so VC must be able to pressurize the CR with either intake isolated. (If both intakes isolate, the Operator opens the least contaminated one.)
This outside and recirculation air from the CR passes through a filter unit consisting of High Efficiency particulato Air (HEPA) filters and a carbon adsorber bed. The flow through this filter unit should be close to the design flow to ensure proper filtration of the air. Flow exiting this filter unit is split as necessary to pressurize the CR and the CR area. Updated criteria for the split of these two flows has been developed by Design Engineering to ensure proper pressuriza-tion of the CR and CR. areas for all operating alignments. Sections of this procedure allow taking measurements and adjusting Manual Volume Dampers (MVDs) as necessary to optimize _all of the above mentioned 8
flows and CR pressurization. If a MVD is moved in one section that could affect the flows in another section, the procedure requires that the affected section is performed OR the sections are performed con-currently, and final data is not taken until all MVDs have been ,
adjusted.
Other sections of this procedure allow taking measurements and adjust-ing MVDs to ensure other areas of the VC system are operating as designed. Total supply flows to the CR, CR area, switchgear rooms, ,
and exhaust flows from the battery room, will be verified to ensure I proper operation. Acceptance criteria for these sections were devel-4 oped using CNTC-1578-VC series of Test Acceptance Criteria Sheets from Design Engineering.
All measurement of flows and adjustment of MVDs are performed while the VC System is in its normal operating alignment. When performing the sections that affect or may affect safety related flows due to MVD
. adjustments, the train of VC under test will be declared inoperable.
2 The opposite train of VC will remain operable during this period as required by Tech. Specs. Therefore, the margin of safety as defined in the bases of Tech. Specs. Will not be reduced. Installa-tion / removal of test instrumentation, securing duct access. doors, and securing any adjusted MVDs are independently verified within section .
12.0 or 13.0 of the procedure. Also, all adjusted MVDs will be marked, and the method of marking will be documented on Enclosure 13.11. For these reasons and the ones stated above, this procedure does not create or increase the probability of a malfunction of equipment important to safety OR create or increase the consequences of an accident. Therefore, an unreviewed safety question does not exist.
PT/1/A/4550/03B Retype #2, Changes 0 to 7 Incorporated Additional changes incorporated:
- All references to obtaining insert identification numbers. The reason for this procedure is to inventory Special Nuclear Materi--
als (SNM--Fuel Assemblies) and inserts are not SNM.
- Step 2.1 -- Deleted reference to Regulatory Guide 5.13 since it is not applicable to an operating power station, and replaced it with 10CFR70.51 since that is applicable.
- Step 2.4 -- Deleted references to itemized Tech.-Specs. since they do not add in the use of the procedure.
- Throughout procedure abbreviated Spent Fuel Pool as SFP and Fuel Assembly as F/A.
- Step 6.11 -- Deleted precaution about moving load of >3000 lbs.
above F/A in SFP since this will never be challenged by the procedure.
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- Section 5.0 ~~ Reorganized test equipment section, but equipment is still the same.
- Section -- 8.0 Deleted redundant steps since the operating Procedure (OP) for declaring manipulator crane verifles the same l conditions.
- Sections 9.0 and 12.0 ~~ Spelled out in more detail what is being ,
l required by procedura now that insert numbers are not being read.
- Step 12.17 -- Added step for removing equipment from SFP.
- Enclosure 13.3 -- At bottom of page, reduced the amount of 3
information recorded by data taker to a simple name and date since other data is unnecessary.
This inventory procedure is not required by Tech. Specs. and is not mentioned in the FSAR, but it is required by 10CFR70.51. Catawba is committed to perform this procedure every six months per Administra-
. tive Policy Manual (APM) 3.9.5.1 and Station Directive 3.9.1. This procedure, while satisfying the Special Nuclear Material (fuel) inventory in the Spent Fuel Pool, does this without placing any equipment in an unusual position from which they are designed. No Tech. Specs, are violated by the use of this procedure.
3 PT/2/A/4550/03B Retype #2, Changes 0 to 5 Incorporated Additional changes incorporated:
- All references to obtaining insert identification numbers. The reason for this procedure is to inventory Special Nuclear Materi-als (SNM--Fuel Assemblies) and inserts are not SNM.
- Step 2.1 -- Deleted reference to Regulatory Guide 5.13 since-it 4
is not applicable to an operating power station,=and replaced it
- with 10CFR70.51-since that is applicable.
- Step 2.4 -- Deleted references to itemized Tech Specs since they do not add in the use of the procedure. _
- Throughout procedure abbreviated Spont Fuel Pool as SFP and Fuel Assembly as F/A.
- Step 6.11 -- Deleted precaution about moving load of >3000 lbs.
above F/A in SFP since this will never be challenged by the procedure.-
- Section 5.0 -- Reorganized test equipment section, but equipment is still the same.
- Section 8.0 -- Deleted-redundant steps since OP for declaring manipulator crane verifles the same conditions. ,
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,. ..a.,_..._._-..._._ . - _ . . _ _ - _ . _ _
- Sections 9.0 and 12.0 -- Spelled out in more detail what is being required by procedure now that insert numbers are not being read.
- Step 12.17 -- Added step for removing equipment from SFp.
- Enclosure 13.3 -- At bottom of page, reduced the amount of information recorded by data taker to a simple name and date since other data is unnecessary.
This inventory procedure is not required by Tech. Specs. and is not mentioned in the FSAR, but it is required by 100FR70.51. Catawba is committed to perform this procedure overy six months per APM 3.9.5.1 and Station Directive 3.9.1. This procedure, while satisfying the Special Nuclear Material (fuel) inventory in the Spent Fuel pool, does this without placing any equipment in an unusual position from which they are designed. No Tech. Specs are violated by the use of this procedure.
TN/2/A/0205/00/01A Initial Issue This procedure provides for implementation of Nuclear Station Modifi-cation (NSM) CN-20205, Rev. O. NSM CN-20205 will replace the person-nel airlock air relief line and associated check valves to provide adequate venting of the airlock in case of-internal pressurization.
The air relief line and check valves are sized to ensure that internal airlock pressure does not exceed the design pressure of 15 psig.
This procedure is for the Unit 2 upper al: lock. The existing sar relief line will be cut out and removed. A special studding outlet and check valves will be installed in the same location. Appropriate notes are included in the procedure to ensure \his work is done in Modes 5,6 and No Mode (i.e., when containment integri?y is not re-quired.) During core alterations, Tech Specs. require at least one door in each airlock to be closed. Notes to ensure this Tech. Spec.
is met are included. After completion of the work, the check valves will be tested to verify their opening pressure. Also, a leak rate test will be performed on both the check valves and the entire airlock. All of the testing will be completed prior to returning the airlock to service. An unroviewed safety question is not created by this procedure.
TN/2/A/0205/00/02A Initial Issue This procedure provides implementation guidelines for NSM CH-20205, Rev. O. NSM CN-20205 will replace the personnel airlock air relief line and associated check valves to provide adequate venting of the airlock in case of internal pressurization. The air relief line and check valves are sized to ensure that internal airlock pressure does 1 not exceed the design pressure of 15 psig. 1 l
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This procedure is for the Unit 2 lower airlock. The existing air relief line will be cut out and removed. A special studding outlet and check valves will be installed in the same location. Appropriate notes are included in the procedure to ensure this verk is done in Mod:ss 5, 6 and No Mode (i.e., when containment integrity is not required.) During core alterations, Tech. Specs. require at least one door in each airlock to be closed. Notes to ensure this 1ech. Spec.
is met are included. Af ter completion of the work, the chtsck valves will be tested to verify their opening pressure. Also, a leak rate test will be performed on both the check valves and the entire airlock. All of the testing will be completed prior to returning the airlock to service. An unroviewed safety question is not created by this procedure.
TN/2/B/0422/00/01A Initial Issue This procedure provides for tne implomontation of NSM CN-20422, Work Unit 01. NSM CN-20422 installs a computer point to monitor the operation of the feeddater isolation valvo's hydraulic motors. The hydraulic motors are used in the valve actuators to pump hydraulic fluid into the cylinder to force the valve open. The motor periodi-cally starts to maintain the hydraulic pressure and the valve's open position. The frequency of motor operation gives an indication of hydraulic leaks in the system. The computer will monitor the motor start relay and alarm if the motor starts more than once every six hours. The motor breaker will be monitored by connecting to auxiliary breaker contacts. The motor and actuator circuits are not being affected by this NSM. Alreo, the safe position of these valves is closed, The motor only operates to open the valve. The safety related valve close circuitry is not being affected by this NSM or procedure.
W0rx in this procedure is required to be completed before entering mode 4. This is because the feedwater isolation valves are required to be stroked before entering mode 4. This procedure tags out the l
hydraulic pump motor and thus must be complete to enable Performance testing. The valve stroke test is required for this modification.
This procedure will also test the relay contact and computer point and has Performance test the computer point timing function.
Work in this procedure will be performed with the unit in an outage in modes 5, 6, and no mode. The only exception is the cable pulling which can be performed with the unit at power. A hold statement is used to indicate the procedure steps that require an outage. Breakers specified for isolation affect the valve controls and hydraulic pump motor only. The clearing of these isolations is required for mode 4 i
and is addressed in the procedure by requiring the paperwork to be l cleared by mode 4.
l l The feedwater isolation valves are required for containment integrity in modes 1 through 4. The work in this procedure and associated paperwork are required to be completed before entering mode 4.
Therefore, the probability of an accident or malfunction of equipment 11 1
~ _
evaluated in the FSAR will neither be increased nor created as a result of this procedure. The margin of safety as defined in the 4
basis to any of the Tech. Specs. will not be reduced. The implementa- i tion of this procedure will not create an unreviewed safety question.
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TN/1/B/1036/00/01A Initial Issue This procedure provides guidance for irplementation of NSM CN-11036.
NSM CN-11036 installs a computer point to monitor the operation of the feedwater isolation valve's hydraulic motors. The hydraulic motors are used in the valve actuators to pump hydraulic fluid into the en inder to force the valve open. The motor-periodically starts to
" .ntain the hydrculic pressure and the valve's open position.
The frequency of motor operation gives an indication hydraulic leaks in the system. The computer will monitor the motor start relay and alarm if the motor starts more than once every six hours. The motor breaker Will be monitored by connecting to auxiliary breaker contacts. The
! motor and actuator circuits are not being affected by this NSM. Also, the safe position of these valves is closed. The motor only_operatec to open the valve. The safety related valve close circuitry is not being affected by this NSM or procedure.
Work in this procedure is required to be completed before entering mode 4. This is because the feedwater isolation valves are required to be stroked before entering mode 4. This procedure tags out the hydraulic pump motor and thus must be complete to enable Performance i
testing. The valve stroke test is required for this modification.
This procedure will albo test the relay contact and computer point and has Performance test the computer point timing function.
l Work in this procedure will be performed with the unit in an outage in modes 5, 6, and no mode. The only exception is-the cable pulling which can be performed with the unit at power. A hold statement is j used to indicate the procedure steps that require an outags. Breakers l
specified for isolation affect the valve controls and hydraulic pump motor-only. The clearing of these isolations is required for mode 4 and is addressed in the procedure by requiring the paperwork to be clea' cod by mode 4.
f The feedwater_ isolation valves are required for containment integrity in modes 1 through 4. The work in this procedure and associated paperwork are required to be completed before entet_ng mode 4. '
Therefore, the probability of an accident or malfunction of equipment evaluated-in the FSAR will neither be increased nor created as a result of this procedure. The margin of safety as defined in the >
basis to any of the Tech. Specs will not be reduced. The implementa-tion of this procedure will not create an unreviewed safety question.
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TN/1/A/1088/00/01A Initial Issue '
This procedure provides for implementation of-NSM CH-11088, Rev' O, .
Work Unit 01. NSM CN-11088, Rev. O, reroutes the valve stem leakoff 13 m y- .-,s,ymy -
,ar-y s w- we-,-a-# qiw- --di*
-e girn-sy.e,w-,- p --wi.+- g.-. y ., y .-r - .i. g ,eurye--m.9yg,"4- T"' "*--=-*'7-C'"t- *
- I i '
l header so it discharges into the Reactor Coolant Drain Tank (NCDT) recirculation line. The new connection to the recirculation line is at the lowest possible point in order to allow gradual entrainment and 4 condensation of the steam. The purpose of this procedure is to provide guidance for the modification to the valve stem leakoff neader and NCDT recirculation piping.
Implementing this procedure will require isolation and draining of the NCDT. Isolating the NCDT will affect the following inputs to the NCDT:
- 1) All Reactor Coolant Pumps No. 2 and 3 Seals
- 2) Chemical and Volume Control (NV) System Excess Letdown Line
- 3) NV System Regenerative Heat Exchanger No. 1 Tube Side Drain
- 4) NV System Excess Letdown Heat Exchanger No. 1 Drain
- 5) Reactor Vessel Head Flange Leak Detection Drain
- 6) Valve Stem Leakoff Header The operations Group will coordinate the isolations necessary to implement this procedure. The piping modification will be implemented during an outage and in Modes 5, 6 and no mode. The reactor coolant pumps and NCDT will be out of service during the piping modifications.
Some post-modification testing will be performed in Mode 3 with the NCDT back in service. Implementing this modification does not affect the operation or function of any systems required for refueling operations.
A temporary station modification will be implemented by the Ms. ?.e-nance Engineering Services Group to monitor vibration of the piping affected by this modification. The monitoring will continue for i several months, and possibly until the next refueling outage. After monitoring is complete, Design Engineering will analyze the data to determine if any unacceptable vibration occurred.
Implementation of the procedure does not create any new accident scenarios and does not affect any new equipment important to safety.
Per this discussion, an unreviewed safety question does not exist.
TN/1/A/1103/00/01A Initial Issue-l This procedure will provide reactor coolant leg temperature to the j safe shutdown facility from reactor coolant pump loop C (1NCRDS910) i instead of loop A (1NCRD5860).
The implementation of the referenced procedure will affect the cold l
leg wido range instrumentation (RTCs) for the reactor coolant system (NC), -with regard to remote shutdown and accident monitoring in the Safe Shutdown Facility (SSF). The RTDs that are modified by this
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- , . . . . - . , _ _ - - , , ,. ~ ~ . . - - _a,__..,..._.- _ . . . - . . - _ ~ .. _ . _ _ . -;
procedure are safety-related (QA-1). The use of individual station control procedures is required to complete the implementation of this 1 procedure.
The wide range temperature instrumentation that is modified by this procedure is required to be operable in Modes 1, 2, 3, and 4. Due to this requirement, the referenced procedure will not be implemented during the modes in which operability is required. Tnir procedure will be completed and signed-off prior to Unit One going up to Mode 4.
This procedure is scheduled to be implemented during the Unit 1 End Of Cycle 4 refueling outage during which the wide range temperature instrumentation for NC is not required.
Each affected instrument will be tested prior to returning the instru-ment to service to ensure proper operation. As a result, the intent of the modification will be fully challenged. A cross-calibration test for RTD loop 1NCRD5910 will be performed independently of the referenced procedure during Modes 3 and 4. No Tech. Spec. Will be affected, as one of the reactor coolant loop cold leg RTDs will be operable prior to Mode 3. No unreviewed safety questions exist.
TN/1/A/2151/CE/01A Initial Issue This procedure installs a guide bushing on Main Feedwater (CF) Regu-lating (Reg.) Valve ICF28 yoke to absorb the sido load that the actuator stem applies on the valve stem. The valve yoke will be removed from the valve so that machine work and part installation can be performed. The machine work on the yoke will be directed by the valve manufacturer who will be on site with proprietary drawings. The manufacturer's representative will also supervise the installation of the bushing. All work will be performed on site in accordance with existing station procedures.
The isolations for this procedure are addressed in the Instrument and Electrical (IAE) procedure for actuator removal. The valve actuator l must be removed to implement this procedure. This is a standard i maintenance practice that is directed by an existing IAE procedure.
The necessary system isolations and clearing of them are covered by the IAE procedure.
The unit mode requirement is that Unit 1 be in an outage in modes 5, 6, or no mode. All work in this procedure is required to be complete !
before entering mode 4. A pre-review of all paperwork will be per- !
formed before entering mode 4. This will allow a leak test to be '
performed on the valve at system temperature and pressure. The testing to be performed to complete this modification will include a packing drag test and stroking the valve to verify free movement over the entire stroke. These two tests are addressed in maintenance procedures and are included in this procedure with a sign-off for completion.
Accordingly, the probability of an accident or malfunction of equip-ment evaluated in the FSAR will not be increased. The probability of I 15 i
an accident or malfunction of equipment different than those evaluated in the FSAR will not be created. The margin of safety as defined in the basis to any of the Technical Specifjcations will not be reduced by the implementation of this procedure. This procedure will not create an unroviewed safety question.
TN/1/A/2151/CE/02A Initial Issue This procedure installs a guide bushing on CF Reg. Valve ICF37 yoke to absorb the sido load that the actuator stem applies on the valve stem.
9 e valve yoke will be removed from the valve so that machine work and palt installation can be performed. The machine work on the yoke will be directed by the valve manufacturer who will be on site with propri-etary drawings. The manufacturer's representative will also supervise the installation of the bushing. All work will be performed on site in accordance with existing station procedures.
The isolations for this procedure are addressed in the IAE procedure for actuator removal. The valve actuator must be removed to implement this procedure. This is a standard maintenanca practice that is directed by an existing IAE procedure. The necessary system isola-tions and clearing cf them are covered by the IAE procedure.
The unit mode requirement is that Unit 1 be in an outage in modes 5, 6, or no mode. All work in this procedure is required to be complete before entering mode 4. A pre-review of all paperwork will be per-formed before entering mode 4. This will allow a leak test to be performed on the valve at system temperature and pressure. The testing to be performed to complete this modification will include a packing drag test and stroking the valve to verify free movement over the entire stroke. These two tests are addressed in maintenance procedures and are included in this procedure with a sign-off for completion.
Accordingly, the probability of an accident cc malfunction of equip-ment evaluated in the FSAR will not be increased. The probability of an accident or malfunction of equipment different than those evaluated in the FSAR will not be created. The margin of safety aa defined in the basis to any of the Technical Specifications will not be reduced by the implementation of this proceJoro. This pro' dure will not create an unreviewed safety question.
TN/1/A/2151/CE/03A Initial Issue l
This procedure installs a guide bushing on CF Reg. Valve ICF46 yoke to absorb the sido load that the actuator stem applies on the valve stem.
The valve yoke will be removed from the valve so that machine work und part installation can be performed. The machine work on the yoke will be airected by the valve manufacturer who will be on site with propri-etary drawings. The manufacturer's representative will also supervise the installation of the bushing. All work N111 be-performed on site in accordance with existing station procedures.
16
The isolations for this procedure are addressed in the IAE procedure for actuator removal. The valvo actuator must be removed to implement this procedure. This is a standard maintenance practice that is directed by an existing IAE proceduro. The necessary system isola-tions and clearing of them are covered by the IAE procedure.
The unit modo requirement is that Unit 1 be in an outage in modes 5, 6, or no mode. All work in this procedure is required to be complete before entoring mode 4. A pre-review of all paperwork will be per-formed before entering mode 4. This will allow a leak test to be performed on the valve at system temperature and "ressure. The testing to be performed to complete this modification will include a packing drag test and stroking the valvo to verify free movement over the entire stroke. Those two tests are addressed in maintenance procedures and are included in this procedure with a sign-off for completion.
Accordingly, the probability of an accident or malfunction of equip-mont evaluated in the FSAR will not bo increased. The probability of an accident or malfunction of equipment different than those evaluated in the FSAR will not be created. The margin of safety ac defined in the basis to any of the Technical Specifications will not be reduced by the implementation of thint procedure. This procedure will not create an unroviewed safety question.
TN/1/A/2151/CE/04A Initial Issuo This procedure installs a guido bushing on CF Reg. Valve ICF55 yoke to absorb the side load that the actuator stem applies on the valve stem.
The valvo yoke will be removed from the valvo so that machine work and part installation can be performed. The machine work on the yoke will be directed by the valve manufacturer who will be on site with propri-etary drawings. The manufacturer's representative will also supervise the installation of the bushing. All work will be performed on uite in accordance with existing station procedures.
The isolations for this procedure are addressed in the IAE procedure for actuator removal. The valve actuator must be removed to implomont this procedure. This is a standard maintenanco practico that is directed by an existing 1AE procedure. The necessary system isola-tions and clearing of them are covered by the IAE procedure.
The unit mode requirement is that Unit 1 be in an outage In modes 5, 6, or no mode. All work in this procedure is required to be complete before entering mode 4. A cro-review of all papet'- ;k will be per-formed before entering mode 4. This will allow a leak test to be performed on the valve at system temperature and pressure. The testing to be performed to complete this aodification will include a packing drag test and stroking the valve to verify free movement over the entire stroke. Theso two tests are addressed in maintenance procedures and are inc.tuded in this procedure with a sign-off for completion.
l I
i 17
Accordingly, the probability of an accident or malfunction of equip-ment evaluated in the FSt.R will not be increased. The probability of an accident or malfunction of equipment different than those evaluated in the FSAR will not be created. The margin of safety as defined in l the basis to any of the Technical Specifications will not be reduced l by the implementation of this procedure. This procedure will not create an unreviewed safety question.
TT/0/A/9100/49 Initial Issue Design study CNDS-189/00 is reviewing Catawba's Emergency Core Cooling System (ECCS) actuation system which requires both trains of the VC system to start on a safety Injection (SS) signal and evaluating an appropciate resolution to reduce the post-Design Basis Accident (DBA) sound level in the Control Room (CR). The purpose of this temporary procedure is to obtait. data on the VC System befo" and after an SS signal has been received to aid in the resolution vi the referenced Design study.
Various data will be taken on the VC system before and after the SS signal has been initiated. In order to initiate an SS signal on B train of VC, a jumper will momentarily be placed in the B train control cabinet. The momentary placement of this jumper will allow resetting B train of the VC system at any time during the test without having to rrlet the Diesel Generator (D/G) sequencer first. Since th=
B train chiiler will not start on the SS signal, the procedure re-l quires that CR temperature be monitored continuously after the SS signal has been initiated until B train is reset. If the temperature
- in the CR approaches 85 dog F, after initiation of the SS signal, the
[
procedure requires tha; B train of VC be reset immediately to reduce CR temperature. Tech. Spec. requires the 09 temperature to be s 90 deg F, so a conservative value of 85 deg F was chosen. If any unusual conditi?ns arise with Unit 1 or Unit 2 during the post SS signal operat ion, the test ma be aborted and A train of VC returned to the normai operating alignment.
Train A of VC is required to be operable by the procedure and will be the selected train in opc ation before the SS signal. Initiation of the SS signal will a.)ign the.VC system to the post-DBA operating alignmer' so that the ma'rgin of safety as defined in the bases of Tech. cec. will not be reduced. All data will be taken while the system is operating, and the measurement of this data will not affect the operation OR operability of the VC system. Placement / removal of all jumpers and installation / removal of all applicable test equipment will be independently verified within sections 12.0 and 13.0 of the procedure. For these reasons, this *.rocedure does not create an un-reviewed safety question.
OP/0/B/6400/08 Change #13 This proccours change addresses the operation of a mobile filter flow onit which will supply filtered water to the Filtered Water (YF) 18
system during times that the YF upflow sand filters are out of ser- .
vice. The mobile filter unit is supplied with a continuous turbidity V analyzer and alarm to ensure the water quality is equal to or greater than the YF Upflow Sand Filters. The only available raw water supp1v with enough pressure to facilitate proper flow rates is the fire protection system (RF) .
The Fire Protection Hose Racks are addressed in Technical Specifica-tion section 3/4.7.10.4 and table 3.7-3. In section 3.7.10.4 it states that " fire hose stations shall be OPERABLE when equipment in the areas protected by fire hose statior.s is required". It also states that "With one or more of the fire hose stations INOPERABLE, provide gated wye (s) on the nearest OPERABLE hose station (s). One outlet of the wye shall be connected to the standard length of hose I provided for the station. The second outlet of-the wye shall be con-nected to a length of hose sufficient to provido coverage for the area ,
left unprotected by the inoperable hose station".
For the purpose of this procedure change, gated wye (s) will be in-stalled at fire %ea. racks located as close as possible to the fire hose stations that will become INOPERABLE. One end of the gated wye will be connected to the standard length of hose provided for the station, and the other end of the wye connected to another fire hose of sufficient length to cover the area normally covered by the INOPER-ABLE fire hose.
The Fire Protection System is mentioned in the FSAR-in sections
_ 1.2.2.11 and 9.5.1. In section 9.5.1.1, it states that " Manual fire fighting equipment (i.e., fire hoses and portable fire extinguishers) 1, or activation of fixed water or gas extinguishing systems provide capability-for control of extinguishment in the-incipient stages of a fire, thereby minimizing potential consequences of fire in these areas." It also states that " Inadvertent operation of or a crack in a-fire suppression system would not preclude safe shutdown of the plant since redundant trains of equipment required for safe shutdown are located in separate rooms."
This procedure change does affect the RF system which is addressed in the FSAR in a significant manner; however, due to the use of gated wye (s) as described in Tech. Spec. section 3.7.10.4 and because of the redundant design of the system which provides three 100% capacity main fire pumps, this change will not compromise plant safety.
OP/1/A/6700/01 Change #164 OP/1/A/6700/01, Unit One Data Book, Tablo 2.2 is used to "ecord the 100% Full Power Calibration Currents (at Axial offsets o. +20%, 0%,
9 and -20%) and the M Pactors for each of :he Power Rang 6 Excore Detec-tors. Data is obtained for this table only by approved procedure, PT/1/A/4600/05A, Incore/Excore Calibration. The data recorded on this table is used by Instrument and Electrical (IAE) to adjust the Axial Flux Difference (AFD) Calculating circuitry and Operator Aid Computer (OAC) programs. It may also be used to manually calculate AFD and 3 19
.m _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
l l
l
)
I l
Quadrant Power Tilt Ratio (QPTR) if the OAC is inoperable. Since AFD is used to dynamically adjust b(th the Overtemperature Differential Temperature (OTDT) and Overpower Diffe.rential Temperature (OPDT).
setpoints, the data herein is safety related. Some.FSAR accidents in Chapter 15 depend upon these setpoints for their mitigation.
This change will not increase the probability or consequences-of any
! accident either analyzed or unanalyzed by the FSAR. No safety signif-icant equipment malfunction either analyzed or unanalyzed by the FSAR will be created. (No equipment other than the Power Range Nuclear l
Instrumentation System is effected by this change.) The margin of I safety defined in the bales of the affected Tech. Specs. will not be l reduced in any way by this change.
1 PT/0/A/4400/08 Change #41 The purpose of this change is to ensure that the handwheels of Nuclear Service Water (RN) valves 1(2)RN291 and 1(2)RN351 are not operated without the valve being in at least the desired position. There is some concern about galling the stem per PIR 0-C89-0092.
This change will not impact the RN System response to an accident.
The Component Cooling (KC) Heat Exchanger outlet valve will still fail open. Therefore, the probability of and consequences of an accident previously evaluated in the FSAR-will not be increased. This change in no way creates-the possibility-of an accident different from any already evaluated in the FSAR.
The probability of a malfunction of equipment important to safety l already evaluated in the FSAR is not increased, as are the consequenc-l es. New malfunction-possibilities cannot be created as a result of this change. The operation of the handwheel does not affect the actual valve movement. The full stroke of the valves is affected, but the purpose of this procedure is to set the handwheel to the proper full stroke position.
Since the safety function of the RN System is not affected by this change, and this change does not in itself make any part of the RN System inoperable, the margin of safety as defined in_the bases of Tech. Specs. is not reduced.
PT/0/A/4400/22B Change #23 l The purpose of this change is to: delete tLe requirement for the opposite train of the Control Area Ventilntion (VC) and Chilled Water (YC) Systems-to be in service and-to prov:de for a' minimum 1 flow path for the A Train RN Pumps, since both of their Component Cooling (KC)
Heat Exchangers will be in " Temp" mode and the train crossover valves will be closed. YC Chiller flow will not change significantly1over the course of the test and, therefore, will have negligible effect on the total pump test flow. Opening the Containment Spray (NS) Heat Exchanger (HX) isolation valves Will be controlled by the Control. Room 20
operators, so they will be cognizant at all times of the status of RN Pump Flow.
This change in no way increases the probability of an accident occur-ring. The B Train RN System will essentially be in the failed mode, since the KC HX outlet valves will be failed open. The opposite train of RN will be supplying cooling water to inservice components (except i
possibly the YC Chiller), so no interaction with the reactor or supporting systems is present. The consequences of an accident i
previously evaluated in the FSAR are not increased, since the opposite train of RN will be in service and supplying cooling water to required components, and the train under test is already essentially in the Engineered Safeguards mode.
This change actually dacreases the probability c> and consequences of a malfunction of safety related equipment by ensuring _that a minimum flow path is provided at all times for the A Train RN Pumps. No new equipment failures or malfunctions are created as a result of this change.
Neither train of RN will be inoperable at the tima of the test.
Technical Specification 3/4.7.4 allows two RN pumps to be out of l service for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with both units in Modes 1, 2, 3, and 4.
Therefore, the margin of safety as defined in the bases of Technical Specifications is not reduced.
TN/1/A/0942/00/01A Initial Issue Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on 1CA007A, 98, 11A, 1FWO27A, 55B, 1KC051A, 54B, i
1ND028A, 1NS001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B, 75A, 76B, l- 77A, ISV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete Motor Operated Valve (MOV) testing of the valves. The MOV testing information included' i' the NSM will supersede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and main-tained correctly to accommodate the maximum differential pressure-expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the-valve will operate during normal and abnormal events by setting limitations on Total Thrust, Differential Pressure (D/P) Thrust, and Packing Load.
This procedure will control work being performed on Auxiliary Feedwater (CA) valve 1CA007A. IAE will perform all work at the valve.
IAE will rewirr, the rotors, set up the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve. Valve ICA007A is the CA Pump No. 1 normal suction isolation valve. Prior to return-ing the valve to service, a functional verification and retest will be performed to verify valve operability.
21
This procedure will be implemented with Unit 1 in Mode 4, 5, 6, or No Mode when CA is not required operable.
No unreviewed safety question exists. Y TN/1/A/0942/00/02A Initial Issue Nuclear Station Modification CH-10942, Rev. O will modify the control circuit wiring on 1CA007A, 9B, 11A, 1FWO27A, 55B, JKC051A, 54B, 1ND028A, 1NS001B, 3B, 20A, 38B, 43A, 1RN250A, 3103, 1SM074B, 75A, 76B, 77A, 1SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of the valves. The MOV testing information included in the NSM will super-sede the old torque switch settjug values and replace them with thrust values. The purpose of the thr t Tlues is to ensure that tr.que switch settings are selected, s i d maintained correctly-to accom-modate the maximum differential ; .aure expected on the valve during both normal and abnerwul cy?nts within the design basis. The new s thrust values ensure the valve will operate during normal and abnormal ,
events by setting limitations on Total Thruet, D/P Thrust, and Packing Load.
This procedure will control work beitrg performed on valve 1CA009B.
IAE will perform all work at the valve. IAE will rewire the rotors, set up the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve. Valve 1CA009B is the CA Pump 1B normal suction isclation valve. Prior to returning the valve to service, a functional verification and retest will be performed to verify valve operability.
This procedure will be implemented with Unit 1 in Mode 4, 5, 6, or No Mode when CA is not required operable. No USQ is associated with performance of this procedure.
TN/1/A/0942/00/03A Initial Issue Nuclear Station Modification CN-10942, Rev. O'will modify the control circuit Wiring on 1CA007A, 9B, 11A, 1FWO27A, 55B, 1KC051A, 54B, 1ND028A, 1NS001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B, 75A, 76B, 77A, ISV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of the valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate-during normal and abnormal events by setting limitations on Total Thrust, D/P Thrust, and Packing Load.
22 I
)
i This procedure will control work being performed on valve 1CA011A.
IAE will perform all work at the valve. IAE will rewire the rotors, set up the switch rotors, verify add-on-pak switch setup, and_ perform l
MOV testing of the valve. Valve icA011A is the CA Pump 1A normal suction isolation valve. Prior to returning the valve to service, a functional verification and retest will be performed to verify valve operability.
This p ocedure will be implemented with Unit 1 in Mode 4, 5, 6, or No Mode shen CA is not required operable. An unreviewed safety question does not exist.
TN/1/A/0942/00/04A Initial Issue Nuclear Station Modification CN-10942, Rev. O will modify the control l
circuit wiring on 1CA007A, 9B, 11A, 1FWO27A, 55B, 1KC051A, 54B, l
1ND028A, 1NS001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B, 75A, 76B,
(
77A, ISV025B, 26B, 27A, and 28A to provide " limit actuated" torque l switch bypass contacts which can be adjusted independently of indica-
. tiens or interlocks and provide data to complete MOV testing of the l
valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace tnem with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve-during both normal and abnormal events within the design basis. The nsw-thrust values ensure the valve will operate during_ normal and abnormal events by setting limitations on Total Thrust, D/P Thrust, and Packing Load.
l This procedure will control work being performed on Refueling Water l (FW) valve 1FWO27A. IAE will perform all work at the valve. Con-l struction/ Maintenance Department (CMD) will perform work remote to the i valve required-to support the torque switch bypass modification. IAE will rewire the rotors, set up the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve. Performance will stroke time the valve.
This procedure may be implemented with Unit 1 in Mode-4, 5, or 6 as long as Train B Residual Heat Removal (ND) is operable. In No Mode, this procedure may be implemented at any time. Valve 1FWO27A is the ND Pump 1A suction from FW Storage Tank (ST). Prior.to returning the-valve to service, a functional verification and retest will be per- i formed to verify valve operability.
An unreviewed safety question does not exist.
TN/1/A/0942/00/05A Initial Issue Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on 1CA007A, 9B, 11A, 1FWO27A, 55B, 1KC051A, 54B, IND028A, 1NS001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B, 75A, 76B, 1
23 i ,
I
77A, 1SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testir " of the valves. The MOV testing information included in the NSM wi super-sede the old torque switch setting values and-replace them ni.h thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, D/P Thrust, and Packing Load.
This procedure will control work being performed on valve 1FWO55B.
IAE will perform all work at the valve. CMD will perform work remote to the valve required to support the torque switch bypass modifica-tion. IAE will rewire the rotors, set up the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve.
Performance will stroke time the valve.
This procedure will be implemented with Unit 1 in Mode 4, 5, or 6 as long as Train A ND 10 operable. In No Mode, this procedure may be implemented at any time. Valve 1FWO55B is the hD Pump 1B suction from FWST. Prior to returning the valve to service, a functional verifica-tion and retest will be performed to verify valve operability.
An unreviewed safety question does not exist.
TN/1/A/0942/00/06A Initial Issue Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on 1CA007A, 9B, 11A, 1FWO27A, 55B, 1KC051A, 54B, IND028A, 1NS001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B, 75A, 76B, 77A, 1SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of the valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that. torque switch settings are selected, set,c and maintained correctly to accom-modate the maximum differential pressure expected-on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, D/P Thrust, and Packing Load.
This procedure will control work being performed on Component Cooling (KC) valve 1KC051A. IAE will perform all work at the valve. IAE will rewire the rotors, set up the switch rotors, verify add-on-pak switch
- setup, and perform MOV testing of the valve.
This procedure will be implemented with Unit 1 in any Mode. In Modes 1, 2, S, and 4, this procedure may be implemented only when KC Train B i
. 24
I is operable (refer to Tech. Spec. 3/4.7.3). ~n Mode 5, this procedure may be implemented with the reactor coolant loops filled (Tech. Spec.
3.4.1.4.2). In M3de 6, this procedure may be implemented when the water level is 2 23 feet abm?: tr.: t:p S # the reactor vessel flange (Tech. Spec. 3.9.8.2). Valve 1KC051A is the Train 1A recirculation line isolation valve. Prior to returning the valve to service, a functional verification and retest will be performed to verify valve operability.
An unreviewed safety question does not exist.
TN/1/A/0942/00/07A Initial Issue Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on 1CA007A, 9B, 11A, 1FWO27A, 55B, 1XC051A, 54B, IND028A, INS 001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B, 75A, 76B, 77A, 1SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of the valves. The MOV testing information included in the NSM will-super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the-valve curing both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, D/P Thrust, and Packing Load.
This procedure will control work being performed on valve 1KC054B.
IAE will perform all work at the valve. IAE will rewire the rotors, set up the switch rotors, verify add-on-pak-switch setup, and perform MOV testing of the valve.
This procedure will be implemented with Unit 1 in any Mode. In Modes 1, 2, 3,_and 4, this procedure may be implemented only when KC_ Train A is operable (refer to Tech. Spec. 3/4.7.3). In Mode 5, this procedure may be implemented with the reactor coolant loops filled--(Tech. Spec.
3.4.1.4.2). In Mode 6, this procedure may be implemented when the water level is 2 23 feet above the top of the-reactor vessel flange (Tech. Spec. 3.9.8.2). Valve 1KC054B is the Train 1B recirculation line isolation valve. Prior to returning the valve to service, a functional verification and retest will be performed to verify valve operability.
An unreviewed safety question does not exist.
TN/1/A/0942/00/08A Initial Issue t
Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on 1CA007A, 9B, 11A, 1FWO27A, 55B, 1KC051A, 54B, 1ND028A, 1NS001B, 3B, 20A, 38B, 43A,-1RN250A, 3108, ISM 074B, 75A, 76B, 25 4
h
,'-_n. , -
77A, ISV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can he adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of the valves. The MOV testing information included in the HSM will super- !
sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ens'are the valve will operate during normal and abnormal events by setting limitations on Total Thrust, D/P Thrust, and Packing Load.
I This procedure will control work being performed on Residual Heat Removal (ND) valve IND028A. IAE will perform all work at the valve.
CMD will perform work remote to the valve required to suppo::t the torque switch bypass modification. IAE will rewire the rotors, set up the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve. Performance will stroke-time the valve.
This procedure will be implemented with Unit 1 in Modes 5, 6, or No Mode when this portion of the ND system is not required operable.
Valve IND028A is the ND Heat Exchanger 1A outlet to Chemical and Volume Control System (NV) pump suction. Prior to returning the valve to service, a functional verification and rotest will be performed to verify valve operability.
An unreviewed safety question does not exist.
TN/1/A/0942/00/09A Initial Issue Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on 1CA007A, 9B, 11A, 1FWO27A, 55B, 1KC051A, 54B, IND028A, INS 001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B,-75A, 76B, 77A, 1SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete ~MOV testing of the valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that1 torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on-the valve dur3ng both normal and abnormal events within the design basis. The new thrust values ensure the-valve will operate during normal and ainormal events by setting limitations on Total Thrust, D/P Thrust, and Packing 1 Load.
This procedure will control work being performed on Containment Spray (NS) valve INS 001B. IAE will perform all work at the valve. CMD will perform work remote to the valve required to support the torque switch bypass modification. IAE will rewire the rotors, set up.the switch rotors, verify add-on-pak switch setup, and perform Mov testing of the valve. Performance will stroke time the valve.
26 j
1 This procedure will be implemented with Unit 1 in Modes 5, 6, or No Mode when the NS system is not required operable. Valve INS 001B is the NS Pump 1B suction from containment sump. Prior to returning the valve to service, a functional verification and retest will be per-formed to verify valve operability.
An unreviewed safety question does not exist.
TN/1/A/0942/00/10A Initial Issue Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on 1CA007A, 9B, 11A, 1FWO27A, 558, 1KC051A, 54B, 1ND028A, INS 001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B, 75A, 76B, 77A, ISV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of the valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace-them with-thrust values. The purpose of the thrust values is to-ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, D/P Thrust, and Packing Load.
This procedure will control work being performed on valve INS 003B.
IAE will perform all work at the valve. CMD will perform work remote to the valve _ required to-support the torque switch bypass modifica-tion. IAE will rewire the rotors, set up--the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve.
Performance will stroke time the valve.
This procedure will be implemented with Unit 1 in Modes 5, 6, or No Mode when this portion of the ND system is not required operable.
Valve INS 003B is the NS Pump 1B suction from FWST isolation valve.
Prior to returning the valve to service, a functional verification and retest will be performed to verify valve operability.
An unreviewed safety question does not exist.
TN/1/A/0942/00/11A Initial Issue Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on 1CA007A, 9B, 11A, 1FWO27A, 55B, 1KC051A, 54B, 1ND028A, INS 001B, JB, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B, 75A, 76B, 77A, 1SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to coniplete MOV testing of the valves. The MOV testing-information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque 27
-i
)- .
switch settings are selected, set, and maintained. correctly to accom- .
modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, D/P Thrust, and Packing Load.
This procedure will control work being performed on valve 1NS020A. -
IAE will perform all-work at the valve. CMD will perform work remote to the valve required to support the torque switch bypass modifica-tion. IAE will rewire the. rotors, set up the switch rotors, verify l add-on-pak switch setup, and perform MOV-testing of the-valve.
Performance will stroke time the valve.
This procedure _will be implemented with Unit 1 in Modes 5, 6, or No Mode when the NS system is not required-operable. . Valve 1HS020A is i s
the NS Pump 1A suction from FWST_ isolation valve. - Prior to returning the valve to service, a functional verification and-retest will be performed to verify valve operability.
An unreviewed safety questior doec not exist.
TN/1/A/0942/00/12A Initial issue Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on 1CA007A, 9B, 11A, 1FWO27A, SSB, 1KC051A, 54B, 1ND028A, 1NS001B, 3B, 20A, 38B,-43A,.1RN250A, 310B,'1SM074B,-75A, 76B, 77A, ISV025B, 2GB, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which-can be adjusted independently-of indica-tions or interlocks and provide data to complete MOV testing of:the valves. The-MOV testing information included in:the NSM will super-sede theJold torque switch setting values and replace them with' thrust values. The purpose of the thrust values is to. ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the_ valve:during both normal and-abnormal events within the design basis. -The new thrust values ensure the valve will operate-during normal-and abnormal events by setting limitations on Total-Thrust,-D/P Thrust, and Packing =
Load.
This procedure will control work-being performed ~on valve INS 038B.
IAE will perform all work at the valve. CMD will perform work remote to the valve required to support the torque switch bypass modifica-tion. IAE will-rewire the rotors, set up the switch rotors,. verify =
add-on-pak switch setup, and perform MOV testing of the valve.
Performance will stroke time the valve.
This procedure will be implemented with Unit 1 in Modes 5, 6, or No Mode when the US system is not required operable. Valve 1NS038B is the ND Pump.1B-discharge to containment apray header-isolation-valve.
Prior to returning the valve to service, a functional verification and retest:will be performed to verify valve operability.
20
An unrevjewed safety question does not exist. I TN/1/A/0942/00/13A Initial Issue Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on ICA007A, 9B, 11A, 1FWO27A, 55B, 1KC051A, 54B, 1ND028A, 1NS001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B, 75A, 76B, 77A, 1SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of the valves. The MOV testing information included in the NSM will super-l sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve _during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, D/P Thrust, and Packing Load.
This procedure will control work being performed on valve INSO43A.
IAE will perform all work at the valve. CMD will perform work remote to the valve required to support the torque switch bypass modifica-tion. IAE will rewire the rotors, set up the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve.
Performance will stroke time the valve.
This procedure will be implemented with Unit 1 in Modes 5, 6, or No Mode when the NS system is not required operable. Valve 1NSO43A is the ND Pump 1A discharge to containment spray header isolation valve.
Prior to returning the valve to service, a functional verification and retest will be performed to verify valve operability.
An unreviewed safety question does-not exist.
l TN/1/A/0942/00/14A Initial Issue Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on 1CA007A, 9B, 11A, 1FWO27A, 55B, 1KC051A, 54B, 1ND028A, 1NS001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B, 75A, 76B, 77A, ISV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of.-indica-tions or interlocks and provide data to complete MOV testing of the
- valves. The MOV testing information included in the NSM--will super __
l sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values.is to ensure that torque switch settings are selected, set, and maintained correctly to accom-
- modate the maximum differential pressure expected on the valve during i both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal
- events by setting limitations on Total Thrust, D/P Thrust, and Packing Load.
29
1 l
l J l
This procedure will control work being performed on Nuclear Service Water (RN) valve 1RN250A. IAE will perform all work at the valve.
IAE will rewire the rotors, set up the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve. Performance will stroke time the valve.
i This procedure will be implemented with Unit 1 in Modes 4, 5, 6, or No Mode when this portion of the RN system is nce required operable.
I Valve 1RN250A is the RN header A to CA pump suction isolation. Prior to returning the valve to service, a functional verification _and retest will be performed to verify valve op ability.
An unreviewed safety question does not exist.
TN/1/A/0942/00/15A Initial Issue l
Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on 1CA007A, 9B, 11A, 1FWO27A, 55B, 1KC051A, 54B, 1ND028A, INS 001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B, 75A, 76B, 77A, ISV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-
! tions or interlocks and provide data to complete MOV testing of the valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that_ torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust,_D/P Thrust, and Packing Load.
This procedure will contro) work being performed on valve 1RN310B.
IAE will perform all work at the valve. IAE will rewire the rotors, set up the switch rotors, verify add-on-pak switch setup, and perform MOV testing of-the valve. Performance will stroke time the valve.
This procedure will be implemented with Unit 1 in Mode- 4, 5, 6, or No Mode when this portion of the RN system is not requh c. operable.
Valvo 1RN310B is the RN header B to CA pump suction isolation-valve.
Prior to returning the valve to service, a functional verification and retest will be performed to verify valve operability.
- _ An unreviewed safety question does not exist.
l TN/1/A/0942/00/16A-Initial Issue Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on 1CA007A, 9B, 11A, 1FWO27A, 55B, 1KC051A, 54B, 1ND028A, 1NS001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 3AM074B, 75A, 76B, 77A, 1SV025B, 26B, 27A, and 28A to provide " limit accuated" torque switen bypass contacts which can be adjusted independently of 30
j l
indications or interlocks and provide data to complete MOV testing of the valven. The MOV testing information included in the NSM will supersede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accommodate the maximum differential pressure expected on the valve during both normal and abnormal events withir the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, D/P Thrust, and Packing Load.
This procedure will control work being performed on Main Steam (SM) valve 1SM074B. IAE will perform all work at the valve. IAE will rewire the rotors, set up the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve. Performance will stroke l time the valve.
I This procedure will be implemented with Unit 1 in any Mode. Valve I 1SM074B is the Steam Generator (S/G) 1D outlet header blowdown control valve. Prior to returning the valve to service, a functional verifi-cation and retest will be performed to verify operability.
t An unreviewed safety question does not exist.
TN/1/A/0942/00/17A Initial Issue Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on 1CA007A, 98, 11A, 1FWO27A, 55B, 1KC051A, 54B, 1ND028A, INS 001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B, 75A, 76B, 77A, 1SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of the valves. The MOV testing informa'. ion included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will coerate during normal and abnormal events by setting limitations on Tota. Thrust, D/P Thrust, and Packing Load.
l This procedure will control work being performed on valve 1SM075A.
I IAE will perform all work at the valve. IAE will rewire the rotors, set up the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve, Performance will stroke time the valve.
This procedure will be implemented with Unit 1 in any Mode. Valve 1SM075A is the S/G 1C outlet header blowdown control valve. Prior to returning the valve to service, a functional verification and retest will be performed to verify valve operability.
An unreviewed safety question does not exist.
31
TN/1/A/0942/00/18A Initial Issue Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on 1CA007A, 98, 11A, 1FWO27A, SSB, 1KC051A, 54 B,-
1ND028A, 1NS001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B, 75A, 76B, 77A, 1SV025B, 26B, 27A, and 28A-to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of the valves. The MOV testing information included in the NSM will super-I sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-l modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, D/P Thrust,fand Packing Load.
This procedure will contro) work being performed on valve 1SM076B.
IAE will perform all work at the valve. IAE will rewire the rotors, set up the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve. Performance will stroke time the valve.
This procedure will be implemented with Unit 1 in any Mode. Valve 1SM076B is the S/G 1B out3et header blowdown control valve. Prior to returning the valve to service, a functional-verification and retest will be performed to verify operability.
An unreviewed safety question does not exist.
TN/1/A/0942/00/19A Initial Issue Nuclear Station Modification CN-10942, Rev. O will modify the control circuit Wiring-on 1CA007A, 9B, 11A, 1FWO27A, 55B, 1KC051A, 54B, IND028A, INS 001B, 3B, 20A, 38B, 43A, 1RN250A, 310B,.1SM074B, 75A, 76B, 77A, 1SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of the valves. The_MOV testing information included in the NSM will super-sede-the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque C switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during-normal and abnormal events by setting limitations on Total Thrust, D/P Thrust, and Packing Load.
This procedure will. control work being performed on valve 1SM077A.
IAE will perform all work at the valve. IAE will-rewire the rotors, set up the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve. Performance will stroke time the valve.
32
i This procedure will be implemented with Unit 1 in any Mode. Valve l 1SM077A is tne S/G 1A outlet header blowdown control valve. Prior to i returning the valve to service, a functional verification and retest I will be performed to verify operability.
An unreviewed safety question does not exist.
TN/1/A/0942/00/20A Initial Issue l Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on ICA007A, 9B,-11A, 1FWO27A, 55B, 1KC051A, 54B,.
IND028A, INS 001B, 3B, 20A, 38B, 43A,-1RN250A, 310B, 1SM074B, 75A, 76B, 77A, ISV025B, 26B, 27A, and 28A to-provide " limit actuated" torque switch bypass con acts which can be adjusted independently of indica-tions or interlocks and provide data to. complete MOV testing of the valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and_ replace them with-thruct values. The purpose of the-thrust values is to ensure that torque switch settings are selected, set, and-maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations.on Total-Thrust, D/P Thrust, and Packing Load.
This procedure will control work being performed on Main Steam Vent to Atmosphere (SV) valve 1SV025B. IAE will perform all work at the valve. IAE will rewire the rotors, set up-the switch rotors, verify
, add-on-pak switch setup, and perform MOV testing of the valve.
! Performance will stroke time the valve.
This procedure will be implemented with Unit 1 in any Mode. Valve 1SV025B is the Main Steam 1D power-operated relief isolation valve.
Prior to returning the valve to service, a functional verification and ,
retest will be performed to verify operability. <
An unreviewed safety question does not exist.
TN/1/A/0942/00/21A Initial Issue l
Nuclear Station Modification CN-10942, Rev. O will modify the control' circuit wiring on 1CA007A, 9B, 11A, 1FWO27A, 55B, 1KC051A, 54B, l 1ND028A, 1NS001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B, 75A, 76B, !
77A, ISV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica- l tions or interlocks and provide data to complete MOV testing of the i valves. The MOV testing information included in the NSM will super- l sede the old torque switch setting valuesJand replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained' correctly to accom-modate the maximum _ differential pressure expected on the valve during ,
both normal and abnormal events within the design basis. The new H l
33
thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, D/P Thrust, and Packing Load.
This procedure will control work being performed on valve 1SV026B.
IAE will perform-all work at the valve. IAE will rewire the rotors, set up the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve. Performance will stroke time the valve.
This procedure will be implemented with Unit 1 in any Mode. Valve 1SV026B is the Main Steam 1C power operated relief isolation valve.
Prior-to returning the valve to service, a functional verification and '
retest will be performed to verify valve operability.
An unreviewed safety question does not exist.
TN/1/A/0942/00/22A Initial Issue Nuclear Station Modification CN-10942, Rev. O will modify the control circuit wiring on 1CA007A, 9B, 11A, 1FWO27A, 55B, 1XC051A, 54B, IND028A, 1NS001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B, 75A, 76B, 77A, ISV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently.of indica-tions or interlocks and provide data to complete MOV testing of the-valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with-thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly"to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during-normal and abnormal events by setting limitations on Total Thrust, D/P Thrust, and Packing Load.
- This procedure will control work being performed on valve 1SV027A.
L IAE will perform all work at the valve. IAE will rewire the rotors, set up the switch rotors,_ verify add-on-pak switch setup, and perform Mov testing of the valve. Performance will stroke-time the valve.
This procedure will be implemented with Unit 1 in any Mode. Valve 1SV027A is the Main-Steam 1A power operatedirelief isolation valve.
l Prior to returning the valve to service, a functional verification and retest will be performed to' verify operability.
An unreviewed safety question does not exist.
TN/1/A/0942/00/23A. Initial Issue Nuclear: Station Modification CN-10942, Rev. O wil-1 modify thefcontrol circuit wiring on 1CA007A, 9B,-11A, 1FWO27A, 55B, 1KC051A, 54B, 1ND028A, 1NS001B, 3B, 20A, 38B, 43A, 1RN250A, 310B, 1SM074B, 75A, 76B, 77A, ISV025B, 26B, 27A, and 28A to provide " limit actuated" torque 34
switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of the valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, D/P Thrust, and Packing Load.
This procedure will control work being performed on valve 1SV028A.
IAE will perform all work at the valve. IAE will rewire the rotors, set up the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve. performhnce will stroke time the valve.
This procedure will be implemented with Unit 1 in any Mode. Valve 1SV028A is the Main Steam 1B power operated relief isolation valve.
Prior to returning the valve to service, a functional verification and retest will be performed to verify valve operability.
An unreviewed safety question does not exist.
PT/2/A/4400/06A Change #2 This change will allow the Refueling Water Storage Tank (FWST) temper-ature to be increased in order to provide a better heat transfer test on Containment Spray (NS) heat exchanger 2A by use of a new Standing Work Request, 10982 SWR. This modification has been performed many times in the past by use of a Temporary Station Modification. The FWST temperature will remain within the range of 70 to 100 dog F as allowed by Technical Specifications. Temperature stratification in the FWST will be avoided by operation of the recirculation pumps, i
PT/0/A/4400/22A Change #23 The purpose of this change is to delete the sign-offs for the "As Found" and "As Left" positions for valves 1RN291 and 2RN291 on Enclo-sure 13.9 of this procedure. Air is failed to these valves in the procedure in order to assure that they are wide open. These valves normally throttle flow, so when air is'later supplied to them, it is not certain that they will return to the initial position. This does not pose any safety problem, however. The KC heat exchanger outlet valves will be assured of returning to the pre-test conditions by l
having air re-supplied by steps in the procedure. The consequences of an accident previously evaluated in the FSAR are not increased, since the valves are in the accident positions while air is failed. Neither train of RN will be inoperable at the time of the test. Technical l Specification 3/4.7.4 allows two RN pumps to be out of service for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with both units in Modes 1, 2, 3, and 4.
35
TT/2/B/9100/52 Revision /0 There is currently a problem with obtaining adequate, consistent flow to Radiation Monitor (EMF) 2 EMF- 3 4 . The flow problem seems to be associated with the plating out of magnetite on the small bore piping and valves in the flow path to the EMF. Efforts have been made to clean the system, but only improve the situation temporarily. A more permanent solution to the flow problem is needed. Exempt Change CE-2505 will remove Nuclear Sampling (NM) valves 2NM-193, 203, and 222 and replace them with a straight run of pipe. Valves 2NM-246 and 291 will be replaced with Whitey brand ball valves. Also, valve 2NM-213 will be removed and replaced with a spool piece equipped with a Linear Kinetic Cell (LKC). This LKC will be used to test whether or not the LKC will be effective in preventing the plating out of particles on the piping and components in the sample line from Steam Generator 2C.
This procedure will be performed at the completion of the Sxempt Change and will be used to gather data required to determine the effectiveness of the modification.
FSAR Section 9.3.2.2.1, Nuclear Sampling System, describes the func-tion of EMF-34 as follows:
"Each line contains an air operated isolation valve which will automatically close on a signal from the radiation monitor."
Due to the flow problems described above, the EMF is currently is91at-ed, rendering this feature inoperable. During the test, only one Steam Generator Sample will be supplied to the EMF at a time. Th!s will not affect plant operations. Provided the test proves that the LKC and associated throttling system is adequate, all four sample lines will be equipped with an LKC and throttle valves. This will enable EMF-34 to be returned to service and restore this feature.
FSAR Section 10.4.8, Steam Generator Blowdown System, states that the system provides a continuouc sample for the measurement of radioactiv-ity of the blowdown liquid. FSAR Section 11.5.1.2.1.2, Steam Genera-ter Water Sample Monitor, describes the same actions. The system as described is currently inoperable, and this test will be used for data collection in order to define necessary actions for returning the system to operable status. EMF-34 is required to be operable per Tech. Spec. 3.3.3.10, Radioactive Liquid Effluent Monitoring Instru-mentation, in order to vent the Blowdown Tank to the atmosphere. Due to the inoperabilit) of EMF-34, the Blowdown Tank cannot be vented to the atmosphere. Data gathered during this procedure will determine if the installed-modification can correct all the flow problems associat-ed with the EMF and allcw the system to be returned to operable status.
The Blowdown Sampling System is designed to alert the operators of small Steam Generator tube leaks and to mitigate the consequences of a l major tube failure. The performance of this test will not increase l the probability or consequences of this type accident. The data I gathered will help restore the system to operable status so that the components can perform as designed.
i 36 l
No new accidents can be created by the performance of this test. Flow paths will be used as designed with only one Steam Generator being tested at any time. The piping and components used for the test are Non-Safety Related, Duke Class G. No interface with systems necessary for safety are involved. Therefore, the probability, consequences, or possibility of equipment failures important to safety will not be '
affected. The test has no affect on any safety margins as defined in the Tech. Specs. or the Bases to any Tech. Spec.
PT/2/A/4450/05A Re-type, Changes O to 18 Incorporated The use of " TEST" selector switches was doloted from all sections, and a jumper was placed (when applicable) to serve a function of the deleted switch. Deleting the above switches operates the system closer to, or in, the designed alignment (see below). When the same terminals at Solid State Protection System (SSPS) -Cabinet 2SSPSA are used in more than one section of this procedure, a jumper with a switch is installed, and the position of the switch is changed to initiate the appropriate action. Using the jumper with a switch reduces the number of times that jumpers are placed in 2SSPSA, so the possibility of a personnel error is reduced. Switches, fans, and dampers are verified once at the beginning of the procedure to be in the correct alignment and then at the end to ensure that they are left in the correct alignment. Required initial test position of.the switches, fans, and dampers are addressed in each section of the procedure. If another section is not to be performed immediately after the section being performed, the procedure requires that the selector switches for each fan are returned to the " Auto" position.
Problems were encountered with the breaker for Hydrogen Skimmer Fan (HSF) 2A tripping during August 1989. If HSF-2A is operated during the performance of this procedure, the procedure requires that the break:ar for HSF-2A is cycled to ensure that the trip latch lever has been fully reset per recommendation of the manufacturer. After the breaker is cycled, section 12.6 verifies that power has been returned to HSF-2A by the completion of Enclosure 13.1.-
Section 12.1 of this procedure-verifies: the auto-start of,HSF-2A-and Air Return Fan (ARF) 2A on a Containment Isolation Phase B (Sp) test signal after the required time delay, the 15 minute run time for each fan, the running speed and current for each fan, and the Containment Pressure Control Circuitry (CPCC) permissives to start and stop ARF-2A. Two jumpers are placed in 2EATC4: one to open the bypass dampers when ARF-2A starts, and one to simulate Containment Air Return and Hydrogen Skimmer System (VX) valve 2VX1A open and allow HSF-2A to start. -Power is removed from the Air Return Fan Discharge Damper (ARF-D-2) to prevent an inadvertent opening of the. ice condenser doors during the operation of ARF-2A. Containment Spray Pressure Transmit-ter 2NSPT5160 is placed in the test position, and the test pot is adjusted inside 2CPCC1 to simulate a > 0.4 psid containment pressure which gives one start permissive for ARF-2A. A jumper in 2SSPSA initiates the Sp signal (timer 2VXTD1 starts), which gives the start signal to both fans after the timer expires. After both-fans start, the required alignment of associated dampers is verified and then the 37 j
speed and current for each fan is measured. In the event of a LOCA during the operation of the fans in section 12.1, ARF-D-2 would not open and violate the 9 minute time delay for operation of ARF-2A as assumed in the Design Basis Accident (DBA), but 2VX1A would open after the 9 minute delay so that HSF-2A would operato as assumed in the DBA.
The Sp test signal is removed before HSF-2A and ARF-2A are shut down to verify fan operation after the Sp signal is removed (as designed).
After the 15 minute run time is met, 2NSPT5160 is returned to normal, and ARF-2A is verified to stop. Then the selector switch for HSF-2A is placed in the "Off" position, and HSF-2A is verified to stop.
Applicable jumpers are then removed, and power is returned to ARF-D-2.
Section 12.6 verifies that power has been returned to ARF-D-2 by the completion of Enclosure 13.1. Except for the closing of 2 ARF-D-2 and the simulation of an open 2VX1A, the above alignment operates HSF-2A and ARF-2A as designed using the proper safety _ circuitry.
Section 12.2 of this procedure verifies the auto-open of the hir Return Fan discharge damper (ARF-D-2) on in Sp test signal after the required time delay and the CPCC permissives to prevent / enable the opening of ARF-D-2. After verifying that the CPCC permissive is not initiated, a jumper is placed in 2EATC4 to simulate an Sp signal for ARF-D-2, and then ARF-D-2 is verified to be closed. The jumper in 2EATC4 is removed, then 2NSPT5170 is placed in the test position, and the test pot is adjusted inside 2CPCC1 to simulate-a > 0.4 psid containment pressure which gives one open permissive for ARF-D-2. A jumper in 2SSPSA initiates the Sp signal (timer 2VXTD31 starts), which gives the open signal to ARF-D-2 after the timer expires. Since there is a possibility that 2VX1A could open if the Sp signal is left in for 9 minutes, a " CAUTION" statement is included to ensure that the Sp test signal is removed before 2VX1A is given the signal to open. The procedure requires that 2 ARF-D-2 is closed immediately after it opens in order to reduce the possibility of Ice condenser bypass leakage in the event of a LOCA during the test. The ARF-2A and HSF-2A are blocked from starting by placing their selector switches in the "OFF" position. This is to prevent bypass leakage in the event of a LOCA and to ensure that the Ice Condenser doors are not opened during the test. The above alignment opens ARF-D-2 as designed using the proper safety circuitry.
Section 12.3 of this procedure verifies the auto-open of the Hydrogen Skimmer Fan (HSF-2A) suction valve (2VX1A) on an Sp test signal, followed by the start of HSF-2A after the required time delay. The design of the system requires that 2VX1A open or begin to open before HSF-2A will start. There are two 9 minute timers associated with the start of HSF-2A on an Sp signal. One is 2VXTD1.which is the timer that gives a start permissive signal to HSF-2A and ARF-2A, and one is 2VXTD11 which gives the open signal to 2VX1A. .This section verifies the correct time for 2VXTD11 (section 12.1 or 12.4 for 2VXTD1). A jumper in 2SSPSA initiates an Sp signal for HSF-2A (timer 2VXTD1 starts), and then at least three minutes later another jumper in 2SSPSA initiates an Sp signal to open 2VX1A (timer 2VXTD11 starts).
This alignment ensures that the valve opening after 2VXTD11 expires is what starts HSF-2A. The procedure requires that HSF-2A is shut down
- immediately after it starts to limit time that HSF-2A operates with l
l l 38 l
l - . _ _ .
l 2VX1A open. A CAUTION statement is included to ensure that HSF-2A is not operated more than 5 minutes with 2VX1A open. ARF-2A is blocked from operation by placing the selector switch in the "OFF" position to ensure that it will not start in the event of a LOCA during this section and violate the 9 minute time delay. The above alignment operates HSF-2A and 2VX1A as designed using the safety circuitry.
The performance of sections 12.2 and 12.3 of this procedure put the VX system in a configuration which deviates from the assumed initial conditions for a Designed Basis Accident (DBA), because the deck leakage area is increased by opening ARF-D-2 and air is moved from
, Lower to Upper containment by running HSF-2A with 2VX1A open. In both of these alignments, a condition exists for Ice condenser bypass leakage during a LOCA. The open damper (ARF-D-2) is addressed in PIR 2-C87-0061, and the conclurion of the PIR was that the effect of the additional steam bypass area on initial compression peak pressure was small. Both sections are required by Tech. Spec. 4.6.5.6 to be d performed. Tech. Spec. 3.6.5.5 allows an equipment hatch to be open for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> which also deviates (in the same manner as above) from the initial conditions for a DBA. Caution statements and notes are included in each section to minimize the time that the system is in an alignment as mentioned above. Since the test duration for both sections 12.2 and 12.3 is on the order of a few minutes, not hours, the margin of safety as defined in Tech. Spec. Will not be reduced.
Section 12.4 of this procedure verifies: the auto-start of the Hydro-gen Skimmer Fan (HSF-2A) on an Sp test signal after the required time delay, the 15 minute run time, and the fan running speed and_ currents with 2VX1A closed. HSF-2A is operated in the same alignment as in section 12.1, using the appropriate jumper in 2EATC4 to simulate 2VX1A open, and the same jumper in 2SSPSA, except ARF-2A is blocked from operating by placing the selector switch in the "Off" position and by not initiating the CPCC permissive. In the event of a LOCA during performance of this section, ARF-2A would not start and violate the 9 minute time delay. This section is included for retest purposes but may be performed instead of section 12.1 if desired.
Section 12.5 of this procedure verifies: the auto-start of the Air Return Fan (ARF-2A) on an Sp test signal after the required time delay, CPCC permissives-to start and stop the fan, the 15 minute run time, and the fan running speed and current with the discharge damper (ARF-D-2) closed. ARF-2A is operated in the same alignment as in section 12.1, using the appropriate jumper to open the bypass dampers when the fan starts and the same jumper in 2SSPSA. In the event of a LOCA during the performance of this section, HSF-2A would operate as designed because 2VX1A would not open and start HSF-2A before the required time delay. This section is also included for retest purpos-es but may be performed instead of section 12.1 if desired.
The procedtre requires that Air Return Fan 2A (ARF-2A) and Hydrogen Skimmer Fan 2A (HSF-2A) are declared inoperable durir; the performance of this test. ARF-2B and HSF-2B will remain operable for the duration of the test as required by Tech. Specs. The system is not placed in any unusual alignments that would increase the probability of an
\
o 39 l
i accident OR create the possibility of an accident different than already evaluated. The margin of safety as defined in the bases of Tech. Specs will not be reduced. Placement / removal of jumpers, opening / closing of breakers and returning switches, fans and dampers to the required positions are independently verified within section 12.0 or 13.0 of the procedure. Power is verified to be returned to equipment that was deenergized by the procedure on Enclosu' a 13.1.
For these reasons and the ones stated above, this procedure does not increase the probability of OR the consequences of a malfunction of equipment important to safety. The possibility of a malfunction of equipment different than already evaluated will not be created and the consequences of an accident will not be increased by this procedure.
Therefore an unreviewed safety question does not exist.
PT/1/A/4450/05A Re-type, Changes 0 to 20 Incorporated The use of " TEST" selector switches was deleted from all sections, and a jumper was placed (when applicable) to serve a function of the deleted switch. Deleting the above switches operates the system closer to or in the designeo alignment (see below). When the same terminals at 1SSPSA are used in more than one section of this proce-dure, a jumper with a switch is installed and the position of the switch is changed to initiate the appropriate action. Using the jumper with a switch reduces the number of times that jumpers are placed in 1SSPSA, so the possibility of a personnel error is reduced.
Switches, fans, and dampers are verified once at the beginning of the procedure to be in the correct alignment and then at the end to ensure that they are left in the correct alignment. Required initial test position of the switches, fans, and dampers are addressed in each section of the procedure. If another section is not to be performed immediately after the section being performed, the procedure requires that the selector switches for each fan are returned to the " Auto" position. Problems were encountered with the breaker for HSF-2A tripping during August 1989. If HSF-1A is operated during the perfor-mance of this procedure, the procedure requires that the breaker for HSF-1A is cycled to ensure that the trip latch lever has been fully reset per recommendation of the manufacturer. After the breaker is cycled, section 12.6 verifies that power has been returned to HSF-1A by the completion of Enclosure 13.1.
Section 12.1 of this procedure verifies: the auto-start of HSF-1A and ARF-1A on an Sp test signal after the required time delay, the 15 minute run time for each fan, the running speed and current for eacn fan, and the CPCC permissives to start and stop the ARF-1A. Two jumpers are placed in 1EATC4: one to open the bypass dampers when ARF-1A starts, and one to simulate IVX1A open and allow HSF-1A to start. Power is removed from ARF-D-2 to prevent an inadvertent opening of the ice condenser doors during the operation of ARF-1A.
1NSPT5160 is placed in the test position, and the test pot is adjusted inside 1CPCC1 to simulate a > 0.4 psid containment pressure which gives one start permissive for ARF-1A. A jumper in 1SSPSA initiates the Sp signal (timer IVXTD1 starts), which gives the start signal to both fans after the timer expires. After both fans start, the 40 l
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required alignment of associated dampers is verified and then the speed and current for each fan is measured. In the event of a LOCA during the operation of the fans in section 12.1, ARF-D-2 would not i open and violate the 9 minute time delay for operation of ARF-1A as j assumed in the Design Basis Accident (DBA), but IVX1A would open after the 9 minute delay so that HSF-1A would operate as assumed in the DBA.
The Sp test ignal is removed before HSF-1A and ARF-1A are shut down to verify fan operation after the Sp signal is removed (as designed).
After the 15 minute run time is met, 1NSPT5160 is returned to normal and ARF-1A is verified to stop. Then, the selector switch for HSF-1A is placed in the "Off" position, and HSF-1A is verified to stop.
Applicable jumpers are then removed, and power is returned to ARF-D-2.
Section 12.6 verifies that power has been returned to ARF-D-2 by the completion of Enclosure 13.1. Except for the closing of ARF-D-2 and the simulation of an open IVX1A, the above alignment operates HSF-1A and ARF-1A as designed using the proper safety circuitry.
Section 12.2 of this procedure verifies the auto-open of the Air Return Fan discharge damper (ARF-D-2) on an Sp test signal after the required time delay and the CPCC permissives to prevent / enable the opening of ARF-D-2. After verifying that the CPCC permissive is not <
initiated, a jumper is placed in 1EATC4 to simulate an Sp signal fcr ARF-D-2, and then ARF-D-2 is verified to be closed. The jumper in 1EATC4 is recoved, then 1NSPT5170 is placed in the test position, and the test pot is adjusted inside 1CPCC1 to simulate a > 0.4 psid containment pressure which gives one open permissive for ARF-D-2. A jumper in ISSPSA initiates the Sp sigral (timer IVXTD31 starts), which gives the open signal to ARF-D-2 after the timer expires. Since there is a possibility that 1VX1A could open if the Sp signal is left in for 9 minutes, a " CAUTION" statement is included to ensure that the Sp test signal is removed before IVX1A is given the signal to open. The procedure requires that 1 ARF-D-2 is closed immediately after it opens in order to reduce the possibility of Ice Condenser bypass leakage in the event of a LOCA during the test. ARF-1A and HSF-1A are blocked from starting by placing their selector switches in the "OFF" posi-tion. This is to prevent bypass leakage in the event of a LOCA and to ensure that the Ice condenser doors are not opened during the test.
The above alignment opens ARF-D-2 as designed using the proper safety circuitry.
Section 12.3 of this procedure verifies the auto-open of the Hydrogen Skimmer Fan (HSF-1A) suction valve (1VX1A) on an Sp test signal, followed by the start of HOF-1A after the required time delay. The design of the system requires that IVX1A open or begin to open before HSF-1A will start. There are two 9 minute timers associated with the start of HSF-1A on an Sp signal. One is IVXTD1, which is the timer that gives a start permissive signal to HSF-1A and ARF-1A, and one is IVXTD11 which gives the open signal to IVX1A. This section verifies the correct time for IVXTD11 (section 12.1 or 12.4 for IVXTD1). A jumper in 1SSPSA initiates an Sp signal for HSF-1A (timer IVXTD1 starts), and then at least three minutes later another jumper in 1SSPSA initiates an Sp signal to open IVX1A (timer IVXTD11 starts).
l This alignment ensures that the valve opening (after IVXTD11 expires) is what starts HSF-1A. The procedure requires that HSF-1A be shut 41 l
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1 down immediately after it starts to limit time that HSF-1A operates with 1VX1A open. A CAUTION statement is included to ensure that HSF-1A is not operated more than 5 minutes with IVX1A open. ARF-1A is blocked from operation by placing the selector switch in the "Off" position to ensure that it will not start in the event of a LOCA during this section and violate the 9 minute time delay. The above alignment operates HSF-1A and IVX1A as designed using the safety circuitry.
The performance of sections 12.2 and 12.3 of this procedure put the VX system in a configuration which deviates from the assumed initial J
conditions for a Design Basis Accident (DBA), because the deck leakage l area is increased by opening ARF-D-2 and air is moved from Lower to l Upper containment by running HSF-1A with IVX1A open. In both of these l alignments, a condition exists for Ice condenser bypass leakage during a LOCA. The open damper (ARF-D-2) is addressed in PIR 2-C87-0061, and the conclusion of the PIR was that the effect of the additional steam bypass area on initial compression peak pressure was small. Both sections are required by Tech. Spec. 4.6.5.6 to be performed. Tech.
Spec. 3.6.5.5 allows an equipment hatch to be open for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> which also deviates (in the same manner as above) from the initial condi-tions for a DBA. Caution statements and notes are included in each section to minimize the time that the system is in an alignment as l mentioned above. Since the test duration for both sections 12.2 and 12.3 is on the order of a few minutes, not hours, the margin of safety as defined in Tech. Specs. Will not be reduced.
Section 12.4 of this procedure verifies: the auto-start of the Hydro-gen Skimmer Fan (HSF-1A) on an Sp test signal after the required time delay, the 15 minute run time, and the fan running speed and currents with IVX1A closed. HSF-1A is operated in the same alignment as mentioned above. Since the test duration for both sections 12.2 and 12.3 is on the order of a few minutes, not nours, the margin of safety as defined in Tech. Specs, will not be reduced.
Section 12.4 of this procedure verifies: the auto-start of the Hydro-gen Skimmer Fan (HSF-1A) on an Sp test signal after the required time delay, the 15 minute run time, and the fan running speed and currents with IVX1A closed. HSF-1A is operated in the same alignment as in section 12.1, using the appropriate jumper in 1EATC4 to simulate IVX1A open and the same jumper in 1SSPSA, except ARF-1A is blocked from operating by placing the selector switch in the "Off" position and by not initiating the CPCC permissive. In the event of a LOCA during performance of this section, ARF-1A would not start and violate the 9 minute time delay. This section is included for retest purposes but may be performed instead of section 12.1 if desired.
Section 12.5 of this procedure verifies: the auto-start of the Air Return Fan (ARF-1A) on an Sp test signal after the required time delay, CPCC permissives to start and stop the fan, the 15 minute run time, and the fan running speed and current with the discharge damper (ARF-D-2) closed. ARF-1A is operated in the same alignment as in section 12.1 using the appropriate jumper to open the bypass dampers when the fan starts and the same jumper in 1SSPC . In the event of a 42
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)
LOCA during the performance of.this section,(ISF-1A would operate as designed because IVX1A would not open and start HSF-1A before the reqaired time delay. This section is also included for retest purpos-es but may be performed instead of section 12.1 if desired.
The procedure requires that Air Return Fan 1A (ARF-1A) and Hydrogen Skimmer Fan 1A (HSF-1A) are declared inoperable during the-performance of this test. ARF-1B and HSF-1B will remain operable for the duration of the test as required by Tech. Specs. The system is not placed'in L any unusual alignments that would increase the probability of an accident OR create the possibility of an accident different than already evaluated. The margin of safety _as defined in the-bases of Tech. Specs. will not be reduced. _ Placement / removal of jumpers, opening / closing of breakers and returning switches, fans and dampers to the required positions are independently verified within section 12.0 or 13.0 of the procedure. Power is verified to be returned to '
equipment that was deenergized by the procedure on Enclosure 13.1.
For these reasons and the ones stated above, this procedure-does not increase the probability of OR the consequences of a-malfunction of equipment.important to safety. The possibility of a malfunction of equipment different than already evaluated will not be created and the consequences of an accident will not be increased by this procedure.
Therefore an unreviewed safety question does not exist.
PT/1/A/4200/01E Change #30 Detailed steps are being added to disconnect-the test equipment upon completion of the door _ seal test (i.e., the small_ seal and large seal tests on both Containment door and Auxiliary door.) The procedure presently instructs the technician to simply disconnect the test equipment upon test completion.
-Disconnecting the test equipment without first closing the seal manual isolation valve would allow the air tank associated with that seal to deferessurize. To-prevent-this, the technician may close the seal-manual isolation valve before disconnecting the test equipment.
Including specific steps (which are independently verified) for the closing and subsequent reopening of this manual valve will ensure that the valve is not inadvertently left closed. Also, visually verifying that the seal is reinflated upon opening the manual valve will provide assurance that the valve plug is not stuck on its seat.
Steps are being-added--to verify that the " Seal Inflated" lights are on prior to the technician leaving the area. This will ensure that the airlock doors are sealed closed upon test completion. There is no USQ created by these changes.
PT/1/A/4200/01F Change #32 Detailed steps are being added to disconnect the test equipment upon completion of the door seal test (i.e., the small seal and large seal tests on both containment door and Auxiliary door.) The procedure 43
_ _ _ __j
l
? l presently instructs the technician to simply disconnect the test l I
equipment upon test completion.
Disconnecting the test equipment without first closing the seal manual isolation valve would allow the air tank associated with that seal to i depressurite. To prevent this, the technician may close the seal
( manual isolation valve before disconnecting the test equipment.
I Including specific stops (which are independently verified) for the closing and subsequent .copening of this manual valve will ensure thst the valve is not ina6;ertently left closed. Also, visually verifying that the seal i; rainflated upon opening the manual valve will provide assurat.ca Laat the valve plug'is not stuck on its seat.
staps are being added to verify that the " Seal Inflated" lights are on prior to the technician leaving the area. This will ensure that the dirlock doors are sealed Closed upon test Completion. There is no USQ
! created by these changes.
PT/1/A/4200/07B Change #33 ,
This restricted change was written to perform a limited flow head I
curve test on Centrifugal Charging Pump la by modifying the IWP test to determine new pump rotor operability. The suction source will remain no" mal letdown, and the discharge flow poth will remain normal chargin v. Since this pump will essentially be operating in recircula-tion through the reactor coolant system (i.e., the pump will not be adding water from the FWST), shut 6own margin and Low Temperature Overpressurization (LTOP) concerns are precluded. Additionally, one
- Pressurizer Power Operated Relief Valve (PORV) is adequate to maintain LTOP protection in the event of a charging pump starting with the Reactor Coolant system water solid. Residual heat removal will not be affec*ed and the minimum number of injection flow paths will be maintalaed.
l l PT/1/A/4400/06A Change #7 The purpose of this change is to control changing the actpoints of the Refueling Water Storage Tank (FWST) heaters for the Containment Spray Heat Exchanger 1A capacity testing using this procedure and a Standing Work Request. This is essentially a Temporary Station Modification (TSM) that has been performed many times in the past. This change will streamline. the process without having to go through the TSM paperwork every time. - The FWST temperature will remain within the l range of 70 to 100 deg F as allowed by Technical Specifications.
l Temperature stratification in the FWST will be avoided by operating the recirculation pumps while the TSM is installed.
l PT/1/A/4400/06B Change #4 The purpose of this change is to control changing the setpoints of the FWST heaters for NS Heat Exchanger 1B capacity testing using this procedure and a Standing Work Request. This is essentially a 44
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Temporary Station Modification (TSM) that has been_ performed many times in the past. This change will streamline the process without having to go through the TSM paperwork every time. The FWST tempera-ture will remain within the range of 70 to 1C0 deg F as allowed by Technical Specifications. Temperature stratification in the FWST will be avoided by operating the recirculation pumps whi3e the TSM is installed.
PT/2/A/4400/06B Change #1 The purpose of this change is to control changing the Tetpoints of the FWST heaters for NS Heat Exchanger 2B-capacity testing using this procedure and a Standing Work Request. This is essentially a Tempo-rary Station Modification (TSM) that has been performed many times in the past. This change will streamline the process without having to go through tha TSM paperwork every time. The FWST temperature will remain within the range of 70 to 100 deg F as allowed by Technical Specifications. Temperature stratification in the FWST will be avoided by operating the recirculation pur.ps while the TSM is in-stalled.
EP/1/A/5000/2E3 Deletion EP/1/A/5000/2E3, Containment High Radiation Lavel, is being deleted as a result of the deletion of Emergency Procedure Guideline (EPG)
FR-Z.3, Response to High Containment Radiation. The deletion of this EPG is part of Duke Power's commitment to satisfy the requirements specified in Supplement 1 of NUREG-0737. This commitment was made in a letter from H. B. Tucker to the NRC dated February 18, 1988.
Monitoring of Containment radiation by the " CONTAINMENT" Critical-Safety Function (CSF) Status Tree is being deleted. This input to the
" CONTAINMENT" CSF was deleted under Safety Parameter Display System (SPDS) modification #3 (Program Request P880020). Under this SPDS modification, the Containment radiation information is_being trans-ferred to a new SPDS status block which is labelled " RADIATION." This new block monitors radiation levels at the following points:
a) Unit Vent b) Main Steam Line c) Condenser Steam Air Ejectors (CSAL) d) Steam Generator (S/G) Water Sample e) Containment 1 Incorporating this change will reflect the as-built condition of the SPDS and the current guidance given in the Emergency Procedure Guide- I i lines.
1 EP/2/A/5000/2E3 Deletion 45 l
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s EP/2/A/5000/2E3, Containment liigh Radiation Lovel, is being deleted as a result of the deletion of Emergency Procedure Guideline FR-Z.3, e Responso to High Containment Radiation. The deletion of this EPG is r*rt of Duke Power's commitment to satisfy the requirements specified in N 1pplement 1 of NUREG-0737. This commitment was made in n letter f re t B. Tucker to the NRC dated February 18, 1988.
Monitoring of Ceatainment radiction by the " CONTAINMENT" Critical Safety Function Status Tree is being deleted. This input to the
" CONTAINMENT" CSF was deleted under SPDS W. Cfication #3 (Program Request P880020). Under this SPDS modiff as'.on, the Containment radiation information is being transferrett to a new SPDS status block which is labelled " RADIATION." This now k' 4 monitors radiation 14vels at the following points:
a) Unit Vent c
b) Main Steam Lino c) CSAE d) S/G Watcr Samplo e) Containment incorporating thia change will reflect the as-built condition of the GFDS and the current guidance given in the Emergency Procedure Guide-lines.
OP/1/h/C700/01 Change #165 This change replaces page 1 of Table 2.2 with a new page. This incorporates new Power Range Nuclenr Instrumentation System (NIS) talibration data obtained per PT/1/A/4600/05D, Interim Incore/Excore Calibration. Table 2.2 is used to record the 100% Full Power Calibra-tion Currents (at Axial Offsets of +20%, 0%, and -20%) and the M factors for each of the Power Range Excore Detectors. Data is ob-tained for this tabic only by approved proceduro PT/1/A/4600/05A, Incore/Excore Calibration, or PT/1/A/4600/05D, Interiw Incore/Excore Calibration. The data recorded on this table is used by IAE to adjust the AFD calculating circuitry and OAC programs. It may also be used to manually calculate AFD and QPTR if the OAC is inoperable. Since
/
AFD is used to dynamically adjust both the OTDT and OPDT setpoints, the data herein is safety related. A number of FSAR accidents depend upon these setpoints for their mitigation. No equipment other than the Power Range NIS is affected by this change. The margin of safety will not be reduced.
OF/1/A/6700/01 Change #166 x
This change replaces page 1 of Table 2.2 with a new page. This incorporates new Power Range NIS Calibration data obtained per PT/1/A/4600/05A, Incore/Excore Calibration. Table 2.2 is used to record the 100% Full Power Calibration Currents (at Axial Offsets of
+20%, 0%, and -20%) and the M factors for each of the Power Range Excore Detectt;rs. Data is obtained for this table only by approved 46 l
v
u i
procedure PT/1/A/4600/05A, Incore/Excore Calibration, or PT/1/A/4600/05D, Interim Incore/Excore f.'ullbration. The data recorded on this table is used by IAE to adjust the AFD calculating circuitry and OAC programs. It may also be used to manually calculate AFD and QPTR if the OAC is inoperable. Since AFD is used to dynamically adjust both the OTDT and OPDT setpoints, the data herein is safety related. A nuniber of FSAR accidents depend upon these setpoints for their mitigation. The change of the data in this table only reflects what is currently used in the NIS and serves as a reference that the instrumentation 16 properly calibrated. No equipment other than the Power Range NIS is affected of this change. The margin of safety will
, not be reduced.
PT/2/A/4200/01E Ch*1ge #12 Detailed steps are oeing added to disconnect the test equipment upon completion of the door seal test (i.e., the small seal and large seal tests on both containment door and Auxiliary door.) The procedure presently instructs the technician to simply disconnect the test equipment upon test completion.
Disconnecting the test equipment without first closing the seal manual isolation valve would allow the air tank associated with that seal to depressurize. To prevent this, the technician may close the seal manual isolation valve before disconnecting the test equipment.
Including specific steps (which are independently verified) for the closing and :ubsequent reopening of this manual valve will ensure that the valve is not inadvertently left closed. Also, visually verifying that the seal is reinflated upon opening the manual valve will provide assurance that the valve plug is not stuck on its seat.
Steps are being added to verify that the " Seal Inflated" lights are on prior to the technician leaving the area. This will ensure that the airlock doors are sealed closed upon test completion. There is no USQ created by these changes.
PT/2/A/4200/01F Change #11 Detailed steps are being added to disconnect-the test equipment upon l
completion of the door seal test (i.e., the small seal and large seal tests on both containment door and Auxiliary door.)- The procedure presently instructs the technician to simply disconnect the test equipment upon test complot!on.
Disconnecting the test equipment without first closing the seal manual-isolation valve would allow the air tank associated with that seal to I
depressurize. To prevent this, the technician may close the seal manual !0olation valve before disconnecting the test equipment.
Including specific steps (which are indapendently verified) for-the closing and subsequent reopening of thid manual valve will ensure that the valve is not inadvertently left closed. Also, visually verifying thnt the seal is reinflated upon opening the manual valve will provide assurance that the valve plug is not stuck on its-seat.
47
,, _. _ -_ _ , _ _ _ _ _ ~ . _ , ,
Steps are being added to verify that the " Seal Inflated" lights are on prior to the technician leaving the area. This will ensure that the airlock doors are sealed closed upon test completion. There is no USQ created by these changes.
PT/1/A/4200/41B Change /19 This procedure change will allow leak rate testing of containment Air Release and Addition (VQ) valve 1VQ16A from outside containment, using 10978 SWR. The Standing Work Request will install a plate in plac3 of a flow rescri: ting orifice to provide a pressure boundary for the leak rate test. The plate will block the recirculation flow path from the discharge of the fans through IVQ15D and IVQ16A. However, this flow path is not used in either the air release mode or air addition mode.
Therefore, the VQ system will still be functional while this blank orifice plate is installed.
Furthermore, the only safety function of the VQ system is containment isolation (1VQ15B and IVQ16A both receive a Phase "A" Containment Isolation signal), which is not affected by this test. Therefore, an unreviewed safety question is not created by this change.
EP/1/A/5000/02 Retype #5 The following changes are included in Retype #5
- 1) Deleted EP/1/A/5000/203, Containment High Radiation Level, and its associated logic from the " CONTAINMENT" Critical Safety Function Status Tree. This Emergcncy Procedure (EP) is being deleted as a result of the deletion of Emergency Procedure Guideline (EPG) FR-Z.3, Response To High Containment Radiation.
The deletion of this EPG is part of Duke Power's commitment to satisfy the requirements specified in Supplement 1 of NUREG-0737.
This commitment was made in a letter from H.B. Tucker to the NRC dated February 18, 1988. The deletion of the Containment Radia-tion input to the Safety Parameter Display System (SPDS) was made under SPDS modification SPD #3 (Program Request P880020).
- 2) Deleted EP/1/A/5000/2E2, High Containment Sump Levo?. and its associated logic from the " CONTAINMENT" Critical Safety Function Status Tree and substituted in its place EP/1/A/5000/2E2, Incom-plete Containment Isolation. This change is in response to the deletion of Emergency Procedure Guideline FR-Z.2, Response to Containment Flooding, and the creation of Emergency Procedure Guideline FR-3.2, Response To Incomplete Containment Isolation.
These changes are also being made to satisfy the requirements in Supplement 1 of NUREG-0737 as mentioned above. The deletion of the Containment Sump Level input to SPDS and the inclusion of inputs for containment Isolation were made under SPDS modifica-tion SPD #3 (Program Request P880020).
48
a v n. & -1%.,22 4 ,a.a.e. e.J--
u,,,wua, w --- a --
w-. -- s-- A +-+ - - 2 - -m-- -
-n , ,
1 I
I
- 3) Modified the symptoms in the procedure to adequately describe its proper implementation. Implementation of this procedure is controlled by EP/1/A/5000/01, Reactor Trip Or Safety Injection; EP/1/A/5000/03, Loss Of All AC Power; and EP/1/A/5000/lA, Reactor '
Trip Response. This change does not alter the intent of this procedure, but will provide the correct symptoms as reflected in >
the guidance given in the other EPs.
- 4) Added Display Group Computer Program information to step 1.a. ,
This information was added under SPDS modification SPD #3. These display groups will give the operator up-to-date status informa-tion on the inputs to the Status Trees.
- 5) Corrected various format errors to ensure compliance with the EP Writer's Guide.
Incorporating these changes will reflect the as-built condition of the SPDS and the current guidance given in the Emergency Procedure Guide-lines.
EP/2/A/5000/02 Retype #3 The following changes are included in Retype #3:
- 1) Deleted EP/2/A/5000/2E3, Containment High Radiation Level, and its associated logic from the " CONTAINMENT" Critical Safety Function Status Tree. This EP is being deleted as a result of the deletion of Emergency Procedure Guideline FR-Z.3, Response To High Containment Radiation. The deletion of this EPG is part of Duke Power's commitment to satisfy the requirements specified in Supplement 1 of NUREG-0737. This commitment was made in a letter from H.B. Tucker to the NRC cated February 18, 1988. The dele-tion of the containment Radiation input to SPDS was made under SPDS modification SPD #3 (Program Request P880020).
- 2) Deleted EP/2/A/5000/2E2, High containment Sump Level, and its associated logic from the " CONTAINMENT" Critical Safety Function Status Tree and substituted in its place EP/2/A/5000/2E2, Incom-plete Containment Isolation. This change is in response to the deletion of Emergency Procedure Guideline FR-Z.2, Response to Containment Flooding, and the creation of Emergency Procedure Guideline FR-Z.2, Response To Incomplete containment Isolation.
These changes are also being made to satisfy the requirements in Supplomont 1 of NUREG-0737 as mentioned above. The deletion of the Containment Sump Level input to SPDS and the inclusion of
( inputs for containment Isolation were made under SPDS modifica-l tion SPD #3 (Program Request '880020).
- 3) Modified the symptoms in the procedure *.o adequately describe its proper implementation. Implementation of this procedure je controlled by EP/2/A/5000/01, Reactor Trip Or Safety Injection; EP/2/A/5000/03, Loss Of All AC Power; and EP/2/A/5000/1A, Reactor Trip Response. This change does not alter the intent of this 49
procedure, but will provide the correct symptoms as reflacted in the guidance given in the othcr EPs.
- 4) Added Display Group Computer Progra.n information to step 1.a.
This information is being added under SPDS modification SPD #3.
i These display groups will give the operator up-to-date status information on the inputs to the Status Trees.
- 5) Cotrected various format errors to ensure compliance with the EP Writer's Guide.
Incorporating these changes will reflect the as-built condition of the SPDS and the current guidance given in the Emergency Procedure Guide-lines.
EP/1/A/5000/2E2 Retype #0 This change deletes EP/1/A/5000/2E2, High Containment Sump Level, Retype #2 and substitutes in its place EP/1/A/$000/2E2, Incomplete Containment Isolation, Retype #0. This change is in response to the deletion of Emergency Procedure Guideline (EPG) FR-Z.2, Response to Containment Flooding, Revision 1, and the creation of Emergency Procedure Guideline FR-Z.2, Response To Incomplete Containment Isola-tion, Revision #0. These EPG changes are part of Duke Power's commit-ment to satisfy the requirements specified in Supplement 1 of NUREG-0737. This Commitment was made in a letter from H. B. Tucker to the NRC dated February 18, 1988.
The deletion of the Containment Sump Level input to SPDS and the inclusion of inputa for Containment Isolation were made under SPDS modification #3 (Program Request P880020).
EP/1/A/5000/2E2, Incomplete Containment Isolation, has been reviewed by the safety Analysis Group in Design Engineering and has been verified for technical correctness and adherence to the guidance given in the EPG.
l Incorporating these changes will reflect the as-built condition of the SPDS and the current guidance given in the Emergency Procedure Guide-lines.
EP/2/A/5000/2E2 Retype #0 This change deletes EP/2/A/5000/2E2, High Containment Sump Level, Retype #1 and substitutes in ite place EP/2/A/5000/2E2, Incomplete containment Isolation, Retype #0.- This change is=in response to the deletion of Emergency Procedure Guideline.(EPG) FR-Z.2, Response'to Containment Flooding, Revision 1, and the creation of Emergency Procedure Guideline FR-Z.2, Response To Incomplete Containment Isola-tion,-Revision #0. These EPG changes are part of Duke Power's commit-ment to satisfy the requirements specified in Supplement 1 of 50-
I NUREG-0737. This Commitment was made in a letter from H. B. Tucker to the NRC dated February 18, 1988. j The deletion of the Containment Sump Level input to SPDS and the inclusion of inputs for Containment Isolation were made under SPDS modification #3 (Program Request P880020).
EP/2/A/5000/2E2, Incomplete Containment Isolation, has been reviewed by the Safety Analysis Group in Design Engineering and has been '
I verified for technical correctness and adherence to the guidance given in the EPG.
Incorporating these changes will reflect the as-built condition of the SPDS and the current guidance given in the Emergency Procedure Guide-lines.
MP/0/A/7550/04 Retype, Changes 0 to 1 Incorporated This safety evaluation is for the reissuing of MP/0/A/7550/04, changes 0 to 1 incorporated, prepared on 12-16-1988. This procedure has been completely rewritten. (Note: This procedure was approved on 12-5-1989.)
Tech. Specs, 3.11.2.1 and 3.11.2.5 may be affected by this procedure.
Operations has the responsibility and the procedures for compliance with these Tech. Specs. Maintenance will be performed on this recombiner when Tech. Specs, allow, per operation's procedures.
This rewrite will clarify and assure that maintenance activities will return the recombiner to as-designed conditions. The changes made by this rewrite have been reviewed against approved vendor manuals, design documents, and station procedures to ensure that the corrective maintenance controlled by this procedure will return the recombiner to l
as-built /as-designed condition. These actions will ensure the recombiner compliance with FSAR accident analysis. Since the recombiner will be returned to as-designed conditions, the possibili-I ty, consequences, or probability of a malfunction will bs. reduced.
Therefore, no USQ exists.
l i OP/0/A/6400/06C Retype #10, Changes 61 to 86 Incorporated The following is a list of the changes made to the retypet
- 1. Added 5 sign to Limit and Precaution 2.6 and reason for caution to ensure the operator is properly informed.
- 2. Added new Limit and Precaution 2.8 per letter from Alex Almaguer, File no.: CN-100.67, dated March 6, 1989.
- 3. Opened all four Component Cooling (KC) Heat Exchanger (HX) inlets from Nuclear Service Water (RN) to ensure RN supply in Enclosure 4.1.
51
- 4. Placed another KC HX discharge valve in Mini-flow on the other train to ensure both trains have a mini-flow path in case of an Auto-start signal in Enclosure 4.1.
- 5. Added Action 29 to the title of Enclosure 4.10. Action 29 was added to Tech. Specs. due to the third RN pit Channel being added.
- 6. Rewrote Enclosures 4.12 and 4.13 to provide the operator more information on where the enclosure is going and how the plant is aligned on loss of an RN pump and/or its associated Diesel ,
Generator (D/G). This did not change the plant alignment, but clarified how tc clign the plant for the different modes. Steps were added to these enclosures to isolate an Essential Header of RN on a unit in Mode 5, 6, or No Mode. Previously, these enclo-sures isolated the non-essential header, a containment Spray (NS)
HX and a train of assured makeup to the Auxiliary Feedwater (CA) system to ensure an RN pump on that train could supply its required load during a LOCA. By adding the option to isolate an essential header, the requirement to write a restricted procedure change during an outage is no longer required.
- 7. Addition of new Enclosure 4.14 for two inoperable RN pumps and/or their associated D/G on one train. This enclosure ensures the appropriate equipment is removed from service to prevent equip-ment damage. No realignment of the RN System is required, since the other Train of RH will separate on an Sp signal from the
- inoperable train of RN.
- 8. Also, other minor changes were made to make the procedure more user friendly.
- 9. Added information to Enclosure 4.13 to ensure the proper actions are taken if both RN pumps _become inoperable during Modes 1-4.
The enclosure requires the unit to shut down under Tech. Spec.
3.0.3, but maintains a train of CA and NS operable on the unit shutting down in case of an accident.
- 10. Added new Enclosure 4.15 to describe the required action when 3 l
RN pumps become-inoperable. This enclosure shuts down both-units-due to Tech. Spec. 3.0.3.
- 11. Incorporated previously approved changes 62-86.
The changes made to this retype of OP/0/A/6400/06C do not affect the FSAR or the evaluations made in the FSAR. These changes do not result in an unreviewed safety question and do not reduce the margin of safety assumed in the FSAR or Tech. Specs.
TN/1/A/1156/00/01A Initial Issue TN/1/A/1156/00/01A will provide the guidelines for in-plant pre-fabri-cation activities for the-Linear Variable Displacement Transducer 52
(LVDT) addition to Main Foodwater (CF) Control and Bypass Control Valves ICF28, ICF30, ICF37, ICF39, ICF46, ICF48, 1CF55, and 1CF57.
The LVDTs will supply signals to the OAC and to the transient monitor for analog information of full valve stroke.
Currently, position indication for Control Valves 1CF28, 1CF37, 1CF46, and 1CF55 is provided by a Fisher Type 304 limit switch. This limit switch provides full open/ full closed and 25% open indication.
Position information for Bypass Control Valves 1CF30, ICF39, ICF48, and 1CF57 is provided by two Namco limit switches. These limit switches provide full open/ full closed indication only. The limit switches will continue tc provide their present position indication, in addition to the new indication provided by the LVDTs.
This procedure mounts two new terminal boxes to be used in conjunction with the LVDTs. It also provides for cable-pulling and for termina-tions at several locations which may be done prior to the Unit 1 End-of-Cycle 4 refueling outage.
This procedure may be implemented with Unit one in any Mode. No work will be done on this procedure that will adversely affect the opera-bility of the plant.
Accordingly, this modification will not incror.se the probability or coneequences of an accident previously evaluated, or different than any already evaluated, in the FSAR. Nor will it increase the proba-bility or consequences of an equipment mclfunction previously evaluat-od, or different than any already evaluatec, in the FSAR. The margin of safety defined in the bases of the Technical Specifications is unaffected. An unreviewed safety question does not exist.
TN/1/A/1156/00/02A Initial Issue TN/1/A/1156/00/02A will provide the guidelines for adding Linear Variable Displacement Transducers (LVDTs) to Main Foodwater Control Valves ICF28, 1CF37, ICF46, and 1CF55. This procedure will also provide the guidelines for connecting the LVDT Trip Arm Linkage to the valve stems of the Feedwater Control Valves, as well as Bypass Control Valves ICF30, 1CF39, ICF48, and 1CF57. The LVDTs will supply signals to the OAC and to the transient monitor for analog information of full valve stroke.
Currently, position indication for Control Valves ICF28, 1CF37, 1CF46, and 1CF55 is provided by a Fisher Type 304 limit switch. This limit switch provides full open/ full closed and 25% open indication.
Position information for Bypass Control Valves ICF30, ICF39, 1CF48, and 1CF57 is provided by two Namco limit switches. These limit switches provide full open/ full closed indication only. The limit switches will continue to provide their present position indication in addition to the new indication provided by the LVDTs.
This procedure mounts the LVDTs on the Main Feedwater Control and Bypass Control valves (1CF28, 1CF30, 1CF37, 1CF39,-1CF46, 1CF48, 53 i
l
l l
I l
l 1CF55, and 1CP57). It also connects the LVDT Trip Arm Linkages to both the Main Feedwater Control Valves and the Bypass control Valves, l Functional verification and retest requirements have been specified in the proceduro. The testing will fully challengo the integrity of all affected components.
This procedure may be implemented with Unit One in Mode 5, 6, or No Mode, when the CF valves are not required by the Technical Specifica-tions to be operable.
Accordingly, this modification will not increase the probability or consequences of an accident prcviously evaluated, or different than any already evaluated, in the FSAR. Nor will it increase the proba-bility or consequences of an equipment malfunction previously evaluat-od, or different than any already evaluated, in the FSAR. The margin of safety defined in the bases of the Technical Specifications is unaffected. An unroviewed safety question does not exist.
TN/0/A/2437/CE/01A Initial Issue This procedure installs seal welded nitrogen and hydraulic fittings on the spare feedwater isolation valvo actuator. The work outlined in this procedure will be performed in accordance with existing station procedures for QA work. The actuator being modified is the spare and, therefore, there will be no effects on the plant when this modifica-tion'is installed. The actuator will be removed from the warehouse, modified, tested, and returned to the warol.ouse as the spare. The actuator will be retested per existing station procedures when it is installed in the plant. Work in this procodure may be performed with either unit in any mode, and will require no isolations. The testing outlined in this procedure will pressure test the hydraulic and nitrogen systems in the actuator to ensure no leaks exist. It will also timo the stroke of the actuator on the bench to ensure it will travel in less than 5 seconds. As this modification is on a spare i
actuator, the probability of an accident or malfunction of equipment l evaluated in the FSAR will neither be increased nor created. The l
margin of safecy as defined in tho basis to any of the Technical l Specifications will not b2 reduced. The implementation of this procedure will not create an unreviewed safety question.
l TN/0/A/2439/CE/01A Initial Issue This procedure installs seal welded nitrogen and hydraulic fittings on the spare feedwater isolation valve actuator. The work outlined in this procedure will be performed in accordance with existing station procedures for QA work. The actuator being modified is the sparc and, therefore, there will be no effects on the plant when this modifica-tion is installed. The actuator will be removed from the warehouse, modified, tested, and returned to the warehouse as the spare. The actuator will be retested por existing station procedures when it is installed in the plant. Work in this procedure _may be performed with 54
l either unit in any modo, and will require rio isolations. The testing outlined in this procedure will pressure test the hydraulic and nitrogen systems in the actuator to ensure no leaks exist. It will I also timo the stroke of the actuntor on the bench to ensure it will travel in loss than 5 seconds. As this modification is on a spare actuator, the probability of an accident or malfunction of equipment evaluated in the FSAR will neither be increased nor created. The margin of safety as defined in the basis to any of the Technical Specifications will not be reduced. The implomontatioli of this procedure will not create an unroviewed safety question.
t l TN/1/A/2698/CE/01A Initial Issuo Exempt Change CE-2698 will modify the control circuit wiring on Component Cooling (KC) valvo 1KC081B to provide " limit actuated" torque switch bypass contacts which can be adjusted indopondently of indications or interlocks. 1XC081D is the inlet valve to the Residual Heat Removal (ND) System Heat Exchanger.
This proceduro will provido guidelines for Instrument and Electrical (IAE) personnel to rowire the switch rotor and set up the switch rotor. This procedure will also require performance to stroke timo 1XC081B.
l During the implomontation of this procedure, Unit one Emergency Core cooling System Train B will be inoperable and Unit Ono will bo in a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement per Toch. Spec. 3/4.5.2. This valve, 1XC081B, is required to open on a Safety Injection signal followed by a low-low Refueling Water Storago Tank (FWST) level, and on a high-high contain-mont pressure signal.
Prior to returning 1KC081B to service, a functional horification and rotest will be performed to verify valve operability.
Accordingly, this procedure will not increase the probability or consequences of an accident previously evaluated, or different than any already evaluated, in the FSAR. Nor will it increase the proba-bility or consequences of an equipment malfunction previously evaluat-od, or different than any already evaluated, in the FSAR. The margin of safety defined in the bases of the Technical Specifications is unaffected. An unroviewed safety quoetion doos not exist.
TN/2/A/2699/CE/01A Initial Issuo Exempt Change CE-2699 will modify the control circuit wiring on valve 2KC056A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indications or interlocks.
2KC056A is the inlet valve to the Residual Heat Removal (ND) System L
Heat Exchanger.
55
This procedure will provide guidelines for IAE to rewiro the switch rotor and set up the switch rotor. This procedure will also requiro Performance to stroke time 2KC056A.
During the implementation of this proceduro, Unit Two Emergongy Core Cooling System Train A will be inoperabic and Unit Two will be in a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement per Tech. Spoc. 3/4.5.2. This valve, 2KC056A, is required to open on a Safety Injection signal followed by a low-low Refueling Water Storago Tank (FWST) lovel, and on a high-high contain-me:t pressure signal.
N 3 sr to returning 2KC056A to service, a functional verification and retest will bo performed to verify valve operability.
Accordingly, this proceduro will not increase the probability or consequences of an accident provicusly evaluated, or differont than any already evaluated, in the FSAR. Nor will it increase the proba- l bility or consequences of an equipment malfunction previously evaluat-ed, or different tha*. any already ovaluated, in the PSAR. The margin of safety defined in the bases of the Technical Specifications is unaffected. An unreviewed safety question does not exist.
PT/0/A/4450/15A Initial Issue i
The Instrument Air (VI) system supplies compressed air to various instruments around the plant. This procedure measures the dew point temperature of the air in the VI system at variouc locatiens in the plant to ensure low moisture air is supplied. The FSAR describos the loss of air to the VI system, which would cause the encumatically operated valves in the station essential for safe shutdown to fail in their safo posiV. ions. Precautions have been included in the procedure to notify Nuclear Control Room Operators of alarms that could be actuated during this test and to open valvos slightly so that VI system precsure will not be affected. The taking of these air sampics will not increase the probability or consequences of an accident or malfunction of equipment. A small amount of air will be taken from the system, which will not change the system in a significant manner.
This procedure will not create an accident or malfunction of safety-related equipment that is not addressed in the FSAR, since the only function that is being performed is the taking of air from the system.
The Instrument Air system is not discussed in Tech. Spoc., and the
, margin of safety as defined in the bases to any Tech. Spec, is not affected. Therefore, this procedure does not reduce the margin of safety as defined in Tech. Spec. or create an unreviewed safety question.
TT/0/A/9100/50 Initial Issuo The purpose of this temporary procedure is to obtain a correlation i between Control Room (CR) penetration hole size and CR pressure. Data obtained will be used to determine the maximum penetration hole size that will be allowed in a CR pressure boundary for Nuclear Station 56
i Modifications or other work that involves a CR pressure boundary.
This will be accomplished by cutting holes in penetrations using a i piece of one, two, and/or three inch diameter PVC pipe. After a hole is cut, another piece of the same size gappp4 PVC pipe will be insert-ed into the same hole. Then the cap (or piece of PVC pipe) will be removed and CR pressure will be measured. The above will be repeated until two heles, cut by a three inch diameter piece of PVC pipe, have been cut into two penetrations OR adequate data has been obtained as determined by Test Coordinator. The two penetrations that were chosen reference the Unit 1 Cable Room, which is slightly positive due the pressurized filter train on Control Area Ventilation (VC), so all doors from the Cable Room to the Service Building are held open for this test (to equalize pressure between the Service Building and Cable Room) in order to give conservative measurements for CR pressure. A fire /tornedo watch will be implemented as required for all the doors that are held open. Train B of VC was chosen to be the operating train during the test because it provides higher pressurization when one intake is isolated.
CR pressure will be monitored any time that the penetration hole size is altered, and if the pressure falls below the Tech. Spec. require-ment, the procedurn requires that appropriate holes be plugged in order to increase pressure above the Tech. Spec. requirement. The penetrations were chosen close to the manometer that is used to measure the CR pressure to aid in the above requirement. CR pressure will also be measured before and after the test to ensure that the pressure boundary of the CR was not degraded by this procedure. If the pressure at the end of the test is significantly less than the pressure at the start of the test, the procedure requires that the train A VC pressurization test be performed. The procedure informs the Shift Supervisor to notify the Test Coordinator in the event of an accident or tornado during the performance of this test, and a CAUTION statement requires that all holes be plugged and doors closed C Ji were altered by this procedure if an accident were to occur duttng the test. Another CAUTION statement requires that all tornado doors are closed in the event of a tornado during this test. After the test is completed, all holes in penetrations will be plugged and then sealed one at a time. If the data obtained indicates that removing a piece of PVC pipe may cause the OR pressure to fall below the Tech. Spec.
requirement, the procedure requires that the hole be plugged on one side (after the pipe is removed) and then sealed as soon as possible.
All data will be taken with VC operating in the normal alignment. No instruments or modifications will be installed / performed that would affect VC in any manner other than the holes in the penetrations.
Control Room doors are held open at times to perform maintenance on the doors, which reduces the pressure in the CR well below the Tech.
Spec. requirement. The CR doors are also opened and closed many times during the day for normal " momentary" access to the CR. The size of l holes created by this test are insignificant compared to an open CR !
door, and the controls on the procedure ensure that any violation to l CR pressure is only momentary. Also, if an accident were to occur J during the test, the procedure requires that the CR pressure boundary l be returned to the As Found condition (as stated above). -A security l 57 e w
officer will be present at all times when a penetration is violated and until it is sealed. For those reasons and the ones stated above, this procedure does not increase the probability cr consequences of an accident previously evaluated in the FSAR OR increase the probability or consequences of a malfunction of equipment important to safety.
Also, this procedure will not create the possibility of an accident or malfunction of equipment impor'. ant to safety dif ferent than already evaluated in the FSAR. The margin of safety as defined in the bases of Tech. Spec. will not be reduced. Therefore, this precedure does not create an unreviewed safety question.
TN/2/A/05"7/00/01A Initial Issue This procedure controls cable pulling activities necessary for the implementation of the new Unit 2 Events Recorder System per NSM CH-20557, Rev. 00.
The cables being pulled under this procedure are Non-Safety cables.
These cables will not be terminated per this p: mo4ure. These cables will be routed such that the separation requirem....s as stated in the FSAR are maintained. The requirement for firewatches to be estab-lished when fire boundaries are breached is written in this procedure.
This procedure will involve the penetration of a Control Room Ventila-tion System (VC) firestop. Because of this, steps, notes, and warn-ings have been incorporated in this procedure to ensure VC operability during the implementation of this procedure. This procedure will not l create a breach of any security boundaries. No other_ systems will be
[
affected by this procedure, ind. as a result, no FSAR changes will be l
created from the implementatu of this procedure. This procedure will not have any effect on the bases for the EMC or VC system as stated in the Tech. Specs. This procedure will not create any unreviewed safety questions.
PT/0/A/4400/22A, Change #22 The-purpose of this change.is to delete the requiroment for the opposite train Control Area ~ Ventilation (VC) and chilled Water (YC)
System to be in service and to provids for a minimum flow path for the B Train Nuclear Service. Water (RN) Pumps, since both of their Compo-nent Cooling (KC) Heat Exchangers will be in " Temp" mode and the train crossover valves will be closed. YC Chiller flow will not change-significantly over the course of_the test and, therefore, will have negligible effect on the total pump test flow. Opening the contain-i ment Spray (NS) Heat Exchanger (HX) isolation valves will be-con-trolled by the control Room Operators so they will be cognizant at all times of the status of RN Pump Flow.
This change in no way increases the probability of an accident occur-ring, whether or not it has been_ evaluated in the FSAR._ _The A Train RN System will essentially be in.the failed mode-since the KC HX outlet valves will be failed open. The opposite train of RN will be supplying cooling water to inservice-components (except1possibly the-I l
58
YC Chiller), so no interaction with the reactor or supporting systems is present. The consequences of an accident previously evaluated in the FSAR are not increased, since the opposite train of RN will be in service and supplying cooling water to required components, and the train under test is already essentially in the Engineered Safeguards mode.
This change actually decreases the probability of and consequences of a malfunction of safety related equipment by ensuring that a minimum flow path is provided at all times for the B Train RN Pumps. No new equipment failures or malfunctions are created as a result of this change.
Neither train of RN will be inoperable at the time of the test. l Technical Specification 3/4.7.4 allows two RN pumpr to be out of !
l service for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with both units in Modeb 1, 2, 3, and 4.
Therefore, the margin of safety as defined in the bases of Technical Specifications is not reduced, l
OP/2/A/6700/01 Change #109 OP/2/A/6700/01 (Unit Two Data Book) Table 2.2 is a table of data for use by plant personnel. The following describes the two sections of the table and how the data is obtained and used.
FULL POWER CURRENTS This section is used to record the 100% Full Power currents for +20, 0, and -20% axial offset and the M factors for each of the Power Range Excore Detectors. Data is obtained for this section only by use of approved procedures such as:
pT/2/A/4600/05A, Incore/Excore Calibration, PT/2/A/4600/05D, Int erim Incore/Excore Calibration, or PT/2/A/4600/05G, Por.t Refueling Incore/Excore Calibration.
other tests may supply data to this section, but in all cases the tests must be approved tests.
The data recorded here is used by Instrument and Electrical personnel to adjust the Axial Flux Difference (AFD) calculating circuitry and inputs to Operator Aid Computer (OAC) programs. It is also used in manual calculations of AFD and Quadrant Power Tilt Ratio (QPTR) (for instance, when the OAC is inoperable).
Since AFD is used to dynamically adjust the Overtemperature Differen-tial Temperature (OTDT) setpoints, the data here is safety related. A number of accidents in Chapter 15 of the FSAR depend on these setpoints for mitigation.
I l TRIP SETPOINTS l
59 l
l - -
The trip sotpoints for both tho Intermediate (N-35 and N-36) and Power Range (N-41, N-42, N-43, and N-44) detectors are recorded on page 2 of the table. These are used by IAE in setting the Reactor Trip setpoints for the detectors. The data here may be obtained by a variety of means.
I Trip setpoints for Intermodlate Range Detector may only be obtained by use of approved procedures to calculate or measure the 25% Full Power Reactor Trip setpoints.
Trip setpoints for the Power Range Detectors may not be deliberately set greater than 109% Full Power. The trip setpoint may .ts set lower than 109% by use of approved procedures, by direction of Tech. Specs, or by direction of the Shift Supervisor. The 309% is set by Tech.
Specs. to ensure operation is bounded by the assumptions used in the FSAR chapter 15 accidents listed above. Any setpoint below 109% may be used for conservatism or to comply with Tech. Specs.
This procedure change updates Page 1 of Table 2.2 -- Excore Detector Data, to reflect the latest calibration data obtained per PT/2/A/4600/05A, Incore/ Excore Calibration. PT/2/A/4600/05A was performed for the regular quarterly surveillance.
Information in OP/2/A/6700/01 (Unit Two Data Book) is changed only by approved proc-adure change. It will not increase the prcbabili-ty/ consequences of an accident analyzed in the FSAR or create an accident not analyzed in the FSAR. No analyzed or unanalyzed malfunc-tion of safety related equipment will be created. The margin of safety as defined in the basen to Tech. Specs, will not be reduced in any way.
OP/2/3/6100/10Y, Change #5 This chant revised the following (1) Various intarlocks and trips associated with Radiation Monitors (EMF) 35, 36, and 37 Hi Rad alarms were detailed in the Automatic Actions of the responses for these alarms.
(2) Supplementary Action 1 was added to the responses for EMF-35, 36, and 37 Hi Rad alarms to allow the Auxiliary Building Ventilation (VA)
System to be returned to service in the Filter Mode. This will allow monitored filtered releases to continue from the Auxiliary Building through the unit vent, rather than allowing the VA System to remain shutdown and having uncontrolled and unmonitored releases through various doors.
(3) Supplementary Action 2 was added to the responses for EMF-35, 36, and.37 Hi Rad alarms to allow the Spent Fuel. Pool Ventilation (VF)
System to be returned to service in the Filter Mode when EMF-42 is inoperable. This will allow the humid fuel pool area air to continue being filtered while the cause of the alare is investigated.
60
( .4 ) Supplementary Acticn 5 was added to the responses for EMF-35, 36, and 37 Hi Rad alarms to ensure tha'; diring accident situations the Health Physics (HP) Dose Assessment coordinator is aware of all critical EMF alarms.
1 l (5) Supplementary Action 3 doloted 2 EMF-33 Condenser Steam Air Ejector 3 from the check list and added the applicable Containment EMF (2 EMF-38, l 39 and 40). 2 EMF-33 does not supply an input to the Unit 2 Unit Vent.
This check list checks the activity levels of the EMPs that monitors j systems providing inputs into the Unit Vent.
l (6) Added a note in the Automatic Action section stating EMF-41 is j inoperable when the VA system is shutdown.
The changes made concerning information on trips, interlocks and alarm defeats all reflect the As-Built condition of the plant as determined from the applicable CNEE's and other controlled references. The changes made to the Supplementary Action section of the Unit Vent Hi Rad alarms permit the une of " BYPASS" switches to defeat normal interlocks to allow releases to continue through tne Unit Vent in a controlled, monitored fashion. This ensures that HP remains aware of the real radiological situation and thot releases are being monitored while the cause of the alarm is investigated.
FSAR Sections 9.4.2 and 9.4.3 have been reviewed; and it has been determined that allowing the VA and VF Systems to continue to operate in the filter mode while there is a Unit Vent Hi Rad alarn does not exceed the design basis of the systems or affect any Chapter 15 Accident Analysis. Tech Specs 3.7.7 and 3.9.11 have been reviewed and the changes made by this change do not reduce the margin of safety.
TN/1/A/0854/00/01A Initial Isnue This procedure provides implomontation guidelines for NSM CN-10854, Rev. O, Work Unit 01. NSM CN-10854 will replace the personnel air lock air relief line and associated check valves to provide adequate venting of the air lock in case of an internal air line leak. This air relief line and check valves are sized to ensure that internal air lock pressure does not exceed the design limit of 15 psig. Tb i. ,
procedure is for the Unit i upper air lock. The existing eir relief line will be cut out and removed. A special studding optiot and check valves will ba installed in the same location. Apprrpriate notes are included in the procedure to ensure this work is done in Modes 5, 6, and No mode (i.e., when containment integrity is not required.)
During core alterations, Tech. Specs. require at least one door in each airlock to be closed. Notes to ensure this Tech. Spec. is met are included. After completion of the work, the check valves will be tested to verify their opening pressure. Also, a leak rate test will be performed on both the check valves and the entire airlock. All of the testing will be completed prior to returning the airlock to service. Based on the above discussion, an unreviewed safety question will not be created by this procedure.
61
% , m ]
Catawba Nuclear Statign Summary o. reaceduro Changes, Tests, and Experiments Completed from 11/1/89 to 9/30/90 -- Volumo 2 TN/1/A/0295/00/01A Initial Issuo This proceduro provides implementation instructions for Nuclear Station Modification (NSM) CN-10295. This NSM will replace 1WLLS5760 and 1WLLS6860 with a single FCI-FR72LLMPL multi-point senso< assembly with three sensing points and remotely mounted electronics. The now instrument will be tagged 1WLLS5760. The now level switch will control both the starting and stopping of the incore instrumentation pit sump pump as well as giving an alarm on Hi-Hi level.
The implementation of this procedure will require pulling two new non-safety cables and rerouting and retagging tNo other non-safety cables. All firestops opened for these cables will be closed such that fire boundarios are not breached without having a fire watch.
All termination points for this modification will be isolated from electrical power. This will cause numerous pieces of equipment to be out of service. However, consideration has been given to the effects of these isolations in the developmer.t of the procedLce. Calibration of the now level switch has boon specified in the procedure to ensure proper operations. This procedure will not create any unroviewod safety questions (USQ).
TN/1/A/100b/01/43A Initial Issue This proceduro provides implementation instructions for Nuclear Station Modification (NSM) CN-11005 Rov. 1. This NSM modifies various piping system math models with the objective of reducing the number of mechanical snubbers required. This procedure provides guidance for the removal of snubbers deleted, as well at the modification of certain other support / restraints on the Diesel Generator Engine Lube oil (LD), Boron Recycle (NB), Boron Thermal Reconoration (NR), Wasto Gas (WG), and Liquid Radwasto (WL) systems.
The on~y safety concern which may arise as a result.of the implementa-tion of this work unit is for the soismic qualification of these system's piping. These systems have been qualified for the present support / restraint (S/R) configuration. Likewise, they have been qualified for the S/R configuration which will be in place after the procedure has been implemented. However, the interim configuration (with some deleted snubbers removed and some still in place) has not been analyzed because the many possible combinations of S/R configura-tions would be difficult to analyze. For this reason, Design has stipulated that the piping systems must be modified within a specific order to consider the piping affected by this work unit to be operable during implementation of this work unit. This procedure will provido the necessary controls to ensure that the work is finished in the i
I i
specified order. For the above reasons, it is concluded that no USQ is created by this procedure.
TN/1/A/1005/01/21A Initial Issue This procedure provid3s implementation instructions for NSM CN-11005 Rev. 1. This NSM modifies various piping system math models with the objective of reducing the number of mechanical snubbers required.
This procedure provides guidance for the removal of snubbers deleted from the auxiliary feedwater system. This work also involves several main feedwater system supports. These supports will.either be deleted from the system or revised to a different configuration, thus reducing the aaber of snubbers remaining.
The only safety concern which may arise as a result of the implementa-tion of this work unit is for the seismic qualification'of the Auxil-iary and Main Feedwater. System piping. These systems have been qualified for the present support / restraint (S/R) configuration.
Likewise, they have been qualified for the S/R configuration which will be in place after the procedure has been implemented. However, the interim configuration (with some deleted snubbers removed and some still in place) has not been analyzed because the many possible combinations of S/R configurations would be difficult to analyze. For this reason, Design has stipulated that the piping systems must be modified within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in order to consider the piping affected by this system model to be operable during the implementation of this work unit. This procedure will provide the necessary controls to ensure that this work unit is finished in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.. If the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit is exceeded, this procedure ensures that the-Technical i specification requiremente are met.
If the piping affected by this modification is not required to be operable during the implementation of this work unit, no time limit l applies. For the above reasons, it is concluded that'no USQ is created by this procedure.
l
~
TN/1/A/1005/01/18A Initial Issue This procedure provides implementation instructions for NSM CH-11005 i Rev. 1. This NSM modifies various piping system math models with the l
objective of reducing the number of mechanical snubbers required.
This procedure provides guidance for the removal of snubbers deleted
! from the Main Steam Vent to Atmosphere (SV) system. These sapports i will either be deleted from the system or revised to a different
- configuration, thus reducing the number of snubbers remaining.
I The only safety concern which may arise as a result of the implementa-
, tion of this work unit is for the seismic _ qualification of the SV
, system piping.- This system has been qualified for the present sup--
port / restraint (S/R) configuration. Likewise, it has been qualified -
- for the S/R configuration which_will be in place after the procedure has been implemented. However, the interim configuration (with some l
sw1ew-,+m -w r - ev , -w- -- ,.*-=w- - --' *,w--e-- , --
deleted snubbers removed and some still in place) has not been ana-lyzed because the many possible combinations of S/R conf.gurations would be difficult to analyze. For this reason, Design has stipulated that the piping systems must be modified within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in order to concider the piping affected by this system model to be operable during the implementation of this work unit. This procedure will provide the necessary controls to ensure that this work unit is finished in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit is exceeded, this procedure ensures that the Technical Specification requirements are met.
If the piping affected by this modification is not required to be operable during the implementation of this work unit, no time limit applies. For the above reasons, it is concluded that no USQ is created by this procedure.
TN/1/A/1005/01/17A Initial Issue This procedure provides implementation instructions for NSM CN-11005 Rev. 1. This NSM modifies various piping system math models with the objective of reducing the number of mechanical anubbers required.
This procedure provides guidance for the removal of snubbers deleted from the Residual Heat Removal (ND) system. This ND system math model also includes several Safety Injection (NI), Containment Spray (NS),
and Refueling Water (FW) system supports. These supports will either be deleted from the system or revised to a different configuration, thus reducing the number of snubbers remaining.
The only safety concern which may arise as a result of the implementa-tion of this work unit is for the seismic qualification of the ND, HI, NS, and FW system piping. These systems have been qualified for the present support / restraint (S/R) configuration. Likewise, they have been qualified for the S/R configuration which will be in place after the procedure has been implemented. However, the interim configura-tion (with some deleted snubbers removed and some still in place) has not been analyzed because the many possible combinations of S/R configurations would be difficult to analyze. For this reason, Design has stipulated that the piping systems must be modified within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in order to consider the piping affected by this system model to be operable during the implementation of this work unit. This proce-dure will provide the necessary controls to ensure that this work unit is finished in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit is exceeded, this procedure ensures that the Technical Specification requirements are met.
If the piping affected by this modification is not required to be operable during the implementation of this work unit, no-time limit l applies. For the above reasons, it is concluded thr.c no USQ is
- created by this procedure, l
i TN/1/A/1005/01/16A Initial Issuo 3
l
i This procedure provides implementation instructions for NSM CN-11005 I i Rev. 1. This NSM modifies various piping system math models with the l
, objective of reducing the number of mechanical snubbers required. ;
This procedure provides guidance for the removal of snubbers deleted from the safety Injection (NI) system. This NI system math model also includes several Residual Heat Removal (ND) and Refueling Water (FW)
- system supports. These supports will either be deleted from the system or revised to a different configuration, thus reducing the number of snubbers remaining.
! The only safety concern which may arise as a result of the implementa-tion of this work unit is for the seismic qualification of-the N1, ND, and FW system piping. These systems have been qualified for the present support / restraint (S/R)-configuration. Likewise, they have been qualified for the S/R configuration which will be in place after the procedure has been implemented. However, the interim configura-tion (with some deleted snubbers removed and_some still in place) has not been analyzed because the many possible combinations of S/R configurations would be difficult to analyze. For this reason, Design has stipulated that the piping systems must be modified within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in order to consider the piping affected by this system model to be operable during the implementation of this vork unit. This proce-dure will provide the necessary controls to ensure that_this work unit is finished in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit is exceeded,-this procedure ensures that the Technical Specification requirements are met.
If the piping affected by this modification is not required to be operable during the. implementation of this work uni %s Oo time limit
~
applies. For the above reasons, it is concluded thd2 no USQ is created by this procedure.
TN/1/A/1005/01/15A Initial Issuo '
l j This procedure provides implementation instructions for NSM CN-11005-l Rev. 1. This NSM modifies various piping system math models with the L objective of reducing the-number of mechanical. snubbers required.
This procedure provides guidance for the removal of snubbers deleted from the component Cooling (KC) system. These supports will either be deleted from the system or-revised to a different configuration, thus reducing the number of snubbers remaining.
. The only safety concern which may arise as a result of the implementa-l tion of this work unit is for the seismic qualification of the KC i system piping. This system-has been qualified for the present sup-port / restraint (S/R)_ configuration. Likewise, it has been qualified for the S/R configuration which will be in place after the procedure has been implemented. However, the interim:configurationz (with some deleted snubbers removed and some still:1n place) has not been ana-lyzed because the many possible combinations-of1S/R configurations would be' difficult to analyze. For this~ reason,. Design has stipulated that the piping systems must be modified within=72 hours in order to consider the piping affected by this system model to be operable 4
..-._--- .- - -.- ..- --- . - - - . . = - -_ -_
during the implementation of this work unit. This procedure will provide the necessary controls to ensure that this work unit is finished in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit is exceeded, this procedure ensures that the Technical Specification requirements are met.
If the piping affected by this modification is not required to be operable during the implementation of this work unit, no time limit applies. For the above reasons, it is concluded that no USQ is created by this procedure.
TN/1/A/1005/01/14A Re-write This procedure provides implementation instructions for NSM CH-11005 Rev. 1. This NSM modifies various piping system math models with f.he objective of reducing the number of mechanical snubbers required. -
This procedure provides guidance for the removal of snubbers deletod from the Residual Heat Removal (ND) system. This ND system math model also includes one safety Injection (NI) system support to be deleted.
These supports will either be doloted from the system or revised to a different configuration, thus reducing the number of snubbers remain-ing.
The only safety concern which may arise as a result of the implementa-tion of this work unit is for the seismic qualification of the HD and NI system piping. These systems have been qualified for the present support / restraint (S/R) configuration. Likewise, they have been qualified for the S/R configuration which will be in place after the procedure has been implemented. However, the interim configuration (with some deleted snubbers removed and somo still in place) has not been analyzed because the many possible combinations of S/R configura-tions would be difficult to analyze. For this reason, Design has stipulated that the piping systems '.nust be modified within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in order to consider the piping affected by this system model to be operable during the implementation of this work unit. This procedure will provide the necessary controls to ensure that this work unit is finished in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit is exceeded, this procedure ensures that the Technical Specification requirements are met.
If the piping affected by this modification is not required to be operable during the implementation of this work unit, no time limit applies. For the above reasons, it is concluded that no USQ is created by this procedure.
TN/1/A/1005/01/13A Initial Issue This proccdure provides implementation instructions for NSM CN-11005 Rev. 1. This NSM modifies various piping system math models with the objective of reducing the number of mechanical snubbers requ:tred.
This procedure provides-guidance for-the removal of snubbers deleted from the Chemical and Volume Centrol (NV) system. These supports will 5
cither be doloted from the system or revised to a different configura-tion, thus reducing the number of snubbers remaining.
The only safety concern which may arino as a result of the implomonta-tion of this work unit is for the seismic qualification of the NV system piping. This system has boon qualified for the present sup-port /rostraint (S/R) configuration. Likewise, it has boon qualified for the S/R configuration which will be in place after the proceduro has been implomonted. However, the interim configuration (with some doloted snubbers removed and some still in place) has not boon ana-lyzed because the many possible combinations of S/R configurations would be difficult to analyzo. For this reason, Design has stipulated that the piping systems must be modified within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in order to considor the piping affected by this system model to be operable curing the implomontation of this work unit. This procedure will provide the necessary controls to ensure that this work unit is finished in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> timo 3'-4 t is excooded, this procedure ensures that the Technical Specif: ation requirements are mot.
If the piping affected by this modification is not required to be operable during the implementation of this work unit, no timo limit applies. For the above reasons, it is concluded that no USQ is created by this procedure.
TN/1/A/1005/01/12A Initini !*suo This proceduro provides implenentation instructions for NSM CN-11005 hov. 1. This NSM modifics various piping system math models with the objective of reducing the nun 0er of mechanical snubbers required.
This procedure provides guidance for the removal of snubbers doloted from the Residual Heat Removal (ND) cystem. These supports will either be doloted from the system or revised to a different configura-tion, thus reducing the number of snubbers romaining.
The only safety concern which may atjao as a result of the implementa-tion of this work unit is for the seimnic qualification of the ND system piping. This system has boon qualified for the present sup-port / restraint (S/R) configuration. Likewise, it has been qualified for the S/R configuration which will be in place after the procedure has boon implomonted. However, the interim configuration (with some deleted snubbers removed and some still in place) has not been ana-lyzed because the many possible combinations of S/R configurations would be difficult to analyze. For this reason, Design has stipulated that the piping systems must be modified within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in order to consider the piping affected by this system model to be operable during the implomontation of this work unit. This proceduru will provide the necessary controls to ensure that this work unit is finished in 7?. hours. If the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit is excooded, this proceduro ensures that the Technical Specification requirements are i met.
)
6
If the piping affected by this modification is not required to be ,
operable during the implementation of this $ork unit, no time limit applies. For the above reasons, it is co:r .uded that no USQ is created by this procedure.
TN/1/A/2655/CE/01A Initial 7ssue This procedure providen implementation instructions for Exempt Change CE-2655. This exempt ebange provides an access hole in the main steam systea piping at steam generator (S/G) IC. This hole will be used to access inside the piping in order to perform a radiographic examina-tion of the weld between the piping and the steam generator. This weld must be inspected in order to comply with the ASME Code Section XI Inservice Inspection requirements. After the examination is complete, the hole will be plugged and seal welded. The purpose of this procedure is to provide guidance for drilling the hole and installing a half coupling and plug.
The Operations Group will coordinate the isolations necessary to implement this procedure. S/G 1C and its sesociated main steam line will be out of service during the implementation of this procedure.
This procedure will be implemented during an outage and in Modes 5, 6, and No Mode. The Operations Group will be notified to ensure contain-ment integrity is maintained while the affected main steam line is open to containment. Testing for Exempt Change CE-2655 will be performed in accordance with the stat'on Post Modification Testing Program. A visual inspection for leans will be performed in Mode 3 with the Main Steam system at normal system temperature and pressure.
Per this discussion, a USQ does not exist.
TN/1/A/1005/01/44A Initial Issue This procedure provides implementation instructions for Nuclear Station Modification (NSM) CN-11005 Rev. 1. This NSM modifies various piping system math models with the objective of reducing the number of mechanical snubbers required. This procedure provides guidance for the removal of snubbers deleted as well at the modification of certain other support / restraints on the component cooling (KC), Instrument Air '
(VI), Chemical and Volume Control (NV), Boron Recycle (NB), and Waste Gas (WG) systems. These supports will either be deleted from the system, or revised to a different configuration, therefore, reducing the number of snubbers remaining.
The only safety concern which may arise as a result of the implementa-tion of this work unit is for the seismic qualification of the piping.
These systems have been qualified for the present support / restraint (S/R) configuration. Likewise, they have been qualified for the S/R configuration which will be.in place after the procedure has been implemented. However, the interim configuration (with some deleted snubbers removed and some still in place) has not been analyzed because the many possible combinations ~of S/R configurations-would be difficult to analyze. For this reason, Design has stipulated that the 7
1 i
l i
piping systems must be modified within a specific order to considnr the piping af fected by this work unit to be operable during implet sn-tation of this work unit. This procedure will provide the necessary controls to ensure that the work is finished in the specified order.
For the above reasons, it is concluded that no USQ is created by this procedure.
TN/1/A/1005/01/45A Initial issue This procedure provides implementation instructions for NSM CN-11005 Rev. 1. This NSM modifies various piping system math models with the objective of reducing the number of mechanical snubbers required.
This procedure provides guidance for the removal of snubbers deleted from the Auxiliary Feedwate" (CA) and Feodwater Pump Turbine Exhaust This modification also includes several CA and TE I
(TE) systems.
system supports. These supports will either be deleted from the system, or revised to a different configuration, therefore, reducing the number of snubbers remaining.
The only safety concern whicr. may arise as a result of the implementa-tion of this work unit is for the seismic qualification of the CA and TE system piping. This math model has been qualified for the present support / restraint (S/R) configuration. Likewise, it has been quali--
fled for the S/R configuration which will be in place after the procedure has been implemented. However, the interim configuration (with some deleteo snubbers removed and some still in place) has not been analyzed because the many possible combinations of S/R configura-tions would be difficult to analyze. For this reason, Design has stipulated that the piping systems must be modified within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in order to consider the piping affected by this system model to be operable during the implementation of this work unit. This procedure will provide the necessary controls to ensure that this work unit is finished in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit is excecded, this procedure ensures that the Technical Specification requirements are met.
If the piping affected by this modification is not required to be operable during the implementation of this work unit, no time limit applies. For the above reasons, it is concluded that no USQ is created by this procedure.
TN/1/A/1005/01/20A Initial Issue This procedure provides implementation instructions for NSM CN-11005 Rev. 1. This NSM modifies various piping system math models with the objective of reducing the number of mechanical snubbers required.
This procedure provides guidance for the removal of snubbers deleted from the CA system math model, This CA system math model also in-cludes several Main Feedwater (CF) system supports. These supports will either be deleted from the system,-or revised to a different configurati(q, therefore, reducing the number of snubbers remaining.
8 j
The only safety ccecern which may arise as a result of the implementa-tion of this work unit is fcr the seismic qualification of the CA and CF system piping. This math modEl has been qualified for the present support / restraint (S/R) configurati9n. Likewise, it has been quali-fled for the S/R configuration which will be in place after the procedure has been implemented. However, the interim configuration (with some deleted snuhbers removed and some still in place) has not been analyzed because the many possible combinations of S/R configura-tions would be difficult to analyze. For this reason, Design has stipulated that the piping systems must be modified within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in order to consider the piping affected by this system model to be operable during the implementation of this work unit. This procedure will provide the necessary controls to ensure that this work unit is finished in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit is exceeded, this procedure ensures that the Technical Specification requirements are met.
If the piping affected by this modification is not required to be operable during the implementation of this work unit, no time limit applies. For the above reasons, it is concluded that no USQ is created by this procedure.
TN/1/A/1005/01/19A Initial Issue This procedure provides implementation instructions for NSM CN-11005 Rev. 1. This NSM modifies various piping system math models with the objective of reducing the number of mechanical snubbers required.
This procedure provides guidance for the removal of snubbers deleted from the CA system math model. This CA system math model also in-cludes several CF system supports. These supports will either be deleted from the system, or revised to a different configuration, therefore, reducing the number of snubbers remaining.
The only safety concern which may arise as a result of the implementa-tion of this work unit is for the seismic qualification of the CA and l
CF system piping. This math model has been qualified for the present support / restraint (S/R) configuration. Likewise, it has been quali-tied for the S/R configuration which will be in place after the trocedure has been implemented. However, the interim configuration (rith some deleted snubbers removed and some still in placC, has not bten analyzed because the many possible combinations of S/R vonfigura-tians would be difficult to analyze. For this reason, Design has sticulated that the piping systems must be modified within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in ordor to consider the piping affected by this system model to be oper able during the implementation of this work uni".. This procedure 1 will orovide the necessary controls to ensure that this work unit is finished in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit is exceeded, this procedtre ensures that the Technical Specification requirements are met. i I
If the piping affected by this modification is not required to be i operable during the implementation of this work unit, no time limit 9
1 l
applies. For the above reasons, it is concluded that no USQ is f created by this procedure. l l
TN/1/A/1005/01/01A Initial Issue This procedure provides implementation instructions for NSM CH-11005 Rev. 1. This NSM modifies various piping system math models with the objective of reducing the number of mechanical snubbers required.
This procedure provides guidance for the removal of snubbers deleted from the Main Steam to Auxiliary Equipment (SA) System math model.
This SA system math model also includes several Main Steam (SM) and Auxiliary steam (AS) system supports. These supports will either be l
deleted from the system, or revised to a different configuration, therefore, reducing the number of snubbers remaining.
The only safety concern which may arise as a result of the implementa-tion of this work unit is for the seismic qualification of tho SA, SM, and AS system piping. This math model has been qualified for tiie present support / restraint (S/R) configuration. Likewise, it has been qualified for the S/R configuration which will be in place after the procedure has been implemented. However, the interim configuration (with some deleted snubbers removed and some still in place) has not been analyzed because the many possible combinations of S/R configura-tions would be difficult to analyze. For this reason, Design has stipulated that the pining systems must be modified within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in order to consider the piping affected by this system Lodel to be operable during the implementation of this work unit. This procedure will provide the necessary controls to ensure that this work unit is finished in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit is exceeded, this procedure ensures that the Technical Specification requirements are met.
If the piping affected by this modification is not requised to be operable during the implementation of this work unit, no time limit applies. For the above reasons, it is concluded that no USQ is created by this procedure.
TN/1/A/1123/00/01A Initial Issue This procedure provides implementation instructions for Nuclear Station Modification (NSM) CN-11123, Rev. O. This NSM installs a flush connection on the stuffingbox of each Unit 1 Nuclear Service Water (RN) pump. Also, a new stuffingbox and packing gland made of stainless steel will be installed on each pump. The new stuffingbox is drilled and tapped for the flush connection as well as the existing connections. These changes are intended to reduce build-up of corro-sion products and facilitate flushing of the stuffingbox upper bearing area. This procedure provides guidance for performing this modifica-tjon on RN Pump 1A.
Implementation of this procedure requires im Pump 1A to be isolated and removed from service. The RN supply to N Pump 1A upper bearing 10_
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i and stuffingbox will be isolated also. This procedure will be per-formed during a refueling outage. Technical specifications allow one train of RN to be out of service while operating in Mode 5 or below.
This procedure requires RN Train 1.B and Train B related components be operable while RH Pump 1A la out of service. This will ensure the RN system can perform its intended function during refueling by supplying cooling water flow to rystems and components necessary to maintain !
safe shutdown. I Prior to restoring M! Pump 1A to service, the stuttingbox and associ-ated piping will be leak couted to assure the system pressure boundary has been restored. Also, a flow balance w).11 be performed to estab-lish proper flow to the pump bearings and stuffingboc.
Based on the considerations above, no USQs are judged to bo involved i or created by this procedure. '
TN/1/A/1123/00/02A Initial Issue This procedure provides implementation instructions for NSM Cf-11123, Rev. O. This NSM installs a flush connection en the stuffingbox of each Unit 1 Nuc] ear Service Water (RN) pump. Also, a new stuffingbox and packing gland made of stainless steel will be installed on each .
pump. The now stuffingbox is drilled and tapped for the flush connec-tion as well as the existing connections. These changes are intended to reduce build-up of corrosion products and facilitate flushing of the stuffingbox upper bearing area. This procedure provides guidance for performing this modification on RN Pump 1B.
Implementation of this procedure requires RN pump 1B to be isolated and removed from service. The RN supply to PN Pump 1B upper bearing and stuf A'ingbox will be isolated also. This procedure will be per-formed during a refueling outage.
- Technical Specifications allow one train of RN to be out of service while operating in Mode 5 or below.
, This procedure requires RN Train 1A-and Train A related components be
( operable while RN Pump 1B is out of service. This will ensure the RN system can perform its intended function during refueling by supplying cooling water flow to systems and components necessary to maintain safe shutdown.
Prior to restoring RN Pump 1B to service, the stuffingbox and assoc 1-ated piping will be leak tested to assure the system pressure boundary-has been restored. Also, a flow-balance will be performed to estab-lish proper flow to the pump bearings and stuffingbox.
Based on the considerations above, no USQs are judged to be involved or created by this procedure.
l TN/2/A/0557/00/01A Change #1 l
i This change deletes step 8.6 which required a Control Room Ventilation (VC) retest prior to any control room firestop breach per Nuclear 11
Station Modification (NSM) CN-20557, Rev. O. No retest was required I because the system is considered operable.
The cables being pulled por this procedure are non-safety cables. One I of the cables will involve the penetration of a VC system firestop. !
Because of this, steps, notes, and warnings have been incorporated in j this procedure to ensure VC operability during the implementaelon of this procedure. The retest requirements of this procedure only require a retest after all work is complete. Thorofore, the retest at the start of the control Room work is not required because the VC system is considered operable. No other systems will be affected by this procedure change. Accordingly, this procedure change will not create any USQs.
TN/2/A/0557/00/02A Initial Issuo This procedure provides work activities necessary for the replacement of the existing Rochester Instruments Sequential Events Recorder on Unit 2 with a Dranetz System 22 model, per Nuclear Station Modifica-tion (NSM) CN-20557, Rev. O. It directs work activities for the cable terminations and wiring changes for the replacement for the existing system, as well as the installation of all components of the new Dranetz system. It also provides the necessary steps for wiring 27 new Operator Aid Computer (OAC) points which were originhlly wired via the Events Recurder.
For these work activities to take place, the entire Unit 2 Events Recorder will be removed from service. This will occur while init 2 is on line. The new OAC points will be wired to devices without any isolation being required. The Event Recorder is not Tech. Speg.
related. This system is used to perform monitoring of plant systems and post-trip analysis. The purpose of the new system is the same.
The new system will be installed and tested before it is returned to service. These tests will ensure that the new system will perform all of its intended functions per the design bases. There will be cables voided and removed from the Control Room. These activities will breach the firestops in the control Room. Because of this, fire.
watches are to be established when fire boundaries are breached. This procedure will involve the penetration of a VC system firestop.
Because of this, steps, notes, and warnings have been incorporated in this procedure to ensure VC operability during the implementation of this procedure. This procedure will not create a breach of any security boundaries. No other systems will be affected by this procedure. Accordingly, this procedure will not create any USQs.
MP/0/A/7650/109 Retype, Changes 0 to 0 Incorporated This procedure revision involves changes to the procedures for clean-ing and maintaining the station's containment spray (NS) heat exchang- i ers. The changes are to ensure that cleaning skid pressure iw cor-rectly maintained. This procedure is used to maintain the heat 12
l exchangers in optimum condition. This change does not involve any USQ.
MP/0/A/7650/56 Retype, Changes 0 to 7 Incorporated This item involves changes to the procedure for cleaning and maintain-ing the station's hoat exchangers. The change added sign-offs and independent verification to log the number of cleaning brushes insert-ed and removed from the heat exchangers. This procedure la used to maintain the *-5t exchangers in optimum condition. this chhnge does not involve arg USQ.
MP/0/A/7150/23 Re-type, Changes 0 to 7 Incorporated This procedure provides a standard method for the retorquing of reactor coolant pump bolts. This procedure rewrite clarifies snd assures that maintenance activities will return the pumps to as-de-signed conditions. Maintenance ./111 be performed on thess pumps when Tech. Specs, allow, per operation's procedures. These ar 'nns will ensure the pumpu' compliance t .h FSAR accident analysis. "crefore, no USQ exists.
EP/1/A/5000/2E2 Laletion This procedure, "High Containment Sump Level" is being deleted and replaced with " Incomplete Ct ainment Isolation", retype 0. This change is in response to the deletion of Emergency _ Procedure Guideline (EPG) FR-Z.2, Response to Containment Flooding, Revision 1 and the creation of Emergency Procedure Guideline FR-Z.2, Rebponse to incam-plete containment Isolation, Revision #0. The EPG changes are part of Duke Power's cemmitment to satisfy the requirements of NUREG-0737 Supplement 1. i ' intion of the Containment Sump Level Input to the Safety Parameter t . System (SPDS) and the inclusion of inputs for Containment Isolation were made under SPDS modification #3 (Program Request P880020.) Incorporating these changes will reflect the as-built condition of the SPDS and the current guidance in the Emer-co-ucy Procedure Guidelines.
EP/2/A/5000/2E2 Deletion This proceaure, "High Containment Sump Level" is be?.ng deleted and replaced with " Incomplete Containment Isolation", retype O. This change is in response to the deletion of Emergency Procedure Guideline (EPG) FR-Z.2, Response to Containment Flooding, Revision 1 and thu creation of Emergency Procedure Guideline FR-Z.2, Response to Incom-plete Containment Isolation, Revision #0. The E's changes are part of Duke Power's commitment ;o satisfy the requirements of NUREC-0737 Supplement 1. The deletion of the Containment Sump Level Input to the Safety Parameter Display System (SPDS) and the inclusion of inputa for Containment Isolation were made under SPDS mc dification #3 (Program 13 a
1
Request P880020.) Incorporating these changes will retlect the as-built condition of the SPDS and the current guidance in the Emer-cency Procedure Guidelines.
TN/1/A/1187/00/01A Initial Issue "his procedure provides guidance for the replacement of valves 1SA002 md 1SA005 under Nuclear Station Modification (NSM) CN-11187, Rev. O.
The gate valves 1SA002 and 1SA005 provide steam isolation for the auxiliary feodwater pump turbine. This application requi.js tight shutoff capability, and the valves must maintain this capability after monthly stroking against operating differential pressure. The exist-ing valves require excessive maintenance and still continue to leak by. This-modification replaces the above valves with new valves that are more suitable for this application.
Implementing this procedure wl;l require isolatior. f valves 1SA002 ar; ISA005. - The operations Grcup will coordinatad th, isolations necessary to implement this procedure. The valve .tolacement will be imple?ented during an outage and in Modes 4, 5, 6, an.;i No Mode.
Auxiliary Feedwater le not required to be operable in'these modes.
The Auxiliary Feedwater PumI Turbine will be cut of cervice during the modification. Testing for this NSM will be >erformed in accordance with the Post-Modification Testing Progr am t . tha station. The following testing is planned.
l Before Mode 3:
Pressure testing of new welda Steam supply line hee; tracing _ operability testing Calibration of 1SA002-and 1SA005 Instrumentation IWV testing of 1SA002 and 1SA005 In Mode 3:
Auxiliary Feedwater Pump Turbine Governor- response testing
- Auxiliary Feedwater Pump head curve ;esting Auxiliary Feedwater Pump IWP testing Auxiliary Feedwater System operability testing This testing will ensure the valves perform their intendad function.
Per this discussion, there are no USQs associated with japlementation of this procedure. .
TN/1/A/1145/CO/01A 771tial Issue ine purpose of the procedure is to provide guidance to add controls for valve 1NV039 ca Auxiliary Shutdown-Panel (ASP) 1ASPA. Presently, j valves 1NV039A and 1NV032B fail to the open position upon transfer of I control from the main control room to the auxiliary shutdown panels.
Nuclear Station Modification (NSM) CN-11145 adds controls.on Train A and Train B auxiliary shutdown panels for valves 1NV039A and 1NV032B respcctively. This will enable the operator to close one of the t
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valves, if necessary, when providing auxiliary prosaurizer spray from the ASP.
No work will begin on this procedure until Unit 1 is in Modes 5, 6, or No Mode. During this time, the equipment affected by this procedure is not required to be operable. No system will be prevented from performing any function important to safety while this work is being performed. All equipment affected by this procedure and the deJign intent of the rodification will be fully tested by Performance under Temporary Procedure TT/1/A/9200/57. This test will also be performed in Modes 5, 6, or No Mode. Accordingly, no USQs are involved with implementation of this procedure.
TN/1/A/1145/00/02A It tal Issue The purpose of the procedure is to provide guidance to add controls for valve 1NV032 on 1ASPB. Presently valves 1NV039A and 1NV032B fall to the open position upon transfer of control from the main control room to the auxiliary shutdown panels. NSM CN-11145 adds controls-on Train A and Train B auxiliary shutdown panels for valves 1NV039A and 1NV032B respectively. This will enable the operator to close one of the valves, if necessary, when providing auxiliary pressurizer spicy from the ASP.
No work will begin on this procedure until Unit 1 is in Modes-5, 6, or No Mode. During the implementation of this procedure, as a result of the electrical isolations, power-operated relief valves 1NCO32B and 1NC036B will fail closed. Per Technical Specification 3.4.9.3, these valves are required to be operable during Modes 5 and.6 with the-reactor vessel head on. To comply with this rnquircment, tnis proce-dure will be implemented with one safety valve (1NC1,_1NC2, or 1NC3) removed, or the Unit i reactor vessel head removed to provide a vent path of 4.5 square inches.
All other equipment affected by this procedure is not required to be operable. No system will be prevented from performing any function important to safety while this, work is being performed. All equipment affected by this procedure and the design intent of the modification will be fully tested by Performance under Temporary Procedure TT/1/A/9200/57. This test will al a be performed j ' Odes 5, 6, or No Mode. Accordingly, no USQs are- Jnvolved with imple: .tation of this procedure.
TN/1/A/0067/01/01A Initial Issue This procedure provides guidelines for installing portions of.the-Post-Accident Liquid Sampling (PALS) Panel piping used_to return samples taken from the reactor coolant'(NC) and residual heat removal (ND) sys#;2ms after an accident to the volume control tank._ These activities are a part of Nuclear Station Modification (NSM) CN-10067, ;
Rev. 1, This procedure will install portions of the return piping, l i
13
s along with the associated valves and hangers, and makes tie-ins to the Volume control Tank (VCT) sample recircalation line.
In order to perform this procedure, the VCT, along with the VCT sample recirculation piping, must be isolated and drained. The purpose of the VCT is to provide surge capacity for part of the NC expansion volume not accommodated by the pressurizer. This proceduro will be performed during No Mode when the NC system is out of service, and the additional surge capacity is not required. This will allow the VCT anu the sample recirculation piping to be drained.
After completion of this procedure, the VCT sample recirculatban piping pressure boundary will be restored, and the VCT can be returned to service to provide surge capacity for the NC system. The piping will be visually inspected at system temperature and pressure to verify piping integrity. Based on the above discussion, no USQ is involved.
TN/1/A/0753/00/03A Initial Issue l This procedure provides guidelines for performing activities under Nuclear Station Modification (NSM) CN-10753, Rev. O. This NSM pro-vides for the replacement of the existing RTD bypass manifold system with fast-response, narrow-range thermowell type RTDs installed in the existing hot and cold leg scoops of each Reactor Coolant (NC) loop.
This procedure provides guidance for mechanical activities associated with deleting the existing a 7 bypass manifold piping, associated l valves, hanger components, ar. rupture restraints.
In order to perform this proc' tre, the NC system, along with the associated RTD system, will bc isolated and drained to allow the RTD piping to be cut loose from the NC system. These activities will be performed during the NC system draindown for Reactor Vessel Head removal, which is a rcutine maintenance activity performed during unit outages. After the RTD piping is cut loose from the NC system, the NC system pressure boundary will be restored by the addition of temporary caps and flanges at the RTD piping connections on each NC loop, and l
' the NC system will be refilled to support fuel load activities. After the NC system is filled, the caos and flanges will be visually in-spected by operations persor.nel to verify no leakage,=and to ensure that NC system integrity is acceptable to support fuel load activi-ties. In the event of leakage from the flanged and capped.RTD piping, make-up capabilities are available from the Refueling Water Storage Tank to provide borated water to maintain NC inventory. Also, failure of the temporary caps and flanges would result in a snall break LOCA.
However, this type of accident has been addressed in the accident analysis for Catawba. By restoring the NC system pressure boundary, the NC system can perform its intended function in Modes 5 and 6, as required.
After the core is unloaded, the NC system will be drained to allow machining of the hot leg scoops to provide for installation of the thermowell type RTDs. These activities will be performed by 16 i
l l
I Westinghouse under procedure MPII 2.7.2 DCP-1. This work will be )
completed and the NC pressure boundary restored prior to refill of the j NC system. This will restore the NC pressure boundary as required by .
Tech. Specs, and the FSAR. Functional testing of the affected por-tions of the NC pressure boundary will be performed during Mode 3 of I
Unit i startup to verify NC system integrity.
i Based on the above discussion, no USQ is involved with or created by tnis procedure.
MP/0/A/7150/04 Re-type, Changea 0 to 9 Incorpcrated This orocedure is for corrective maintenance of the component cooling iumps, to ensure chct Lne pumps are niaintained in their original condition. There are :wo redundant trains of component cooling, which allows one train to be removed from service for maintenance activi-ties. This procedure provides instruction, based upt aanufacturer's literature, to perform corrective maintenance, and return the pump to its as-built condition. Performance testing is required after any major maintenance activity to ensure that the pump meets its desired performance. Thus, no USQ is created by this procedure. .
MPII 2.7.2 DCP-1 Initial Issue This procedure provides implementation instru' ' lons for Nuclear Station Modification (NSM) CN-10753, Rev. O. The purpose of this procedure is to provide details of mechanical installation of thermowells for individual resistance temperature detectors into the prin.ary system loop piping to achieve RTD bypass elimination installa-tion. It includes the removal of RTD bypass piping by usa Jf conven-t tional machining, Metal Disintegration Machine (MDM), welding, and the l nondestructive examinations to the hot leg scoops, crossover leg 3" l
RTD manifold return nozzles, cold leg 2" RTD manifold connection, and new thermowell boss penetrations in the Loop hot leg and each cold leg pipe for elimination of the present RTD bypass system.
In order to perform this procedure, the Reactor Coolant (NC) system, along with the associated RTD system, must be isolated and drained and all fuel removed from the core. This Mill be ; performed during No Mode, which will allow the NC loops to ch drained to a level low enough to allow the required machining vnd welding on the NC loops.
Technical Specifications allow the NC systcn to be removed from service and the NC loops drained to suppc1% anit outage activities.
After the NC loops have been machined and the required nozzle caps and bosses installed, thermowells will be installed at each RTD location to restore the NC system pressure coundary. This will allow the NC system to be refilled.
The thermowells, bosses, and nozzle caps were shop fabricated and inspected to meet the requir ements to ASME Section III, 1983 Edition.
All installations and inspections will be performed ir accordance with ASME Code Section XI requirements and the Catawba Quality Assurance 17
program applicable to safety related equipment. All welds will be inspected at NC system temperature and pressure during Mode 3 to verify NC system integrity. Failure of the thermowells would result in a small break LOCA. However, this type of accident has been addressed in the accident analysis for Catawba. By restoring the NC system pressure boundary, the NC system can perform its intended function during fuel load activities and subsequent power escalation as required by Technical Specification. The integrity of the NC pressure boundary is maintained by adhering to the applicable ASME code sections and the NRC general design criteria. Tho pressure retaining capability and fracture prevention characteristics of the piping is not compromised by implementation of-this procedure. Based on the considerations above, no USQs are created by this proceduro.
PT/2/A/4200/09A Change #66 This restricted change is being made to allow verification of slave relay contact status for valves 2Rdd39A and 2RN841B to meet testing requirements, without actual valve motion. These valves are tagged in their safety position (closed) with the piping drained for freeze protection. The procedure is also being changed to verify the slave relay contact status for 2NV11A and to keep 2NV13A from closing when it is tested. 2NV11A is presently tagged closed (its safety position) due to excessive stroke time. 2NV13A is being kept open to prevent loss of letdown flow. When the slave relay dest device is installed on 2NV13A, the valve is inoperable since it is not capable of closing on a Phase A Containment Isolation signal. 2NV15B will remain capable of closing to isolate the penetration. Based on the above, no USQ is created by this procedure change.
MP/0/A/7300/03 Re-type, Changes 0 to 1 Incorporated The purpose of this procedur; is to perform scheduled preventive maintenance inspection and servicing on diesel generator starting air system dryers and associated equipment. This change adds two new sections to this procedure. These new' sections provide a method of procedural documentation and guidance for removal, replacement, and corrective maintenance of the dryer backpressure regulators. Incorpo-ration into this procedure is to consolidate inspections and activi-ties under one procedure for this component. Steps for veri *ying the set pressure or resetting the backpressure regulator-pressure have been added. This addition verifies or restores the backpressure regulator back to its original material and operating condition. This change does not involve a USQ.
PT/1/A/4200/01C Re-type, Changes 0 to 54 Incorporated This procedure is performed to measure the leak rates of all contain-ment isolation valves required by Catawba FSAR table 6.2.4-1 and to satisfy Technical Specification Surveillance Requirements. There are two changes being made. The first involves altering the valvo lineups 18
l l to satisfy containment closure requirements. Closure is always assured by either the inboard or outboard containment isolation valve i and applicable vents and drains. The validity of the type C test is (
not compromised.
j The second change revise 7 the acceptance criteria to higher values.
l These changes are based on documented owner specified leakage rates l which will now be utilized for type C leak rate testing acceptance criteria. It does not affect the test method or procedure execution in any way.
For the above reasons, this procedure change does not represent a USQ.
PT/0/A/4450/08 Change #10 This procedure was changed to obtain the control room pressure and I pressurization flow in all three outside air intake alignments. This chcnge will allow measuring the control room pressure and pressuriza-tion flow with either intake isolated and both open to ensure proper positive pressure in the control room and proper pressurization Ilow for all outside air intake alignments. Another changa was made to allow measuring the control room recirculation flow and the cross train flow. Measurement of the cross train flow will provide a check to ensure adequate scaling of dampers on the opposite train. These measurements are being added for information only. The last change was made to enable using a different range manometer for measurement of flows.
l l No new operating alignments were created by this change tisides the j isolation of one outside air intake, and an intake is normally isolat-ed whenever the radiation monitor associated'with the intake is inoperable. The measurement of the new flows are taken while the system is operating in its normal mode and will not affect the system operation. Therefore, a USQ does not exist.
PT/2/A/4600/05E Change #3 This procedure change added the Design Starruo and Operational Re* t as a reference document. Steps were added and/or changed to calc. 4ee the now Intermediate Range currents based 4.n predicted data and the first full power full core nap from the previous cycle, rather than the last map from the previous cycle. Tne beginning of cycle values will be more representative of core conditions than when the old F3tpoints were determined. The precautions have been changed to indicate that the intermediate rangt tripo should not be relied upon until adjustments are made to the setpoints. Additional editorial changes were made.
l The procedure is perfert .a before the beginning of each fuel cycle to ensure that the margin to Peactor Trip setpoints for the power ranges and intermediate ranges is not adversely affected by changes in the core radial power distribution as a result of the new core loadinc.
19
The procedure only affects the Nuclear Instrumentation (ENB) system. I l Calibration data for each power range and intermediate range channel I is calculated based on the change in radial power distribution as determined from predictions for the new cycle and measurements from the last cycle. Calibrations are performed using approved Instrument Procedures by qualified Instrument and Electrical (IAE) personnel.
There is no change to the function of the ENB system.- Changes made to ;
the procedure enhance the calculation of the Intermediate Range Setpoint data. No USQ is created by this change.
l l MP/0/B/7700/05 Retype This procedure controls maintenance activities on the main and aux 11-lary condensers at Catawba. The major change to this procedure adds sign-offs and independent verification to count and verify that brushes are not left in the condensors after cleaning activities. No USQ is created by this procedure change.
OP/1/A/6200/04 Change #44 The purpose of this change is to increase the value of " normal" Residual Heat Removal System flow from 3000 gpm to batween 3300 and 3500 gpm, and to make the steps agree with the system elignment.
The change is a result of a meeting between Westinghouse and Design Engineering. No USQ is created by this change..
TN/1/A/1051/00/03A Initial Issue This procedure provides implementation instructions for Nuclear Station Modification (NSM) CN-11051 Rev. O. This NSM will install in-line connectors in the RG-11/U triaxial cables near the bottom of the Nuclear Instrumenration cabinets. RG-59 triaxial cable ansemblies will be installed from the in-line connectors'to the cable termination points in the cabinets. This procedure will control installing the in-line connectors and RG-59 cable for Intermediate Ranga Detector N35.
This procedure will be performed during the refueling obtage when Intermediate Range Detector N35 is not required to be operable per Technical Specification-3/4.3.1. No systems will be prevented from performing any function important to safety while this work is being-done. For these reasons; no USQ exists.
TN/1/A/1051/00/04A Initial Isrue This procedure pro'.*idas implcmentation instructions for NSM_CN-11051 Rev. O. This NSM will install in-line connectors in the RG-11/U triaxial cables near the bottom of the Nuclear Instrumentation'cabi-
.' nets. RG-59 triaxial cable assemblies will be installed from the in-line connectors to the cable termination points in the cabinets.
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i This procedure will control installing the in-line connectors and RG-59 cable for Intermediate Range Detector N36.
This procedure will be performri during the refueling outage when Intermediate Range Detector N36 is not required to be operable per Technical Specification 3/4.3.1. No systems will be prevented from performing any function important to sr.fety while this work is being done. For these reasons, no USQ exist.s.
TN/1/A/1051/00/05A Initial Issue This procedure provides implementation in:=tructions for NSM CN-11051 Rev. O. This NSM will install in-line connectors in the RG-11/U triaxial cables near the bottom of the Nuclear Instrumentation cabi-nets. RG-59 triaxial cable assenblies will be installed frc the in-line connectors to t* cable -.erminstion points in the caulnets.
This procedure will ca: ' 1 instti ng th; in-line connectors and RG-59 cable for Power ..ange Detector N41.
This procedure will be performed during the refueling outage when Power Range Detector N41 is not required to be operable per Technical Specification 3/4.3.1, No systems will be prevented from performing any function important oc safety while this work is being done. For these reasons, no USQ exists.
TN/1/A/1051/00/06A Initial Issue This procedure provides implementation instructions for NSM CN-11051 Rev. O. This NSM will install in-line connectors in the RG-11/U triaxial cables near the bottom of the Nuclear Instrumentation cabi-nets. RG-59 triaxial cable assemblies will be installed from the in-line connectors to 'he cable termination points in the cabinets.
This procedure will control installing the in-line connectors and RG-59 cable for Power Range Detector N42.
This procmlu > will be performed during the refueling outage when Power Rania ,x tector N42 is not required to be operable per Technical Specificatn c 3/4.3.1, No systems will be prevented from performing any function important to safety while this work.is being done. For these reasons, no USQ exists.
TN/1/A/1051/00/07A Initial Issue This procedure provides implementation instructions for NSM CN-11051 Rev. O. This NSM will install in-line connecters in the RG-11/U triaxial cables near the bottom of the Nuclent Instrumentation cabi- 4 nets. RG-59 triaxial cable assemblies will be installed from the l in-line connectors to the cable termination points in the cabinets.
This procedure will control installing the in-line connectors and RG-59 cable for Power Range Detector N43.
l 21 l
l l
This procedure will be performed during the refueling outage when Power Range Detector N43 is not required to be operable per T3chnical Specification 3/4.3.1. No systems w!11 be prevented from performing any function important to safety while thPs work is being done. For these reasons, no USQ exists, i l
TN/1/A/1051/00/08A Initial Issue ,
This procedure provides implementation instructions for NSM CH-11051 Rev. O. This-NSM will install in-line connectors in the RG-11/U triaxial cables near the bottom of the Nuclear Instrumentation cabi-nets. RG-59 triaxial cable assemblies will be installed from the in-line connectors to the cable termination points in the cabinets.
This procedure will control installing the in-line connectors and RG-59 cable for Power Rango Detector N44.
This procedure will be performed during the refueling outage when-Power Range Detector N44 is not required to be operable per Technical Specification 3/4.3.1. _No systems will be prevented from performing any function important to safety while this work is being done. For these reasons, no USQ exists.
TN/1/A/0753/00/02A Initial Issue This procedure gives implementation instructions for Nuclear Station Modification (NSM) CN-10753 Rev. O. This NSM provides for the re-placement of the existing RTD Bypass Manifold system with fast-re-sponse, narrow range thermowell-type RTDs installed in the existing hot and cold leg scoops of each Reactor Coolant loop. This procedure provides guidelines for performing electrical and instrument work associated with this modification.
No work will begin on this procedure until Unit:1.is in Mode 5, 6, or No Mode. During the implementation of this procedure, as a result of the electrical isolations powering down the Process Control-Cabinets I t
through IV, the Low Temperature overpressurization Protection '(UDOP) will be inoperable. Per Technical-Specification 3.4.9.3, this func-tion is required to be operable during Modes 5 and 6 with the Reactor vessel Head on. To comply with this requirement, this procedure'will be implemented with one safety valve (INC1, INC2, or 1NC3) removed, or when the Reactor Vessel Head is removed to provide a vent path-of 4.5 square inches.
All other equipment affected by this procedure is not required to be
- operable during Modes 5, 6, or No Mode.- No system will be prevented i from performing any function luportant~to safety while this work is L being performed. All equipment affected by this procedure and-the-design intent of the modification will be completely tested by this
, procedure. Construction Maintenance Department craft will perform a functional on equipment affected by the isolaticns only in Mode 5,-6, or No Mode. Nuclear Production Instrument and Electrical (IAE) personnel will perform applicable calibrations of T/Tavg protection 22
for all four Reactor Coolant channels, of steam flow and feed flow loops, and a cold Shutdown Response Time Tust for the new RTDs in-stalled per this modification in Modes S, 6, or No Mode. During Mode 3, another Response Time Test and a Cross Calibration of the RTDs will be performed by IAE to ccmplete the functional testing of this modifi-cation. Based on the above, no USQ exists. !
PT/1/A/4200/55 Retype, Changes 0 to 3 Incorporated This change added section 12.20 to group-test Auxiliary Feedwater (CA) check valses icA8, ICA10, and ICA12 simultaneously. Steps were added
, to requiru Auto Open Valves, after closure, to have their control room switches placed in the " CLOSED" position. These switches have three positions -- OPEN, CLOSED, and AUTO. A spurious signal could cause a valve to open should it be left closed, but in AUTO mode. Additional "
changes were made to conform the format to the Procedure Writer's Guide.
This procedure measures leakage through check valves 1CA8, ICA10, and.
ICA12. These check valves provide the boundary to prevent safety grade water from the Nuclear Service Water (PV) system from spilling through a possible pipe break upctream of the check v'Ives and not be g available to the CA pumps. This procedure will be performed in either Mode 4, 5, 6, or No Mode. The CA system is not requicci to be opera-( ble when this procedure is performed.- Prior to performing the valve lineup for the test, the breakers for the pump motors will be racked out open and the isolation valves for the steam to the CA pump turbine will be closed._ The valve lineup will isolate suction to all CA purps. Utilization of the isolation measures of this procedure will ensure that the CA pumps will be unable to start during the perfor-mance of the procedure. This will prevent the pumps from starting without a suction source. No USQ is created by this procedure.
MP/0/A/7400/36 Change #5 and Re-type This re-type is being upgraded to the new proceoure format. The procedure has been changed from a several size, same manufacturer pump procedure back to a single size pump procedure, using the originally approved procedure. This was done to eliminate the confusion in the disassembly and assembly of the pump and drive _ assembly. The proce-dure title was changed to denote the single use application. Two steps were adde; to enhance the already existing visual inspections.
They are beitig added to the manufacturer's instruction manual. Their -
function is to evaluate material condition and to ensure excellent material condition remains. Two enclosures have been upgraded to coincide with component installed configuration. These changes do not represent a USQ.
TN/1/A/1214/00/01A Initial Issue 1
23 e,
This procedure will replace steam generator ID Blowdown (BB) Isolation Valve 1BB008A with a new gate valve, item #06H-210. The flow path from Steam Generator 1D will be out of service during the replacement of 1BB00LA. The containment isolation valves downstream of 1BB008A i
will be used to satisfy Tech. Spec. requirements for control of penetrations that have direct access to outside atmosphere for con-tainment integrity / closure during core alterations, fuel movement, and Reactor Coolant (NC) system mid-loop operations.
Additionally, the work associated with 1 PENT 0115 will act break the pressure boundary of the penetration. Breakers F04B in 1EMXS and F04B in IEMXK will be opened and Red Tagged to ensure valve 1BB008A and plug #7 in 1 PENT 0115 are de-energized for electrical work. Breaker F04B in 1EMXK also supplies power to the motor operator for contain-ment isolation valve INM190A. Operations has responsibility for this valve to ensure containment integrity / closure.
BB is not required to be operable in Modes 4, 5, 6, or No Mode. Red Tags will be removed and breakers F04B in 1EMXS and F04B in 1EMXK will be closed for valve set up, verification of remote position indica-tion, Motor Operated Valve testing, and stroke time testing.
Upon completion of electrical work, a Leak Rate Test will be performed on 1 PENT 0115 to ensure its pressure boundary has been maintained.
Testing of the new valve will consist of performing Motor Operated Valve (MOV) testing, and verifying all remote position indication and status light indications. Stroke time tests will be performed prior to Mode 3, and again in Mode 3. A differential pressure test will also be performed on Valve 1BB008A. Hydrostatic and appropriste Non-Destructive Examinations will be performed by station and Quality Assurance Procedures. Based on the above discussion, it is determined that a USQ does not exist.
TN/1/A/1214/00/02A Initial Issue This pro 7edure will replace steam generator 1B Blowdown Isolation Valvo 1BE 019A with a new gate valvo, item #06H-210. The flow path from Steam Generator 1B will be out of service during the replacement of 1BB019A. The containment isolation valves downstream of 1BB019A will be used to satisfy Tech. Spec. requirements for control of penetrations that have direct access to outside atmosphere for con-tainment integrity / closure during core alterations, fuel movement, and NC system mid-loop operations.
Additionally, the work associated with 1 PENT 0115 will not break the pressure boundary of the penetration. Breakers F04C in 1EMXS and F04B in 1EMXM will be opened and Red Tagged to ensure valve 1BB019A and plug #5 in 1 PENT 0115 are de-energized for electrical work. Breaker F04B in 1EMXK also supplies power to the motor operator for contain-ment isolation valve 1NM190A. Operations has responsibility for this i valve to ensure containment integrity / closure.
24 3
BB is not required to be operable in Modes 4, 5, 6, or No Mode. Red Tags will be twmoved and breakers F04C in 1EMXS and F04B in 1EMXK will be closed for valve set up, verification of remote position indica-tion, Motor Operated Valve testing, and stroke time testing.
Upon completion of electrical work, a Leak Rate Test will be performed on 1 PENT 0115 to ensure its pressure boundary has been maintained.
Testing ot the new valve will consist of performing MOV testing, and verifying all remote position indication and status light indications.
Strcke time tests will be performed prior to Mode 3, and again in Mode
- 3. A differential pressure test will also be performed on Valve 1BB019A. Hydrostatic and appropriate Non-Destructive Examinations will be performed by station and Quality Assurance Procedures. Based on the above discussion, it is determined that a USQ does not exist.
TN/1/A/1214/00/03A Initial Issue This procedure will replace steam generator 1A Blowdown Isolation Valve 1BB056A with a new gate valvo, item #06H-210. The flow path from Steam Generator 1A will be out s' service during the replacement of 1BB056A. The contain' ment isolation valves downstream of 1BB056A will be used to satisfy Tech. Spec. requirements for control of penetrations that have direct access to outside atmosphore for con-tainment integrity / closure during core alterations, fuel movement, and NC system mid-loop operations.
Additionally, the work associated with 1 PENT 0115 will not break the pressure boundary of the penetration. Breakers FOSA in 1EMXS and F04B in 1EMXK will be opened and Red Tagged to ensure valve 1BB056A and plug #7 in 1 PENT 0115 are de-energized for electrical work. Breaker F04B in 1EMXK also supplies power to the motor operator for contain-ment isolation valve 1NM190A, Operations has responsibility for this valve to ensure containment integrity / closure.
BB is not required to be operable in Modes 4, 5, 6, or No Mode. Red Tags will be removed and breakers F05A in 1EMXS and F04B in 1EMXK will be closed for valve set up, verification of remote position indica-tion, Motor Operated Valve testing, and stroke time testing.
Upon completion of electrical work, a Leak Rate Test will be performed on 1 PENT 0115 to ensure its pressure boundary has been maintained.
Testing of the new valve will consist of performing MOV testing, and i verifying all remote position indication and status light indications.
Stroke time tests will be performed prior to Mode 3, and again in Mode 3 A differential pressure test will also be performed on Valve 1BB056A. Hydrostatic and appropriate Non-Destructive Examinations will be performed by station and Quality Assurance Procedures. Based on the above discussion, it is determined that a USQ does not exist.
TN/1/A/1214/00/04A Initial Issue 25
. . . . .~
This procedur s will replace steam generator 1C Blowdown Isolation Valve 1BB069A with a new gate va;ve, item /06H-210. The flow path from Steam Generator 1C will be out of service during the replacement of 1BB060A. The containment isolation valves downstream of 1BB060A will be used to satisfy Tech. Spec. requirements for control of penetrations that have direct access to outside atmosphere for con-tainment integrity / closure during core alterations, fuel movement, and NC system mid-loop operations.
Additionally, the work associated with 1 PENT 0115 will not break the pressure boundary of the penetration. Breakers F05B in IEMXS and F04B in 1EMXK will be opened and Red Tagged to ensure valve 1BB060A and plug #7 in 1 PENT 0115 are de-energized for electrical work. Breaker F04B in 1EMXK also supplies power to the motor operator for contain-ment isolation valve 1NM190A. Operations has responsibility for this valve to ensure containment integrity / closure.
BB is not required to be operable in Modes 4, 5, 6, or No Mode. Red Tags will be removed and breakers F05B in 1EMXS and F04B in 1EMXK will
&2 closea Tor valve set up, verification of remote position indica-tion, Mot r Operated Valve testiny, and stroke time testing. ,
Upon completion of electrical work, a Leak Rate Test will be performed on 1 PENT 0115 to ensure its pressure boundary has been maintained.
Testing of the new valve will consist of performing MOV testing, and veritying all remote position indication and status light indications.
Stroke time tests will be performed prior to Mode 3, and again in Mode
- 3. A differential pressure test will also be performed on Valve ;
1BB060A. Hydrostatic and cppropriate Non-Destructive Examinations will be performed by stat -a and Quality Assurance Procedures. Based on the above discussien, it is determined that a USQ does not exist.
l TN/1/A/1122/00/01A Initie.1 issue This procedure provider implementation instructions for Nuclear Station Mcdification (NSM) CN-11122 Rev. O. This NSit adds a steam siphon system between the Auxiliary Feedwater Turbine Driven Pum e (CATDP) and the Auxiliary Feedwater Turbine Driven Pump Sump. This j procedure provides guidance for the installation of a Jiquia eductor on the Auxiliary Feedwater Turbine Driven Pump lube oil cooler line.
This eductor will utilize the lube oil cooler cooling water discharge as the motive fluid to entrain the turbine exhaust condensate. Other related piping and instrument modifications are also covered by this procedure.
The work associated with this procedure is to be performed with Unit 1 in Modes 4, 5, 6, or No Mode. The CATDP is not required to be opera-ble during this period of time. The isolations required to perform this modification are only associated with the Auxiliary Feedwater Pumps and the CATDP turbine exhaust system. These systemn are not required to be operable during Modes 4, 5, 6, or No Mode. The fire barrier for the CATDP will be penetrated during the installation.
Since fire barriers are required to be operable at all times, the 26 n --
requiraments of Tech. Spec. 3.7.11 will be met until the fire barrier is returned to service. This procedure, along with detailed work j eentrol procedures, will adequately govern the return to service of all components associated with this modification. Based on the above discussj on, a USQ is not created by this procedure.
TN/2/A/1122/00/02A Initial Issue This procedure provides guidunce for performing the post-modification testing associated with NSM CN-11122. Thia NSM adds an eductor on the Auxiliary Feedwater Turbine Driven Pump lube oil cooler line. This procedure theroughly governs the following: (1) installation and removal of test equipment, (2) test and normal operation positioning of valves, (3) removal and return to service of the auxiliary feeawater turbine driven pump sump pumps, (4) monitoring of the auxiliary feedwater turbine driven pump sump level during test, and (5) meeting the test acceptance criteria values for the affected equipment.
All systems and components associated with this procedure v 11 be considered operable and ready for service following completion of this procedure, based on the controls provided by this procedure, other station procedures, and obtaining appropriate test acceptance criteria values. Based on the above discussion, there are no USQs associated with this procedure.
TN/1/A/ J 186/00/01A Initial '.esue This procedure provides implementation instructions for Nuclear Station Modification (NSM) CN-11186, Rev. O. This modification will replace Auxiliary Feedwater (CA) Discharge Isolation Valve ICA66B with a new gate valve. This is being done to assure'that there is suffi-cient torque available to meet design requirements. Also, new valve ICA288 is being added between the bonnet for valve ICA66B and the upstream process piping. This will prevent bonnet _overpressurization on ICA66B. This procedure provides guidance for the remov-l a.'/rcinstallation of valve 1CA66B, and the installation of new valve iv.288.
A blank is being installed in 1 CAFE 5090, by procedure TN/1/**1186/00/01A to ensure that the CA flow path to Steam Generator l
1A is isolate; during the replacement of valve ICA66B. This blank will also be urad to maintain containment integrity / closure during l core alterations, fuel movemnot, and NC System mid-loop conditions.
I Breaker F08A in 1EMXL will bc opened and Red Tagged under Block Tagout CA1VLV to ensure valve 1CA66B is de-energi 9d for electrical work.
This does not present a concern for safet; operation of the CA system because the system is not required to be operable in Modes 4, 5, 6, or No Mode. The Red Tag for breaker F08A will be lifted from block tagout CA1VLV for valve sat up, verification of remote position indication, motor operator valve (MOV) testing, and stroke time testing.
27
i Testing of the new valve will consist of performing MOV testing and i
verifying all remote position indications. Stroke time tests will be l performed prior to mode 3, and again in mode 3. A differential l pressure test will not be performed on valve 1CA66B due to it being i one of the discharge isolation valves for the turbine driven CA pump, j and sufficient differentia) pressure cannot be obtained. Hydrostatic and appropriate Non-Destructive Examinations will be performed by station and Quality Arsurance Procedures. Based on the above discus-sion, a USQ does not exist.
'N/1/A/1186/00/02A Initial Issue This procedure provides implementation instructions for NSM CN-11186, Rev. O. This modification will replace ?.ux1 iary Feedyater Discharge Isolation Valve ICA62A with a new gat) velve. This is being done to assure that there is sufficient torque available to meet design requirements. Also, new valve 1CA287 is being added between the bonnet for valve 1CA62A and the upstream process piping. This will prevent bonnet overpressurization on 1CA62A. This procedure provides guidance for the removel/ reinstallation of valve ICA62A, and the installation of new valve ICA287.
A blank is being installed in 1 CAFE 5090 by this procedure to ensure that the CA flow path to Steam Generator 1A is isolated durine the replacement of valve ICA62B. This blank will also be used t maintain co.itainment integrity / closure during core alterations, fuel movement, i and NC System mid-loop conditions. Breaker F08A in 1EMXK will be opened and Red Tagged under Block Tagout CA1VLV to ensure valve ICA62B it. de-energized for electrical work. This does not present a concern fsr safe operation of the CA system because the system is not required to be operable in Modes 4, 5, 6, or No Mode. The Red Tag for breaker F08A will be lifted from block tagout CA1VLV for valve set up, verifi-cation of remote position indication, motor operM or valve (MOV) testing, and stroke time testing.
Testing of the new valve will consist of performing MOV testing and -
verifying all remote position indications. Stroke time tests will be performed prior to rode 3, and again in mode 3. A differential pressure test will also be performed on valve 1CA62B.- Hydrontatic and appropriate Non-Destructive Examinations will be performed by station' and Quality Assurance Procedures. Based on the above discussion, a USQ does not exist.
l TH/1/A/1186/00/03A Initial Issue This procedure provides implementation instructions for NSM CN-11186, Rev. O. This modification will replace Auxiliary Feedwater Discharge Isolation Valve 1CA58A with a new gate valve. This is being done to assure that there is sufficient torque available to meet design requirements. Also, new valve 1C3286 is being added between the bonnet for valve ICA58A and the u; stream process piping. This will prevent bonnet overpressurizatior on 1CA58A. This procedure provides 28
guidance for the removal / reinstallation of valve ICA58A, and the installation of new valve 1CA286.
A blank is being installed in 1 CAPE 5100 by this procedure to ensure that the CA flow path to Steam Generator 1B is isolated during the replacement of valve ICA58A. This blank will also be used to maintain containment integrity / closure during core altorations, fuel movement, and NC System mid-loop conditions. BIeaker R03C in 1EMXI will be opened and Red Tagged under Block Tagott CA2VLV to ensure valve ICA58A is de-energized for electrical work. This does not present a concern for safe operation of the CA system because the system is not required to be operable in Modes 4, 5, 6, or No Mode. The Red Tag for breaker
, R03C will be lifted from block tagout CA2VLV for valve set up, verifi-cation of remote position indication, motor operator valve (MOV) testing, and stroke time testing.
Testing of the new valve will consist of performing MOV testing and verifying all remote position indications. Stroke time tests
- ill be performed prior to mode 3, and again in mode 3. A differential pressure test will also be performed on valve ICA58A. Hydrostatic and appropriated Non-Destructive Examinations will be performed by station and Quality Assurance procedures. Based on the above discussion, a UCQ does not exist.
4 TN/1/A/1186/00/04A Initial Issue This procedure provides implementation Instrattions for NSM CN-11186, Rev. O. This modification will rep?. ace Auxiliary Feedwater Discharge Isolation Valve 1CA54B with a new gate valve. This is being done to '
assure that there is sufficient torque available to meet design requirements. Also, new valve ICA285 is being added between the bonnet for valve 1CA54B and the upstream process piping. This will >
prevent bonnet overpressurization on 1CAG4B. This procedure provides guidance for the removal / reinstallation of valve ICAS4B, and the installation of new valve 1CA285.
A blank is being installed in 1 CAFE 5iOO by procedure TN/1/A/1186/00/03A to ensure that the CA flow path to Steam Gsnerator 1B is isolated during the replacement of valve 1CA54B. This blank will also be used to maintain containmant integrity / closure during core alterations, fuel movement, and NC System mid-loop conditions.
Breaker Rv3C in 1EMX'd will be opened and Red Tagged under Block Uagout CA2VLV to encure valve ICA54B is de-energized for electrical work.
This does not present a concern for safe operation of the CA system because the system is not required to be operable in Modes 4, 5, 6, or No Mode. The Red Tag for breaker R03C will be lifted from block tagout CA2VLV for valve set up, verification of remote position indication, motor operator valve (MOV) testing, and stroke time j testing, t
Testing of the new valve will consist of performing MOV testing and verifying all remote position ;ndications. Stroke time tests will be p erformed prior to mode 3, and again in mode 3. A differential 29
- v x , J
E pressure test will not be performed on volve 1CAS4B because it is one of the discharge isolation valves for the turbino driven CA pump and sufficient differential pressure cannot be obtained. Hydrostatic and appropriate Non-Destructive Examinations will be performed by station and Quality Assurance Procedures. Based on the above discussion, a USQ does not exist.
TN/1/A/1186/00/05A Initial Issue This procedure provides implementation instructions for NSM CN-11186, Rev. O. This modification will replace Auxiliary Feedwater Discharge Isolation Valve ICA50A with a new gate valvo. This is being done to assure that there is sufficient torque available to meet design
, requirements. Also, new valve 1CA284 is being added between the bonnet for valve ICA50A and the upstream process piping. This will prevent bonnet overpressurization on 1CA50A. This procedure provides guidance for the removal / reinstallation of valve 1CA50A, and the installation of new valve ICA284.
A blank is being installed in 1 CAFE 5110 by this procedure to ensure that the CA flow path to Steam Generator 1C is isolated during the replacement of valve ICA50A. This blank will also be used to maintain containment integrity / closure during core alterations, fuel movement, and NC System mid-loop conditions. Breaker R04A in 1EMXS will be opened and Red Tagged under Block Tagout CA3VLV to ensure valve 1CA50A is de-energized for electrical work. This does not present a concern for safe operation of the CA system because the system is not required to be operable in Modes 4, 5, 6, or No Mode. The Red Tag for breaker F04A will be lifted from block tagout CA3VLV for valve set up, verifi-cation of remote position indication, motor operator valve (MOV) testing, and stroke time testing.
Due to physical interferences, breaker F01C in 1MXK, sliding links H-3
, and H-4 in 1TBOX0100 will be opened, and cables 1CA541 and 1CF555 disconnected from valves 1CASV1512 and 1CF51, respectively. Solenoid valve 1CASV1512 allows remote operation for valve ICA151, and will fail to its closed position when the breaker is opened. Cables will be reconnected upon installation of valve ICA50A, re-energized, and stroke time tested.
Tecting of the new valve will consist of performing MOV testing and varAfying all remote position indications. Stroke time tests will be performed prior to mode 3, and again in mode 3. A differential pressure test will not be performed on valve ICA50A because it is one of the discharge isolation valves for the turbine driven CA pump and sufficient differential pressure cannot te obtained. Hydrostatic and appropriate Non-Destructive Examinatio:.s will be performed by station and Quality Assurance Procedures. Based on the above discussion, a USQ does not exist.
TN/1/A/1186/00/06A Initial Issuo
)
30
< __ _ _A
-- .- - .- __ - . . . ~ ,. . . _ _ . _ . -
This procedure provides implementation instructions for NSM CN-11186, Rev. O. This modification will replace Auxiliary Feedwater Discharge Isolation Valve ICA46B with a new gate valve. This is being done to assure that there is sufficient torque available to meet design requirements. Also, new valve 1CA283 is being added between the bonnet for valve ICA46B and the upstream process piping. This will prevent bonnet overpressurization on ICA46B. This procedure provides guidance for the removal / reinstallation of valve 1CA46B, and the installation of now valve 1CA283.
A blank is being installed in 1 CAFE 5110 by procedure TN/1/A/1186/00/05A to ensure that the CA flow path to Steam Generator 1C is isolated during the replacement of valve 1CA46B. This blank will also be used to maintain containment integrity / closure during core alterations, fuel movement, and NC System mid-loop conditions.
Licaker R03A in IEMXB will be opened and Red Tagged under Block Tagout CA3VLV to ensure valve 1CA46B is de-energized for electrical work.
This does not present a concern for safe operation of the CA system because the system is not required to be operable in Modes 4, 5, 6, or No Mode. The Red Tag for breaker R03A will be lifted from block tagout CA3VLV for valve set up, verification of remote position indication, motor operator valve (MOV) testing, and stroke time testing.
Testing of the new valve will consist of performing MOV testing and verifying all remote position indications. Stroke time tests will be performed prior to mode 3, and again in . node 3. A differential pressure test will be performed on valve ICA46B. Hydrostatic and appropriate Non-Destructive Examinations will be performed by station and Quality Assurance Procedures. Based on the above discussion, a USQ does not exist.
l TN/1/A/1186/00/071. Origir.al This procedure provides implementation instructions for NSM CN-11186, Rev. O. This modification will replace Auxiliary Feedwater Discharge Isolation Valve 1CA42B with a new gate valve. This is being done to assure that there is sufficient torque available to meet design requirements. Also, new valve 1CA282 is being added between the bonnet fcr valve ICA42B and the upstream process piping. This will prevent bonnet overpressurization on ICA42B. This procedure provides guidance for the removal / reinstallation of valve 1CA42B, and the installation of new valve 1CA282.
A blank is being installed in 1 CAFE 5120 by procedure TN/1/A/1186/00/08A to ensure that the CA flow path to Steam Generator 1D is isolated during the replacement of valve 1CA42B. This blank will also be used to maintain containment integrity / closure during l core alterations, fuel movement, and NC System mid-loop conditions.
Breaker F08B in 1EMXL will be opened and Red Tagged under Block Tagout CA4VLV to ensure valve ICA42B is de-energized for electrical work.
This does not present a concern for safe operation of the CA system because the system is not requ ed to be operable in Modes 4, 5, 6, or 31
No Mode. The Red Tag for breaker F08B will be lifted from block tagout CA4VLV for valve set up, verification of remote position indication, motor operator valve (MOV) testing, and stroke time testing.
Testing of the new valve will consist of performing MOV testing and verifying all remote position indications. Stroke time tests will be performed prior to mode 3, and again in mode 3. A differential pressure test will be performed on valve 1CA42B. Hydrostatic and appropriate Non-Destructive Examinations.will be performed by station and Quality Assurance Procedures. Based on the above discussion, a USQ does not exist.
TN/1/A/1186/00/08A Original This procedure provides implementation instructions for NSM CN-11186, Rev. O. This modification will replace Auxiliary Feedwater Discharge Isolation Valve ICA38A with a new gate valve. This is being done to assure that there is sufficient torque available to meet design requirements. Also, new valve ICA281 is being added between the bonnet for valve 1CA38A and the upstream process piping. This will prevent bonnet overpressurization on 1CA30A. This procedure provides guidance for the removal / reinstallation of valve ICA38A, and the installation of new valve ICA281.
A blank is being installed in 1 CAFE 5120 oy this procedure to ensure that the CA flow path to Steam Generator 10 is isolated during the replacement of valve ICA38A. This blank will also be used to maintain containment integrity / closure during core alterations, fe.el movement, and NC System mid-loop conditions. Breaker F08B in 1EMXK will be opened and Red Tagged under Block Tagout CA4VLV to ensure valve ICA38A is de-energized for electrical work. This does not present a concern for safe operation of the CA system because the system is not required to be operable in-Modes 4, 5, 6, or No Mode. The Red Tag for breaker l F08B Lill be lifted from block tagout CA4VLV for valve set up, verifi-cation of renote position indication, motor operator valve (MOV) l i testing, and stroke-time testing.
Testing of the new valve will consist of performing MOV testing and verifying all remote position indications. Stroke time tests will be performed prior to mode 3, and again in mode 3. A differential pressure test will not be performed on valve 1CA38A because it is one of the turbine drive CA pump discharge valves and sufficient 1differen-tial pressure cannot be obtained. Hydrostatic and apprcpriate Non-De-structive Examinations will be performed by station and Quality Assurance Procedures. Based on the above discussion, a USQ does not exist.
TN/1/A/1051/00/01A Original This procedure provides implementation instructions for NC1 CN-11051 Rev. O. This NSM will install in-line connectors on the RG-11/U 32 l
l . .- . - - - ..
F triaxial cables near the bottom of the Nuclear Instrumentation cabi-nets. RG-59 triaxial cable assemblies will be installed from the in-lino connectors to the cable termination points in the cabinets.
This procedure will control installing the in-line connectors and RG-59 cable for Source Range Detector N31.
This procedure will be performed during the refueling outage when Source Range Detector N31 is not required to be operable per Technical Specification 3/4.3.1 or 3/4.9.2. No systems will be prevented from performing any function important to safety while this work is being done. For these reasons, no USQ exists.
TN/1/A/1051/00/02A original Thic procedure-provides implementation instructions for NSM CN-11051 Rev. O. This NSM will install in-line connectors on the RG-11/U triaxial cables near the bottom of the Nuclear Instrumentation cabi-(
nett. RG-59 triaxial cable assemblies will be installed from the in-line connectors to the cable termination points in the cabinets.
This procedurn will control installing the in-line connectors and RG-59 cable for Source Range Detector N32.
This procedure will be performed during the refueling outage when Source Range Detector N32 is not required to be operable per Technical Specification 3/4.3.1 or 3/4.9.2. No systems will be prevented from performing any function important to safety while this werk is being done. For these reasons, no USQ exists.
PT/1/A/4200/09 Re-type, Changes 0 to 73 Incorport. mW Enclosure 13.21 was added to provide additional guidance on the placeme.nt of the jumpers on the blackout load centers during sections 12.3 and 12.6. The procedure was changed so that both trains of Solid State Prc tection System (SSPS) are assumed to be in TEST prior to the
, test. A CAUTION statement was added to ensure that the test does not start after 2345. This change was made because the Response Time' Test (RTT) program on the Operator Aid Computer (OAC) will not function properly if the test extends through 0000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. Steps have been added to check the circuitry for the hydrogen recombiners in the Blackout /LOCA sections of the test. This is to ensure that the recombiners get the permissive to start on Load Group 13. The re-quirement for closing the supply breakers in.the system _ alignment was deleted since the recombiners will not be started. Only the permis-sive will be verified. Valve 1RN67A was added to enclosures 13.1.1 and 13.2.1, and valve 1RN69B was added to enclosures 13.4.1-and 13.5.1. These valves were inadvertently left out of these enclosures previously. The alignment for the Spent Fuel Pool Cooling (KF) system is now specified so that the KF pump on the train under test may be started and left running for the duration of the test. The breaker for the Unit 2 Nuclear Service Water Pump Structure -Ventilation - (VZ) fans will be taken to the OFF position in the Blackout section so that the Unit 1 fans will start during the test. In sections 12.9 and 33 4
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12.10, steps have been added to verify the trip circuitry on Chemical and Volume Control System (NV) pump #1. The change is being made so '
that all components on relays K648 and K649 will be tested and l PT/1/A/4200/09A, Auxiliary Safeguards Test Cabinet Periodic Test, will not have to be performed to verify the trip circuitry. In sections 12.9 and 12.10, valves 1KF101B and 1KF103A will be verified to reposi-tion during the test. This will require that 1KF101B initially be open during section 12.10 and that 1XF103A bc open during section 12.9. These valves are normally closed. The change is being made so
- hat all components on relays K648 and K649 will be tested and PT/1/A/4200/09A, Auxiliary Safeguards Test Cabinet Periodic Test, will l not have to be performed to verify that the valves will reposition.
l Jumpers have been added so that all 4 steam generator Power Operated
! Relief Valves (PORVs) will be kept open by a simulated high steam line pressure signal. The change is being made so that all PORVs will close when the main steam isolation signal, from either train, is received. On enclosures 13.47 and 13.6.7, and in section 2.0, " Refer-ences Needed to Perform Procedure," the Chemical Volume Control System Pump Head Curve was changed due to the pump replacement during Novem-ber 1989.
In sections 12.1 through 32.6, the opposite train of the one under test will remain operable during the test. All pumps on the train under test are inoperable (due to alignment to recirculation flow path) except for the Nuclear Service Water (RN) and Component Cooling (KC) pumps (in see, tion 12.2 and 12.5), both of Jhich are aligned for normal operation. The Rh' and KC pumps are inoperabic in sections 12.1, 12.3, 12.4, and 12 6 due to the Diesel Generator (D/G) being l inopernble. The opposite train is capable of performing any safety l functions. The discharge of the Chemical and Volume Control System
! (NV) Pump and the Safety Injection (NI) pump on the train under test are isolated from the Reactor Coolant (NC) system by two isolation valves to provide low temperature overpressure protection (LTOP) as i required by Tech. Specs. In addition, as pracaution to avoid inadver-tent actuation of a Reactor Coolant (NC) Power Operated Relief Valve l
(PORV), the prerequisites to the test specify that the NC system cannot be water solid. Tech. Spec. 3.1.2.1 requires that one boron injection flow path be operable. One boron injection flow path will remain operable since the NV pump on the opposite train of the one being tested is not rendered inoperable by this test. Radiation Monitors (EMFs) 38, 39, and 40, which provide interlocks to terminate i containment air release through the Containment Purge (VP) and Con-l tainment Air Release and Addition (VQ) systems, will be rendered inoperable by the conduct of this test when the sample lines are l isolated by a Phase A containment isolation signal. However, both VP and VQ operation will be terminated upon the receipt of the Phase A containment isolation signal.
The Spray / Sprinkler system will become inoperable due to the closure of the containment isolation valve to the annulus sprinklers. The test should be completed within the one hour in which action must be taken. If the test cannot be finished within one hour, the action statement can be easily satisfied. A Limit and Precaution has been added to the procedure states that portions of Fire Protection (RF) l 34
will be inoperable due to valves-1RF389E 'EV44'B, and 1RF457B receiv-ing a Phase A containment isolation sic' t*vtions 12.1, 12.3, 12.4, and 12.6, the D/G on the train b .c - 1 inoperable due to the accelerated sequence boing defeat . 'G runs for an extended period of time, the D/G becc.+; ' - c due to fuel oil-l level dropping below the minimum allow..d tec .pecs. However, l only one D/G is rwquired to be operab)2 '
. nodes 5 and 6. Assuming an initial fuel oil supply of 80,000 gallons, tne D/G would have to be run for greater than 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> before becoming inoperable due to fuel oil supply. The test should not run the D/G tor more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The procedure requires the D/G on the opposite train to be operabita.
Sections 12.7 and 12.8 do not make the Cold Leg Accumulators inopera-ble since they are only required in Modes 1 through 3, and the test is done in Modes 4, 5, or 6. Sections 12.9 and 12.10 will ea:t make the Residual Heat Removal (ND) train under test inoperable since the auction piping is drained. These sections will be done when the unit is below Mode 4 when only one train of ND is required. Only one train of ND will be tested at a time. Section 12.11 does not make Main Steam inoperable, and is performed in Mode 5 when Main Steam Isolation is not required.
Based on the above discussion, no USQ is created by this procedure.
Th/2/A/2406/CE/01A Initial Issue This procedure provides implementation instructions for Exempt Change CE- 2 4 0 6. This Exempt Change replaces Valcor Solenoid Valves Model V79900-21-3 with model V0900-39-3 for valves 2CASV1850, 2CASV1851, 2CASV1860, 2CASV1061, 2CASV1870, 2CASV1871, 2CASV1880, and 2CASV1881.
Cne procedure is a generic procedure for replacement of all eight valves.
i construction Maintenance Department _ personnel will replace the sole-l noid valves and verify that the affected control valve will stroke by actuation of each solenoid which is replaced. Performance will stroke time the affected control valve using each solenoid valve which is j replaced, l
l This procedure may be implemented with Unit 2 in any mode. Valves 2CA185, 2CA186, 2CA187, and 2CA188 are containment isolation valves and are required to be operable OR closed in Modes 1, 2, 3, and 4 per Tech. Spec. 3.6.3. No USQ is created by this procedure.
TN/5/A/0411/00/02A Initial 3as"A J This procedure provides implementation instructions for Nuclear I Station Modification (NSM) CN-50411, Rev. O. This NSM deletes i unused Nuclear Service Water (RN) pump lube injection crossover piping i between RN train A and trajn B containing valve 1RN026. This modifi-cation is required to facilitate removal of RN pump 1A for mainte-nance. This procedure provides guidance for cutting and capping the 1
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RN pump lube injection crossover piping ca the train A side of the crossover isolation valve IRN026.
) During implementation of this procedure, RN train A will be considered inoperable. This procedure will be performed with Unit 1 in Mode 5 or below, and Unit 2 operating in any-mode. The procedure requires RN train B to be operable and RN train 2A to be returned to service within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This situation is allowed by Tech. Spec. 3.7.4.
Prior to returning A train RN to service, the piping will be capped and hydrostatically tested to ensure RN system integrity. Restoring the RN pressure boundary and providing pump lube injection flow to RN pump 2A will return RN train 2A to-service and meet Tech. Spec.
requirements. This will ensure that RN can perform its intended function. Based on the considerations above, no USQs are judged to be created by this procedure.
TN/1/A/0675/00/02A Initial Issue This procedure provides implementation instructions for Nuclear Station Modification (NSM) CN-10675 Rev. O. This NSM will revise the control circuitry to fail open valves 1CA048 and 1CA052 by de-energiz-ing their respective solenoid valves at their respective Safe Shutdown Facility (SSF) disconnect enclosures. Valve ICA050 will have its control circuit modified to open the valve upon transfer of plant controls to the SSF. Valve 1NV101A will have its control circuit modified to fail the valve closed by de-energizing the solenoid valve at the respective SSF disconnect enclosure. All electrical controls will be deleted from valva 1TE033A. This procedure will control the removal and return to service of all equipment affected by this modification.
This procedure will be implementec with Unit 1 it; Modes 5, 6, or No Mode. The equipment affected by this procedure ir not required to be operable during:these modes. No systems will be prevented from performing any function important to safety while this work is being performed. All equipment affected by this_ procedure and the design
, intent of the modification will be completely tested by this proce-dure. Performance will originate temporary test procedure TT '/A/9200/57 which will control the retest to be performed. This pr.cedure will control whcn the retect is procedure is to be complet-ed, and documents completion or that procedure. The retest measures will ensure that the modified valve circuitry will function per the Design Basis when the SSF is activated. Based on the above, no USQ exists.
TN/1/A/1186/00/05A Rotype 1, Change 1 Inco porated This procedure provides implementaticr .astructions for NSM CN-11186, Rev. O. This modification will replac<> Auxiliary Feedvater (CA)
Discharge Isolation Valve-1CA50A with a new gate valve. This is being done to assure that there is sufficient torque available to meet design requirements. Also, new valve ICA284 is b41ng added between 36 J
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the bonnet for. valve 1CA50A and the upstream process piping. This will prevent bonnet overpressurization on 1CA50A. This procedure
- provides guidance for the removal / reinstallation of valve 1CA50A, and
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the installation of new valve 1CA284. Retype #1 is to incorporate changes made in change #1, and the following: 1) Correct typographi-cal errors, 2) add support 1R-CA-1679 to steps 8.11 and 8.25 for removal and restoration, 3) change step 8.8 to say install rather than replace, 4) add interference removal to steps 8.9 and 8.10, and 5) delete original step 8.11.
l A blank is being installed in ICAFE5110 by this procedure to ensure that the CA flow path to Steam Generator 1C is isolated during the replacement of valve ICA50A. This blank will also be used to maintain containment integrity / closure during core alterations, fuel movement, and Reactor Coolant System mid-loop conditions. Breaker F04A in 1EMXS will be opened and Red Tagged under Block Tagout CA3VLV to ensure valve 1CA50A is de-energized for electrical work. This does not present a concern for safe operation of the CA system because the system is not required to be operable in Modes 4, 5, 6, or No Mode.
The Red Tag 'or breaker F04A will be lifted from block tagout CA3VLV for valve set ap, verification of remote position indication, motor operator valve s . '07 ) testing, and stroke time testing.
Due to physical interferences, breaker F01C in 1MXK, sliding links H-3 and H-4 in 1TBOX0100 will be opened, tad cables 1CAS41 and 1CF555 disconnected from valves 1CASV1512 and 1CF51, respectively. Solenoid valve 1CASV1512 allows remote operation for valve 1CA151, and will
(
l fail to ite closed position when the breaker is oper.ed. Cables will be reconnected upon installation of valve 1CA50A, re-energized, and stroke time tested.
Testing of the new valve will consist of performing MOV testing and verifying all remote position indications. Stroke time tests will be performed prior to mode 3, and again in mode 3. A differential pressure test will not be performed on valve 1CA50A because it is one cf the discharge isolation valves for the turbine driven CA pump and sufficient differential pressure cannot be obtained. Hydrostatic and appropriated Non-Destructive Examinations will be performed by station and Quality Assurance Procedures. Based on the above discussion, a USQ does not exist.
MP/0/A/7200/05 Re-type, Changes 0 to 2 Incorporated Section 11.0 of this procedure was completely re-written, and the affected data sheets on enclosure 13.1 were changed. Also, enclosures 13.2 and 13.4 were added. This re-write clarifies and assures that the maintenance activities will return the turbines to as designed conditions. The changes have been reviewed against approved vendor manuals, design documents, and station procedures to ensure that the corrective maintenance will return the turbines to as-built, as-de-signed conditions. These actions ensure the turbines will comply with the FSAR accident analysis. Tech. Spec. 3.7.1 is affected by this procedure. Operations has the procedures and the responsibility to 37
i ensure that maintenance is performed on this turbine when Tech. Specs.
allow. Thus, no USQ exists.
PT/1/A/4150/17 Re-tr7e #3, Changes o to 8 Incorporated This re-type for ~~he end of cycle (EOC) Catawba 1 cycle 4 (ClC4) reload includes a change to check out both the Source Range (S/R) and t.ke Boron Dilution Mitigation System (BDMF) detectors with the secon-dary source assemblies.and arrays of'twice burned fuel assemblies. !
. This is to see whether during future relouds_the Secondary Sources may-be left out, and whether the BDMS detectors can be used instead of the source range detsetors.- The reference section was changed to add references to Tech._ Spec. 3/4.7.6 and 3/4.3.3 on control room ventila-tio*a r.nd radiation monitoring and Station Directive 3.1.29 on fuel reliability.- Step 2.12 was deleted, dueLto not using the old off-load pattern.. . Other changes included adding references to the Westinghouse Specification on fuel handling, Reg. Guide 5.29, FSAR chapter 15.4.7
' on Inadvertent Fuel Loading, and Memo by G. A. Harbin on Fuel Unload-ing.- The Pre-Mode 6 Surveillance Periodic Test (PT) and Refueling Component Checkout PT were added'to_the Prerequisite test section._
The limits and precautions section was changed to helt erre altera-tions if Residual Heat Removal (ND) temperature e.xceeds 140-degrees per the revised Tech. Spec, other changes included adding-reference to Employee Training and Qualification System-(ETQS) qualifications of fuel handlers, and precautions on' control room ventilation and cen-
- i trifugal charging pump. Prerequinite system conditions were changed to include containment. Purge (VU) operability, and delete reference to 4 the operating Procedure. Step 10.1 was. changed to require both
, refueling canal-and ND Boron concentrations:be greater than 2000 ppm.
The procedure steps were changed to delete recording the Spent Fuel
. Pool boron every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. other changes included adding a step to notify the. shift supervisor that Core Alterations are;about to begin, and to allow marking a etep N/A if the. source range; detectors are not required-to be operable, Some enclosures.were changed as well. END temperature and flow requirements were changed:to reflect the revised
- Tech. Specs. A step-was changed to allow temperature detectors to be p used-on refueling bridges. A-step was added on the boron injection =
r flow path. Enclocure 13.5-was changed to the EOC C1C4 pattern for.
i unload and allowed checkout of S/R and BDMS. Enclosure 13.7'was added for the-recommended core unload pattern. The bow:and twist' reference dilection was changed in enclosure 8. The core pattern for'C1C4 was t
added to-enclosure 13.9. A data sheet for S/RLand BDMS-data was added to. enclosure 13.10, and a multiple step sign off form wassadded to-2 t enclosure 13.11. A number of enclosures were changed to reflect-o changes made elsewhere-in the body of the procedure. A number of
- _ editorial changes were made!for clarification.
H This procedure is designed to unload-the core in a safe and orderly i, manner.- ThisLprocedure createdono accident scenarios that are not -
i
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already analyzed. The only. loads being moved are fuel _ assemblies.
~ FSAR section=15.7.4, Fuel Handling Accidents in'the Containment Building, is bounding. There'is no off-normal-operation of safety equipment.. No USQ is created by this-procedure.
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TN/5/A/0411/00/01A Initial Issue This procedure _provides implementation instruction for Nuclear Station Modification (NSM) CN-50411. This NSM will dolete unused Nuclear Service Water (RN) pump lube injection crossover piping between RN train A and Train B containing the valve 1RN026. This piping inter-feres with the removal of RN Pump 1A for maintenance. This procedure covers the cutting and capping of the crossover piping on the train B side of the crossover isolation valve 1RN026.
During implementation of this procedure, RN train B will be considered inoperable. This procedure will be performed with Unit 1 in Mode 5 or below, and Unit 2 operating in any mode. The procedure requires RN train 2A to be operable and RN train B to be returned to service within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This situation is allowed by Tech. Spec. 3.7.4. A freeze seal will be installed to allow the Train B side of the. cross-over piping to be capped without having to drain the Train B pump lube
, injection piping. The freeze seal will provide-isolation while work is being performed and serve as a Hydrostatic boundary. Failure of the freeze seal will not prevent RN from performing its function, since RN train B is considered inoperable. Prior to releasing the freeze seal, the piping will be capped and hydrostatically tested to ensure RN system integrity. Restoring the RN pressure boundary and providing pump lube injection flow to RN pumps 1B and 2B will return RN loop B to service and meet Tech. Spec. requirements. This will ensure that RN can perform its intended function. The cutting and capping of the B train crossover piping will leave the remaining piping connected to Train A free-ended. Design Engineering has evaluated this configuration and determined that Train A seismic integrity will not be affected. Based on the considerations above, no USQs are judged to be created by this procedure.
TN/1/A/0910/01/02A Initial Issue This procedure provides implementation instructions for Noticar Station Modification (NSM) CN-10910 Rev. 1. This NSM wjAl complete all the electrical work remaining with the Upper Head Irjection (UHI) system deletion. It will delete the controls and cables from_various cabinets throughout the plant. The purpose of this procedure is to control the removal of. wiring which is associated with tLe removal of the UHI system. It deals with the wiring in the Auxiliarf luilding.
The isolations required by this procedure are needed so thac deleted cables may be removed from cabinets. No equipment will be isolated until the plant is in a condition in which the equipment 14 not required to be operable. Prior to completion of the procelure, functionals and/or retests will be performed to ensure that the equipment affected by this procedure has not been adversely 'ffected.
During Section 8.11, Unit 1 will not be able_to provide Nuclear Service Water (RN) miniflow protection.- Notes and steps ar e included in the procedure to ensure that RN miniflow is.being provicad by Unit
- 2. Based on the previous discussion, a USQ will not be create?.
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1 l
l TN/1/A/0753/00/03A change #2 This change allows Construction Mainter.ance Department (CMD) personnel to make the RTD piping severance cuts at each reactor coolant loop.
This will allow portions of RTD piping to be removed prior to PCI starting the RTD modifications activities to reduce area radiation levels. These cuts will be made using porta-band saws.
As a consequence of the cutting operation, metallic debris will be formed which can be postulated to enter the recctor coolant system.
Westinghouse has evaluated this concern and determined that the potential debris does not constitute a USQ.
TN/1/A/1214/00/04A Change #3 This change is to add a support for removal / reinstallation to aid in the replacement of valve 1BB60A. The support will be removed and reinstalled using approved procedures. The support will be rein-stalled prior to any mode in which it is required for component / system operability. Based on this discussion, a USQ does not exist.
MPII 2.7.2 DCP-1 Change #2 This change allows PCI to install the Reactor Coolant (NC) loop A, B, C and D bosses prior to draining the NC system in support of No Mode activities. This will allow certain activities which can be performed with the loops filled to begin prior to draining the NC loops to avoid impacting the outage schedule. This change also incorporated the new loop B hot leg boss locations as specified by FCN-DCP-40507E.
Evaluation of the boss installation prior to draining the NC loops has !
been done by Westinghouse in SECL-90-067A, and determined not to-represent a USQ. Relocation of the Loop B Hot Leg bosses does not affect the original safety _ evaluation. Based on the Westinghouse evaluation, these changes do not represent a'USQ.
TN/1/A/0753/00/03A Change #1 Change #1 to this procedure allows PCI to install the Reactor Coolant (NC) loop A, B, C, and D bosses prior to draining the 1U0 system in support of No Mode activities. This will allow certain activities which can be performed with the loops filled to begin prior-to drain -
ing the NC loops to avoid impacting _the. outage. schedule. It also allows Construction Maintenanco Department personnel to install travel stops while the NC system draining is in progress.
Evaluation of the boss installation prior to draining the NC loops has been done by Westinghouse in SECL-90-067A, and determined not to represenc a USQ.
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OP/1/A/6150/06 Change #22 i
This change provides operator instructions for injecting nitrogen into the steam generators to aid in draining the reactor coolant system.
The reactor coolant system will be drained to 23-24% level.1 Nitrogen will be injected into the reactor coolant system through the high-pressure impulse tubing for the reactor coolant system flow transmit-ters. The instrument taps are located on the crossover legs nearly directly under the steam generator tube bundle. This ensures that nitrogen migrates to the steam generator U-tubes and not into the residual heat removal suction piping._ The reactor coolant system will be vented to atmosphere, eliminating any potential for overpressurization. The nitrogen injection flow rate (~10 cfm) will be commensurate with the drain down rate (~75 gpm) which will maintain the reactor coolant syctem level virtually constant. Two nitrogen bottles will be injected into each steam generator, which will reduce i
the actual level in the U-tubes to ~28% reactor coolant system level.
l This higher level will ensure _an excess of nitrogen is not injected.
Any level decrease to 22% reactor coolant system level will.be used as i
an indication that the U-tubes have drained completely, and-injection
, will be secured. The steam generators are drained sequentially, l lessening the chance of level changes taking place as the tubes drain.
l The cooling effect of the nitrogen injected was determined to be
! minimal in relation to the heat removal by Residual Heat Removal System cooling.
A loss of Residual Heat Removal during drained down conditions.is discussed in FF,AR section 5.4.7. This change will reduce the amount of level fluctuations in the reactor coolant system when reliable level is the most critical. This will minimize the potential for equipment malfunction. For these reasons, no USQ is judged to be created.
PT/1/A/4200/02A Change #23 l This change deletes step 3.1.2, step 8.1 and changes step 8.0 to none.
Section 12.0 has been changed to verify that the Containment Purge (VP) train A and B Upper and Lower Containment Valve's enable switches
- are in the " Block closed" position andLthat the key is removed.
- Also, section 12.0 contains steps to verify _the Incore VP Train A and-Train B enable key switches are in the normal position with the " Block Closed" lights-lit. These changes are beingimade because the existing
, key switches have been modified to disable the VP system without pulling fuses. No USQ exists.
j TN/1/A/1122./00/01A Change #1.
. This change is to ensure that the bench test of valve 1CA280 is 4
performed prior to installation. 'This does not change'the safety-evaluation of the original procedure.- No.USQ exists.
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1 OP/0/B/6200/16 Change #29 This change replaced pages 1 through 3 af the procedure and Enclosures 4.4 and 4.5 to allow ice blowing during core alterations. Modifica-tions to the ice delivery system consist of adding a full port ball valve in the flexible delivery line (M-394) for use during fuel movement, and not using the air return piping through M-395 during fuel movement, thus leaving this penetration capped. The operating process for ice loading must also bc altered to have a worker at the ball valve to close the valve on containment isolation, stoppage of the containment purge system, or any time the ice delivery system is stopped. The valve being used is identical to the one tested at McGuire to insure that the operator can close the valve even with full ice blockage. Supply air for the ice blowing equipment will be taken from the auxiliary building and exhaust will be discharged into the reactor building and then through the containment purge exhaust ducts.
The system would only be operated when the containment purge system is in operation.
Tech. Specs. require that containment integrity be maintained during any fuel movements. When the ice delivery syetem is in operation, containment integrity is provided by a six pound positive pressure into containmant through.the ice delivery system. The ball valve installed will maintain the penetration closed when the ice delivery system is not in service. This system will only be operable in Mode 6 when containment pressurization from an accident is not significant.
The hose between the containment penetration and the ball valve will be made of fire retardant material. The ice delivery system is in place temporarily and is not connected to any operating safety sys-tems. It does not affect the quality of ice delivered to the ice l condenser. The only connection to safety related equipment is at the containment penetration. The penetration will be closed by the ball valve in any period when the ice delivery system is not in operation.
The failure of the ball valve will not create-a leak path with conse-quences greater than those considered in the current accident analysis for a fuel handling accident. In the event of a loss of decay heat removal, the penetration may be closed using the ball valve, or by disconnecting the ice delivery system and closing the penetration with a blind flange. For these reasons, there are no USQs associated with this procedure change.
l l TN/1/A/1214/00/02A Change #1 This change is for the addition of one support which requires remov-I al/ reinstallation for the removal / reinstallation of valve 1BB019A.
This does not change the conclusion of the initial evaluation that no USQ is involved.
TN/1/A/1214/00/04A Change #1 This change is for the addition of one support which requires remov-al/ reinstallation for the removal / reinstallation of valve 1BB060A.
42
This does not change the conclusion of the initial n"aluation that no USQ is involved.
TN/1/A/1214/00/01A Change #1 This change is for the addition of one support which requires remov-al/ reinstallation for the removal / reinstallation of valve 1BB008A.
This does not change the conclusion of the initial evaluation that no USQ is involved.
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TN/1/A/1186/00/0GA Changu #1 This change corrects a typographical error ir. step 8.9. This does not change the conclusion of the initial ev 1.uation that no USQ is in-volved.
PT/1/A/4450/03C Change #9 This restricted procedure change was made in order to obtain pressure distribution measurements and other data in the annulus as required by Design Engineering for possible elimination of the annulus ventilation (VE) system train A lower ring header. The data will be taken after removal of two-elbows located at the bottom of each drop of A train return air duct. Evaluation of this data will determine if the train A duct lower ring header can be removed without significantly affect-ag the performance of VE A train. Pressure measurements will be taken at four different locations in the annulus by routing tygon tubing from the annulus side of IVEPT5010 to the different locations.
A manometer will be connected to the Auxiliary building side of 1VEPT5010 (located downstream of a vent valve) to determine the the pressure in the annulus at the four points. Data will also be ob-tained on the air temperatures at each point in the annulus, air temperature at the auxiliary building side of 1VEPT5010, outside air temperature, barometric pressure, and fan motor amps. All of the above data wi]l be obtained while VE train A is operating in the
~
normal alignment per OP/1/A/6450/02. This restricted test is required by this change to be run while unit 1 is in Mode 5, 6, or No Mode.
Connections made to IVEPT5010 are the only significant part of this change, and the removal of these connections are independently veri-fled within this change. For these reasons, a USQ does not exist.
PT/1/A/4200/01T Change #25 This change to the Containment Penetration Valve Injection Water (NW) system performance test procedure changes three aspects of the proce-dure. The NW surge tank pressure is increased for some penetrations to account for any pressure that may still be on the valve as a result of the penetration not being completely drained. Other penetrations have had the lineups changed to ensure that they are completely drained. Steps and CAUTION statements have been added to the 43
procedure to ensure that containment closure and/or containment integrity requirements are not violated. The test pressure is being increased by up to 2 psig to account for any-pressure on the valva due to water not being drained. Some lineups were changed so that the 4 penetrations are "open" on both sides of the containment vessel, but steps have been added to ensure that these lineups are not performed when containment closure / integrity is required. The test pressure is higher than the pressure assumed when Nuclear Service Water is supply-ing the tank, thus giving conservativo leakage values. Some of the penetrations being tested have had' lineups changed to ensure complete draining, but they are still out of service es before. For these reasons, a USQ is not created by this change.
PT/1/A/4200/01T Change #26 This change affects two aspects of NW leakage tests on various valves.
"ae penetration alignment steps have been modified to allow the use of the Operatione vont and drain procedure, OP/1/A/6200/20, on many penetrations. The reason for the change is that OP/1/A/6200/20 drains the penetrations more completely, and use of the one alignment for NW and type C tests will result in fewer containment entries, which is an ALARA concern. OP/1/A/6200/20 also addressen containment closure requirements for each penetration. NW surge tank pressure is in-creased for some penetrations to account for any pressure that may still be on the valve as a result of piping not being completely drained on both sides of the valve. The test method is_not changed.
The test pressure is higher than the pressure assumed when Nuclear Service Water is supplying the tank, thus giving conservative leakage values. For these reasons, a USQ is not created by this change.
MP/0/A/7150/72 Change #1 ,
This procedure provides a method for verifying tne main steam safety relief valves relieve at the proper pressure. This change affects step 11.10, the determination of the differential pressure. The correction-factor used for calculation was changed from 0.339 to 0.352 as recommended by the valve _ manufacturer. - The manufacturer made this change based on empirical data from numerous tests. This procedure maintains the valves in ranges for which they-were designed to operate and pro 'es verification of operation within Tech. Spec. 3.7.1.1.
This chtty3 does not create a USQ. ;
MPII 2.7.2 DCP-1 Change #1-This change revises the-procedure to allow Westinghouse-to fabricate a-pipe and cap assembly for capping the Loop B cold leg nozzle (0 degree location) outside of containment and install the cap as an assembly. ;
This will allow Westinghouse-to perfczm critical welds and weld inspections outside of containment whicP will reduces-personnel dose. t The cap assembly will consist of a piece of 1" SS, SA-376/ Type 304,
- Schedule 160, class A pipe, approximately 5" long with a butt welded t
44 i
pipe cap. All welds will be inspected prior to installation of the cap assembly and in accordance with American Society of Mechanical Engineers (ASME) Code Section XI requirements and the Catawba Quality Assurance program applicable to safety related equipment. The assem-bly will be installed during NO MODE in conjunction with Westinghouse RTD modification activities. These activities are addressed in the original evaluation for MPII 2.7.2 DCP-1. After installation, the welds will be inspected at Reactor Coolant system temperature and pressure to verify system integrity. The integrity of the pressure boundary is maintained by adhering to the applicable ASME code sec-tions and the NRC general design criteria. Thus, no USQ is created by this change.
MP/0/A/7200/04 Re-type, Changes O to 1 Incorporated Step 11.4.11 was revised to give epocific instructions on how to bleed air from the governor. Step 11.5 was added to provide instructions on how to manually override the governor for overspoed trip testing of the auxiliary feedwater pump turbine. Enclosure 13.1 was revised to add appropriate sign-offs.
Tech. Spec. 3.7.1 may be affected by this procedure. Operations has the responsibility and the procedures for compliance with this Tech.
Spec. Maintenance will be performed on this governor when Tech.
Specs. permit. This revision will clarify and assure that maintenance activities return the governor to as-designed conditions. Therefore, no USQ exists.
PT/1/A/4150/18 Re-type #3, Changes O to 7 Incorporated This procedure is designed to provide a safe transfer of fuel assembly inserts from the Cycle 4 fuel assemblies to their Cycle 5 fuel assem- i blies. The changes for tais retype included changing step 3.2 to say ETQS. Enclosure 13.6 was revised to add cautions and steps to handle bent thimble plugs. All other changes to the procedure were made to update the procedure to go from Cycle 4 to Cycle 5 fuel assemblies.
This following steps and enclosures were modified as a result: 1.1, 2.1, 12.1, Enclosures 13.1, 13.3, and 13.4.
This procedure is bounded by the accident analysis performed by FSAR Chapter 15.4.7, Inadvertent Loading and Operation of a Fuel Assembly in an Improper Location. This chapter also analyzes operation and loading of a fuel assembly with an incorrect insert. This procedure requires movement of inserts and fuel handling tools over the spent fuel in the Spont Fuel Pool, but these inserts and tools weigh less than a fuel assembly. Thus, the accident analysis in FSAR chapter 15.7.4, Fuel Handling Accidents, is bounding. This procedure does not require off-normal operation of safety equipment. Thus, no USQ is created by this procedure.
OP/1/A/6100/09A Retype #6, Changes 17 to 19 Incorporated i
45
This procedure, Annunciator Response for 1A Diesel Generator (D/G)
Panel, was revised in an effort to make all four D/G Annunciator Response manuals as close to identical as possible. This involved incorporating all existing retypes, and making other changes based on design and procedure changes. These changes in the annunciator response are primarily due to changes in OP/1/A/6350/02, Diesel Generator Operation. Including the required actions and precautions for filter and strainer swaps in the Diesel OP eliminated confusion and probable errors. This accounts for the changes to windows A1, A2, B10, C3, and C4. The changes in Windows D5, D6, D7, and D8 involve determining whether adequate or excessive Nuclear Service Water (RN) flow exists to cool the Diesel Generator Engine Cooling Water (KD) heat exchanger. These flow rates are only available from computer points, which may be accessed in the Control Room. The changes to Windows AS, B1, and E9 were as a result of previously evaluated equipment changes which were installed by the Nuclear Ctation Modifi-cation (NSM) process. The changes to Window B9 were necessary as a result of mounting the overspeed and D/G speed pickups on the same accessory drive. Although normal indication and overspeed control may be lost through failure of the accessory drive, normal frequency (governor) and voltage control may still be active. Combined with manually starting the DC Fuel Oil Pump, the diesel may be acceptable for emergency service. The changes to windows E7 and F4 added trou-bleshooting information 5hich were because of information received from Operator and Maintenance experience. The change in Window F9 to include possible failure of the non-seismic automatic drain traps for the Diesel Generator Starting Air (VG) aftercoolers, was to allow the operator to use seismically mounted valves for isolation of the non-seismic traps should they fail. The other changes were because of a recent Tech. Spec. Interpretation which defined at what pressure, and under what conditions, the VG tanks became inoperable. Additional changes were made to correct grammar and format, or to improve clari-ty. These changes do not create a USQ.
OP/1/A/6100/09B Retype #6, Changes 16 to 18 Incorporated This procedure, Annunciator Response for 1B D/G Panel, was revised in an effort to make all four D/G Annunciator Response manuals as close to identical as possible. This involved incorporating all existing retypes, and making other changes based on. design and procedure-changes. These changes in the annunciator response are primarily due to changes in OP/1/A/6350/02, Diesel Generator Operation. Including the required actions and precautions for filter and strainer swaps in
,. the Diesel OP eliminated confusion and probable errors. This' accounts l for-the changes to windows A1, A2, B10, C3, and C4.- The changes in i Windows DS, D6, D7, and D8 involve determining whether adequate or-excescive RN flow exists to cool the KD heat exchanger. These flow rates are only available from computer points, which may be accessed in the Control Room. The changes to Windows.A5, B1, and E9 were-as a result of previously evaluated equipment changes which were. installed by the NSM process. The changes to Window B9 were necessary as a result of mounting the overspeed and D/G speed pickups on the same accessory drive. Although normal indication and overspeed control may
! 46
1 l be lost through failure of the accessory drive, normal frequency l (governor) and voltage control may still be active. Combined with manually starting the DC Fuel Oil Pump, the diesel may be acceptable for emergency service. The changes to windows E7 and F4 added trou-bleshooting information which were because of information received
, from Operator and Maintenance experience. The change in Window.F9 to l include possible failure of the non-seismic automatic drain traps for the VG aftercoolers was to allow the operator to use seismically mounted valves for isolation of the non-seismic traps should they ,
fail. The other changes were because of a recent Tech. Spec.-Inter-pretation which defined at what pressure, and under what conditions, the VG tanks became inoperable. Additional changes were made to correct grammar and format, or to improve clarity. These changes do not involve a USQ.
OP/2/A/6100/09B Retype #4, Changes 11 to 13-Incorporated This procedure, Annunciator Response.for 2B D/G Panel, was revised in ,
an effort to make all four D/G Annunciator Response manuals as close '
to identical as possible. This involved incorporating all existing retypes, and making other changes based on design and procedure changes. These changes in the annunciator response are primarily due to changes in OP/2/A/6350/02, Diesel Generator Operation. Including the required actions and precautions'for filter 1and strainer swaps in the Diesel OP eliminated confusion and probable errors. This accounts for the changes to windows A1, A2, B10, C3, and C4. The changes in Wirdows DS, D6, D7, and D8 involve determining whether adequate or excessive RN flow exists to cool the KD heat exchanger. These flow rates are only available from computer points, which may be accessed in the Control Room. The changes to Windows AS, B1, and E9 were as a result of previously evaluated equipment changes which were installed by the NSM process. The changes to Window B9 were necessary as a result of mounting the overspeed and D/G speed pickups on the same accessory drive. Although normal indication and overspeed control may be lost through failure of the accessory drive, normal frequency i (governor) and voltage control may still be active. Combined with-manually starting the DC Fuel Oil Pump, the diesel may be acceptable for emergency service. The changes to windows E7 and F4 added trou-bleshooting information which were because:of information received from Operator and Maintenance experience. The change in Window F9 to _
include _possible failure of the non-seismic automatic drain traps-for the VG aftercoolers was to allow the-operator to use-seismically mounted valves for isolation of the non-seismic traps should they-fail. The other changes were because of a recent Tech. Spec. Inter-pretation which defined at what pressure, and under what conditions,-
the VG tanks became inoperable. The changes to Window D8 are as a result of a Design Engineering evaluation concerning the* ability to cool the Unit 2 D/Gs with Fire Protection Water (RF) in a situation where RN would not be available.- This evaluation' led to the: develop-ment of Enclosure 4.24 of OP/2/A/6350/02, Alignment of RF to KD Heat Exchanger for Alternate Cooling, and this change refers the operator to that procedure. Additional changes were made to correct grammar l and format, or to improve clarity. These enanges do not create a USQ.
-47 I - _ __ _ _ _ _ _ ._ . . - -
i l
OP/2/A/6100/09A Retype #4, Changes 11 to 14 Incorporated This procedure, Annunciator Response for 2A D/G Panel, was revised in an effort to make all four D/G Annunciator Response manuals as close to identical as possible. This involved incorporating all existing retypes, and making other changes based on design and procedure changes. These changes in the annunciator response are primarily due to changes in OP/2/A/6350/02, Diesel Generator Operation. Including the required actions and precautions for filter and strainer swaps in the Diesel OP eliminated confusion and probable errors. This accounts for the changes to windows A1, A2, B10, C3, and C4. The changes in Windows DS, D6, D7, and D8 involve determining whether adequate or excessive RN flow exists to cool the KD heat exchanger. These flow rates are only available from computer points, which may be accessed in the Control Room. The changes to Windows A5, B1, and E9 were as a result of previously evaluated equipment changes which were installed by the NSM process. The changes to Window B9 were necessary as a result of mounting the overspeed and D/G speed pickups on the same accessory drive. Although normal indication and overspeed control may be lost through failure of the eccessory drive, normal frequency (governor) and voltage control may still be active. Combined with manually starting the DC Fuel _ Oil Pump, the diesel may-be acceptable for omergency service. The changes to windows E7 and F4 added troubleshooting information which were because of information received from Operator and Maintenance experience. -The change in Window F9 to include possible failure of the non-sejamic automatic drain traps for the VG aftercoolers, was to allow the operator to use seismically mounted valves for isolation of the non-seismic traps should they fail. The other changes were because of a recent Tech. Spec.- Inter-pretation which defined at what pressure, and under what conditions the VG tanks became inoperable. The changes to Window D8 are as a result of a Design Engineering evaluation concerning the ability to cool the Unit 2 D/Gs with'RF in a situation where RN would not be available. This evaluation led to the development of Enclosure 4.24 of OP/2/A/6350/02, Alignment of RF to KD Heat Exchanger for Alternate Cooling, and this change refers the operator to'that procedure.
Additional changes were made to correct grammar and format, or to improve clarity. These changes do not involve a USQ.
PT/1/A/4200/02E Change #5 This change adds Nuclear Sampling (NM) valve 1NM489 to Enclosure-13.2.
Vent valves 1NM489 and 1NM490 were added upstream and downstream of valve INM022A per Nuclear Station Modification'CN-11086 in order to facilitate the drainage of containment penetration M-310 for_ Type-C leak rate testing. 1NM489 in an inside containment isolation valve required to be verified closed when taking credit for closure on the containment side of penetration M-310. No USQs are created.
1 OP/1/A/6200/20 Change #39 l
[
48
F Vent valves 1NM489 and 1NM490 were added upstream and downstream of valve 1NM022A por NSM CN-11086.in order to facilitate the drainage of containment penetration M-310-for Type C leak rate testing. Enclo-sures 4.27 and 4.28 of OP/1/A/6200/20 have been rewritten to include these vents in the procedure for isolating, draining, refilling, and realigning containment penetration M-310. No USQ is created.
i MP/0/A/7600/115 Retype -
1 This procedure provides a method for disassembly, inspection, reassem-
! bly, and corrective maintenance for Valcor Solenoid Valves. The l procedure was changed in order to address one additional item number, j The technical information contained within this procedure was not j changed in a significant manner. No USQ is created by this change.
5 PT/1/A/4150/22 Retype #4 The majority of the changes made to this procedure were to update it to use data for Catawba 1 Cycle 5 (CICS) loading plan, to make the Unit 1 procedure match the Unit 2 procedure, and to allow the use of the Boron Dilution Mitigation System (BDMS) Detectors instead of the Source Range Detectors. A detailed list of changes follows.
Step chance 2.0 Section changed to match Unit 2 procedure 2.16 Changed to specify reference to CICS loading plant 3.1 Added time-duration of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 3.3 Changed to specify ETQS qualified 4.5 Changed time to 3 months to match actual performance period of procedure 4.5-4.12 Added to match unit 2 procedure 5.5 Deleted -- only one scalar timer will be used 6.0 Revised to match Unit 2 procedure 6.1.9 Changed Residual Heat-Removal (ND)- temperature to 140 deg F due to-new Tech Spec. -
- 6. l i21 Added centrifugal charging pump operable requirement 6.19 Added precaution-on use of Operations'new assembly shoehorn to ease fuel assembly into location 6.17 Clarified 8.0 Changed to match Unit 2 procedure 8.17- -Deleted Source Range to allow use of BDMS 8.19 Added notification of Shift Supervisor 10.1 Deleted time specification to allow use of BDMS 10.2 Deleted time specification to allow use of BDMS 10.5 Deleted step as PAO data is not required 11.1 Deleted acceptance criteria on ICRR. This was not a real acceptance criteria but a limit and precaution to stop loading fuel.
11.2 Deleted acceptance criteria-on ICRR. This-was not a real acceptance criteria but a limit and precaution to stop loading fuel.
49 l
- es 12.0 Changed to match Unit 2 procedure 12.2 Changed ND temperature to 140 deg F j 12.7 Changed the order to update the core map after core !
verification. This is due to a change in the core Verification procedure. l 13.1.6 Deleted PAO table l i
13.3 Added date/ time column and added note to use the ND heat l exchanger inlet temperature i 13.4 Revised to use CIC5 fuel assemblies and locations '
i 13.5 Revised to match the Unit 2 procedure l 13.6 Changed to allow use of BDMS I 13.7 Changed to allow use of BDMS l 13.8 Revised to match Unit 2 procedure and to allow use of BDMS '
13.9 Changed to match unit 2 procedure 13.10 Changed to CICS core loading pattern and to load on BDMS l side 13.12 Changed to match Unit 2 procedure 13.15 Revised to match unit 2 procedure and to allow use of BDMS l l
This procedure is designed to load the core in a safe and orderly manner. This test is not described in the FSAR. The closest descrip-tion is found in FSAR Section 9.1.4, which describes the process used in an incore fuel assembly / insert shuffle. This procedure complies i
with the loading description found there. It is different in that the core is completely unloaded at the beginnin3 while the FSAR section leaves fuel assemblies in the core at all times. FSAR Table 14.2.12-2 (page 2), Initial Fuel Loading Abstract, describes the fuel loading procedure used at Catawba. This procedure complies with this descrip-tion except for the use of temporary detectors. In addition, this procedure allows the use of the Boron Dilution Mitigation System (BDMS) detectors instead of just the Source Range Detectors. The BDMS detectors are just as accurate and were determined to be OPERABLE with only Secondary Source Assemblies during the Catawba 1 End of Cycle 4 unload. The use of BDMS Will allow N-31 and N-32 Containment evacua-tion setpoint to be reset without stopping the loading. In addition, the BDMS detector will initiate automatic boration if a dilution or subcritical multiplication accident should occur.
One other difference in this cycle loading is that during the Rod Control Cluster Assembly (RCCA) Eddy Current / Ultrasonic Testing, 4 4
RCCAs were found to have rodlet wear greater than allowable. Two RCCAs were replaced with spares that were onsite. The other two RCCAs were replaced with un-irradiated replacements obtained from Watts Bar.
These RCCAs are identical to the rest of the RCCAs. The core loading plan document will be revised to reflect the new control rod ids.
Instructions are included to avoid formation of unanalyzed clusters of fuel assemblies. This procedure creates no accident scenarios not already analyzed. The only loads being moved are fuel assemblies.
FSAR section 15.7.4, Fuel Handling Accidents in the Containment Building, is bounding. The accident in FSAR Chapter 15.4.7, Inadver-tent Loading and Operation of a Fuel Assembly in an Improper Location is also bounding for this procedure. To prevent this, the Fuel assemblies are identified as they are loaded. PT/1/A/4550/03C, Post 50
Refueling Core Verification, is performed after all fuel assemblies are loaded.
There ic no off-normal operation of safety equipment. No USQ is created by this procedure.
TN/2/A/0504/00/01A original This procedure provides implementation instructions for Nuclsar Station Modification (NSM) CN-20504, Rev. O. This NSM adds a flush connection to the stuffingboxes for the Unit 2 Nuclear Service Water (RN) Pumps. This modification installs a now stuffingbox and packing gland made of stainless steel. The new stuffingbox is drilled and tapped for the flush connections as well as existing connections.
These changes are intended to reduce build-up of corrosion products and facilitate flushing of the stuffingbox upper bearing area. This procedure provides guidance for performing this modification on RN pump 2A.
Implementation of this procedure requires RN Pump 2A to be isolated and removed from service. The RN supply to RN Pump 2A upper bearing and stuffingbox will be isolated also. Technical Specifications allow RN Pump 2A to be out of service for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during normal unit operation (Mode 4 or above). This procedure will be performed during a refueling outage; therefore, only Mode 5, 6, and No Mode require-monts apply. While Unit 2 is in Mode 5 and 6, only one RN pump is required to be operable. If this pump is lost, the Technical Specifi-cation Requirements are such that fuel movement will be delayed until the pump can be restored. In No Mode, no RN pumps are required to be operable. However, RN Pump 2B is to be maintained in an operable condition during this modification work. It is necessary to ensure that the RN system can perform its intended function by supplying cooling water flow to systems and components necessary for plant
. safety during normal operation and accident conditions. As described I in the FSAR, sufficient RN pump capacity is included to provide the cooling water to safely shutdown each unit, and valves are arranged in such a way that the loss of one train does not jeopardize the entire system. If both RN pumps 2A and 2B were out of service during this modification work, this demand could be met by RN pumps 1A and 1B. If unit 1 loses one of its pumps, the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement would apply. In this case, one RN pump could carry the Engineered Safe-guards loads on the operating unit and the decay heat loads on the non-operating unit. Sufficient redundancy of piping and components j
is provided to ensure that cooling is maintained to. essential loads at all times.
Prior to restoring RN Pump 2A to service, the stuffingbox and associ-ated piping will be leak tested to assure the system pressure boundary has been_ restored. Also, a flow balance will be performed to estab-lish proper flow to the pump bearings and stuffingbox. Therefore, temporarily removing RN pump 2A from service does not constitute a USQ.
51
_ . _ _ __ _ . _ _ ______ _._ _ ____ _ ~ . __ ._. .._
TN/2/A/0504/00/02A Original This procedure provides implementation instructions for NSM CN-20504, Rev. O. This NSM adds a flush connection to the stuffingboxes for the Unit 2 Nuclear Service Water (RN) Pumps. This modification installs a new stuffingbox and packing gland made of stainless steel. The new stuffingbox is drilled and tapped for the flush connections as well as existing connections. These changes are intended to reduce build-up of corrosion products and facilitate flushing of tne stuffingbox upper bearing area. This procedure provides guidance for performing this modification on RN pump 28.
Implementation of this procedure requires RN Pump 2B to be isolated and removed from service. The RN supply to RN Pump 2B upper bearing and stuffingbox will be isolated also. Technical Specifications allow RN Pump 2B to be out of service for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during normal unit operation (Mode 4 or above). This procedure will be performed during a refueling outage; therefore, only Mode 5, 6, and'No Mode require-ments apply. While Unit 2 is in Mode 5 and 6, only one RN pump is required to be operable. If this pump is lost, the Technical Specifi-aation Requirements are such that fuel movement will be delayed until the pump can be restored. In No Mode, no RN pumps are required to be operable. However, RN Pump 2A is to be' maintained in an operable-condition during this modification work. It is necessary to ensure that the RN system can perform its intended
- function by supplying cooling water flow to systems and components necessary for plant.
safety during normal operation and accident conditions. _As described in the FSAR, sufficient RN pump capacity is included to provide the cooling water to safely shutdown each unit, and-valves are arranged in such a way that the loss of one train does not jeopardize the entire system. If both RN pumps 2A and 2B were out of service during this modification work, this demand could be met by RN-pumps 1A and 1B. If Unit i loses one of its pumps, the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement would apply _ In this case, one RN pump could carry-the Engineered Safe-guards loads on the operating unit and the decay heat loads on the non-operating unit. Sufficient redundancy of piping and components is provided to ensure that cooling is maintained to essential loads at all times.
Prior to restoring RN Pump 2B to service,Ethe stuffingbox and associ-ated piping will be leak' tested to assure the system pressure boundary has been restored. Also, a flow balance will be performed to estab-lish proper flow to the pump bearings and stuffingbox. Therefore, temporarily removing RN pump 2B from service does not constitute a USQ.
OP/1/A/6250/03B Change #16 This change deletes valves 1BB-162 and 1BB-164 and adds valve 1BB-248 in enclosures for filling 1A and 1D steam generators per Nuclear l
Station Modification (NSM) CN-10588. This change will permit the refill of 1A and 1D steam generators without the_NSM paperwork being completed. Upon completion of the NSM paperwork, these changes will 52 i
be made permanent. This procedure change will not change the function l of the Steam Generator Blowdown (BB) or Steam Generator Wet Layup (BW) l systems. This change only affects the non-safety portion of the BB system in the Turbine Building. It does not involve or affect any i
part of a system, including the containment isolation portion of the l BB System, associated with any accident discussed in the FSAR. Thus, l this change does not involve a USQ.
MP/0/A/7600/107 Change #2 The changes were made in order to clarify the process of setting the
, disc lift stop (steps 11.8.24 and 11.8.25.) The change consisted of the addition of a note clarifying steps 11.8.24-25 and also providing an acceptance tolerance as provided by the manufacturer. The above changes should ensure proper lift stop setting during valve reassem-bly, and maintain the valve within its original design specification.
No USQs are involved.
MP/1/A/7150/42 Change #13 This procedure changed .evised the tolerance for stud elongation readings on step numbers 9.1, 11.5.58, 11.5.60.2, and 11.5.61.15 to read 0.47 (+.008/ .002) .045 to .055. The change was made to increase the stud clongation tolerance used for reactor vessel head tensioning l
per Exempt Change Variation Notice CE-2532. This change was evaluated under the Exempt Change and determined not to constitute a USQ.
PT/2/A/4450/03C Change #5 This restricted change was written to allow the performance of the Annulus Vacuum Decay Time test, using the revised criteria, in re-sponse to NRC Information Notice 90-02, PIR C-90-0054 which addresses inoperability of Control Area Ventilation (VC) due to the Auxiliary Building Ventilation (VA) interface, and an incorrect er.ror adjustment to the acceptance criteria. The vacuum decay time test measures the time that it takes the Annulus pressure to decay from -3.5 inches of water to -0.5 inches of water with Annulus Vertilatich (VE) shutdown.
The purpose of this test is to ensure less than 2,0*> scfm air in-leakage into the Annulus. The 2,000 scfm in-leakage is the assumed value used in CANVENT and ACTDOF to calculate dose for the post-LOCA conditions. This test is not required directly by Technical Specifi-cations, but in-leakage outside the above mentioned value may increase the dose above the appropriate 10CFR limits. This change only affects section 12.9 of this procedure.
Section 12.9 was revised to allow repeating the vacuum decay time test under different Auxiliary Building Ventilation (VA) and Fuel Pool Ventilation (VF) alignments in order to determine the affects of VA/VF ali oments on the decay time. An option to align VA train A in the LO n exhaust mode was also included. One momentary jumper is placed in 2DGLSA-2 that initiates the signal to VA. If an accident occurs 53
. _ _ . . - - _ =. . . .
i during this period, VA will function as designed because the LOCA alignment will already be established for VA (A Train.) VF is only required to be operable during fuel movement or with any overhead load above the fuel pool. A prerequisite to the test ensures that there is ;
no work in progress or planned during the performance of this test I that may require the operability of Unit 2 VF. Therefore, VA and Unit '
2 VF may be aligned as necessary for the test.
During the test, annulus pressure transmitter (2VEPT5001), which sends a signal to the pressure controller, is failed to allow drawing a negative annulus pressure greater than 4 inches of water. Compensato-ry measures were added to ensure that this transmitter is returned to l
service in a timel, lanner in the event of an accident or unusual event during the pr-?ormance of the test. In the event of an acci-dent, there will be to exhaust flow to the unit vent (after returning the transmitter to service) until the annulus pressur decays back down to the setpoint of -1.5 inches of water. Pulli g the Annulus to a pressure greater than 4 inches water vacuum is assumed not to be harmful because the penetrations are rated for pressures in the Psi range, and the annulus fans are not capable of pulling a vacuum greater than 1 psig. The time that it takes to decay,was reduced from 90 to 84 seconds based on calculation CNC-1240.00-00-0009. The limits of the decay test were error adjusted incorrectly on past tests, and this change implements the correct error adjustment which will give more conservative results.
Additional changes were made to include independent verifications for installing / removing and modifying instrumentation. Train A of VE was also chosen as the operating train because the A train transmitter that is failed is located beside the installation point of the test manometer. This further ensures that the transmitter can be returned to normal in a timely manner as mentioned in the above compensatory measures. The test manometer is installed at the vent for 2VEPT5000 because an isolation valve exists to isolate the manometer if neces-sary. Any alignment of the VE system will be under the appropriate Operations procedure, and the VE operation alignment was not modified other than discussed above. The VA system will be alignea using the Operations procedure except when the signal is simulated to establish the LOCA exhaust mode. If VA is aligned to the LOCA exhaust mode, it will be retuned to normal after completion of the VE vacuum decay time test. Therefore, a USQ does not exist.
TN/1/A/1214/00/02A Change #3 This change adds a hanger interference removal / reinstallation. It also moves the valve actuator step for setup prior to lifting red tags for testing. These changes will improve accessibility and also prevent any accidental damage to the actuator prior to the valve being properly setup. Approved procedures will be used for the hanger and valve work. The support will be reinstalled prior to any Mode in which it will be required for component or system operability. Thus, this change does not involve a USQ.
54
TN/1/A/1214/00/04A Change #4 This change adds one support for removal and reinstallation to aid in the replacement of valve 1BB060A. This support / restraint (S/R) will 1
be removed / reinstalled using approved procedures. It also allows electrical connections to be performed in conjunction with installa-tion of valve 1BB060A. This S/R be removed and reinstalled using
+
i apigoved procedures. The S/R will be reinstalled before any mode ,
change in which it is required for component / system operability.
j Based on the above, no USQ exists.
TN/1/A/1214/00/04A Change #5 i This change adds a hanger interference removal / reinstallation. It 3
also moves the valve actuator step for setup prior to lifting the red i tags for testing. These changes will' improve accessibility to the valve during installation and also will prevent any accidental damage
< to_the actuator prior to the valve being properly setup. Approved l procedures will be used for the hanger and valve work. The support-
- will be reinstalled before any mode change in which it is required for j component / system operability. Based on the above, no USQ exists.
l l PT/1/A/4550/09 Change #3 l This change replaces all of Enclosure 13.3 with a new. Enclosure _13.3.
The new enclosure gives the' location of all fuel. assemblies for Unit 1 i
Cycle 5. Blank spaces indicate which rack locations to inspect to
, ensure proper completion of PT/1/A/4150/18,. Fuel Assembly Insert
! Shuffle Procedure, that shuffles the inserts from the Unit 1 Cycle 4 "e
fuel assemblies to the Unit 1 Cycle-5 fuel assemblies. The change is needed to update'the locations for Cycle 5. This procedure is-de-4 signed to verify that all fuel assemblies that1are going back into the
- core for Cycle 5 are in the proper Spent Fuel Pool locations and.
- contain the proper Cycle 5 inserts. To-do this, each fuel assembly-is inspected _and its location, Region Reference Number,-and' insert ID.is i recorded. Then this record is compared'to-theLafter'shuffleLmap found in PT/1/A/4150/18. -If_any discrepancies are
- found, they are resolved-f before completion of the procedure. This procedure is designed:to
] prevent the accident analyzed in FSARLChapter 15.4.7, Inadvertent-l Loading and Operation of a Fuel-Assembly ~in an Improper Location.-
j While loads are transported across the spent fuel in_the pool, the loads consist only;of an underwater camera, support poles,-and cables.
. This is bounded by the accident analysis in FSAR chapter 15.'7.'4, Fuel j Handling Accidents. Thus, thic' procedure _ change does'not create a
, USQ.
i PT/1/A/4150/18 Change #8
] This change corrects a typographical _ error on Enclosure 13.1. The
, location incorrectly read G-38 instead of the correct location C-38.
i It does not create a USQ.
55 i . _ . . - _ _ - __ ._ _.
MPII 2.7.2 DCP-1 Change #3 This change allows the 1.313 i .015 counterbore at the Loop A Hot Leg 0 degree scoop position to be machined to a depth of 2.583" (.075" out of tolerance). Westinghouse has evaluated this change and determined that this nonconformance is acceptable. The nozzle spacing and reinforcement calculations were conservatively-based on the assumption that the 1.313" counterbore extended through_the reactor coolant pipe wall. Since the calculation allowed penetration through the pipe wall, this nonconformance has no affect on the original calculation and is acceptable. Thus, this change does not create a USQ.
MPII 2.7.2 DCP-1 Change #4 This change allows Westinghouse-to machine the Loop B Hot Leg thermowells to permit fit-up to the Loop B bosses. This is required due to a machining error in the Loop B boss configuratio.. Westing-house has evaluated this change-and determined that machining the Loop B thermowells is acceptable. The original 10CFR50.59 evaluation-performed by Westinghouse is not affected by this change. Thus, based on Westinghouse evaluation, this change does not represent a USQ.
PT/1/A/4450/01 Retype #6, Changes 6 to 7 Incorporated.
This retype changes the surveillance-test method for_which heater operation is verified. This retype will ennble Operations to use help from Instrument and Electrical (IAE) to actually energize all of the heaters. IAE will then take currents across the heater banks to verify all of the heaters are e.nergized. This_ method replaces the previous method which only verified that the heaters worked in the-automatic mode. During low humidity, this did not verify-that the '
heaters energized. IVPP-5690 is used for controlling the_ heaters.
1VPP-5670 and IVPP-5680, which are_used for tripping the Containment Purgo (VP) fans on high humidity, are not affected by the IAE signal.
To energize the heaters, IAE will input a signal between 4 and 20 mA, which will simulate a humidity which is above the 1VPP-5690 setpoint, to ensure that all heaters will remain energized. After 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of operation, the signal will be adjusted to simulate _a humidity which is below the 1VPP-5690 setpoint. This will de-energize the heaters, thus verifying the automatic operation of the heaters.E This retype put the VP system in compliance with Tech. Spec. surveillance 4.9.4.2.A. It does not create a USQ.
TT/1/A/9100/29 Initial Issue This procedure determines the number of starts that can be obtained from-one Diesel-Generator Starting Air (VG) tank precsurized to 235
- psig. The FSAR states that the VG system ha8 capacity for five starts on the Diesel Generator (D/G) with two tanks pressurized. As a result of the FSAR statement and-a proposed Tech. Spec. interpretation, the D/G is now declared inoperable if a VG tank drops below 235 psig. If i
56 1
s
- r-,--- ---w
one VG compressor is out of service, the o90*.abl9 compressor has to be crotts-tied in order to get two tanks abovt 235 psfg. This test is to be run in order to collect data on VG air consumption so that Design Engineering may revise their calculation on air consumption and possibly change the design requirements.
The D/G will be inoperable during this test due to one VG tank being isolated. The D/G will be started by the normal operating procedures except that one VG tank will be isolated for this test. The isolation of one tank reduces the volume of air that is available for starting and the number of starts that can be expected. Based on the above considerations, this temporary test does not create a USQ.
PT/0/A/4150/26 Initial Issoa This procedure,. Rod Control Cluster (RCCA) Assembly Ultrasonic / Eddy Current Testing, is being issued to direct all activit is associated with the aetup of Ultrasonic / Eddy Current examination equipment, use of this equipment to examine RCCAs, and the dismantling and removal of it following completion of examination. The testing apparatus em-ployed performs both Ultrasonic Testing (UT), which assesses external RCCA defects, and Eddy Current Testing (ECT) which quantifies MCCA clad cracking induced internally. All examination data is obtained and analyzed by approved Vendor procedure (which is incorporated as an enclosure of this procedure.) Such examinations are essential to the accurate assessment of the RCCA's cladding integrity..
The equipment affected by this procedure is the RCCAs themselves. All assumptions in the existing FSAR analyses concerning RCCA manipula-tions will be unaffected because the RCCAs Vf.21 at all times be handled in the Spent _ Fuel Pool'with the shae handling apparatus-that is used during normal Fuel Assembly Component shuffles. . Movement of l the RCCAs is performed under approved procedure at all_ times. The examination fixture is designed to fit into a Spent Fuel Pool storage
-rack location. It has been-designed to be compatible with the RCCA Change Tool. The conduits through which the RCCAs rodlets are lowered and raised are poly coated to preclude damage to external cladding-surfaces. Additionally, these conduits are sized such that they are larger-than the most restrictive diameter of a Fuel Assembly Guide Thimble. Verification of proper _ replacement of RCCAs into their designated fuel assemblies will be performed per PT/0/a/4150/26. This will ensure that all RCCAs will reside in-the proper core locations following core reloading.
Accuracy of examination data is validated by a rigorous pre-examina-tion benchmark using a Vendor supplied RCCA'with internal and external clad defects. Proper calibration of UT and ECT equipment is verified by correct identification of the manufactured defects. These examina-tions have no effect upon latching and manipulation of the RCCAs using the Rod Control System Hardware. Accident Analyses' based upon assump .
tions concerning stuck rods or misoperation of the Control-Rod System are therefore unaffected.
57
,- s -
,__....,r- y~ ,
This Non-Destructive Examination Program will ensure that the RCCAs with degraded cladding are replaced before failures during power operation can result in loss of absorber and thus shutdown margin.
This procedure does not create a USQ.
j PT/u/A/4400/08 Retype, Changes 0 to 42 Incorporated The purpose of this procedure is to ensure that all safety related l components receive adequate cooling water during a faulted Engineered )
Safety Feature (ESP) situation from the Nuclear Service Water (RN) l System. This procedure retype does NOT incorporate any changes not !
previously reviewed. Thus, it does not create a USQ.
1 PT/1/A/4350/02E Retype, Changes 0 to 47 Incorporated
SUMMARY
OF CHANGES The primary purpose of this retype is to provide.the Test Coordinator with a workable, more efficient, and safer procedure than the previous one. To fulfill this objective, the Procedure Writer's Guide was utilized, as well as past experience by Performance, and subject matter knowledge of systems and e7mponents by personnel from other affected station groups. As a secondary objective, ATWS/AMSAC testing is added to the procedure as Sections 12.10 and 12.11. A thorough review of all ATWS drawings, Instrument and Electrical (IAE) proce-dures, and 10CFR50.62 illustrated the need to perform ATWS/AMSAC testing on a periodic basis as presented in this procedure.
The following is a description of the specific changes involved in achieving these objectives. Under Section 1.0, the purpose of ATWS/AMSAC testing is stated. Under Section 2.0, references were i added which were needed to prepare the ATEL/AMSAC testing portions of the procedure. Under Section 3.0, Operations lineup time was added as 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. To comply with'the new Station Directive 3.0.12, ELECTRICAL i
CIRCUIT ISOLATION TAGGING, specific instructions were added in section 6.5. Section 7.0 was changed to have individual sign-offs for each procedure section. E:eps were added with SRO sign-off to verify that l the condensate-(CM) and Main Feedwater (CF) system will not fill the l steam generators when CF and Auxiliary Feedwater (CA) valves are
! aligned for each specific test. NOTES were added in certain subsec-tions of Section 8.0 to state that CA Pump 1A and/or CA. Pump 1B
( " Auto-Start Defeat" is non-functional unless its respective train of l
the Solid State Protcetion System (SSPS) is in OPERATE Mode. These NOTES will serve to alert the Test Coordinator to the fact that-
" Auto-Start Defeat" is non-functional based on SSPS conditions, so that other means may be utilized for effective pump operation-isola-tion.
Train 1A of SSPS must be-in OPERATE Mode to perform Sections 12.1.1 and 12.1.2. This has been added as a prerequisite to Subsections:8.1 and 8.2. The preferred method to block unwanted steam generator HI-HI and LO-Lo signals le to use plant scheduling as a means of ensuring 58
that either placement of specific trains of SSPS into TEST Mode or establishing non-alarm steam generator levels occurs as a prerequisite to those sections affected by steam generator alarms. The following steps of 8.0 were modified to reflect the individual requirements of the procedure sections in 12.0 which they represent: 8.1.8, 8.2.9, 8.9.3, 8.10.3, 8.11.3, 8.12.3, 8.14.1, 8.15.1, 8.18.1, and 8.19.1.
Steam Generator (S/G) Level Alarm requirements were removed from 8.3 and 8.4. Steps 8.3.5, 8.4.6, 8.7.6, and 8.8.7 were added to verify that the speed control potentiometer for CA Pump /1 is set on maximum prior to its 2/4 Low-Low S/G level Auto-Start and Blackout Start.
CAUTION statements in 8.7 and 8.8 were modified to state that the Auto-Start capability of CA Pumps 1A and 1B is being made inoperable by the electrical modifications of those steps--not the actual pumps themselves. The pumps will still have manual start capability.
Applicable steps have been modified to address blockage of P-14. Any mention of " Auto-Start Defeat" will use RESET and DEFEAT to instruct the Test Coordinator as to what modes " Auto-Start Defeat" should be in. In order to open 1CF10 and 1CP17, two jumpers must be placed to bypass ICM137 and 1CM140 open interlocks. These have been added to the procedure.
Link sliding steps and jumper placement steps have been renumbered so that minimal in-the-field backtracking will occur. Also, entries into the Radiation Control Area (RCA) have been minimized by placing these modification steps in a logical order. The Reactor Trip Breakers 1RTA and 1RTB cannot be racked in CLOSED unless the Digital Rod Position Indicator System (DRPI) is Operable. The alternative to this condi-tion is to white tag open Feeder Breakers 1LXC5B and 1LXD53 if DRPI is not operable. This logic is incorporated into the procedure. In Section 8.0, wherever interlocks would prevent Operations from per-forming a valve lineup, NOTES have been added to inform the Test Coordinator of the need to complete the other prerequisites to defeat these interlocks before proceeding with the valve lineup. Prerequi-site subsections were changed to use different sliding links to block CF Pump trip signals. The sliding links to be utilized are L-15 and L-19 in 1EATC4, L-25 and L-20 in 1EATC2. Changes'were also made in the restoration steps to reflect the inclusion of these new sliding links.
, ATWS/AMSAC prerequisites were added to the procedure as Subsections 8.20 and 8.21. In them, the Motor-Driven CA Pumps (CA Pump 1A and 1B) are racked open to TEST, thereby allowing only a simulated simultane-ous start of both CA pumps on an ATWS signal. All other trip signals are blocked so that only an ATWS nctuation will occur. Prerequisite steps to doenergize relays AH. *
,, 3F, AD, AC, and AB are included to simulate both CF Pumps' ATWS te S:y is reset so that an ATWS signal can demonstrate positively the -
Lam's ability to-start both CA pumps and trip the Main Turbine. In osuulon 12.10, the signal will be two out of three pressure switches tripped on two out of two CF Pumps. In Section 12.11, the signal will be three out of four CF flow paths blocked. For Section 12.11 the ATWS/AMSAC circuitry must have its turbine power 2 40% interlock bypassed. This is performed in prereq-uisite 8.21.8, where sliding link C-1 in 1ELCP0323 is opened and the RESET pushbutton for "AMSAC FOR CF VALVES" is depressed.
59
For the addition of ATWS/AMSAC procedure sections 12.10 and 12.11, 9.11 and 9.12 will be added to describe the test method utilized. For purpoces of explanation, they are quoted here 9.11 Section 12.10 The Main Turbine and both CF Pumps will be reset. The Motor-Driven CA Pump Breakers will be racked to TEST. Both CF Pumps will be tripped using ATWS/AMSAC circuitry. The simulated start of both Motor-Driven CA Pumps and the trip of the Main Turbine will be verified.
9.12 Section 12.11 The Main Turbine will be reset. The Motor-Driven CA Pump Break-ers will be racked to TEST. A loss-of 3/4 CF Flow Paths signal will be simulated. The simulated start of both Motor-Driven CA Pumps and the trip of the Main Turbine will be verified.
Section 10.0 will be modified to include Trip of Main Turbine and Motor-Driven CA Pump starts on ATWS/AMSAC' signals.
Acceptance Criteria for this procedure have been reviewed. It was decided to split the acceptance criteria into three separate specifi-cations. All starts for CA Pumps 1A and 1B will remain under Accep-tance Criterion 11.1. For CA Pump 11 (the Turbine-Driven CA Pump)
Auto-Start on 2/4 Low-Low Steam-Generator Level will be covered by 11.2, and Blackout Start will be covered by 11.3. The remaining acceptance criteria were renumbered in Section 11.0. 1As a part of the addition of ATWS/AMSAC testing to the procedure, the following specif-ic acceptance criteria were added:
- 1. Acceptance Criterion 11.10 was added to verify relays AH, AG, AF, AD, AC, and AB function properly from their ener-gized to deenergized status.
- 2. Acceptance Criterion 11.11 was added to verify that the Main Turbine trips and both Motor-Driven CA Pumps simulate start upon the simulated trip of both CF Pumps using ATWS circuit-ry.
- 3. Acceptance Critorion'11.12 was added to verify that the Main Turbine trips and both Motor-Driven CA' Pumps simulate start upon a loss of 3/4 CF Flow Paths signal. !
CAUTION statements throughout Section 12.0-involving manipulationc of sliding links and placing jumpers were deleted, since their purpose is fulfilled by the general statement in the Limits and Precautions regarding modifications for testing,-Step 6.4. In all procedure sections, whenever a positive actuation occurs, a NOTE has been placed to alert the Test Coordinator of the impending-action. In.all proce-dure sections, steps have been added to instruct Test Coordinator to log the test out of the Test Logbook if procedure section is complete.
In all procedure sections, instructions for valve and breaker lineup 60
restoration by operations has been reworded to be more specific in how lineups should be restored. Tho logic of closing sliding links and removing jumpers in Section 12.0 is designed so there will be minimal backtracking and RCA entries by Performance personnel during this test. Notes have been added to allow testing personnel to restore sliding links after sections with identical sliding link and jumper modifications are performed, subject to the provision of Limit and Precaution 6.4.
The following changes were implemented regarding Main Turbine Stop ^
Valves, Combined Intercept Valves, and Combined Interradiate Stop Valves. " Combined" was deleted; wrile an adequate reflection of the nature of these two components, it still does not reflect the_ plant labeling and nomenclature. They were therefore changed to be Inter-cept Valves and Intermediate Stop Valves. Abbreviations, which are a used on the Control Room Main Turbine Panel, are added after any I reference to the Main Turbine Stop Valves (MSVs), Intercept Valves (IVs), and Intermediate Stop Valves (ISVs). Instructions included on valve lineups provide instructions to the Control Room Operator as to how to open the Main Turbine Stop Valves and the Intercept Valves.
Sections 12.10 and 12.11 were added to test ATWS/AMSAC Actuation Circuitry. 'The means of test of Section 12.10 is to block all non-ATWS/AMSAC trip signals and then apply the actuation signal. This signal is generated by deenergizing relays AH, AG, AF, AD, AC, and AB such that 2/3 relays are deenergized on ecch CF pump, which results in a "both CF Pumps tripped signal" which initiates ATWS/AMSAC. The procedure is designed to test three different Test Plan combinations.
of 2/3 relays (one combination each time the procedure is perforned) such that all relays are eventually tested every three times the-procedure is performed. Previous test date-blanks are provided to enable the Test Coordinator to maPa the appropriate choice for Test Plan.
In Section 12.11, as in 12.10, a rotating test plan is also set up--this one involving practical combinations of valves which together give a "3/4 CF Flow Paths Blocked" ATWS/AMSAC initiation signal. The plans are so designed as to completely test all involved valves every four times-the procedure is performed.
4 The restoration circuitry of Sections 12.10 and 12.11 is designed so that no additional CA Auto-Starts:will occur, i.e., relays becoming reset and inadvertently actuating again.
Enclosure 13.20 was renumbered to' Enclosure 13.31. Enclosure 13.21 was deleted. Enclosures 13.27, 13.28,.13.29, Ord 13.30 were-added to 1 the enclosures as part of ATWS/AMSAC testing. All enclosures were i renumbered to reflect these changes. I Valves 1SM1, 1SM3, 1SMS, and 1SM7 were added as a means of isolating ,
the main condenser when the Main Turbine Stop Valves (MSV's) are- l opened for testing.
Additional changes were made of an editorial nature.
61
__ . _ _ . _ ~. __ _ . . _ . _ - . _ _ _ . _ _ ___
SAFRT's EVALUATION 1 i
Motor Driven CA Pumos Auto-Starts In these sections, each pump is aligned to recirculate to the Upper Surge Tank (UST) separately and is response time tested for various ,
I automatic start signals. Actuation of associated flow control valves is also tested in response to these signals. Accident flow rate is set up in advance by Operations. These sections are tested in either Mode 4, 5, 6, or No Mode, when CA is not required to be operable. The pumps are operated well within their mechanical design limits.
Turbine Driven CA Pumo Auto-Starts In these sections, CA Pump #1 is aligned to the UST and is response time tested for various automatic start signals. Actuation of associ-ated flow control valves is also tested in response to these signals.
Accident flow rate is set up in advance by Operations. These sections must be performed in Modes 1, 2, or 3 since Main Steam must be avail-able to run the pump. However, at least one motor driven CA pump will still be operable, and both will be available during this testing.
The pump will be run in recirculation mode and started by placement of a jumper which affects no other components. The pump will not be operated outside of design parameters. The required number of CA pumps will remain available.
Main Feedwater Isolation These sections will test Main Feedwater (CF) Isolation on Hi-Hi Doghouse Level and Hi-Hi Steam Generator Level. Steam Cenerator Level will be simulated high by manipulation of Process Control Cabinet logic cards and placing one train of SSPS in test. This testing will-be performed in Modes 4, 5, 6, or No Mode.
Placing jumpers for the Hi-Hi Doghouse Level causes feedwater isola-
! tion of valves pertinent to the affected doghouse. These actuations fail these valves to their safe position, and thus do not cause inoperability of any comoonents, regardless of mode. Process Control Cabinet modifications end placement of one train of SSPS in test are allowable in Modes 5 r,r 6. In Mode 4, however, one train of SSPS will be inoperable due t o being in test with the opposite train still remaining operable. In addition, manipulation of Process Control Cabinet logic cards will reverse the logic of the S/G level interlocks if levels are initially in the Hi-Hi state. This_will result in all four channels of S/G 1evel being failed low. The reactor will already be tripped, and the CA system is not required in Mode 4. The only equipment operation is the normal stroking of valves.
Main Turbine Trio The main turbine will be tripped on the following signals: trip of both CF pumps, reactor trip, Hi-Hi S/G level, and manual trip. The main turbine is not safety related, and since these trips will be performed with no steam passing through the turbine by having the Main Steam Isolation Valves (MSIV) closed, there will be no transient effect on the reactor coolant system, and no interaction with any safety related system.
62
t s ATWS/AMSAC Actuations The main turbine will be tripped and both Motor Driven CA pumps will simulate start on the following ATWS/AMSAC signals: 2/2 CF pumps tripped, and 3/4 CF flow paths isolated. These sections will be performed in Modes 4, 5, 6, or No Mode. Since the ATWS/AMSAC system is not a Technical Specification required system, operability is not a concern, except that in the test procedure, blocking of some safety related trip signals and interlocks is necessary to ensure positively that the test actuation is caused by the ATWS/AMSAC system. Although the main turbine will trip similar to Sections 12.5, 12.6, 12.8, and 12.9, the Motor Driven CA pumps will be racked Open to TEST, thus ensuring that these pumps will not mechanically start. Their breakers will close to verify actuation of ATWS/AMSAC. The valves manipulated 1 for section 12.11 are not required for containment closure, since the S/Gs are assumed to be adequate barriers to potential releases of radioactivity. In any case, the test may be aborted and all-CF valves closed immediately, should this be required. None of these components is required in Modes 4, 5, 6, or No Mode.
Based on the above discussion, this retype of PT/1/A/4350/02E does not constitute a USQ.
OP/2/A/6350/02 Retype #9, Changes 42 to 44 Incorporated The changes in OP/2/A/6350/02 (Diesel Generator Operation) Retype 9 are as follows: Corrected acceptable lube oil and cooling water temperatures from 140-160 plus or minus 10 Deg F while shutdown to 140 to 150 Deg F. This allows for a 5'F range from the 145'F setpoint.
Additicaally, the low temperature lube oil and jacket water annunciators alarm at 140*F decreasing, indicating that a problem exists at that temperature.
Renamed Enclosures 4.10 and 4.11 to eliminate confusion about remote and local starts and instead termed them " Starts from the D/G Room" and " Starts from the Control Room". Added a new section to Enclosure 4.17 concerned with swapping lube oil strainers and changed the Enclosure name as requested by Quality Assurance (QA -- QA CN-89-32.)
Incorporated procedure changes which added the Diesel Generator Starting Air (VG) Cross-tie 2A and VG Cross-tie 2B enclosures to the procedure. Added a new enclosure to swap fuel oil filters and strain-era due to procedural enhancements requested by QA. (QA CN-89-32)
Added a new enclosure to align the Fire Protection (RF) system to cool the Diesel Generator Engine Cooling Water (KD) Heat Exchanger when Nuclear Service Water (RN) is not available and the operation of the D/G is essential. Incorporated the procedure change which added the VG 2A and the VG 2B Cross-tie enclosures into the procedure.
In ENCLOSURES 4.6, 4.7, 4.8, and 4.9, corrected lube oil and jacket water temperatures in Steps 2 and 3. Raised upper level of lube oil sump tank to 105%, or one-half inch above the STATIC FULL mark on the dipstick per Maintenance recommendation (Step 17). Changed-the means of determining low jacket water level to the use of the low level annunciator. The Jacket Water Low level alarm is a Safety Related, j
63
calibrated instrument (Step 13). Corrected the location of the Run/Stop laob. Provided an additional means of determining whether greatar than the Tech. Spec, required volume of fuel oil is in the main fuol oil tanks by verifying the HAIN FO TANK TECH SPEC annunciator is not in alarm. If the alarm is lit, the Operator may then go tr - fuel oil tanks local gauge to verify that the volume is between 77, and 77,850 gallons to meet the Tech. Spec. requirement (Step 33).
In ENCLOSURES 4,10 and 4.11, corrected Lube Oil and Jacket Water temperature requirements in the initial conditions section. Reas-signed step numbers 2.11 through 2.15 as higher order steps. Placed note to remind Operator to fill out D/G Operational Parameters Period-ic Test.
In ENCLOSURE 4.12, corrected the location of the run/stop knob. In ENCLOSURE 4.13, added Step 2.2 to notify th1 Control Room Senior Reactor Operator (SRO) to make Tech. Spec. Log entries. Re-wrote Step 2.6 to give clear directions for the cases of taking a D/G out of set / ice with and without the opposite unit's same train D/G operable.
Added Step 2.15 to notify the Control Room SRO that the D/G is re-turned to service. In ENCLOSURE 4.14, changed the procedure steps to refer to the shutdown portions of enclosures 4.10 and 4.11 due to changes in those procedures.
In ENCLOSURE 4.17, re-wrote Stop 2.1 and Stop 2.2 to address venting the filters befere placing them in service, better describe the actual operation of the 3-way valve pairs, verify that the oil pressures are what are expected after the swap, and issue of Standing Work Requests (SWR) to replace the filter cartridges. Developed new Steps 2.3 and 2.4 to direct the Operator through the process of swapping strainers.
In ENCLOSURE 4.19, added Steps 2.1 and 2.23 to notify the Contrci Room SRO that the D/G is being placed in Maintenance Mode for the barring.
In ENCLJSURE 4.20, added sign-offs for D/G, date, and initials. In l
ENCLOSURES 4.21 and 4.22, incorporated procedure changes which added the enclosures to the procedure, and also made procedure changes to l
address the change in configuration of the cross-tie hose. (It now contains a bleed off valve.) Also made a correction to have the operator remain in the D/G 1A or 1B Room during the procedure's entirety.
Developed ENCLOSURE 4.23 to remove Diesel Generator Fuel Oil (FD) strainer swaps from the Annunciator Response Manual and place them in a procedure like other swaps, and to better describe the swapping process. The procedure also addresses the issuance of SWRs to clean the dirty filters and strainers. The enclosure gives an additional CAUTION on listening to the D/G while swapping filters or strainers and returning the filter / strainer to its original position if the D/G sounds like it is losing power due to the swap. Steps 2.3 and 2.4, concerned with swapping the fuel oil strainers, also includes the trending of differential pressures across the strainer and logging them on a logsheet before making the swaps. This is a current prac-tico covered by Tech. Mono, now incorporated into a procedure.
64
Developed ENCbOSURE 4.24 to align either of the Unit 2 D/Go KD Heat exchangers to the RF (Firo Protection System) in the event that the Nuclear Service Water System is unable to supply water to cool the system. This was a response to a scenario in which the NRC postulated a loss of RN. Design Engineering determined that 3870 KW could be generated by a single D/G when cooled by RF.
All changes have boon originated because of previous procedural unclarity, because of changes in another proceduro, or changes to the D/G by the NSM process. There are no now situations where the proba-bility of the loss of a D/G is increased. As the loss of one D/G is a previously analyzed event, thoso changes do not croato a USQ.
TT/3/B/9200/44 Initial Issue The purpose of this procedure is to perform testing to docume.t the natural frequencios of Quality Assurance (QA) condition 4 cioinet(s),
with particular emphasis on the lowest fundamental natural frequency.
Operating experience item VIL-W/87-29 expresses concerns for potential seismic intaractions of Class 1E electrical cabintcs. Most concerns have boon climinated by survey and ar,alytical worn. To complete work on this item, it is necessary to determine natural frequencies of two QA condition 4 cabinets which are next to Class lE electrical cabi-nets. The cabinets of interest are the Loose Parts Monitor and the Incore Instrumentation Cabinet, both of which are located in the control Room. To test for the natural frequencies of these cabinets, it is necessary to open the cabinot, attach instrumentation to the cabinet wall or framo, induce froo vibration, and measure the natural frequencies of that vibration, Freo vibration will be induced using a rubbor tip, instrumented force hammer. The stresses induced by stubbing the cabinet will not harm the cabinet. Post test operation and calibration checks will bo i performed in accordance with applicable technical specification requirements. The cabinots to be tested will be deenergized, and the plant is in Mode 6 during the test. .
This test is performed while the equipment is deenergized and not required by Technical Specifications. The test does not leave any
, permanent changes after completion, and the cabinets are performance tested for operability, prior to being placed into use. The cabinets being tested are not important to safety, and the tests wi?.1 not cause i
sufficient stress to adversely affoct those cabinots into structural i failure. This test does not add, dolote, or permanently alter the l affected equipment in any way. No new failure modes are added. No l plant set points, safety limits, or design paramoters will be adverse-l ly affected. For those reasons, it is concluded that no USQ exists.
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l PT/1/A/4350/02E Chango #47 This change has been written to allow installation of needed instru-montation to monitor critical data on Auxiliary Feedwater Pump Turbine 65
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(CAPT) /1. Specific parameters to be measured in addition to the CAPT
- 1 response time to an error adjusted blackout signal aret (1) pump pressure versus time and (2) pump speed versus time. These parameters will be recorded simultaneously by a test visicorder. The visicorder will receive its start signsi from a double polo switch. One set of leads from this switch will be connected across the affected link, which directly opens 1SA2 or 1SAS (CAPT steam supply valves). The other set of leads will be connected across the visicorder to give it the start signal. The tachometer output, likewise, will be placed across sliding link M-5 to measure the signal to the AFWPTCP RPH process meter. This meter is not used by the control room, and measurement by this procedure of the same will not affect the control room RPM reading, since it is a separate circuit. Finally, the pump outlet pressure transmitter is totally separate from process instru-mentation, except for its hookup to the pressure tap prior to running the pump.
All modifications are independently verified to be installed and removed. The test will align the CAPT to the Upper surge Tank. Pump operability will not be affected. The pump may be made operable by closing 1CA68, even if the required test equipment is still installed.
For these reasons, this procedure change does not create a USQ.
OP/1/A/6350/02 Retype #13, Changes 58 to 60 Incorporated The changes in OP/1/A/6350/02 (Diesel Generator Operation) Retype 13 are as follows: Corrected acceptable lube oil and cooling water temperatures from 140-160 plus or minus 10 Dog F while shutdown to 140 to 150 Dog F. This allcws for a 5 Dog F range from the 145 Deg F setpoint. Additionally, the low temperature lube oil and jacket water annunciators alarm at 140 Deg F decreasing, indicating that a problem exists at that temperatura.
Renamed Enclosures 4.10 and 4.11 to eliminate confusion about remote and local starts and instead termed them " Starts from the 0/G Room" and " Starts from the Control Room". Added a new section to Enclosure 4.17 concerned with swapping lube oil strainers and changed the Enclosure name as requested by Quality Assurance (QA-- QA CN-89-32.)
Incorporated procedure changes which added the Diesel Generator Starting Air (VG) Cross-tie enclosure to the procedure. Added a new enclosure to swap tuel oil filters and strainers due to procedural enhancements requested by QA. (QA CN-89-32) Incorporated the proce-dure change which added the VG 1A and the VG 1B Cross-tie enclosures into the procedure.
In ENCLOSURES 4.6, 4.7, 4.8, and 4.9, corrected lube oil and jacket l water temperatures in Steps 2 and 3. Raised upper level of lube oil I sump tank to 105%, or one-half inch above the STATIC FULL mark on the dipstick per Maintenance recommendation. (Step 17) Changed the means of determining Ir.V jacket water level to the use of the low level annunciator. (The Jacket Water Low level alarm is a Safety Related, calibrated instrument.) Corrected the location of the Run/Stop knob.
Provided an additional means of de'ermining whether greater than the 66
. _ _ - . - _. - ~ . . _. _ . _ _ . ._. .-
Tech. Spec required volume of fuel oil is in the main fuel oil tanks by verifying the " MAIN FO TANK TECH SPEC" annunci6 tor is not in alarm.
If the alarm is lit, the Operator may then go to the fuel oil tanks local gauge to verify that the volume is betwoon 77,100 and 77,850 gallons to meet the Tech. Spec. requirement. (Step 33)
In ENCLOSURES 4.10 and 4.11, corrected Lube Oil and Jacket Water J temperature requirements in the initial conditions section. Reas- :
signed stop numbers 2.11 through 2.15 as higher order steps. Placed I I
note to remind Operator to fill out the Diesel Generator (D/G) Opera-tional Parameters Periodic Test (PT). In step 2.40, eliminated the duplicate sign-off for closure of valves 1FD-96 and -98.
In ENCLOSURE 4.12, corrected the location of the run/stop knob. In ENCLOSURE 4.13, added Step 2.2 to notify the Control Room Senior Reactor Operator (SRO) to make Tech. Spec. Log entries. Re-wrote Step 2.6 to give clear directions for the cases of taking a D/G out of service with and without the opposito unit's same train D/G operable.
Added Step 2.15 to notify the Control Room SRO that the D/G is re-turned to service. In ENCLOSURE 4.14, changed the procedure steps to refer to the shutdown portions of enclosures 4.10 and 4.11 due to changes in those proceduros.
In ENCLOSURE 4.17, re-wrcte Step 2.1 and Step 2.2 to address venting the filters before placing them in service, better descrioe the actual operation of the 3-way valve pairs, verify that the oil pressures are what are expected after the swap, and issue Standing Work Requests (SWRs) to replace the filter cartridges. Developed new Steps 2.3 and 2.4 to direct tha Operator through the process of swapping strainers.
This was to remove these actions from the Annunciator Response Manual and place them in a procedure, similar to other filter and strainer swaps.
In ENCLOSURE 4.19, added Steps 2.1 and 2.23 to notify the Control Room SRO that the D/G is being placed in Maintenance Mode fcr the barring.
In ENCLOSURE 4.20, added sign-offs for D/G, date, and initials. In ENCLOSURES 4.21 and 4.22, incorporated procedure changes which added the enclosures to the procedure, and also made procedure changes to address the change in configuration of the cross-tie hose (it now contains a bleed off valve.) Also made a correction to have the Operator remain in the D/G 1A or 1B Room during the procedure's entirety.
Developed ENCLOSURE 4.23 to remove Diesel Generator Fuel Oil (FD) filter and strainer swaps from the Annunciator Response Manual and place them in a procedure like other swaps, and to better describe the swapping processes. The procedure also addresses the issuance of SWRs to clean the dirty filters and strainers. The enclosure gives an l additional CAUTION on listening to the D/G while swapping filters or i
strainers and returning the filter / strainer to its original position if the D/G sounds 11ke it is losing power due to the swap. Steps 2.3 and 2.4, concerned with swapping the fuel oil strainers, also include the trending of Differential Pressures across the strainer and logging them on a logsheet before making the swaps. This is a current 67
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practico covered by Tech. Memo which has now been incorporated into a j proceduro.
All changes have boon originated because of previous procedural unclarity, because of changes in another proceduro, or changes to the D/G by the Nuclear Station Modification process. There are no new situations created in which the probability of the loss of a Diesel Generator is increased. The loss of one D/G is a previously analytod event. Thus, those changes do not create a USQ.
OP/1/A/6250/01 Change #68 This restricted change was made to provide guidanco for full Differen-tial Pressure (D/P) stroko verification of Auxiliary feedwater (CA) valves 1CA-149, 150, 151, and 152. These verifications were performed in Mode 5. Those changes were made because both trains of the Solid State Protection Nystem (SSPS) wore in test at the time. In this condition, CA auto start is defontod, and P-14 is alno defeated. The procedure has been revised to moet this condition. Main Feodwater (CF) pump speed will remain protected by an overspeed trip. CF pump miniflow will remain with the CF pump recirculation control valve in automatic. CF pump runout is not expected. Steam generator (S/G) overpressurization is not expected because vacuum will be lined up on each S/G being tested. S/G overfill should not occur because the feedwater regulating valve bypass valvo, which is the only path feeding the S/G, will be closed at 65% Narrow Range Level. If closing the valve does not stop the level increase, the CF pump will be tripped at 80% level by the control Room operator. S/G chemistry will be maintained as normal operation. Thus, this chango does not create a USQ.
OP/1/A/6450/10 Change #21 This chango has been generated as a procedure enhancement as a result of PIR 0-C89-0372. This PIR addressed the potential impact of the Containment Hydrogen Control Systems (VY) on the Annulus Ventilation System (VE), and the possibility that the operation of VY would increase the annulus pressure to greater than -1.5 inches of water vacuum. The results of the PIR determined that the VE system would quito adequatoly control the annulus pressure at the -1.5 inches of water setpo'it with VY in service, but a precautionary statement to insure that the pressure remained at sotpoint was in order. There-foro, the CAUTION before stop 2.4 was added.
Additional research revealed that in an accident scenario in which hydrogen concentration reached the levels requiring the use of the VY system, several factors would preclude its use. Hydrogen buildups in containment due to the combined offects of Zirconium-Water reaction, radiolysis of the core and sump, corrosion of metals, and primary coolant hydrogen release to the concentration requiring a hydrogen release would also be accompanied by high radioactive noble gas concentrations. Unless the presibility of containment failure due to 68
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internal pressure combined with an uncontrolled hydrogen burn pressure spike existed, there would be no release of containment atmosphere through VY. If this condition existed, the decision to use VY would only come through direction by the Technical Support Center (TSC).
This would only come as a last possibility, after failure of the redundant, safety related Hydrogen Recombiners and the Hydrogen Igniters. The release could only be performed under the direction of a Gaseous Waste Release (GWR) Package, in which the calculations for offsite Dose had been performed. Due to these factors, changes were made to place the procedure r :eps that allowed entry into the proco-dure only at the direction of the TSC. References to required hydro-gen concentrations were deleted because concentrations above 3.0% may be required to be present based on the ability of the containment vessel to survive a hydrogen burn, as opposed to a release. The requirement for a GWR was placed in the proceduro as an INITIAL CONDITION. References to these requirements were also noted in the LIMITS AND PRECAUTIONS section of the procedure. These changes do not create a USQ.
OP/2/A/6450/10 Change #9 This change has been= generated as a procedure enhancement as a result of PIR 0-C89-0372. This pIR addressed the potential impact of the containment Hydrogen Control Systems (VY) on the Annulus Ventilation system (VE), and the possibility that the operation of VY would increase the annulus pressure to greater than -1.5 inches of water vacuum. The results of the PIR determined that the VE system would quite adequately control the annulus pressure at the -1.5 inches of water setpoint dth VY in service, but a precautionary statement to insure that the pressure remained at setpoint was in order. There-fore, the CAUTION before step 2.4 was added.
Additional research revealed that in an accident scenario in which hydrogen concentration reached the levels requiring the use of the VY system, several factors would preclude its use. Hydrogen buildups in l containment due to the combined effects of Zirconium-Water reaction, radiolysis of the core and sump, corrosion of metals, and primary coolant hydrogen release to the concentration requiring a hydrogen release would also be accompanied by high radioactive noble gas concentrations. Unless the possibility of containment failure due to internal pressure combined with an uncontrolled hydrogen burn pressure-spike existed, there would be no release of containment atmosphere through VY. If this condition existed, the decision to use VY would only come through direction by the Technical Support Center (TSC) .
This would only come as a last possibility, after failure of the redundant,. safety related Hydrogen Recombiners and the Hydrogen Igniters. The release could only be performed under the direction of a Gaseous Waste Release (GWR)-Package, in which the calculations for offsite Dose had been performed. Due to these factors, changes were made to place the procedure steps that allowed entry into the proce-dure only at the direction of the TSC. References to required hydro-gen concentrations were deleted because concentrations above 3.0% may be required to be present based on the ability of the containment i
69
l vessel to survive a hydrogen burn, as opposed to a release. The requirement for a GWR was placed in the procedure as an INITIAL l l
CONDITION. References to these requirements were also noted in the LIMITS AND PRECAUTIONS section of the procedure. These changes do not create a USQ.
l MP/0/A/7200/01 Change #4 Steps 11.5.15 and 11.5.18 on the data sheet were revised to show the new torque value (400 ft. Ib.) provided by the Vendor and included in l the vendor manual by Exempt Change CE-2517. Step 11.5.17.1 was added to provide a torque value (17-19 ft. Ib.) and step for torquing the stem connector and special nut. This information was also provided by the Vendor and included in the vendor manual. Enclosure 13.1 was revised to add the appropriate sign-offs. The changes made by this revision have been provided by the vendor and included in the vendor manual by Exempt Change CE-2517. The corrective maintenance con-trolled by this procedure will return the governor valve to as-built conditions. These actions will ensure the governor valve's compliance with FSAR accident analysis. Thus, no USQ exists.
TN/5/A/0375/00/01A Original
, This procedure providen guidance for implementation of Nuclear Station l Modification (NSM) CH-F0375 Rev. O. This NSM adds an oil reclaim
, system to each of the chilled Water (YC) control room area chillers.
l The intent of this modification is to ensure lubrication oil is l returned to the compressor during low chiller load operation and prevent underfilling, which results in subsequent alarms and trips on low oil level, thus making the chiller unit more reliaLc.e. This specific procedure will install an oil reclaim line for YC Control Room Area chiller 1CRA-C-1.
The Control Room Area Chilled Water System (YC) is an engineered safety feature consisting of two 100% capacity water chillers, pumps, piping, and control systems. Each train of Control Area Ventilation (VC) and chilled Whter (YC) must have its chiller operable to be considered operable. At least one VC system is required to be opera-( ble at all times. Each redundant VC/YC train is served from a sepa-l rate train of emergency class 1E power. Therefore, with Chiller 1CRA-C-1 out of service for implementation of this procedure, this assures the integrity and availability of one train of VC in the event of any single active failure. 1CRA-C-1 may be isolated and field work performed during any mode as long as B Train of VC/YC remains opera-ble, which is a requirement of this procedure. This procedure re-quires that ICRA-C-1 be returned to service within 7 days as required by t}e Limiting Condition for Operation of Tech. Spec. 3/4.7.6. For those reasons, this procedure aoes not create a USQ.
TN/SlA/0375/00/02A Original 70
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)
This procedure provides guidance for implementation of NSM CN-50375 Rev. O. This NSM adds an oil reclaim system to each of the YC control room crea chillers. The intent of this modification is to ensure lubrication oil is returned to the compressor during low chiller load operation and prevent underfilling, which results in subsequent alarms and trips on low oil level, thus making the chiller unit more reli-able. This specific procedure will install-an oil reclaim line for YC l Control Room Area chiller 2CRA-C-1.
The Control Room Area Chilled Water System (YC) is an engineered safety feature consisting of two 100% capacity water chillers, puraps, piping, and control systems. Each train of Control Area Ventilatlon (VC) and chilled Water (YC) must have its chiller operable to be considered operable. At least one VC system is required to be opera-ble at all times. Each redundant VC/YC train is served from a sepa-rate train of emergency class 1E power. Therefore, with Chiller 2CRA-C-1 out of service for implementation of this procedure, this !
assures the integrity and availability of one train of VC in the event of any single active failure. 2CRA-C-1 may be isolated and field work l j
performed during any mode as long as A Train of VC/YC remains opera- I ble, which is a requirement of this procedure. This procedure re- '
quires that 2CRA-C-1 be returned to service within 7 days as required by the Limiting condition for Operation of Tech. Spec. 3/4.7.6. For-these reasons, this procedure does not create a USQ.
TN/1/A/2761/CE/01A Original This procedure provides implementation instructions for Exempt Change CE-2761. This exempt change will reposition the operating elevation of the fully withdrawn Rod Control Cluster Assemblies (RCCAs) from 230 steps to 225 steps. In addition, the bank overlap will be changed to 112 steps. This will be accomplished by changing the thumbwheel settings in the IRE System Logic Cabinet (1ERCC0006) . No isolations are required, and the changes will be made while the unit is in Modes 3, 4, 5, 6, or No Mode. Functional verifications will be done by both Instrument and Electrical (IAE) and Performance personnel. This will consist of Rod Drop Time Testing and verifications of overlap set-tings, respectively. This testing has been discussed with the system expert and has been determined to adequately test the modification.
This change has been evaluated under Exempt Change CE-2761, and determined not to create a USQ.-
TT/1/A/9200/58 Initial Issue This procedure will serve as a retest per Post Modification Testing letter for Nuclear Station Modification (NSM) CN-10675. This proca-dure will stroke valves 1CA48, 1CA52, and 1NV101A whose circuitry was modified to include Safe. Shutdown Facility (SSF) disconnect pins on the SSF transfer plugs, which will function to open 1CA48 and 1CA52-and to close_1NV101A when respective transfer disconnect' plugs are transferred from PLANT MODE to SSF MODE. Once in their fail safe position, the inability to manipulate these valves by_non-SSF means 71
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Will be verified. The other valves affected by manipulating the SSF transfer disconnect plugs may be in any position to perform this section of the test; however, if those valven are open, they will fail to their fail safe position, which is closed. CAUTION statements and a prerequisito requiring the Test Coordinator to notify operations that manipulating the transfer disconnect plugs will cause these valves to close, if they are not closed already, has been included.
Valvo ICA50 had its circuitry modified by NSM CH-10675 to open and remain unable to be closed by non-SSF means on a transfer of control power from " Normal Incoming Breaker Fed from MCC 1EMXA" (Breaker 1EMXS-F01A) to "Alternato Incoming Dreaker Fed from MCC SMXG" (Breaker 1EMXS-F03A). This valve will be aligned closed and stroked to open as directed by the procedure. All other valves affected by transferring control upon swapping breakers 1EMXS-F01A and 1EMXS-F03A may be in any position; however, upon swapping those breakers, they will fail closed. A CAUTION statement, as well as a prerequisite requiring the Test coordinator to inform operations that the valves will close, should they be open, has been included. Additionally, direct control of a number of valves will be transferred to the SSF by swapping the breakers. A CAUTION statement, as well as a prerequisite requiring the Test Coordinator to inform Operations that the valves will be controllable only from the SSF once control power is transferred to the SSF, has been included.
Stroko tests of all valves affected by NSM CN-10675, as have been considered to be testable by TT/1/A/9200/57 and NOT the implementation procedure, will bo performed either by IWV test procedures, or by Operations personnel or their designees from the control Room /NI Test Panel. No stroke times are considered, since this is not a factor in determining the soundness of affected valves' circuitry. Whether or not a valve electrically moves la the only criteria needed to verify that the valves circuitry has not been adversely affected by the performance of this modification.
' The SSF functions to take Unit 1 to a safe shutdown condition, and this test will only be performed in Modo S, 6, or No Mode. Operations is notified of the status and impending results of affected components before control is transferred to the SSF. The safety function of these valves (control by non-SSF means) is specifically tested by this procedure. Based on the above, no USQ exists.
OP/1/A/6250/01 Change #67 This restricted change provides procedural guidance on performing a Full Differential pressure Stroke Verification for the Main Feedwater Bypass to Auxiliary Feedwater Nozzlo Valves (1CA-149, -150, ~151, and
-152.) The information obtained in this test will be used to deter-mine the valve operability by design.
These verifications will be performed in Mode 5. Main Feodwater (CF) pump spood will remain protected by an overspeed trip. CF pump miniflow will remain with the CF pump recirculation control valve in 72
automatic. .CF pump runout is not expected. Steam generator (S/G) overpressurization is not expected because vacuum will be lined up on cach S/G being tested. S/G overfill should not occur because the ,
feedwater regulating valve bypass volve, which is the only path feeding the S/G, will be closed at 70% Narrow Rangn Level. If closing i the vcive does not stop the level increase, the CF pump will be tripped by P-14 Feedwater Isolation Signal. S/G ch3mistry will be maintained as normal operation. Thus, this change does not create a USQ. i PT/1/A/4600/05E Reissue, Changes 0 to 3 Incorporated -
This procedure is performed before the beginning of each fuel cycle to ensure that the margin to reactor trip setpoints for the power ranges .
and the intermediate ranges is not adversely affected by changes in I the core radial power distribution as a result of the new core load- ;
ing.
i The significant technical ghange included in this reissue changes the i source for the " previous Po for intermediate range calculation from >
thefullcoremaptakenwitN"thelastincore/execrecalibrationtothe i first full power full core map of the previous cycle. This change was :
made because intermediate range setpoints are determined at the [
beginning of a cycle during the initial startup of the cycle. The first full core map is more representative of the power distribution which existed when the intermediate range setpoints were datermined I than a full core map near the end of the cycle. The method used to l calculate the new intermediate range setpoints is not affected.
I changes were made to the procedure to allow the use of the Startup and !
Operational Report (SOR) provided by Duke Design for the predicted Fa The SOR is a controlled, Quality Assurance (QA) condition 1 maNu.al which provides information needed-for the performance of this !
test, as well as for physics tests and reactivity calculations. ;
i l Changes made to the procedure affect the Nuclear Instrumentation i System only. Changes made ensure that the calibration currents for i the new cycle are calculated properly. . Calibrations are performed j under approved Instrument Procedures by-qualified Instrument and Electrical personnel. There is no physical change to the system and ;
no change to the intended function of the system. This change does i not create a USQ. ;
r AP/1/A/5500/06 Retype #9, Changes O to 0 Incorporated The following changes were made in Retype #9. The_ reference to .
procedure EP/1/A/5000/2A1, Nuclear Power Generation /ATWS, was deleted from Step C.1 Response Not Obtained-(RNO) column in Case I. This i action is already covered in EP/1/A/5000/01, Reactor Trip or Safety Injection, and does nc.t need to be referenced in this RNO.
t 73 l
A new caution was added prior to stop C.1 in Case II, and an RNO ctatement was added to the same stop. This RNO directs the operator to close a manually operated valve (ICS-69) if 1CA-6 cannot be closed.
If this procedure is used during a loss of offsite power, and the 1A Diesel Generator should fail to start, then 1CA-C will not have power.
If this is the case, it is easier to isolate the Auxiliary Feedwater Condensate Storage Tank (CACST) using 1CS-69, sinco this valve is 1 easier to access. The caution which was added prior to this step warns the operator that closing ICS-69 will isolate the CACST from the Unit 1 and Unit 2 Auxiliary Feedwater (CA) purps. This was added to ensure that the operator is aware of the consequences of this action.
A now step D.1 was added to Case II to take into account the possibil-ity that a Hotwell pump may not be available for transferring water from the Hotwell to the Upper Surge Tank (UST). Two new cautions were added prior to step D.3 in Case III to warn the operator of the automatic actions which could occur if the CA pumps are aligned to the Hotwell. This information was recently discovered during the perfor-mance of a test with the CA pumps aligned to the Hotwell. Valve ICA-178 was added to step D.6 in Case II. This valve must be open to ensure a flowpa~h from the Condenser Circulating Water (RC) piping to the CA pump suction. This valve is normally aligned in the open position. Steps D.7 through D.13 in Case II and Enclosures 1 and 2 were added to provide guidance to the operators in aligning the CA pump suction source when the Hotwell pumps are not available for transferring wrter from the Hotwell to the UST.
Additional changes were made to correct various labelling errors, typographical errors, and format errors. None of the above changes create a USQ.
AP/2/A/5500/06 Retype #5, Changes 0 to 0 Incorporated The following changes were made in Retype #5. The reference to EP/2/A/5000/2A1, Nuclear Power Generation /ATWS, was deleted from Step C.1 Response Not Obtained (RNO) column in Case I. This action is already covered in EP/2/A/5000/01, Reactor Trip or Safety Injection, and does not need to be referenced ir this RNo.
A new caution was added prior to step C.1 in Case II, and an RNO statement was added to the same step. This RNO directs the operator to closa a manually operated valve (ICS-69) if 2CA-6 cannot be closed.
If this procedure is used during a loss of offsite power, and the 2A Diesel Generator should fail to start, then 2CA-6 will not have power.
If this is the case, it is easier to isolate the CACST using 1CS-69, since this valve is easier to access. The caution which was added prior to this step warns the operator that closing ICS-69 will isolate the CACST from the Unit 1 and Unit 2 CA pumps. This was added to ensure that the operator is aware of the consequences of this action.
A new step D.1 was added to Case II to take into account the possibil- !
ity that a Hotwell pump may not be available for transferring water from the Hotwell to the Upper Surge Tank (UST). Two now cautions were 9
1 74
added prior to step D.3 in case III to warn the operator of the automatic actions which could occur if the CA pumps are aligned to the Hotwell. This information was recently discovered during the perfor-mance of a test with the CA pumps aligned to the Hotwell. Steps D.7 through D.13 in Case II and Enclosures 1 and 2 were added to provide guidance to the operators in aligning the CA pump suction sources when the Hotwell pumps are not available for transferring water from the Hotwell to the UST.
Additional changes were made to correct various labelling errors, typographical errors, and format errors. None of the above changes create a USQ.
TT/1/A/9200/59 Initial Issue The data obtained by this procedure will be used to benchmark Safety Analysis' Chemical and Volume Control (NV) system computer model which is used to verify setpoints used in Emergency Operating Procedures (EOPs). This test requires that the NV system is operating in normal letdown and charging mode. During normal operation, valves 1NV294 and 1NV309 Arc approximately 50% open to maintain stable NV and Reactor Coolanc (NC) system operation. The test instructs the operators to fully open these valves. (These are " fail open" valves.) This action will result in an increase in charging flow with some change in Reactor Coolant Pumps' seal flow rates. The increase in charging flow will result in increasing Pressurizer level and decreasing Volume control Tank (VCT) level.
These transients can be mitigated by opening both the 45 and 75 gpm letdown orifices if the Control Room Operator so chooses, per the subject procedure, in accordance with the NV Operation Procedure. The subject procedure provides limit and precaution guidance to the Operator. Pressurizer and VCT minimum and maximum level, and Pressur-izer temperature rates of change, NC system temperature rates of change, and Reactor Coolant Pump minimum and maximum seal limits are outlined in the procedure. If any of these limits are exceeded, the procedure instructs the Test Coordinator to abort the test and in-struct the Control Room Operator to manipulated the above mentioned valves to achieve the desired system conditions.
These limits are within alarm setpoints for Pressurizer and VCT levels and administrative limits for temperature rates of change for the i pressurizer and NC system. The allowable seal flows are within the normal operational limits for the individual reactor coolant system pumps and ensures that the Tech. Spec, total seal flow rate criteria is not exceeded.
In the required mode for this test (Mode 3), all Pressurizer Power Operated Relief Valves (PORVs) are required to be operable per Tech.
Specs. No Emergency Core Cooling System (ECCS) subsystem will be rendered inoperable as a result of this test. Since the charging pump l will essentially be operating in recirculation through the NC system with no boron dilution occurring, and since adequate shutdown margin 75
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l is ensured at Mode 3 boration levels at 0 Cycle Burnup to below Mode 6 NC temperatures, any shutdown margin concerns are precluded. Thus, this test does not create a USQ.
PT/1/A/4550/03A Retype #2, Changes 0 to 3 Incorporated The reference section was changed to add a reference to Section 4.3 of the Catawba Nuclear Station Performance Manual, and to change the reference to Regulatory Guide 5.13 to be 5.29. Regulatory Guide 5.29 is the correct Regulatory Guide for Commercial Power Reactors. The note from the end of section 3.0 was moved to the beginning, and the phrase "after each refueling of the reactor" was added. This reflects the requirements of Regulatory Guide 5.?9. The phrase "and after each refueling of the reactor" was added to step 6.1, per Regulatory Guide 5.29. Steps 8.1 and 8.2 were deleted because Spent Fuel Pool Ventilation (VF) and Radiation Monitors (EMFs) EMF 20 and EMF 21 are not required to be operable for this inventory. The note in section 8.0 was deleted because it is no longer needed. Step 12.9 and the note following were deleted because these requirements are being satisfied by step 12.10. In enclosure 13.3, the required information for step 12.10 was deleted, but the information for 12.9 was changed to include "Dete/ Time".
This inventory procedure is not described in the FSAR, but it is mandated per Regulatory Guide 5.29, Nuclear Material Control Systems for Nuclear Power Plants. This procedure does not put any safety features in an abnormal configuration, and in no way does it compro-mise the safety evaluation in the FSAR for the new fuel storage vault.
Thus, these changes do not create a USQ.
PT/1/A/4250/14 Initial Issue
, The purpose of this procedure is to response time test the Auxiliary Feedwater Pump Turbine (CAPT) #1/ Auxiliary Feedwater (CA) Pump #1 l
Governor and any equipment which could directly or indirectly affect its performance. This will verify proper operation during pump starts and normal operation. The pump will be run in recirculation to the Upper Surge Tank. In this test alignment, CAPT #1 will not be avail-able for emergency actuation without local operator actions. This condition exits for sections 12.1 and 12.2, in which flow is set up to
! a desired.value to the Upper Surge Tank. In Sections 12.3 and 12.4, l however, CAPT #1 is run in miniflow to the Upper Surge Tank, and is thus still capable of providing required flow to the Steam Generators in the event of a CA autostart or. blackout. The Operations Coordina-tors Group schedules CA pump testing such that the required number'of CA pumps remain operable-for the duration of this test. Also, the operations Shif t Unit Senior Reactor Operator (SRO) will verify that prope* CA system operability is maintained before allowing this test to begin.
The electrical circuit modifications needed to perform this. test-only_
affect CAPT #1 and do not affect the other two CA pumps. These 76
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. 1 l
modifications are independently verifind during placement and removal.
During Sections 12.1 and 12.3, CAPT #1 is started by utilizing a test box switch to deenergize solenoid valves for 1SA2 and 1SAS (the CAPT
- 1 steam supply valves) from 1AFWPTCP by opening the solenoid power -
circuits directly. When this test box switch is in the closed posi-tion, the control Room Operators will still have full use of the CAPT ,
, #1 START /STOP switch. Once the test box switch is opened, however, CAPT #1 will start, and the control Room Operators will not be able to stop CAPT #1 using the CAPT #1 START /STOP switch, unless the test box switch at 1AFWPTCP is reclosed. This action, by itself, will stop CAPT #1. At any time, however, the Control Room Operators may stop
- the pump by closing 1CA145, CAPT #1 Trip and Throttle Valve. Proce-dural CAUTION and NOTE statements have been included to inform the Control Room Operators of these CAPT #1 control function details.
CAPT #1 will be run in recirculation mode to the Upper Surge Tank, and started by approved modifications which affect no components other
- than CAPT #1 specific ones. CAPT #1 will hot be operated outside design parameters. The required number of CA pumps will remain available. Thus, this test does not create a USQ.
1 PT/1/A/4200/19 Reissue, Changes 0 to 18 Incorporated 1
The format of this procedure was changed to agree with the Unit 2 procedure. The test method remains the ssme. Steps were added to '
document the positions of the Fuel Oil Day Tank Fill isolation and bypass valves to ensureLthat the valves are returned to their "As i
Found" positions.
This procedure stroke times Diesel Generator (D/G) Fuel Oil (FD) valves 1FD22 and 1FD62. If the bypass valve is not opened during performance of the procedure, the D/G will become inoperable due to the cover being removed from 1FD22 (1FD62). The other train is i
unaffected-by the test and will remain capable of-performing any safety function. This procedure could, lead to one inoperable D/G during this test; however, only one train is required to mitigate an accident. The valve is still capable of operating with the cover ,
removed, and the D/G, if it were to start, hae enough fuel in.the day tank to run for one hour. This will allow an operator time to either shutdown the D/G or open the bypass valve if it-had not been opened prior to the test. For these reasons, this procedure does not create i a USQ.
TN/2/A/0436/00/01A Original This procedure provides implementation instructions for Nuclear Station Modification (NSM) CH-20436 Rev. O, Work Unit 01. The modifi-cation will install in-line connectors in_the RG-11/U triaxial cables near the bottom of the Nuclear Instrumentation Cabinets. RG-59 triaxial cable assemblies will be installed from the in-line connec-tors to the cable termination points in the cabinets. This-procedure l
77
{. . . _ . _ . _ _ _ .
Will control installing the in-line connectors and RG-59 cable for Source Range Detector N31.
This procedure will be performed during a refueling outage when Source Range Detector N31 is not required to be operable por Technical Specification 3/4.3.1 or 3/4.9.2. No systems will be prevented from performing any function important to safety while this work is being done. For this reason, there are no USQs associated with this proce-dure.
TN/2/A/0436/00/02A original This procedure provides implementation instructions for NSM CH-20436 Rev. O Work Unit 02. The modification will install in-line connectors in the RG-11/U triaxial cables near the bottom of the Nuclear Instru-mentation Cabinets. RG-59 triaxial cable assemblies will be installed from the in-line connectors to the cable termination points in the cabinets. This procedure will control installing the in-line connec-tors and RG-59 cable for Source Range Detector N32.
This procedure will be performed during a refueling outage when Source Range Detector N32 is not required to be operable per Technical Specification 3/4.3.1 or 3/4.9.2. No systems will be prevented from performing any function important to safety while this work is being done. For this reason, there are no USQs associated with this proco-dure.
TN/2/A/0436/00/03A Original This procedure provides implementation instructions for NSM CN-20436 Rev.0 Work Unit 03. The modification will install in-line cont.ectors in the RG-11/U triaxial cables near the bottom of the Nuclear Instru-mentation Cabinets. RG-59 triaxial cable assemblies will be installed from the in-line connectors to the cable termination points in the cabinets. This procedure will control installing the in-line connec-tors and RG-59 cable for Intermediate Range Detector N35.
This procedure will be performed during a refueling outage when Intermediate Range Detector N35 is not required to be operable per Technical Specification 3/4.3.1. No systems will be prevented from performing any function important to safety while this work is being done. For this reason, there are no USQs associated with this proce-dure.
{
I TN/2/A/0436/00/04A Original This procedure provides implementation instructions for NSM CN-20436 Rev.0 Work Unit 04. The modification will install in-line connectors in the RG-11/U triaxial cables near the bottom of the Nuclear Instru-mentation Cabinets. RG-59 triaxial cable assemblies will be installed from the in-line connectors to the cable termination points in the 70
I i
cabinets. This procedure will control installing the in-line connec- ;
tors and RG-59 cable for Intermediate Range Detector N36. l This procedure will be performed during a refueling outage when Intermediate Range Detector N36 is not required to be operable per Technical Specification 3/4.3.1. No systems will be prevented from performing any function important to safety while this work is being done. For this reason, there are no USQs associated with this proce-dure.
TN/2/A/0436/00/05A Original ;
This procedure provides implementation instructions for NSM CN-20436 Rev. O Work Unit 05. The modification will install in-line connectors in the RG-11/U triaxial cables near the bottom of the Nuclear Instru-mentation Cabinets. RG-59 triaxial cable assembliL6 will be installed from the in-line connectors to the cable term nction points in the i cabinets. This procedure will control installing the in-line connec-tors and RG-59 cable for Power Range Detector N41.
This procedure will be performed during a reft.wling outage when Power Range Detector N41 is not required to be operable per Technical Specification 3/4.3.1. No systems will be prevented from performing any function important to safety while this work is being done. For this reason, there are no USQs associated with this procedure.
TN/2/A/0436/00/06A Original This procedure provides implementation instructions for NSM CN-20436 Rev. O Work Unit 06. The modification will install in-line connectors in the RG-11/U triaxial cables near the bottom of the Nuclear Instru-mentation Cabinets. RG-59 triaxial. cable assemblies will be installed from the in-line connectors to the cable termination points in the cabinets. This procedure will control installing the in-line connec-tors and RG-59 cable for Power Range Detector N42.
This procedure will be performed during a refueling-outage when Power Range Detector N42 is not required to be operable per Technical Specification 3/4.3.1. No systems will be prevented from performing any function important to safety while this work is being done. For this reason, there are no USQs associated with thic procedure.
1 TN/2/A/0436/00/07A Original This procedure provides implementation instructions for NSM CN-20436 Rev.0 Work Unit 07. The modification will install in-line connectors in the RG-11/U triaxial cables near the bottom of the Nuclear Instru-mentation Cabinets. RG-59 triaxial cable assemblies will be installed from the in-line connectors to the cable termination points in the cabinets. This procedure will control installing the-in-line connec-tors and RG-59 cable for Power' Range Detector N43.
J 79
This procedure will be performed during a refueling outage when Power Range Detector N43 is not required to be operable per Technical 'q Specification 3/4.3.1. No systems will be prevented from performing any function important to safety while this work is being done. For this reason, there are no USQs associated with this procedure.
TN/2/A/0436/00/08A Original This procedure provides implementation instructions for NSM CN-20436 Rev.0 Work Unit 08. The modification will install in-line connectors in the RG-11/U triaxial cables near the bottom of the Nuclear Instru-mentation Cabinets. RG-59 triaxial cable assemblies will be installed from the in-line connectors to the cable termination points in the ;
cabinets. This procedure will control installing the in-line connec-tors and RG-59 cable for power Range Detector N44.
This procedure will be performed during a refueling outage when Power Range Detector N44 is not required to be operable per Technical ,
Specification 3/4.3.1. No systems will be prevented from performing any function important to safety while this work is being done. For this reason, there are no USQs associated with this procedure.
TN/2/B/0420/00/02A Original The purpose of this procedure is to perform cable tie-in activities necessary to implement Nuclear Station Modification (NSM) CN-20420, Rev. O. This NSM re-routes Safe Shutdown Facility (SSF) cables outside of the control complex to meet Design Criteria DC-1.02-1. At present, this criteria is not being met by the cables associated with this NSM. This criteria states that a single fire which damages cables to the Control Room and could render the control Room inopera-ble will not damage cables to the SSF and render it inoperable also. l At present, this criteria is not being met for the cables associated with this NSM. This NSM will correct this discrepancy. The circuits and equipment affected by this procedure are valves 2NV876, 2NV877, '
2NC250A, 2NC253A, 2CA174, 2CA175, 2SA145, 2SAS, SSF NC Cold Leg Temperature RTD Loops 2NCTT5861 and 5881, SSF Pressurizer Heater Bank 2D, and SUF annunciators on the SSF control board.
The changos performed by this NSM will involve a cable re-route only.
The existing cables will-be deleted, and a new cable with'the same cable number as the existing cable will be installed. The new cables will be terminated back to the affected equipment the exact same way in which the old cables were terminated.
This procedure will be implemented during modes 5, 6, and No Mode of the Catawba Unit 2 End of Cycle 3 Refueling Outage. The circuits affected by this procedure will be isolated such that safety and non-safety equipment will not be adversely affected by the implementa-tion of this procedure. The circuits being modified are all non-safe-ty.
80
Each ,2ffected circuit will be tested prior to returning the circuit to service to ensure proper operation. As a result, the intent of the modification will be mot, and the bases for tho affected systems as stated in the Toch. Specs. has boon satisfied. This procedure will not pronto a USQ.
TN/5/A/0404/00/01A Original This procedure provides implomontation instructions for Nuclear Station Modification (NSM) CH-50404, Rev. O Work Unit 1. This NSM will provido a more definitive method of testing control room pressure to ensure compliance with Technical Specification Section 4.7.6.e.3.
A reference log from each adjacent area will be provided and connected to a valved manifold. The manifold will then be connected to a control room manomotor. This procedure will provide guidelines to make core drills, to run copper tubing from adjacent areas to the ,
control room, and to verify that the new manometer is working proper-ly.
During the implomontation of this proceduro, control Room wall pene-trations will be opened. Prior to opening any penetrations, Opera-tions will perform a Control Room Pressure Vorification (OP/0/A/6450/11) with both the existing and now manometers. This functional will verify the now manomotor is operating properly and that Control Room pressure is greater than or equal to 0.2 inches Water Gaugo (W.G.) prior to opening any Control Room penetrations.
This proceduro requires that the craft inform Operations before the penetration is opened and after it is sealed. This procedure will allow one penetration at a time to be oponod, and Operations will log the ponotration size in the Toch. Spoc. Logbook. Toch. Spec.
4.7.6.e.3 requires Control Room pressure to be greater than or equal to 1/8 inch (W.G.). To comply with this spec., operations may not allow more than two Control Room penetrations of 3 inches each to be opened at any one time. Operations will control this with the Tech.
Spec logbook.
While tubing is being run, the open end of the tubing line will be capped so an open air path from the Control Room will.not be created.
After all copper tubing is run from the adjacent rooms to the Control Room, a leak test on the copper tubing w!11 5c performed. The tubing lines will be pressurirod to 20-25 pounds end shall not leak more than '
2 pounds in 10 minutes.
An obstruction test will also be performed to verify no obstruction exists betwoon the manifold and each of the open ended copper tubing lines. This procedure will not allow an open air path from the control Room during this test. This procedure provides guidelines to l
connect the air supply to the test connection prior to opening any valves to the open ended copper tubing lines. Guidelines are also provided to assure that all valves for the open onded copper tubing lines are closed prior to removing the nitrogen / air supply.
81 1
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After all penotrations in the Control Room Wall have been opencd and scaled, Operation will perform a Control Room Pressure Verification (OP/0/A/6450/11) with the now manomotor to verify Control Room pres-sure is greater than 1/8 inch W.G. as required by Tech. Spoca.
For the abovo reasons, this proceduro does not create a USQ.
PT/2/A/4450/03A Rotype #1 The reference section was changed to change the reference to Regulato-ry Guido 5.13 to be 5.29. Regulatory Guido 5.29 is the correct Regulatory Guido for Commercial Power Reactors. The note from the end of section 3.0 was moved to the beginning, and the phrase "after each i refueling of the reactor" was added. This reflects the requirements of Regulatory Guido 5.29. The phrase "and after each refueling of the reactor" was added to stop 6.1, por Regulatory Guido 5.29. Step 7.1 was moved to be new stop 12.1 because thiu step is better suited for section 12.0. Steps 8.1 and 8.2 were doloted becauso Spent Fuel Pool Ventilation (VF) and Radiation Monitors (EMrs) EHF20 and EMP21 are not required to be operable for this inventory. The note in section 8.0 was doloted because it is no longer nooded. Steps 12.2 and 12.3 vero reworded to be like Unit l's 12.2 and 12.3. Step 12.9 and the note following woro deleted because these requirements are being satisfied by step 12.10. In enclosure 13.3, the required information for stop 12.10 was doloted, but the information for 12.9 was changed to include "Dato/Timo".
l This inventory procedure is not describad in the FSAR, but it is mandated por Regulatory Guide 5.29, Nuclear Material Control Systems for Nuclear Power Plants. This proceduro does not put any safety features in an abnormal _ configuration, and in no way does it compro-mise the safety evaluation in the FSAR for the now fuel storage vault.
Thus, those changes do not creato a USQ.
TN/2/A/0272/00/01A Original This procedure provides for implomontation of Nuclear Station Modifi- ,
cation (NSM) CN-20272, Rev.0, Work Unit 1. This NSM will add an ,
analog computer point from instrument loop 2NVLT5760 to monitor Volume Control Tank (VCT) lovel. This procedure will provido the necessary work activities to implement the NSM.
These work activities will affect instrument loop 2NVLT5760. This instrument loop is not required to be operable in the modes in which this p~ocedure r will bo implemented. This new operator Aid Computer (OAC) instrument loop is non-safety. The isolations to instrument loop 2NVLT5760 will affect an input signal-to the process control cabinets, which will affect Chemical and Volume Control System (NV) valvus 2NV172A, 2NV180A: 2NV.109B, 2NV252A, and 2NV253B. These valves will be failed to their safe positions. Notifications are stated in this procedure which will 3ei operations know what the equipment effects are, Other indicGtions of VCT level are available in the 82
\
i l
l 'Jontrol Room for Operator use.
These indications will not be affected I ay this procedure. Instrument loop 2NVLT5760 will be calibrated and i functionally verified prior to its return to service. For these rtasons, this procedure will not create a USQ.
i l TN/2/A/2732/CE/01A original This procedure provides implementation instructions for Exempt change CE-2372. This Exempt Change will modify the control circuit wiring on i Component Cooling (KC) system valve 2KC081B to provide torque switch bypass contacts which can be adjusted independently of indications and i
interlocks. 2KC081B is the inlet valve to the Residual Heat Removal System Heat Exchanger. This procedure will provide guidelines for rewiring the switch rotor and setting up the switch rotor.- This procedure will also require that 2KC0828 be stroke timed. Prior to returning valve 2KC018B to service, a functional verification and retest will be performed to verify valve operability.
During the implementation of this procedure, Unit 2 Emergency Core
- Cooling System (ECCS) Train B will be inoperable,- and Unit 2 will be in a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Tech. Spec. action statement per Tech. Spec. 3/4.5.2.
This valve, 2KCOSIB, is required to open on a-Safety Injection-signal followed by a low-lov Refueling Water Storage Tank level and on a high-high containment pressure signal. This is within the margin of safety as defined in the bases of the Tech.- Specs. For these reasons, this procedure does not create a USQ.
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i t
p.gigyba Nuclear Station Summary of Procedure Changes, Tests, and Experiments
- Completed from 11/1/89 to 9/30/90 -- Volume 3 ,
TN/1/A/1122/00/02A Retype #1 1
This implementation procedure provides guidance for the Post-Modifica- ,
tion Testing (PMT) associated with the installation of an eductor on the Auxiliary Feedwater Turbine Driven Pump (CATDP) lube oil cooler line. Retype #1 to this procedure merely clarifies and rearranges the order of some of the steps to better support implementation during the outage. The basis for this Post Modification Testing (PMT) has been
' outlined in the Catawba Unit 1 Test objectives / Acceptance criteria for valves 1CA-36, ICA-48, ICA52, and 1CA64. This PMT implementation procedure thoroughly governs the followings--(1) installation and removal of test equipment, (2) test and normal operation positioning of valves, C3) removal and return to service of the CATDP sump pumps, (4) monitoring of CATDP sump level during testing, and (5) meeting l
test acceptance criteria values for the affected equipment.
1 4
The isolations required for' implementation of this PMT will'not increase the probability of an accident or malfunction of equipment important to safety previously discussed in the FSAR. It will not create the possibility of an accident or malfunction of equipment important to safety different from any evaluated in the FSAR.- None-of l the plant operating parameters arc being permanently.affected by this procedure. Thus, the margin of safety will not be reduced, and there l are no unreviewed safety questions (USQ) associated with this proce-dure, l
i TN/2/A/0485/00/01A original This procedurn provides implementation instructions for Nuclear Station Modification (NSM) CN-20485,-Rev. O. This NSM provides Reactor Coolant Cold Leg Temperature to the Safe Shutdown Facility from Reactor Coolant Pump Loop C (2NCRD5910) instead of: loop:A-p -(2NCRD5860). The implementation-of the-procedure will affect the cold leg wide range instrumentation (RTDs) for the reactor coolant l
system (NC), with regard to remote shutdown: and accident monitoring in the Safe Shutdown Facility- (SSF). The RTDs that are modified by this l - procedure art safety related-(QA-1). The use.of individual station i
j control procedures is required to complete the implementation of this-procedure.
This proceduro is scheduled to.be implemented-during the Catawba Unit 2 End of Cycle 3 refueling outage during which the NC wide range RTDs are not requ i red .- Each-affected instrument will-be tested prior to being returned to-service. The intent of the. modification will be fully challenged.- No USQ exists, a
1
MP/0/A/7150/05 Re-type, Changes 0 to 6 Incorporated The following changes were made to the procedure during the re-type.
Step 4.3.4 was changed to reduce the :naximum allowable basket lif ting force from 4000 to 3000 pounds, based on Westinghouse recommendations.
Steps 4.3.7 and 4.3.8 were added as additional safety considerations.
Step f,.4 waz added to ensure the ice basket weighing rig calibrations blocke are not taken into containment when the unit is in Mode 4 or aboves. This change comes as a result of PIR 2-C89-0103. Section 8.0 has Leon enanged by moving the list of special tools from the body of the procedure to Enclosure 13.4. This section now also includes a step and sign-off for inspecting lifting devices. S9ction 9.0 was changed by deleting the acceptable weights on an ice basket with a cable cruciform and a plastic bag. The plastic bags are not used. A note was added to Step 11.2 to ensure the ice basket weighing rig calibration blocks are not taken into containment when the unit is in Mode 4 or above. Note 2 after stop 11.3 was added to provide guidance on when data sheets are to be completed. Note 3 after Step 11.3 was added to help ensure that no ice basket cable cruciform top plates are removed while the ice condenser is required to be operable in Mode 4 and above. Step 11.3.1 and sub-steps 11.3.1.1 and 11.3.1.2 were added to provide guidance to the technicians when finding baskets that cannot be weighed. The method described by these steps will preserve i the random nature of determining the chosen basket, as described by Tech. Spec. 4.6.5.1.b.2, by starting with the designated basket and generically moving in one direction. Step 11.3.2 directs the techni-clans back to the responsible Maintenance Engineering Services (MES)
Mechanical Engineer should no baskets in a row of a bay be able to be weighed. Step 11.3.2 was added to provide an additional sign-off to 4
ensure the baskets meet the acceptance criteria of Section 9.0. The l numbers for Steps 11.3.3 through stop 11.3.2.8 have been changed during the procedure revision. Other changes to these steps include minor wording changes and a note added to Step 11.3.3.5 about filling -
out data sheets for stuck baskets. The caution following Step 11.3.4 I
has been changed to reduce the maximum allowable basket lifting force-from 4000 to 3000 pounds por Westinghouse recommendations. Step 11.3.4.4 has been changed to delete the water soluble plastic bag.
- Step 11.4.1 was added to formalize the standard practice of having the j data sheets reviewed by the responsible MES mechanical engineer. The step numbers of the subsequent steps have been changed to allow the insertion of this new step. Section 13.0 was changed to add Enclosure i
13.4, Tool List. The following changes were made to Enclosure 13.1:
(1) a sign-off was added for Step 11.3.1.3 on page 1 of 5, (2) the
, acceptance criteria for basket weights (Section 9.0) were added on i Page 2 of 5, (3) all mention of the water soluble-plastic bag was removed from pages 2, 3, and 4 of 5, and (4) a sign-off was added for l'
Step 8.2 on page 4 of 5. Enclosure 13.3 was rewritten to include the orientations of basket columns and rows.
- This procedure is based on Tech.-Specs, and the FSAR to ensure that 4
the actions it controls will comply with established surveillance and design requirements. This procedure will ensure that accurate ice basket weights are taken on randomly selected baskets which will be used to determine ice bed operability. Thus, no USQ exists.
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2
PT/1/A/4150/21 Re-type, Changes 0 to 13 Incorporated This reissue of PT/1/A/4150/21 iniolves the following substantive changes:
- 1) Deletion of PT/1/A 46tJO/05F, Running Incore/Excore Calibration, which was performed during power escalation from 65% to 80% Full Power. Nuclear Instrumentation System (NIS) Calibration inaccu-racies inherant to the plant conditions of this test method have made it desirable to establish the following testing sequence for calibration of the Power Range NIS during post refueling power escalations o A second performance of PT/1/A/4600/05D, Interim Incore/Excore CalibrLiion, is now required at 76% power if the Incore/Excore calibration check performed concurrently with the full core flux map obtained at this power level demonstrates marginal results. The M factors (and hence the f 3 (delta I) functions to the OT delta T setpoints (Tech.
Spec. Table 2.2-1) would not be affected by this "intermedi-ate" recalibration of the NIS. Control Bank D is required to be positioned at >210 steps withdrawn (wd) during the full core flux map used for this calibration to enhance the accuracy of results.
o PT/1/A/4600/05A, Incore/Excore Calibration, is required to be performed at 76% power if the Incore/Excore calibration check yields unacceptable results. This will ensure that power ascension beyond 90% power is performed with a valid Excore NIS Calibration in place. Power escalation between 76% and 90% power is permissible per accepted Westinghouse Excore NIS Calibration methodology, o once equilibrium Xenon conditions at full power are achieved, an Incore/Excore Calibration check (per PT/1/A/4600/05D) is to be performed concurrently with the full power flux map. Unless the normal Incore/Excore Calibration Surveillance procedure (PT/1/A/4600/05A) was performed at 76% power, it will be required that this procedure be performed to formally calibrate the' Power Range NIS. Due to performance of this test at optimum conditions (100% power with Control Bank D positioned at >210 steps wd initially, moving in to approximately 190 steps Wd to achieve desired A.O. oscillation) the resulting calibration data is the best achievable. Performance of this test at
>75% power complies with Tech. Spec. 3/4.3.1, while the improved calibration enhances compliance with the surveil-lance requirements of Tech. Specs. 3.2.1 and 3.2.4.
o In the event that PT/1/A/4600!95 was performed at 76% power, three courses of action existt (1) the Incore/Excore Calibration check indicates a difference of <1%, which would mean no further Excore NIS Calibration is necessary, (2) the calibration check yields a difference of between 1% and 2%,
3
requiring that an Interim HIS Calibration be performed per PT/1/A/4600/05D, or (3) the calibration check yields a difference >2% Which would require that PT/1/A/4600/05A be repeated at full power.
- 2) The " Intermediate Power" flux map (as defined in FSAR Section 14.3.3.2) has been moved from 80% to 76% power. This is permis-sible per the Startup Physics Test Program outlined in Section 14.3 of the FSAR since " intermediate power" is defined as being between 50% and 80% of rated thermal power. The reason for obtaining this map at a lower power level is the desire to perform a second interim NIS Calibration (if necessary) as soon as possible above 30%, yet at a high enough power level to allow accurate projection of margins to core peaking factor limits at full power. It is no longer necessary to perform this flux map at the previously prescribed ">80% power" since entry into Base Load operation is no longer a consideration (Base Load w(z)
Factors are no longer provided).
- 3) Instructions for establishment of Zero Power Physics Testing (ZPPT) test band have been changed such that a full decade is not necessarily required. This allows test band to be set at a higher flux level in the event nuclear heat addition is observed near the low end (specifically between 1.7 and 2.0) of decade, while still maintaining a factor of 2 margin to the point of adding heat. The ability to establish a higher test band enhanc-es the flux signal from power range NIS channel to reactivity computer thereby improving the accuracy of ZPPT reactivity
, measurements. The original discussion of ZPPT (FSAR Table l 14.2.12-2) merely specifies the test band upper limit be approxi-mately 1 full decade below nuclear heat. The recent revision to l the FSAR, addressing Startup Physics testing (Section 14.3) does not mention this exercise at all.
[
l 4) A new enclosure has been added to perform a functional check of Rod Control System logic to verify proper sequence and overlap of control Banks as they are withdrawn. This also serves as " Post-
! Mod Testing," verifying that modification performed to axially reposition control rods (from fully withdrawn position of 230 steps wd to 225 steps wd) was implemented correctly.
- 5) Changes have been introduced to improve the reactivity computer i checkout at the beginning of ZPPT. Previously, a chronometer has
! been employed to measure the doubling and halving times associat-ed with prescribed reactivity insertions to perform this check-out. These times have been used to obtain the corresponding Reactor Period for use in calculating the theoretical (predicted) reactivity insertions for each of the checkout trials. The l calculated theoretical reactivity has been compared to the i reactivity indicated on the strip chart recorder (the amount of which was determined per Test Coordinator analysis). The new method for performing this checkout employs the IBM AT Reactivity i Computer exclusively, which discerns doubling / halving times and l measures actual inserted reactivities with high precision. The 4
errors associated with tanual operation of a chronometer and human interpretation of flux halving / doubling and quantification of reactivity incartions are thereby eliminated, enhancing the accuracy of this checkout. Calculation of actual and theoretical reactivity by the IBM AT is permissible due to the rigorous Software Controls imposed on the associsted programming. The Cycle 5 design reactivity data input to the computer is obtained from approved design document CNNE-1553.05-00-0005 (Catawba Unit 3 Cycle 5 startup and operational Report). It is incorporated via approved procedure PT/0/A/4150/20C, Westinghouse Digital Reactivity Computer Checkout. The external exponential test routine used to perform the previously described calculations are also validated by this procedure.
- 6) A number of testing activities have been deferred from the 60%
testing plateau to the newly created 76% testing plateau. The first of these is acquisition of Power Range (P/R) and Intermedi-ate Range (I/R) NIS data. This is done to verify full overlap between P/R and I/R NIS and is redundant since this data is also acquired at 30% and 76% power. The second is Unit Load Steady State. The data obtained per this procedure is also acquired at 30% and 76%, making the 50% data unnecessary. In fact, the NC Loop Delta Temperatures extrapoleted from such a low power level (50% power) are of little benefit during power ascension. This data obtained at 76% is much more valuable.
i This same argument I can be applied to deletion of the Thermal Inputs outputs Reli-ability check and Nuclear Steam Supply System (NSSS) Thermal outputs test at 50%. Thermal Inputs outputs Reliability Check will have already been performed at $10% power (to validate inputs to Primary Heat Balance) and at $50% power (to validate inputs to Secondary Heat Balance) allowing early detection of erroneous inputs to OAC TOP Program. Subsequent performance of this PT and NSSS Thermal outputs at $76% power will afford a l follow-up check of all inputs (and their processing via OAC Heat Balance Calculations) at a power level conservatively below full power (thereby precluding potential power escalation beyond 100%
power).
- 7) A more analytical evaluation of Shutdown Margin during Rod Swap is adopted in this reissue. The predicted Referenco Bank worth (increased by 15% for conservatitm) is first deducted from the Total Available Rod Worth less 6 tuck Rod decreased by 10% for conservatism). Then an additional 40 pcm is deducted to account for the insertion of a rod bank, other than the Reference Bank, in the event that some overshoot occurs during the Reference Bank measurement by dilution. These deductions are from the total rod worth (as opposed to rod worth above Hot Zero Power insertion limits) since rod swap is conducted with all rods out except for the Reference Bank which is approximately fully inserted. All rod worths used in this evaluation are obtained from approved Startup and Operational Report supplied by Duke Power Design Engineering.
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- 8) The enclosure which provides power escalation guidelineF to the operators has boon revised to enhance che projected values of Moderator Temperature coefficient (MTC) throughout power ascen-sion. Previously, cruder estimates of MTC behavior were obtained due to consideration of only effects from control rod insertion and increasing power level. Per the new methodology, " increasing power level" effects are broken down into two contributorst T-AVG increase and NC Boron Concentration Change. Prediction data from the Startup and Operational Report are used to deter-mine these terms at various power levels. The overall result of this chs7ge is enhancement of the awareness of the operators of expected plant response during power escalation.
Changes dealing with the enhancements of the reactivity compJter checkout and verification of Shutdown Margin during Rod Swap address the validation of assumed available shutdown margin (discussed in FSAR Section 4.3.1.5). Changes dealing with modification of the method by which Power Range NIS Calibration is performed during power escalation affect the spectrum of accidents which have been analyzed on the basis of axial core power distribution being within prescribed Tech. Spec.
limits. They also specifically impact accidents which take credit for the OTDT trip function, since this is affected by the f (delta I) functions generated by the Excore NIS calibration. These changes in no way invalidate these functions.
Changes made to enhance power escalation guidelines provided for the operatorc are aimed at assuring unit operation within an analyzed configuration at all times. The Post-Modification Testing of the Rod Control System during approach to criticality verifies that this plant equipment functions as required.
Extra conservatism assigned to control rod worths used in Shutdown Margin evaluations assures observance of associated safety margins.
Changes made to Excore NIS Calibration methodology enhance compliance with operating limits on AFD and QPTR, and improve accuracy of inputs to the OTDT function. The data trending of NSSS Thermal Outputs and Unit Load Steady State tests is performed safely below full power to ,
allow extrapolations to predict any reduced margin of safety at 100% l power. No USQ is involved with these changes. l PT/1/A/4450/03C Change #10 )
l This restricted procedure change was written to perform the retest for l Nuclear Station Modification (NSM) CH-11217, to modify section 12.9 in l response to NRC Information Notice 90-02, and to update criteria for sectians 12.1 and 12.2 in response to Catawba Exempt Change CE-2466 in order to complete required testing on the Annulus Ventilation (VE) system prior to Mode 4 in Unit 1 End of Cycle 4 Outage. The retest for NSM CN-11217 will consist of measuring pressure at four different locations in the Annulus by routing tygon tubing from the annulus side of IVEPT5010 to the different locations. A manometer will be connect-ed to the Auxiliary Building side of IVEPT5010 (located downstream of a vent valve) to determine the pressure in the Annulus at four points.
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Connections made to IVEPT5010 from the Annulus and Auxiliary Building and the removal of these connections are independently verified. Data will also be obtained on the air temperatures at each point in the Annulus, air temperature at the Auxiliary Building side of IVEPT5010 , l outside air temperature, barometric pressure, and fan motor amps. The pressures obtained will be compared with criteria provided by Design Engineering. All of the above data will be obtained while train A of Annulus Ventilation is operating in the normal alignment per OP/1/A/6450/02. Restricted change #8 was re-instated to allow per-forming the draw-down test using proper criteria and to add criteria for annulus pressure stabilization. Change #8 is a previously ap-proved change, and no additional changes were mado to the sections referenced by Change #8 except for the requiremont for Unit 1 to be in Modos 5, 6, or No Mode. NRC Information Notice 90-02 was written concerning interactions between different ventilations systems whon performing tests. Teats performed on Unit 2 on 3/1/90 have shown that Spent Fuel Pool Ventilation (VF) and Auxiliary Building Ventilation (VA) have an effect on the annulus vacuum decay time test. Therefore, prerequisites for Section 12.9 were added to shut down VF (both trains) on unit 1 and shutdown VA (both trains) on Units 1 and 2.
Other prerequisites were added to notify Radiation Protection of effects of shutting down VA, onsure that a Containment Air release is not in progress while VA is shutdown, and ensure that VF on Unit 1 is not required to be operable during the performance of section 12.9.
The acceptance critoria fce section 12.9 was also changed because the past error adjustment was incorrect. The new error adjusted accep-tance criteria will give more conservative results.
Unit 1 is required by this change to be in Mode 5, 6, or No Mode. VE is only required to be operable in Modes 1 through 4. Connections made to IVEPT5010 and the removal of these connections are indepen-dently verified. VA and VF will be aligned por the Operations proce-dure for Section 12.9, and will remain operable. For these reasons, no USQ exists.
PT/1/A/4250/03E Change #11 This change is to provide a new flow versus pressure curve for Auxil-inry Feedwater Pump #1 since the pump internals of this pump have been replaced. The head verification test PT/1/A/4250/13C determined that this pump is a 100% performance pump. The new curve was derived from acceptanca criteria data supplied by Design Engineering in CNSD-1223.42-00-11, Revision 3. No USQ exists.
PT/1/A/4250/03C Change #46 This change is to establish new acceptance criteria for Auxiliary Feedwater pump #1 following replacement of the pump internals and implementation of a modification changing the bearing cooler flow. A full head curve test was performed on the pump verifying that the pump was acceptable, and this change will establish a baseline for future IWP testing of this pump. No USQ exists.
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PT/1/A/4250/03A Change #27 l
This restricted procedure change will verify the capability of a motor driven Auxiliary leedwater (CA) pump to take suction from the condens-er hotwell while the hotwell is nnt under vacuum. This capability is described in the FSAR, and credit is taken for this condensate grade water. Since this is a manual start of the pump, suction will not be swapped to the Nuclear Service Water system in case suction from the hotwell cannot be obtained. Also during the test, suction vill be continuously monitored and action taken to stop the pump or realign suction to the Upper Surge Tank if suction pressure decreases. The suction pressure switches will automatically stop the pumps on low pressure as designed.
Therefore, no new malfunctions are created. This test will be per-formed in Mode 5 or 6 when CA is not required. Thus, no USQ is involved.
PT/1/A/4250/03A Change #28 This restricted procedure change modified PT/1/A/4250/03A and Re-stricted Change #27 in regards to the CA pump 1A Manual start low suction pressure pump trip circuitry. During testing performed on 3/7/90, CA Pump 1A would only develop 105 gpm of total flow using the hotwell as its sole suction source. During the test, as pump dis-charge flow was increased, the pump tripped slightly above the value for miniflow. Design Engineering has determined that the lower safe limit of pump suction pressure is 3.5 PSIG. This change will allow blocking of CA pumps 1A manual start low suction pressure trip, and will modify pump circuitry so that flow may be adjusted up'Jard so long as pump suction pressure remains 2 4.0 PSIG. Suction pressure Will se monitored continuously throughout the pump test. In addition, a CAUTION statement will be added to reduce flow should any cavitation be detected by-test personnel. No USQ is involved.
PT/1/A/4200/41A Change #22 This restr acted- change will allow testing of Containment Purge (VP) -
valve IVP2A from inside containment as well as from the annulus. It provides for additional leakage data to be obtained to assist in Catawba's response to PIR 0-C89-0197. The change will allow the use of a blind flange installed on the containment side of-1VP2A so that leakage date can be obtained on the valve from the containment side of the valve seat. The installation and removal of the blind flange, if required,His independently-verified. Normally leakage is only ob-tained from the annulus side of this valve. The test method for-verification of Tech. Spec. leakage will not be changed; only addi-tional data will be obtained. This change does not affect the ability of the Containment Isolation valves on this penetration to close or otherwise seal. No USQ exists.
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PT/1/A/4600/16 Change #17 This change is the result of the Tech. Spoc. Amendments 73 (Unit 1) and 67 (Unit 2) to the Reactor Coolant System Heatup Rato. This lowered the limits to 5 60 degrees F. The Administrative Limit was changed to provide a buffer between the Administrative Limit and the Technical Specification Limit. The change is the result of the analysis of Capsule Z of Unit 2. No USQ exists.
PT/2/A/4200/28A Change #5 Fire Protection System (RF) Valve 2RP447B is the Containment Isolation valve for the Reactor Coolant (NC) Pump Motors, the Containment Ventilation (VV) Carbon Filter Units, and the pipe chase sprinklers.
This is a " dry pipe" header, meaning that 2RF447B is normally closed, and thoro is no water in the supply piping downstream. Stroke testing this valve necessitates opening 2RF447B. Prosently, this header is isolated upstream of 2RF447B prior to stroking it. This change will vont off any residual pressure in the supply piping using 2RF954 to ensure that the " dry pipe" stays dry. This change ensures that no water enters the " dry pipe" as a result of this test and leaks through the sprinhlers onto the equipment being protected.
The primary purpose of 2RF447B is to isolate the containment in the event of a LOCA and this function has not changed. Since the test mothed has not changed, and is within Tech. Spec. limits, no USQ is involved.
PT/1/A/4200/28A Change #11 Valve IRF447B is the containment Isolation Valve for the NC Pump Motors, the VV Carbon Filter Units, and the pipo chase sprinklers.
This is a " dry pipe" header, meaning that 1RF447B is normally closed, and there is no water in the supply piping downstroam. Stroke testing this valvo necessitates opening 1RF447B. Presently, this header is isolated upstream of 1RF447B prior to stroking it. This change will vent off any residual pressure in the supply piping using 1RF979 to ensure that the " dry pipe" stays dry. This change ensures that no water enters the " dry pipe" as a result of this test and leaks through the sprinklers onto the equipment being protected.
, The primary purpose of 1RF447B is to isolate the containment in the l event of a LOCA and this function has not changed. Since the test method has not changed, and is within Tech. Spec. limits, no USQ is involved.
PT/1/A/4200/13G Change #21 This change adds flexibility to the way the procedure may be por-formed. The following is the order in which the procedure Eng per-formed:
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- 1. Sliding link is opened to prevent valve from opening,
- 2. Valve breaker is energized.
- 3. Valve position indications are verified.
- 4. Valve breaker is doenergized.
- 5. Sliding link is closed.
This change adds guidance as to whether the sliding link needs to be opened. The maximum cover pressure of the accumulator is 678 psig.
This change specifies that at Reactor Coolant (NC) pressure of >800 psig, the sliding link will not be opened. This allows one Perfor-mance Technician to perform the test from the Control Room. The change also allows the Unit Supervisor to N/A the step that has power removed from the valve. The Cold Log Accumulator (CLA) isolation valves open automatically when either NC pressuro exceeds the P-11 setpoint OR on safety irejection. The P-11 setpoint is 1955 psig. By ensuring that the NC pressure is greater than CLA pressure, the chance of accuuulator blowdown is removed. Therefore, there is no need for opening the link to defeat the valve opening circuit. This change also allows the Unit Supervisor to N/A the step that has power removed from the valve. This may be necessary if other testing is to be performed after IWV testing. No USQ is involved with this change.
OP/1/A/6100/01 Change #125 This change is the result of Tech. Spec. Amendment 73 (Unit 1) to the Reactor Coolant (NC) System Heatup Rate. This lowered the limits to S 60 degrees F. The Limits and Precautions were changed to provide a buffer between the Administrative Limit and the Technical Specifica-tion Limit. The change is the result of the analysis of Capsule Z of Unit 2. No USQ is involved.
OP/2/A/6100/01 Change #55 This change is the result of Tech. Spec. Amendment 67 (Unit 2) to the NC System Heatup Rate. This lowered the limits to 5 60 degrees F.
The Limits and Precautions were changed to provide a buffer between the Administrative Limit and the Technical Specification Limit. The change is the result of the analysis of Capsule Z of Unit 2. No USQ is involved.
PT/1/A/4200/13E Change #41 This restricted change allows the stroke testing of the Motor Driven isolation valves from Auxiliary Feedwater (CA) pumps 1A and 1B to the steam generators (ICA58A, ICA62A, ICA42B, and 1CA46B) against full differential pressure. This will be accomplished by injecting flow into a depressurized steam generator and closing the appropriate valve. This test is being conducted to obtain data for the valves which have been replaced as a result 7f Nuclear Station Modification CN-11186. Flow will be taken from the upper surge tank and delivered to the steam generatora at an initial rate of approximately 300 gpm.
Once isolatior c2 the ' alve under test has occurred, a minimum flow 10
path will still be available. An additional Limit and Precaution was included to alert 'e Control Room Senior Reactor Operator and Opera-tor of the need to hionitor Steam Generator Levels, so as not to reech Hi-Hi Setpoints. The previously run procedure (Change #35) had a similar precaution, but this addition makes its intent more clear.
This test will be performed in Mode 4, 5, 6, or No Mcde. The CA System is not required at this time. No USQ is involved with this change.
PT/1/A/4200/13A Change #35 Change 35 adds additional steps to enclosures which test the Safety Injection (NI) Pump Hot Leg Isolation Valves to close the NI Pump Cold Leg Isolation Valves for the train under test prior to cycling the Hot Leg Valve. This change will prevent the possibility of affecting the operability of the opposite train by eliminating the cross-train flow
,4th to the Hot Leg which could invalidate ECCS flow balance. This change also clar..fies the Limit and Precaution 6.2 which states that the train under test is rendered inoperable during this test to give the reason for the train inoperability. '
Tech. Specs. allow one train of Emergency Core Cooling Systems thCC.'.)
to be inoperable for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This change does not affect the l NI pumps minimum flow path. The train under test will be inoperable, and the Limiting Conditions for Operations for ECCS will be adhered to. Thus, no USQ exists.
PT/1/A/4200/09 Change #81 1
This change deletes the use of the " ARF-1B TEST" switch. The jumper l added in 1EATC2 serves the function of the deleted test switch (open-i ing the bypass dampers on the start of Air Return Fan (ARF) 18.) The jumper in IIC2 is not imeded since the test switch is no longer being used. The Containment Pressure Control Circuitry permissive for ARF-D-4 (1NSPT5250) is verified to be in the normal position to ensure that ARF-D-4 will not open during the performance of the test. This change will operate Hydrogen Skimmer Fan 1B and ARF-1B closer to the l
designed accident alignment because the deleted test switch bypasses some of the safety circuitry for each fan.
This test will be run in Mode 5. The Containment Air Return and Hydrogen Skimmer System is only required in Modes 1 through 4. No USQ exists.
PT/1/A/4200/09 Change #83 Normally on a Safety Injection Signal, Safety Injection (NI) valvos 1NI9A and 1 nil 0B will open to allow the Chemical and Volume Control (NV) System pumps to inject borated water into the core. This proce-dure tests this interlock. On the trajn under test, the NV pump is 11
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isolated from the Reactor Coolant (NC) system. However, the opposite train NV pump may be in service. This results in water being injected into the NC system and causes an increase in the pressurizer level.
This level change can cause a significant temperature transient that may or may not violate Technical Specification 3.4.9.2. In the interest of compliance, these vulves will not be tested during the actual Engineered Safety Features test, but will be tested separately.
These valves will not be de-energized at the same time, so that the opposite train will always be available to inject water into the core if necessary. The required number of safety systems are still opera-ble during this test. No USQ is involved.
PT/1/A/4200/09 Change #84 This restricted change is to allow retesting of valves 1 nil 0B, 1RN404B, and 1VQ15A. 1RN404B and IVQ15A did go to position during Engineered Safety Features (ESF) testing, but the stroke time was not recorded by the Operator Aid Computer (OAC). 1 nil 0B was red tagged closed during the test, and, therefore, was not stroke timed for the procedure. The test for each of 1RN404B and IVQ15A consists of placing the valve in the OPEN position and closing the corresponding contacts on the Solid State Protection System (SSPS) output relays to close the valve. The valve is timed from the closing of contacts to i the closed indication. 1 nil 0B is timed in the same way, except that initial position is closed and final position is open. The valve is timed from closing of contacts to open indication. Only the valve being tested is affected by closing contacts on an SSPS output relay.
Testing of all three valves is essentially the same as IWV testing.
The valve is taken to its non-safety position and stroked to the safety position. In each case, the valve may be returned as soon as the stroke is complete. The valves will not be made inoperable at any time for the test. No USQ is involved.
PT/1/A/4200/09 Change #85 This change involves only the Cold Leg Accumulator (CLA) Isolation Valve sections of this procedure. The present prerequis.ites call for the Reactor Coolant (NC) system pressure to be less the'. 1900 psig and .
for the NC pressure to be from 0 to 100 psig above tne CLA pressure.
This unnecessarily limits the conditions under wPich the procedure sections may be performed. The NC pressure must be abova the CLA pressure in order to prevent accumulator blowdcan. This change has Operat'ons verify that NC system pressure is at least 100 psig greater than C i Leg Accumulator Pressure (to prevent accumulator blowdown.) I Specif;;ng a 100 psig difference is conservative. Ensuring that the P-11 setpoint is not exceeded is covered under a previous step in the procedure. No USQ is involved with this change.
1 PT/1/A/4150/05 Change #27 12
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1 i
This change incorporates the SNC CORE Theoretical Factor Files for use during Cycle 5 operation into the periodic test procedure used for acquisition and analysis of incore detector data for the purpose of ensuring compliance with Tech. Spec. limits of core peaking factors and axial and radial power distribution. The ascribed theoretical factor files have been generated by approved codes and methodologies by Nuclear Engineering in the Design Engin,aring Department. -Their use of approved Shanstrom Nuclear Associates incore code ensures full compliance with all affected aspects of nuclear safety. No USQ is involved with this change.
OP/1/A/6250/02 Change #43 This change added Auxiliary Feedwater (CA) valve ICA-279 and changed the name and position of valve ICA-215. These changes allow the modification CN-11122 to operate properly. No USQ is involved.
OP/1/A/6250/01 Change #69-This restricted change is being made to provide guidance for full differential pressure stroke verification of Auxiliary Feedwater (CA) valves 1CA-149, 150, 151, and 152. It allows opening the main feedwater (CF) regulating valves to obtain the desired feedwater pressure. During the previous test, the steam generators filled up faster than expected. To allow for this rate, the steam-generator level will be lowered to 20% wide range prior to the test. These verifications will be performed in Mode 5. Equipment safety will not be affected. CF pump speed will remain protected by the overspeed trip. CF pump miniflow will' remain operable. _CF pump runout is not expected. Steam generator level will be controlled procedurally. The CF regulating valves and regulating valve bypass valves will be closed at 65% narrow range level. If necessary, the operator will trip the CF pump at 80% narrow range level. S/G chemistry will be maintained as normal operation. No USQ is involved.
PT/1/A/4150/133 Retype, Changes 0 to 23 Incorporated The changes made to this procedure account.for system changes made per Nuclear Station Modification (NSM) CN-10753. -This NSM removed the RTD l bypass manifolds and placed the Reactor Coolant (NC) System hot and l cold leg RTDs directly in the hot and cold leg piping.- The NC system now has three active hot log RTDs and one active and one spare-cold leg RTD.- Changes made to the procedure now allow data to be taken from all cold leg and hot leg RTDs. No USQ is involved.
PT/1/A/4150/11B Retype, Changes 0 to 8 Incorporated I
This change added guidelines for the. performance of rod swaps starting with the Reference Bank fully inserted.and the. rod bank of lowest worth slightly-inserted. This initial _ condition can result from boron 13
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dilution overshoot during the integral rod worth measurement of the l Reference Bank. Normally rod swaps are conducted from an initial condition of Bank RF <40 pcm of the fully inserted position and all other rod banks fully withdrawn. In this case, the critical position l of Bank RF (with all other banks withdrawn) is noted prior to, and l
following, sequential exchange of the seven banks with Bank RF in order to infor their exact reactivity worths. The initial and final critical positions of Bank RF are then evaluated and the change (i.e.,
reactivity drift) is equally applied over the inferred worths of the seven banks by either incrementally adding (in-the event of inadver-tent dilution of the NCS) or deductin, (in the event of. inadvertent' boration of the NCS) reactivity from ti. inserted integral worth of Pank RF (with all other banks withdrawn) before each of the seven rod swaps. ,
In order to handle the new situation, the inserted worth of the lowest worth bank (Bank 1) is measured prior to commencing rod exchange.
Following completion of the exchanges, the reactivity difference (drift) between the final Bank 1 position and the initial critical position is measured. This reactivity worth is then equally applied, in conjunction with the initial inserted worth.of Bank 1,-over the seven bcnks measured by rod exchange to adjust for the drift noted over the test interval. This adjustment involves addition of the reactivity contributions of the insertion of Bank i to the Total Measured Integral Worth of the Reference Bank. Addition of these new guidelines will save up to two hours of critical path time, which would have been expended by borating Bank 1 out before initiation of rod exchanges. These guidelines ensure that accuracy of inferred rod worths by rod exchange is enhanced by drift adjustment. This is important to the accurate validation of predicted core design parame-ters.
Changes were also made associated with the control rod repositioning (from 230 steps withdrawn to 225 steps withdrawn.) This change had I previously been evaluated separately.
Adequate Shutdown Margin is assured (as required by-Tech. Spec.
l 3.1.1.1) with the additional insertion of Bank-1 per the preliminary l evaluation performed by Enclosure 13.2 of PT/1/A/4150/21. This Enclosure deducts 40 pcm (in addition to the Total Worth of the reference bank, plus 15% to account for uncertainty) from the total available rod worth (minus 10% for uncertainty) to verify that avail-able Shutdown Margin is >1300 pcm (1.3% delta K/K) during rod Swap l testing. No USQ is involved.
i TN/1/A/1005/01/22A Retyp.e #1 Nuclear Station Modification (NSM) CN-11005 Rev. 1 modifies various piping system analyses with the objective of reducing the number of mechanical snubbers. This work unit provides guidance for the removal of pipe attachments and structural attachments for snubbers deleted I from a rigorous analysis model on the Residual Heat Removal-(ND) )
- system. The snubbers were removed by work unit 17. This work unit is 14
merely a " clean-up" work unit. The work unit removes support steel, repairs abandoned concrete anchor holes, and touches up coating for Support / Restraints (S/Rs) deleted by NSM CH-11005 Rev. 1. The S/R support steel is no longer physically attached to the Residual Heat Removal, Safety Injection, Containment Spray, and Refueling Water systems and serves no function. The pipe clamps are also removed by*
the work unit.
This work unit does not affect or impact any equipment, system, or structure necessary for the safe operation or the safe shutdown of the plant. Removing the pipe clamps has no affect on the system (s). No USQ is involved with the implementation of this work unit.
TN/1/A/1005/01/23A Retype #1 NSM CN-11005 Rev. 1 n.odifies various piping system analyses with the objective of reducing the number of mechanical snubbers. This work unit provides guidance for the removal of pipe attachments and struc-tural attachments for snubbers deleted from a rigorous analysis model on the ND system. The snubbers were removed by work unit 14. This work unit is merely a " clean-up" work unit. The work unit removes support steel, repairs abandoned concrete anchor holes, and touches up coating for S/Rs deleted by NSM CN-11005 Rev. 1. The S/R support steel is no longer physically attached to the Residual Heat Removal and Safety Injection systems and serves no function. The pipe clamps are also removed by the work unit.
This work unit do?.s not affect or_ impact any equipment, system, or structure necessary for the safe operation or the safe shutdown of the plant. Removing the pipe clamps has no affect on the system (s). :No USQ is involved with the implementation of_this work unit.
TN/1/A/1005/01/24A Retype #1 2
NSM CN-11005 Rev. 1 modifies various piping system analyses with the objective of reducing the number of mechanical snubbers. This work unit provides guidance for the removal of pipe attachments and struc-tural attachments for snubbers deleted from a rigorous analysis model on the ND system. The snubbers were removed by work unit 12. This work unit is merely a " clean-up" work unit. The work unit removes support steel, repairs abandoned concrete anchor holes, and touches up-coating for S/Rs deleted by NSM CN-11005 Rev. 1. The S/R support-steel is no longer _ physically attached to the ND system and serves no function. The pipe clamps are also removed by the work unit.
This work unit does not affect or impact any equipment, system, or structure necessary for the safe operation or the safe shutdown of the plant. Removing the pipe clamps has no affect on the system (s). No USQ is involved with the implementation of-this work unit.
TN/1/A/1005/01/35A Original 15
NSM CN-11005 Rev. 1 modifies various piping system analyses with the objective of reducing the number of mechanical snubbers. This werk unit provides guidance for the removal of pipe attachments and struc-tural attachments for snubbers deleted from a rigorous analysis ~model on the Chemical and Volume control (NV) system. The snubbers were removed by work unit 13. This work unit is merely a " clean-up" work unit. The work unit removes support steel, repairs abandoned concrete anchor holes, and touches up coating for S/Rs deleted by NSM CN-11005 Rev. 1. The S/R support steel is no longer physically attached to the NV system and serves no function. The pipe clamps are also removed by the work unit.
This work unit does not affect or impact any equipment, system, or structure necessary for the safe operation or the safe shutdown of the plant. Removing the pipe clamps has no affect on the system (s). No USQ is involved with the implementation of this work unit.
TN/1/A/1005/01/36A Original NSM CN-11005 Rev. 1 modifies various piping system analyses with the objective of reducing the number of mechanical snubbers. This work '
unit provides guidance for the removal of pipe attachments and struc-tural attachments for snubbers deleted from a rigorous analysis model on the Component Cooling (KC) system. The snubbers were removed by work unit 15. This work unit is merely a " clean-up" work unit. The work unit removes support steel, repairs abandoned concrete anchor holes, and touches up coating for S/Rs deleted by NSM CN-11005 Rev. 1.
The S/R support steel is no longer physically attached to the KC system and serves no function. The pipe clamps are also removed by the work unit.
This work unit does not affect or impact any equipment, system, or structure necessary for the safe operation-or the safe shutdown of the plant. Removing the pipe clamps has no affect on the system (s). No USQ is involved with the implementation of this work unit.
TN/1/A/1005/01/37A Original NSM CN-11005 Rev. 1 modifies various piping-system analyses with the objective of reducing the number of mechanical. snubbers. This work unit provides guidance for the removal of pipe attachments and struc-tural attachments for snubbers deleted from a rigorous analysis model on the Safety Injection (NI) system. The snubbers were removed by work unit 16. This work unit is merely a " clean-up" work unit. The work unit moves support steel, repairs abandoned concrete anchor holes, and touches up coating for S/Rs deleted-by NSM CN-11005 Rev. 1.
The S/R support steel is no longer physically attached to the NI, l Residual Heat Removal, and Refueling Water systems and serves no function. The pipe clamps are also removed by the work unit.
This work unit does not affect or impact any equipment, system, or structure necessary for the safe operation or the safe shutdown of the 16
plant. Removing the pipe clamps has no affect on the system (s). No USQ is involved with the implementation of this work unit.
TN/1/A/1005/01/38A Original NSM CN-11005 Rev. 1 modifies various piping system analyses with the objective of reducing the number of mechanical snubbers. This work unit provides guidance for the removal of pipe attachments and struc-tural attachments for snubbers deleted-from a rigorous analysis model on the Main Steam Vent to Atmosphere (SV) system. The snubbers were j removed by work unit 18. This work unit is merely a " clean-up" work unit. The work unit removes support steel, repairs abandoned concrete anchor holes, and touches up coating for S/Rs deleted by NSM CH-11005 Rev. 1. The S/R support steel is no longer physically attached to the SV and Main Steam systems and serves no function. The pipe clamps are also removed by the work unit.
This work unit does not affect or impact any equipment, system, or structure necessary for the safe operation or the safe shutdown of the plant. Removing the pipe clamps has no affect on the system (s). No USQ is involved with the implementation of this work unit.
TN/1/A/1005/0:f39A Original NSM CN-11005 Rev. 1 modifies various piping system analyses with the objective of reducing the number of mechanical snubbers. This work unit provides guidance for the removal of pipe attachments and struc-tural attachments for snubbers deleted from a rigorous analysis model on the Auxiliary Feedwater (CA) system. The snubbers were removed by work unit 19. This voi.e unit is merely-a " clean-up" work unit. The work unit removes supt. art steel, repairs abandoned concrete anchor holes, and touches up coating for S/Rs deleted by_NSM CN-11005 Rev. 1.
The S/R support steel is no longer physically attached to the CA and Main Feedwater (CF) systems and serves no function. 'The pipe clamps are also removed by the work unit.
l This work unit does not affect or impact any-equipment, system, or structure necessary for the safe operation or the safe shutdown of the plant. Removing the pipe clamps has no affect on the system (s).- No USQ is -involved with the implementation of this work unit..
TN/1/A/1005/01/40A Original NSM CN-11005 Rev. 1 modifies various piping system analyses with the p objective of reducing the number of mechanical snubbers. This work l unit provideo guidance for the removal of pipe attachments and struc-tural attachments for snubbers deleted from-a rigorous analysis model on the CA system. The snubbers were removed by work unit 20. This ,
work unit is merely a " clean-up": work unit. -The work unit removes- i support steel, repairs abandoned concrete anchor holes, and touches up j coating for S/Rs deleted by NSM CN.005 Rev. 1. The S/R support-17 ,
1 l
l
steel is no longer physically attached to the CA and CF systems and serves no function. The pipe clamps are also removed by the work unit.
This work unit does not affect or impact any equipment, system, or structure necessary for the safe operation or the safe shutdown of the plant. Removing the pipe clamps has no affect on the system (s). No USQ is involved with the implementation of this vork unit.
TN/1/A/1005/01/41A Original NSM CN-11005 Rev. 1 modifies various piping system analyses with the objective of reducing the number of mechanical snubbers. This work unit provides guidance for the removal of pipe attachuents and struc-tural attachments for snubbers deleted from a rigorous analysis model on the CA system. The snubbers were removed by work unit 21. This work unit is merely a " clean-up" work unit. The work unit removes support steel, repairs abandoned concrete anchor holes, and touches up coating for S/Rs deleted by NSM CN-11005 Rev. 1. The S/R support steel is no longer physically attached to the CA and CF systems and serves no function. The pipe clamps are also removed by the work unit.
This work unit does not affect or impact any equipment, system, or structure necessary for the safe operation or the safe shutdown of the plant. Removing the pipe clamps has no affect on the system (s). No USG is involved with the implementaticn of this work unit.
- 1/1/A/1005/01/42A Original NSM CN-11005 Rev. 1 modifies various piping system analyses with the objective of reducing the number of mechanical snubbers. This work unit provides guidance for the removal of pipe attachments and struc-tural attachments for snubbers deleted from a rigorous analysis model on the Main Steam Supply to Auxiliary Equipment (SA) system. The snubbers were removed by work unit 01. This work unit is merely a
" clean-up" work unit. The work unit removes support steel, repairs abandoned concrete anchor holes, and touchos up coating for S/Rs deleted by NSM CN-11005 Rev. 1. The S/R support steel is no longer physically attached to the SA and Main Steam systems and serves no function. The pipe clamps are also removed by the work unit.
This work unit does not affect or impact any equipment, system, or structure necessary for the safe operation or the safe shutdown of the plant. Removing the pipe clamps has no affect on the system (s). No USQ is involved with the implementation of this work unit.
TN/1/A/1005/01/46A Original NSM CN-11005 Rev. 1 modifies various piping system analyses with the objective of reducing the number of mechanical snubbers. This work 18 1
! unit provides guidance for the removal of pipe attachments and struc-tural attachments for snubbers deleted from a rigorous analysis model on the CA system. The snubbers were removed by work unit 45. This work unit is merely a " clean-up" work unit. The work unit removes support steel, repairs abandoned concrete anchor holes, and touches up coating for S/Rs deleted by NSM CN-11005 Rev. 1. The S/R support steel is no longer physically attached to the CA, SA and Main Feedwater Pump Turbine Exhaust systems and serves no function. The pipe clamps are also removed by the work unit.
This work unit does not affect or impact any equipment, system, or structure necessary for the safe operation or the safe shutdown of the plant. Removing the pipe clamps has no affect on the system (s). No USQ is involved with the implementation of this work unit.
EP/2/A/5000/1C2 Retype #6 The following changes were made in Re-type #6. Step 12 was reworded to make it clear to the operator that only non-faulted steam genera-tors should be used in the subsequent cooldown. The wording of guidance given on Enclosure 1 item F, " Characteristics of Natural Circulation was modified. Action now states "IF all NC pumps off, THEN monitor the following parameters at 10 to 15 minute intervals."
l This statement on natural circulation is not applicable unless all l Reactor Coolant (NC) pumps are off. The wording was modified to make
, it clear that this is the case. A step to monitor annulus pressure
! between -1.4 and -1.8 inches water column (WC) at 30 minute intervals was added to Enclosure 1 (Item H). Also, Enclosure 4 was added to provide operator actions if annulus pressure is not within the re-l quired band. These changes were made as a temporary measure to satisfy the requirements given in PIR 0-C89-0283. Per discussions with Design Engineering, it was determined that the monitoring of the Annulus Ventilation (VE) System is only required during a high energy line break inside containment. A step was added to Enclosure 1 (Item I) to ensure that proper minimum flow requirements are met for the Residual Heat Removal (ND) pumps. This information was added as a result of a re-evaluation of ND pump minimum flow requirements which was performed by Design Engineering and Ingersoll-Rand. Substep 5.c was added to Enclosure 2. This information was added to provide additional guidance to the operator in case the previous steps failed to start the fans. Step 6 on Enclosure 2 was reworded to say " Ensure NS pumps and Containment Air R0 turn fans operate as Containment pressure changes." The previous wording had the statement "VX fans."
Only the Containment Air Return fans will start and stop as Contain-ment pressure changes. The Hydrogen Skimmer fans will run continuous-ly until secured. Various typographical and format errors were corrected to ensure compliance with the Emergency Procedure (EP)
Writer's Guide. The changes do not involve a USQ.
EP/1/A/5000/1C2 Re-type #6 19
-- -- - . - - - - -~ . . - - . - - .- - . ..
The following changes are included-in Re-type #6. The caution prior to steps S and 12.d was deleted. It stated that Wide Range Reactor Coolant (NC) temperature should be used when monitoring loop tempera-tures. In the past, if all NC pumps were off, Harrow Range (N/R)
T-Hot and T-Cold were inaccurato because these temperature detectors were located on the RTD manifold. This condition was changed by Nuclear Station Modification (NSM) CN-10753. This NSM placed the new N/R temperature detectors on the loops, and the old detectors on the RTD manifolds were deleted. As a result, N/R NC T-hot and T-Cold are now accurate (and therefore T-Avg is accurate) with the NC pumps on or off. Therefore, this caution is no longer needed. Step-12 was >
reworded to make it clear to the operator that only non-faulted steam generators should be used in the subsequent cooldown. The references to NC W/R T-Hot and NC T-Cold were deleted. Statements now say NC T-Hot and NC T-Cold. This change was made to. reflect the labelling of the Hot and Cold Leg temperature indicators located in the Control Room. The wording of guidance given on Enclosure 1-item F, " Character-istics of Natural Circulation was modified. Action now states "IF all NC pumps off, THEN monitor the following parameters at 10 to 15 minute intervals." This statement on natural circulation is not applicable unless all NC pumps are off. The wording was modified to make it clear that this is the case. A step'to monitor annulus pressure between -1.4 and -1.8 in. WC at 30 minute intervals was added to Enclosure 1 (Item H). Also, Enclosure 4 was-added to provide operator actions if annulus pressure is not within the required band. These changes were made as a temporary measure to satisfy-the requirements given in PIR 0-C89-0283. Per discussions with Design Engineering, it was determined that the monitoring of the VE System is only required during a high energy line break inside containment. A step was added to Enclosure 1 (Item I) to ensure that proper minimum flow require-ments are met for the ND' pumps. This information was added as a result of a re-evaluation of ND pump minimum flow requirements which was performed by Design Engineering and Ingersoll-Rand. Substep 5.c was added to Enclosure 2. This information was added to provide additional guidance to the operator in case the previous steps failed to start the fans. Step-6 on Enclosure 2 was reworded to say " Ensure NS pumps and-Containment Air Return: fans operate as Containment pressure changes." The previous wording had the statement "VX fans."
Only the Containment Air Return fans will start and stop as Contain-ment pressure changes. The Hydrogen Skimmer fans will run continuous-ly until secured. Various typographical and format errors were corrected to ensure compliance with the EP Writer's Guide. These changes do not involve a USQ.
EP/1/A/5000/1C, Retype #10 The following changes are included-in Re-type #10. Step 6 was reword-ed to say " Check for indication of a small or intermediate LOCA." This wording was changed to correct a human factors deficiency. Step 11 gas reworded to say " Verify all NC T-COLDS have remained above 400 F." Per Design Engineering, the intent of this step is to verify ghat all Reactor Coolant System (NC) T-COLDS- have remained above 400 F. Also, the wording in the step was changed from NC W/R T-COLD to 20 l
NC T-COLD. This change was made to reflect the labelling of the Cold Leg Temperature indicators located in the Control Room. The setpoint for verifying proper Annulus Ventilation (VE) system operation was modified. The old setpoint was r. ore negative that -1.5 inches water column. The new setpoint is the band from -1.4 to -1.8 inches water column. This setpoint is mentioned in step 16 and Enclosures 1 and 4.
This new setpoint was obtained from Design Engineering. The new-setpoint band takes into_ account the tolerances associated with the VE System instruments. A step was added to Enclosure 1 (Item H) to monitor Annulus pressure between -1.4 to -1.8 inches water column at 30 minute intervals. Also, Enclosure 4 was modified to provide operator actions if Annulus pressure-is more negative that -1.8 inchec water column. These changes were made as a temporary measure to satisfy the requirements given in PIR 0-C89-0285. Per discussions with Design Engineering, it was determined that monitoring of the VE System is only required during a high energy line break inside con-tainment. A step was added to Enclosure-1 (Item I) to ensure that proper minimum flow requirements are met for the Residual Heat Removal (ND) pumps. This information was added as a result of a re-evaluation of ND pump minimum flow requirements which was performed by Design Engineering and Ingersoll-Rand. Substep 5.c was added to Enclosure 3.
This information was added to provide additional guidance to the operator in case the previous steps failed to start the fans. Step 6 on Enclosure 3 was reworded to say " Ensure NS pumps and Containment Air Return fans operate as Containment pressure changes." The previ-ous wording had the statement "VX fans." Only the Containment Air Return fans will start and stop as Containment pressure changes. The Hydrogen Skimmer fans will run continuously until secured. Substep 16.b was added to dispatch operator to place the hydrogen recombiners in service. This change was requested by Design Engineering. This change is being made to ensure that the Emergency Procedure (EP) properly reflects the guidance given Emergency Procedure Guideline E-1, "High Energy Line Break Inside Containment. The position of dampers 1AVS-D-3 and 1AVS-D-8 listed on Enclosure 14 was changed from closed to open. If Annulus pressure is more positive than -1.4 inches l water column, then the position of these two dampers should be in the open position. Various typographical and format errors were corrected to ensure compliance with the EP. Writer's Guide. No USQ is created by these changes.
TN/2/A/0568/00/02A Original This procedure provides implementation instructivns for Nuclear Station Modification (NSM) CN-20568, Rev. O. This NSM replaces Auxiliary Steam (SA) gate valves 2SA002 and 2SA005, item number 6H-204, with new item number 6H-211. These gate valves provide steam isolation for the auxiliary feedwater pump turbine. The existing valves are not suitable for the operating conditions experienced and require excessive maintenance. The purpose of this procedure is to provide guidance for replacing valve =2SA002 and 2SA005 with valve' item i number 6H-211. (Note: If 2SA005 has already'been replaced, then this procedure will involve 2SA002 replacement only.) Testing for NSM ,
CN-20568, Rev. O will be performed in accordance with the j l
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post-modification testing program at the station. The following testing will be performed:
Before Mode 3 Up:
Pressure Testing of New Welds Steam Supply Line Heat Tracing operability Testing Calibration of 2SA002 and 2SA005 Instrumentation IWV testing of 2SA002 and 2SA005 In Mode 3 Up:
Auxiliary Feedwater Pump Turbine Governor Response Testing Auxiliary Feedwater Pump Head Curve Testing Auxiliary Feedwater Pump IWP Testing Auxiliary Feedwater System Operability Testing Implementing this procedure will require isolation of valves 2SA002 and 2SA005. The Operations group will coordinate the isolations. The valve (s) may be replaced during an outage in Modes 4, 5, 6, or No Mode, or in Modes, 1, 2, and 3 during a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement. The auxiliary foodwater pump turbine will be out of carvice during this modification. Based on the above discussion, no USQ is involved with the implementation of this procedure.
TN/2/A/0471/00/01A Original This procedure provides implementation instructions for Nuclear Station Modification (NSM) CN-20471, Rev. O, Work Unit 01. NSM l
CN-20471 Rev. O reroutes the valve stem leakoff header and a portion i
of the Reactor Coolant Drain Tank (NCDT) recirculation line. The valve stem leakoff header connection is relocated from the NCDT drain line (normally stagnant) to the NCDT recirculation piping (normally flowing). The valve stem leakoff header ties into the NCDT recircula-tion piping at the lowest possible point to allow gradual entrain-ment / condensation of the steam. The purpose of this procedure is to provide guidance for rerouting the leakoff header to the new connec-tion at the NCDT recirculation piping. Testing for NSM CN-20471, Rev.
O will be performed in accordance with the Post Modification Testing program at the station. A Flow Verification Test will be performed to verify unobstructed flow through the new piping modification. This test will be performed before the NCDT is returned to service. In addition, the following tests will be performed in Mode 3 with the NCDT in service and the system at normal temperature and pressure:
(1) A Leakage Test (MP/0/A/7650/88) of the piping. modification and (2)
A Visual inspection for leakage at the quick disconnect fittings that were disconnected during the Flow Verification Test. Also, a tempo-rary station modification will be implemented by the Maintenance Engineering Services Group to monitor vibration of the piping affected by this modification. The monitoring will ccatinue for several months and possibly until the next refueling outage. After monitoring is complete, Design Engineering will analyze the data to determine if any unacceptacle vibration has occurred.
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Implementing this procedure will require isolation and draining of the NCDT. The Operations group will coordinate the isolations necessary to implement this procedure. The piping modification will be imple-mented during an outage in Modes 5, 6, or No Mode. The reactor coolant pumps and NCDT will be out of service during the piping modifications. Implementing this procedure does not affect any systems required for refueling operations. Per this discussion, no USQ exists.
PT/1/A/4200/04E Re-type, Changes 0 to 5 Incorporated This procedure retype added requirements to ensure the Reactor Vessel Head is installed if fuel is in the core to ensure during flow testing that any Sust/ debris does not enter the Reactor-Coolant system. The unit status change ensures the proper scheduling of the test. No USQ is involved with this change.
MP/0/A/7600/94 Re-type, Changes O to 3 Incorporated This procedure provides a method for disassembly, inspection, reassem-bly, and corrective maintenance for Dresser Butterfly Valves. Techni-cal information was obtained from drawings CNM 1205.12-010, 011, and 012. This procedure also contains information specific to Catawba.
The changes made to this procedure were made as part of the procedure upgrade process. All changes were editorial in nature and no signifi-cant technical information was changed. The purpose of this procedure is to correct and improve the performance of these valves within their original design requirements and specifications. No USQ is involved.
MP/0/A/7150/10 Re-type The following procedure changes were made during the rewrite of this -
procedure. Section'2.0 was expanded to include reference to Station Directive 3.11.2, Housekeeping Requirements. Section 4.0 was expanded to include the standard safety precautions currently incorporated in all newly revised ice condenser procedures. These additional state-ments provide guidance on the use of the " buddy" system and being sure everyone is out before turning the lights off. Section 6.0 was revised to include the currently used statements and sign-offs associ-ated with verification and re-verification of procedure working copies. Section 8.0 was changed to include a more-comprehensive list of special tools. Section 11.0 was revised by breaking the steps into smaller, single task steps. The appropriate hold points were added for data sheet sign-offs. There is no change to the intent of the steps to be performed. The responsible Maintenance Engineering Services engineer is tasked with choosing.the flow passages to be inspected now, instead of the technicians performing the work. This change will ensure the channels to be inspected are randomly chosen, as is the intent of the Tech. Specs. In Section 13.0, Enclosures, the data sheets have been changed to reflect the changes to the prerequisites in the body of the procedure.
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i This procedure has been compared with Tech. Specs, and the FSAR to ensure that the actions controlled by the procedure will comply with established surveillance requirements. This procedure will ensure the ice condenser flow channels are maintained in their operable condi-tion. Therefore, no USQ exists.
TN/2/A/0524/00/01A Original This procedure provides implementation instructions for Nuclear ;
Station Modification (NSM) CN-20524, Rev. 0.- This NSM adds control on Train A and Train B Auxiliary Shutdown Panels (ASP) for valves 2NV039A and 2NV032B, respectively. This will enable the operator to close one of the valves, if necessary, when providing Auxiliary Pressurizer Spray from the Auxiliary Shutdown Panel. The purpose for this proce-dure is to provide guidance to add controls for valve 2NV039A on 2ASPA.
No work will begin on this procedure until Unit 2 is in Modes 5, 6, or No Mode. During these modes, the equipment affected by this procedure is not required to be operable. All equipment affected by this procedure and the design intent of this modification will be fully tested by Performance under Temporary Procedure TT/2/A/9200/61. This test will clso be performed in Modes 5, 6, or No Mode. Thus, no USQ is involved.
TN/2/A/0524/00/02A Original This procedure provides implementation instructions for NSM CN-20524, Rev. O. This NSM adds control-on Train A and Train B Auxiliary Shutdown Panels for valves 2NV039A and 2NV032B, respectively.- This will enable the operator to close one of the valves, if necessary, when providing Auxiliary Pressurizer Spray from the Auxiliary Shutdown Panel. The purpose for this procedure is to provide guidance to add controls for valve 2NV032B on 2ASPB.
No work will begin on this procedure until Unit 2 is in Mode 5, 6, or No Mode. During the implementation of this procedure, as a-result of-the electrical isolations, power operated relief valves 2NC032B and 2NC036B will fail closed. Per Toch. Spec. 3.4.9.3, these valves are required to be operable during Modes 5 and 6 withLthe Reactor Vessel ,
Head on. 1 To comply with Tech. Spec. 3.4.9.3, this procedure will be implemented with one safety valve (2NC1, 2NC2, or 2NC3) removed or the reactor head removed to provide a vent path of 4.5 square inches. ;
All other equipment affected by this procedure is not required to-be i operable during Modes 5, 6, or No Mode. All equipment affected by )
this procedure and the design intent of the NSM will be fully tested )
by Performance under Temporary Procedure TT/2/A/9200/61. This test I will also be performed in Modes 5, 6, or No Mode. No USQ exists.- l 1
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PT/2/A/4400/09 Change /11 This procedure change revises the acceptance criteria for the Contain-ment Spray (NS) and Diesel Generator Engine Cooling Water (KD) heat exchangers. The revised criteria are based on the previous results of the test. The results were checked against the heat capacity and flow balance test results to ensure that at this minimum value, the compo-nents were operable. Failure of the test once revised would result in the responsible engineer making a determination to clean the heat exchanger or to perform a flow balance and/or heat capacity test. The
- procedure change also deletes the acceptance criteria for the Auxilia-l ry Shutdown Panel (ASP) Heating, Ventilation, and Air Conditioning (HVAC) units and the Diesel Generator Starting Air (VG) aftercoolers.
The deletion of acceptance criteria for the VG Aftercoolers and the ASP HVAC Units will have no affect on any operability. These results will be used for trending purposes only. The test method is not changed by these changes to the acceptance criteria. No USQ is involved.
PT/1/A/4400/09 Change #24 This procedure change revises the acceptance criteria for the NS and KD heat exchangers. The revised criteria are based on the previous results of the test. The results were checked against the heat capacity and flow balance test results to ensure that at this minimum value, the components were operable. Failure of the test once revised would result in the responsible engineer making a determination to clean the heat exchanger or to perform a flow balance and/or heat capacity test. The procedure change also deletes the acceptance criteria for the ASP HVAC units and the-VG aftercoolers. The deletion of acceptance criteria for the VG Aftercoolers and the ASP HVAC Units will have no affect on any operability. These results will be used for trending purposes only. The test method is not changed by these changes to the acceptance criteria. No USQ is involved.
PT/0/A/4400/08 Change #44 This restricted change affects Enclosure 13.8. This enclosure of the
- procedure is used to verify pump house flows following some type of l
maintenance, without having to perform the entire flow balance. RN l pump discharge pressure is reduced to the pressure recorded during the l last flow balance by using the Containment Spray heat exchangers in I order to simulate the same plant conditions. At times, all four heat exchangers are required to be opened to get pressure low enough. At this time, only three heat exchangers are available. Component Cooling (KC) Trains 1B and 2A are in service. Therefore, it is desirable to use the non-operating KC heat exchangers to lower system pressure. This change will allow this to be done.
The KC Heat Exchanger Flow Control valves have a safety position of ,
open, so failing these valves open will not have a detrimental affect I on safety. Pump flow limits will not be exceeded at any time. No 25 l
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part of RN is made inoperable or is put into any degraded stace. No USQ is involved. 4 OP/1/A/6700/01 Change #176 This change updates curve 1.2.1 in the Unit 1 Data Book. This curve supplies the Operators with a reference to the results of PT/0/A/4150/21, Temporary Rod Withdrawal Limits Determination, per-formed on 04/23/90 for Catawba Unit 1 Cycle 5. This curve is required by Tech. Spec. 3.1.1.3.a to ensure that the Moderator Temperature Coefficient (MTC) is negative at 100% full power (and less negative i than required by Tech. Spec. Figure 3.1-0 at other power levels.) The l
results of PT/0/A/4150/21 provide for Temporary Rod Withdrawal Limits only at 100% and 95% full power. At all other power levels the MTC will be within Tech. Spec. Limits. The limiting Rod Position (Control Bank D) versus Boron Concentration is:
Boron (ppm) Control Bank D (steps) 1363 225 1370 202 1380 182 1392 162 Data from CNNE 1553.05-05 (Catawba Unit 1 Cycle 5 Startup and Opera-l tional Report) (C105 SOR) Figure 9 indicates that the predicted Hot Full Power (HFP) Boron Concentration for 4 Effective Full Power Days (EFPD) is 1047 ppmB. Since this is 316 ppmB less than the most limiting boron concentration, Temporary Rod Withdrawal Limits will no lorger be necessary beyond 4 EFPD. No USQ is involved.
OP/1/A/6700/01 Change #178 This change replaces pages 1 and ? of Table 2.2. This table is a table of data for use by plant personnel. This change incorporates
- new Power Range Nuclear Instrumentation System (NIS) Calibration Currents obtained per PT/1/A/4600/05D and new Intermediate Range NIS 20% rod stop and 25% reactor trip setpoints obtained per PT/1/A/4150/21.
Information in OP/1/A/6700/01 (Unit One Data Book) is changed only by approved procedure change. Data is obtained by use of approved test procedures. No USQ is created by these changes.
OP/1/A/6700/01 Change #180 This change replaces page 1 of Table 2.2. This change incorporates new Full Power Calibration Currents for Power Range NIS Detectors obtained per PT/1/A/4600/05A.
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Information in OP/1/A/6700/01 (Unit One Data Book) is changed only by approved procedure change. Data is obtained by use of approved test procedures. No USQ is created by these changes.
PT/0/A/4400/01A Change #29 Per the National Fire Protection Association (NFPA), the proper method for testing a fire pump is to test at three points on the head curve.
These points are shutoff head (zero flow), design flow, and 150% of design flow. Technical Specification 4.7.10.1.f.2 also requires that the pump be tested at three points on the head curve. Presently this procedure tests the pumps at the design point and 150% design point, but not at shutoff head. This discrepancy was noted by the NRC Resident Inspector on deviation notice 413/89-16-03. Design Engineer-ing issued new Test Acceptance Criteria (TAC) Sheets to allow the relief valve setpoint verification to be performed in lieu of a shutoff head test point. The purpose of this change is to allow the procedure to test to the new TAC sheet data. This change will have test personnel take the data at design flow (approximately 2500 gpm) and then 150% design flow. After the final data point, flow will be reduced until the appropriate pump discharge relief valve lifts. The pressure at which this occurs will be recorded. If this pressure is not within the required tolerance, the relief valve will be adjusted by throttling the handwheel until the pressure is set correctly.
Only one Exterior Fire Protection (RY) pump is made inoperable at any one time. The required number of RY pumps will be operable at all times. Thus, no USQ is involved with this procedure change.
PT/0/A/4400/01S Change #17 In December 1987, PIR 0-C87-0366 was initiated due to low flow /high differential pressure in the Yard and Auxiliary Building sections of the Fire Protection system (RF and RY) piping. The data was evaluated by Design Engineering, and the RF system was deemed operable. Design Engineering did, however, request additional data to be taken in order to better model the system and to learn where the problems actually were occurring in the piping. This additional data consisted of pressure measurements at several different locations in the flow loop, some additional flow data, and an additional flow path through the Residual Heat Removal (ND) header of the RF piping. This change will allow this data to be obtained. In order to acquire this data, PT/0/A/4400/01S has been revised to incorporate the additional infor-mation. The flow paths are essentially the same, with only the ND sprinkler header being added. The data normally taken by this proce-dure will be taken, so that this surveillance may also be met. This change also installs additional instrumentation.
Although the RF/RY System is not directly nuclear safety related, it does serve equipment that is. At no time, however, is any equipment l left unprotected. The RF/RY System is arranged in a loop, so that l when one side of the loop is isolated, as it is in this test, to l
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l prevent short-circuit flow paths, the equipment can still be protect-ed. Three pumps are provided in the syctem, and two of the pump breakers are racked out for this test. One pump is all that is necessary for the system to operate as designed. The test flow path can be isolated at any time by an Operator stationed at the flow control valve. In addition,the Fire Detection System (EFA) will be operational and will still be able to detect a fire anywhere in the plant and alert Operations.of any problem. Fire hoses will be discon-nected temporarily for pressure readings during the test, but can be immediately reconnected at a moments notice by personnel stationed at l these locations. Pressure readings taken at alarm check valves will use existing test connections and will not affect the operation of these valves.
Technical Specification 3/4.7.10.1 allows operation of the Fire Protection system for a period of seven days with one RY Pump opera-ble. Technical Specification 3/4.7.10.2 requires that the listed sprinkler systems be operable. The operation of these valves is not affected by the test. In addition, Technical Specification 3/4.7.10.4 requires that the listed hose stations are operable. While the fire hose is disconnected for pressure readings, the station will be manned at all times in order that the hose could immediately be reconnected if necessary. Therefore, these hose rack stations will remain opera-ble at all times. Since all conditions of the test do not violate Technical Specifications, then the margin of safety as defined in the bases is not reduced. No USQ is involved.
OP/1/A/6700/01 Change #175 1 This change modifies the High Trip Setpoints on page 2 of Table 2.2 for Power Range Detectors N41, N42,.N43, _and N44 from 25% to 109%.
This documents resetting of the Power Range High Trip Setpoints to 109% to allow power escalation above 25% after the completion of Zero Power Physics Testing. Operation with these setpoints at 109% is bounded by the assumptions in FSAR Chapter 15 accident analysis. No USQ is involved.
IP/0/A/3162/05 Re-type, Changes 0 to 14 Incorporated l
( This procedure is used to verify proper operation of the Control Area Ventilation system chlorine detectors, and to provide guidance for maintaining the chlorine detectors. The chlorine detectors are used in the system to mitigate the consequences of' chlorine gas entering the Control Room. The affected intake isolation valve is closed should the detector sense the presence of chlorine.
Control of removing Chlorine Detectors from service to perform this procedure is maintained by the Technical Specification Action Items Log entries. This position of the system is not addressed in the FSAR Safety Analysis. This procedure verifies proper operation of the system as required by Tech. Specs. A USQ does not exist as a result of this procedure.
28
l TN/2/A/2700/CE/01A Original This procedure provides implementation guidelines for Exempt Change CE-2700. This Exempt Change provides an access hole in the Main Steam (SM) piping at Steam Generator (S/G) 2C. -This hole will be used-to access the inside of the piping in order to perform a radiographic examination of the weld between the piping and the S/G. This-inspec-tion is necessary in order to comply with the ASME Code Section XI Inservice Inspection Requirements. After the examination is complete, !
i the hole will be plugged and seal welded. The purpose of this proce-dure is to provide guidance for drilling the hole and installing a half coupling and plug.
The Operations group will coordinate the isolations necessary to implement this procedure.- S/G 2C and its associated SM line will be out of service during the installation phase of.this procedure. _This work will be. performed in Mode 5, 6, and No Mode. The Operations group will be notified to ensure containment integrity is maintained while the affected SM line is open to containment.
Testing will be performed in accordance with the Station Post-Modifi- %
cation testing program. A visual inspection.for leaks will be per-formed in Mode 3 with the affected SM piping at normal system tempera-ture and pressure. This inspection will net affect the function or operation of the SM system. Per this discussion, there are no USQs involved with the-implementation of this proceduro.
PT/2/A/4400/06C Change #O This test is_run to determine the heat removal capability of the Component Cooling (KC) Heat Exchanger 2A. It does not affect its 4 operability during the test. KC train 2A will be aligned for normal operation. Nuclear Service Water-(RN)_will be aligned for normal operation. If KC and RN temperatures are not stable,'2RN291 will be throttled by use of a regulated air supply in an effort to stabilize-temperatures. Both KC and RN will remain operable during the test.
No significant temperature transients are expected to occur. No USQ exists.
MP/0/A/7150/32 Re-type, Changes 0 to 2 Incorporated The changes to the procedure are as follows.- Section 2.0 was. updated and reformatted. Section 11.0 was rewritten to include steps _for removal and reinstallation of the gear motor. Also, the initial
" NOTE" in this section was' revised by making the last'two sentences l into steps. Enclosure.13.1 data sheet was revised to include a sign-off for-reinstallation of the gear motor. The changes made by this rewrite have been reviewed against approved vendor manualn, design documents, and station procedures to ensure that the corrective maintenance controlled by the procedure will return the strainer to as-built /as-designed condition. These actions will ensure the strain-er's compliance with FSAR accident analysis. Operations will ensure 29
that maintenance is performed on these strainers when allowed by Tech.
Specs. No USQ is created. '
l MP/n/A/7150/87 Initial Issue This new procedure provides a method of disassembly, cleaning, correc-tive maintenance, and reassembly for the Nuclear Service Water Lube Injection (S. P. Kenny) Strainer. This new procedure will assure that maintenance activities will return the strainers to as-designed conditions. This procedure has been reviewed against approved vendor manuals, design documents, and station procedures to ensure that the corrective maintenance controlled by this procedure will return the strainer to as-built /as-designed conditions. These actions will ensue that strainer's compliance with FSAR accident analyses. Maintenance will be performed on these strainers when Tech. Specs, allow, per Operations procedures. No USQ is created.
TN/1/A/1105/01/14A Re-type #2 Nuclear Station Modification (NSM) CN-11005 Rev. 1 modifies various piping system math models with the objective of reducing the number of mechanical snubbers required. This work unit covers math model NDF.
This model includes supports in the Residual Heat Removal (ND) and Safety 73jection (NI) systems. These supports will either be deleted from the system or revised to a different configuration. Instructions are provided such that this procedure may be worked in any mode of unit operation.
There are ng system isolations required to implement this procedure.
The only concern is the seismic qualification of the affected systems' piping during implementation of this procedure. The Math Model has been qualified for the present support / restraint configuration. It has also been qualified for the support / restraint configuration which will be in place after this procedure has been implemented. However, the interim configuration (with some deleted snubbers removed and some still in place) has not been analyzed because the many possible l combinations of Support / Restraint (S/R) configurations would require l numerous analyses. For this reason, Design has determined this work may be done while the affected system (s) are operable provided all the support modifications for the entire math model are completed within the 72 houro allowed by the technical specification for snubbers. In order to avoid exceeding the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limit and declaring the affected system (s) inoperable, Design Engineering has performed analyses that I
l indicate the affected piping could be qualified under any combination of snubbers removed provided the modifications to the existing sup-ports have been completed. This procedure is written to complete all modifications before deleting any supports. However, if the support modifications cannot be completed in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit, Design Engineering will be contacted to perform an analysis of the affected piping to determine operability. Design Engineering has stated they could justify operable status of any configuration of modified sup-ports provided the affected piping is adequately supported. Station 30
maintenance procedures provide guidance to ensure piping is adequately supported. This procedure will provide the necessary controls to ensure compliance with all technical specification requirements.
Based on this discussion, there are no unreviewed safety questions associated with the implementation of this procedure.
TN/1/A/1105/01/16A Re-type #1 NSM CN-11005 Rev. 1 modifies various piping system math models with the objective of reducing the number of mechanical snubbers required.
This work unit covers math model NIC. This model includes supports in the NI, ND, and Refueling Water (FW) systems. These supports will either be deleted from the system or revised to a different configura-tion. Instructions are provided such that this procedure may be worked in any mode of unit operation.
There are DQ system isolations required to implement this procedure.
The only concern is the seismic qualification of the affected systems' piping during implementation of this procedure. The Math Model has been qualified for the present support / restraint configuration. It has also been qualified for the support / restraint configuration which will be in place after this procedure has been implemented. However, the interim configuration (with some deleted snubbers removed and some still in place) has not been analyzed because the many possible l
combinations of Support / Restraint (S/R) configurations would require numerous analyses. For this reason, Design has determined this work may be done while the affected system (s) are operable provided all the l support modifications for the entire math model are completed within the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed by the technical specification.for snubbers. In order to avoid exceeding the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limit and declaring the affected system (s) inoperable, Design Engineering has performed analyses that indicate the affected piping could be qualified under any combination of snubbers removed provided the modifications to the existing sup-ports have been completed. This procedure is written to complete all modifications before deleting any supports. However, if the support modifications cannot be completed in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit, Design Engineering will be contacted to perform an analysis of the affected piping to determine operability. Design Engineering has stated they could justify operable status of any configuration of modified sup-
- ports provided the affected piping is adequately supported. Station maintenance procedures provide guidance to ensure piping is adequately supported. This procedure will provide the necessary controls to ensure compliance with all technical specification requirements.
Based on this discussion, there are no unreviewed safety questions associated with the implementation of this procedure.
TN/1/A/1105/01/17A Re-type #2 NSM CN-11005 Rev. 1 modifies various pipino system math models with the objective of reducing the number of me.hanical snubbers required.
31
This work unit covers math model NDE. This model includes supports in the ND, NI, Containment Spray (NS), and FW systems. These supports will either be deleted from the system or revised to a different configuration. Instructions are provided such that this procedure may be worked in any mode of unit operation.
There are Dg system isolations required to implement this procedure.
The only concern is the seismic qualification of the affected syste 5 8 piping during implementation of this procedure. The Math Model has been qualified for the present support / restraint configuration. It has also been qualified for the support / restraint configuration which will be in place after this procedure has been implemented. However, the interim configuration (with some deleted snubbers removed and some still in place) has not been analyzed because the many possible combinations of Support / Restraint (S/R) configurations would require
} numerous analyses. For this reason, Design has determined this work may be done wnile the affected system (s) are operable provided all the support moaifications for the entire math model are completed within the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed by the technical specification for snubbers. In order to avoid exceeding the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limit and declaring the affected system (s) inoperable, Design Engineering has performed analyses that indicate the affected piping could be qualified under any combination of snubbers removed provided the modifications to the existing sup-ports have been completed. This procedure is written to complete all modifications before deleting any supports. However, if the support modifications cannot be completed in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit, Design Engineering will be contacted to perform an analysis of the affected piping to determine operability. Design Engineering has stated they could justify operable status of any configuration of modified sup-ports provided the affected piping is adequately supported. Station maintenance procedures provide guidance to ensure piping is adequately supported. This procedure will provide the necessary controls to ensure compliance with all technical 9pecification requirements.
Based on this discussion, there are no unreviewed safety questions associated with the implementation of this procedure.
TN/1/1005/01/21A Re-type #1 NSM CN-11005 Rev. 1 modifies various piping system math models with the objective of reducing the number of mechanical snubbers required.
This work unit covers math model CAG. This model includes supports in the Auxiliary Feedwater and Main Feedwater systems. These supports will either be deleted from the system or revised to a different configuration. Instructions are provided such that this procedure may be worked in any mode of unit operation.
There are Do system isolations required to implement this procedure.
The only concern is the seismic qualification of the affected systems' piping during implementation of this procedure. The Math Model has been qualified for the present support / restraint configuration. It has also been qualified for the support / restraint configuration which will be in place after this procedure has been implemented. However, 32 ;
I the interim configuration (with some deleted snubbers removed and some still-in place) has not been analyzed because the many possible combinations of Support / Restraint (S/R) configurations would requir-numerous analyses. For this reason, Design has determined this work may be done while the affected system (s) are operable provided all the support modifications for the entire math model are completed within the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed by the technical specification for-snubbers. In order to avoid exceeding the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limit and-declaring the affected system (s) inoperable, Design Engineering has performed analyses that indicate the affected piping could be qualified under any combination of snubbers removed provided the modifications to the existing sup-ports have been completed. This procedure is written to complete all modifications before deleting any supports. However, if the support modifications cannot be completed in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit, Design Engineering will be contacted to perform an analysis of the affected piping to determine operability. Design Engineering has stated they could justify operable status of any configuration of modified sup-ports provided the affected piping is adequately supported. Station maintenance procedures provide guidance to ensure piping is adequately supported. This procedure will provide the necessary controls to ensure compliance with all technical specification requirements.
Based on this discussion, there are no unreviewed safety questions associated w_ the implementation of this procedure.
TN/2/A/0330/00/01A Original Nuclear Station Modification (NSM) CN-20330, Rev. O will modify the control circuit wiring for valves 2CA007A, 9B, 11A, 2FWO27A, 55B, 2KC051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN250A, 310B, 2SM074B, 75A, 76B, 77A, 2SV025B, 26B, 27A,-and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted indepen-
- dently of indications or interlocks and provide data to complete Motor
! Operated Valve (MOV) testing of these valves. The MOV testing infor-mation included in the NSM will supersede the old torque switch )
setting values and replace them with thrust values. The purposelof the thrust values is to ensure that torque switch settings are select-ed, set, and maintained correctly to accommodate the maximum differen-tial pressure expected on the valve during both normal and abnormal l
events within the design basis. The new thrust values ensure the l valve will operate during normal and abnormal events by setting l
limitations on Total Thrust, Differential Pressure Thrust, and Packing Load. This procedure will control work being performed on valve 2CA007A. Instrument and Electrical (IAE) will perform all work at the valve. IAE will rewire the-rotors, setup the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve. Valve 2CA007A is the Auxiliary Feedwater (CA). Pump #2 normal suction isola-tion valve. Prior to returning the valve to service, a functional verification will be performed to verify valve operability. This procedure will be performed in Modes 4, 5, 6, or No Mode when Auxilia-ry FeedWcter is not required to be operable. A USQ does not exist.
33
1 I
TN/2/A/0330/00/02A Original Nuclear Station Modification CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWO27A, 55B, 2KC051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN250A, 310B, 2SM074B, 75A, 76B, 77A, 2SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque l switch settings are selected, set, and maintained correctly to accom-modate the maximun differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting ifmitations on Total Thrust, Differential Pressure Thrust, and Packing Load. This procedure will control work being performed on valve 2CA009B. IAE will perform all work at the valve.
IAE will rewire the rotors, setup the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve. Valve 2CA009B is the Auxiliary Feedwater Pump 2B normal suction isolation valve. Prior to returning the valve to service, a functional verification will be performed to verify valve operability. This procedure will be per-formed in Modes 4, 5, 6, or No Mode when Auxiliary Feedwater is not required to be operable. A USQ does not exist.
TN/2/A/0330/00/03A Original Nuclear Station Modification CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWO27A, 558, 2KC051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN250A, 310B, 2SM074B, 75A, 76B, 77A, 2SV02SB, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted indep'endently of indica-tions or interlocks and provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal i events by setting limitations on Total Thrust, Differential Pressure Thrust, and Packing Load. This procedure will control work being performed on valve 2CA011A. IAE will perform all work at the valve. ,
IAE will rewire the rotors, setup the switch rotors, verify add-on-pak j switch setup, and perform MOV testing of the valve. Valve 2CA011A is '
the Auxiliary Feedwater Pump 2A normal suction isolation valve. Prior to returning the valve to service, a functional verification will be 4 performed to verify valve operability. This procedura will be-per-formed in Modes 4, 5, 6, or No Mode when Auxiliary Feedwater is not l l required. A USQ does not exist.
l 34 1
TN/2/A/0330/00/04A Original Nuclear Station Modification CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWO27A, 55B, 2KC051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN250A, 310B, 2SM074B, 75A, 76B, 77A, 2SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and-provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values.is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve'during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, Differential Pressure Thrust, and Packing Load. This procedure will control work being performed on valve 2FWO27A. IAE will perform all work at the valve.
Construction Maintenance Department (CMD) will perform work remote to the valve required to support the torque switch bypass modification.
IAE will rewire the rotors, setup the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve. Performance will stroke time the valve. Valve 2FWO27A is the Residual Heat Removal (ND) Purp 2A suction from the Refueling Water Storage Tank (FWST).
Prior to returning the valve to service, a functional verification and a retest will be performed to verify valve operability.
This procedure may be implemented in Modes 4, 5, or 6, as long as Train B ND is operable. In No Mode, this procedure may be implemented at any time. A USQ does not exist.
TN/2/A/0330/00/05A Original Nuclear Station Modification CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWO27A, SSB, 2KC051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN250A, 310B, 2SM074B, _75A, 76B, 77A, 2SV025B, 26B, 27A, and 28A to. provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of these l va)ves. The MOV testing information included in the NSM will super-srJe the old torque switch setting values and replace them with thrust salues. The purpose of the thrust values is to ensure that-torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, Differential Pressure Thrust, and Packing Load. This procedure will control work being performed on valve.2FWO55B. IAE will perform all work at the valve.
CMD will perform work remote to the valve required to support the torque switch bypass modification. IAE will rewire the rotors, setup the switch rotors, verify add-on-pak switch setup, and perform MOV testing of t c valve. Performance will stroke time the valve. Valve 35
2FWO55B is the ND Pump 2B suction from the FWST. Prior to returning the valve to service, a functional verification and a retest will be performed to verify valve operability.
This procedure will be implemented on Unit 2 in Modes 4, 5, or 6 as long as Train A ND is operable. In No Mode, this procedure may be implemented at any time. A USQ does not exist.
TN/2/A/0330/00/06A Original Nuclear Station Modification CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWO27A, 55B, 2KC051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN250A, 310B, 2SM074B, 75A, 76B, 77A, 2SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to e7sure that torque switch settings are selected, set, and maintaint.d correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the desija basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, Differential Pressure Thrust, and Packing Load. This procedure vill control work being performed on valve 2KC051A. IAE will perform all work at the valve.
IAE will rewire the rotors, setup the switch rotors, verify add-on-pak switch setup, and perform MOV testing of the valve. Performance will stroke time the valve. Valve 2KC051A is the Component Cooling (KC)
Train 2A recirculation line isolation valve. Prior to returning the valve to service, a functional verification and a retest will be performed to verify valve operability.
l This procedure may be implemented with Unit 2 in any Mode. In Modes 1, 2, 3, and 4, this procedure may be implemented only when KC Train U is operable. In Mode 5, this procedure may be implemented with the reactor coolant loops filled. In Mode 6, this procedure may be implemented when the water level is greater than or equal to 23 feet above the top of the reactor vessel flange. A USQ is not created by this procedure.
TN/2/A/0330/00/07A Original l Nuclear Station Modification CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWU27A, 55B, 2KC051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN250A, 310B, 2SM074B, 75A, 76B, 77A, 2SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will super-sede the old torque switch setting valuas and replace them with thrust values. The purpose of the thrust values is to ensure that torque 36
switch settings are selected, set, and maintained corredtly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, Differential Pressure Thrust, and Packing Load. -This procedure will control work being performed on valve 2KC054B. IAE will perform all work at the valve.
IA will rewire the rotors, setup the switch rotors, verify add-on-pak ewitch setup, and perform MOV testing of the valve. Performance will stroke time the valve. Valve 2KC054B is the Train 2B recirculation line isolation valve. Prior to returning the valve to service, a functional verification and a retest will be performed to verify valve operability.
This procedure may be implemented with Unit 2 in any Mode. In Modes 1, 2, 3, and 4, this procedure may be implemented only when KC Train A is operable. In Mode 5, this procedure may be implemented with the reactor coolant locps filled. In Mode 6, this procedure may be implemented when the water level is greater than or equal to 23 feet above the top of the reactor vescel flange. A USQ is not created by this procedure.
TN/2/A/0330/00/08A Original Nuclear Station Modification CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWO27A, SSB, 2KC051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN200A, 310B, 20M074B, 75A, 76B, 77A, 2SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque r
switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new l thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, Differential Pressure ;
Thrust, and Packing Load. This procedure will control work being
- performed on valve 2ND028A. IAE will perform all work st the valve. ;
l CMD will perform work remote to the valve required to support the torque switch bypass modification. IAE will rewire tre rotore, setup ,
the switch rotors, verify add-on-pak switch setup. Tne MOV setup data (
l for 2ND028A has been deleted from this modification due to the un" l l availability of test equipment. IAE will setup the torque switches as l specified in CNM-1205.00-1997 As-Built. Performance will stroke t ime the valve. Valve 2ND028A is the Residual Heat Removal Heat Excha*.ger 2A outlet to Chemical and Volume Control System pump suction valve.
Prior to returning the valve to service, a functional verification and a retest will be performed to verify valve operability.
l l
i 37 l
l l
This procedure will be implemented in Modes 5, 6, or No Mode when this portion of the Residual Heat Removal System is not required to be operable. A USQ is not created by this proce uro-IN/2/A/0330/00/09A original Nuclear Station Modification CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWO27A, 55B, 2KC051A, 54B, 2ND028A, 2NS001B, 3D, 20A, 3bB, 43A, 2RN250A, 310B, 2SM074B, 75A, 76B, 77A, 2SV0258, 263, 27A, and 28A to provide " limit actuated" torque switch. bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, Differential Pressure Thrust, and Packing Load. This procedure will control work being performed on valvo 2NS0018. IAE will perform all work at the valve.
CMD will perform work remote to the valve required to support the torque switch bypass modification. IAE will rewire the rotors, setup the switch rotors, verify add-on-pak switch setup, and perform MOV testing on the valvo. Performance will stroke time the valve. Valve l 2NS001B is the Containment' Spray (NS) Pump 2B suction from the Con-tainment Sump. Prior to returning the valve to service, a functional verification and a retest will be performed to verify valve operabili-ty. ,
This procedure will be implemented with Unit 2 in Moden 5, 6, or No Mode when Containment Spray is not required to be operable. A USQ is not croated by this procedure.
TN/2/A/0330/00/10A Original Nuclear Station Modification CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWO27A, SSB, 2KC051A, 54B, i 2ND028A, INS 001B, 33, 20A, 38B, 43A, 2RN250A, 310B, 2SM074B, 75A, 76B, l 77A, 2SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and. provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will super-t sede the old torque switch setting values and replace them with thrust i values, The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-l modate the maximum differential pressure expected on the valve during i
both normal and abnormal events within the design basis. The now thrust values onsure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, Differential Pressure Thrust, and Packing Load. This procedure will control work being 38
l 1
performed on valve 2NS003B. IAE will perform all work at the valve.
CMD will perform work remote to the valve required to support the torque switch bypass modification. IAE will rewiro the rotors, setup the switch rotors, verify add-on-pak svitch setup, and perform MOV testing on the valve. Performance will stroke time the valve. Valve 2NS003B is the Containment Spray (NS) Pump 2B suction from Refueling Water Storage Tank (FWST) isolation velve. Prior to returning the valve to servico, a functional verification and a retest will be performed to verify valve operability.
This procedure will bo implemented with Unit 2 in Modes 5, 6, or No Mode when Containment Spray is not required to be operable. A USQ is not created by this procedure.
TN/2/A/0330/00/11A original Nuclear Station Modificativ. CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWO27A, 55B, 2KC051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN250A, 3108, 2SM074B, 75A, 76B, 77A, 2SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjustcc independently of indica-tions or interlocks and provido data to complete MOV testing of these valves. The MOV testing information 'ncludof in the NSM will saper-sode the old torque switch sotting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, Differential Pressure Thrust, and Packing Load. This procedure will control work being performed on valve 2NS020A. IAE will perform all work at the valve.
CMD will perform work remote to the valve required to support the torque switch bypass modification. IAE will rewira the rotors, setup I the switch rotors, verify add-on-pak switch setup, and perform MOV l testing on the valve. Performance will stroke t!ae the valvo. Valve 2NS020A is the NS Pump 2A suction from FWST isolation valve. Prior to returning the valve to service, a functional verification and a retest will be performed to verify valve operability.
This procedure will be implemented with Unit 2 in Modes 5, 6, or No Mode when Containment Spray is nc~ required to be operable. A USQ is not created by this procedure.
TN/2/A/0330/00/12A Original Nuclear Station Modification CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWO27A, 55B, 2KC051A, 54B, 2ND026A, 2 NS 001'J , 3B, 20A, 38B, 43A, 2RN250A, 310B, 2SM074B, 75A, 76B, 77A, 2SV0258, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of those 39 l
l
valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations en Total Thrust, Differential Pressure Thrust, and Packing Load. This procedure will control work being performed on valve 2NS038B. IAE will perform all work at the valve.
CMD will perform work remote to the valve required to support the torque switch bypass modification. IAE will rewire the rotors, setup the switch rotors, verify add-on-pak settch setup, and perform MOV testing on the valve. Performance will stroke time the valve. Valve 2NS038B is the Residual Heat Removal (ND) Pump 2B discharge to con-tainment spray header isolation valve. Prior to returning the valve to service, a functional verification and a rotest will be performed to verify valve operability.
This procedure will be implemented with Unit 2 in Modes 5, 6, or No Mode when containment Spray is not required to be operable. A USQ is not created by this procedure.
TN/2/A/0330/00/13A Original Nuclear Station Modification CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWO27A, 55B, 2KC051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN250A, 3108, 2SM074B, 75A, 76B, 77A, 2SV025B, 268, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will super-sede the old torqua switch setting values and replace them with thrust .
values. The purpose of the thrust values is to ensure that torque i switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valvo during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitatior.s on Total Thrust, Differential Pressure Thrust, and Packing Load. This procedure will control work being performed on valve 2NSO43A. IAE will perform all work at the valve.
CMD will perform work remote to the valve required to support the torque switch bypass modification. IAE will rewire the rotors, setup the switch rotors, verify add-on-pak switch setup, and perform MOV testing on the valve. Performance will stroke time the valvo. Valve 2NSO43A is the HD Pump 2A discharge to containment spray header isolation valve. Prior to returning the valve to service, a function-al verification and a retest will tr performed to verify valve opera-bility.
This procedure will be implemented with Unit 2 in Modes 5, 6, or No lode when Containment Spray is not required to be operable. A USQ is not created by this procedure.
40 t
1 l TN/.2/A/0330/00/14A Original Nuclear Station Modification CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9D, 11A, 2TWO27A, 55B, 2KC051A, 548, 2ND028A, 2NS001B, 3B, 20A, 30B, 43A, 2RN250A, 310B, 2SM074B, 75A, 76B,
- 77A, 2SV025B, 268, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure thst torque nvitch settings are selected, set, and maintained correctly to accom-molate the maximum dif ferential pressure expcoted on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total-Thrust, Differential pressure Thrust, and Packing Load. This procedure will control work being performed on valve 2RN250A. IAE will perform all work at the valve.
IAE will rewirc the rotors, setup the switch rotors, verify add-on-pak nwitch setup, and perform MOV testing on the valve. Performance will stroke time the valve. Valve 2RN250A is the Nuclear Service Water (RN) header A to Auxiliary Feedwater (CA) pump suction isolation valve. Prior to returning the valve to service, a functional verifi-cation and a retest will be performed to verify valve operability.
This procedure will be implemented with Unit 2 in Modes 4, 5, 6, or No Mode when this portion of the RN system is not required to be opera-ble. A USQ does not exist.
TN/2/A/0330/00/15A Original Nuclear Station Modification CN-20330, Rev. O will modify the control circuit Wiring for 2CA007A, 9B, 11A, 2FWO27A, SSB, 2KC051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN250A, 3108, 2SM074B, 75A, 76B, l
77A, 2SV025B, 2GB, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set,-and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal - ' Tbnormal events within the design basis.- The new thrust valt aasure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, Differential Pressure Thrust, and Packing Load. This procedure will control work being performed on valve 2RN310B. IAE will perform all work at the valve.
l IAE will rewire the rotors, setap the switch rotors,-verify add-on-pak switch setup, and perform Mov testing on the valve. Performance will stroke time the valve. Valve 2RN310B is the RN header B to CA pump suction isolation valve. Prior to returning the valve to service, a functional verification and a retest will be performed to verify valve operability.
41
.___._u__. _ _ _ _ _ _
This procedure will be implemented with Unit 2 in Modes 4, 5, 6, or No Mode when this portion of the RN system is not required to be opera-ble. A USQ does not oxist.
TN/2/A/0330/00/16A Original I Nuclear Station Modification CN-20330, Rev. O will modify the control circuit wiring fcr 2CA007A, 9B, 11A, 2FWO27A, 55B, 2Kc051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN250A, 310B, 23M074B, 75A, 76B, 77A, 2SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provido data to complete MOV testing of these valves. The MOV testing information ir.cluded in the NSM will super-sede the old torque switch sotting values and replace them with thrust values. The purpose of the thrust values is to enoure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operato during normal and abnormal events by setting limitations on Total Thrust, Differential Pressure Thrust, and Packing Load. This procedure will control work being performed on valve 2SM074B. IAE will perform all work at the valve.
IAE will rowire the rotors, setup the switch rotors, verify add-on-pak switch setup, and perform MOV testing on the valvo. Valve 2SM074B is the Steam Generator (S/G) 2D outlet header blowdown control valve.
Prior to returning the valve to service, a functional verification will be performed to verify valve operability.
This procedure may be implemented with Unit 2 in any Mode. A USQ does not exist.
TN/2/A/0330/00/17A Original Nuclear Station Modification CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWO27A, 55B, 2KC051A, 54B, l,
2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN250A, 310B, 2SM074B, 75A, 76B, 77A, 2SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, Differential Pressure Thrust, and Packing Load. This procedure will control work being performed on valve 2SM075A. IAE will perform all work at the valve.
IAE will rewire the rotors, setup the switch rotors, verify add-on-pak switch setup, and perform MOV testing on the valve. Valve 2SM075A is the S/G 2C outlet header blowdown control valve. Prior to returning l
l 42 l
1
( the valve to service, a functional verification will be performed to l Verify valve operability.
This procedure may be implemented with Unit 2 in any Mode. A USQ does l
not exist.
l TN/2/A/0330/00/18A original Nuclear Station Modificatior. CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWO27A, 55B, 2KC051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN250A, 310B, 2SM074B, 75A, 7 0b ,.
77A, 2SV025B, 26B, 27A, and 28A to provide H1{g{t gggggggdH torggg switch bypass contacto which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, Differential Pressure Thrust, and packing Load. This procedure will. control work being performed on valve 2SM076B. IAE will perform all work at the valve.
IAE will rewire the rotors, setup the switch rotors, verify add-on-pak switch setup, and perform MOV testing on the valve. Valve 2SM076B is the S/G 2B outlet header blowdown control valve. Prior to returning the valve to service, a functional verification will be performed to verify valve operability.
This procedure may be implemented with Unit 2 in any Mode. A USQ does not exist.
TN/2/A/0330/00/19A Original Nuclear Station Modification CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWO27A, 55B, 2XC051A, 54B, 1
2ND028A, 2NS001B, 38, 20A, 38B, 43A, 2RN250A, 310B, 2SM074B, 75A,.76B, 77A, 2SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within-the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, Differential Pressure .'
Thrust, and Packing Load. This procedure will control work being performed on valve 2SM077A.- IAE will perform all work-at the valve.
IAE will' rewire the rotors, setup the switch rotors, verify add-on-pak 43
.,,,,.,m... , . . , . - < .e-r..,e .,ng,-
,7.- ._.
_.w. -. .g - , , . . .,-,,my.. , . , ~ , , , -.w , , .,. - - , , . , - . . ~ . + . , ,
I switch setup, and perform MOV testing on the valve. Valve 2SM077A is i
the S/G 2A outlet header blowdown control valve. Prior to returning the valve to service, a functional verification will be performed to verify valve operability.
This procedure may be implemented with Unit 2 in any Mode. A USQ does not exist.
TN/2/A/0330/00/20A original Nuclear Station Modification CH-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWO27A, 55B, 2KC051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN250A, 310B, 2SM074B, 75A, 76B, 77A, 2SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch bypass contacts which can be adjusted independently of indica-tions or interlocks and provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, Differential Pressure Thrust, and Packing Load. This procedure will control work being performed on valve 2SV025B. IAE will perform all work at the valve.
IAE t ill rewire the switch compartment, setvo the switch rotors, and perform MOV testing on the. valve. Valve 2SV025B is the Main Steam 2D power operated relief isolation valve. Prior to returning the valve-to service, a functional verification will be performed to verify valve operability.
This procedure may be implemented with Unit 2 in any mode. A USQ does not exist.
TN/2/A/0330/00/21A original Nuclear' Station Modification CN-20330, Rev. O will modify the control circuit wiring for 2CA007A, 9B, 11A, 2FWO27A,_SSB, 2KC051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN250A, 310B, 2SM074B,-75A, 76B, 77A, 2SV025B, 26B, 27A, and 28A to provide " limit actuated" torque switch. bypass contacts which can be adjusted-independently of indica-tions or interloc.ks and provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will super-sede the old torque switch setting values and replace them with thrust' values. The purpose of the thrust values is to ensure that torque switch settings are selected, set, and maintained correctly to accom-modate the maximum differential pressure expected on the valve during both normal and abnormal events within the design basis. The new thrust values ensure the valve will operate during normal and abnormal events by setting limitations on Total Thrust, Differential Pressure Thrust, and Packing Load. ThisLprocedure will control work being l
l i
44 l
l _ _ _ _ . __ .. . _ _ , . _ .
i t
I performed on valve 2SV026B. IAE will por. form all work at the valve. l IAE will rewire the switch compartment, setup the switch rotors, and
. perform MOV testing on the valve. Valve 38V026B is the Main Steam 2c
! power operated relief isolation valve. Prior to returning the valve
! to service, a functional verification will bn performed to verify i valve operability.
3 This procedure may be implemented with Unit 2 in any mode. A-USQ does not exist.
]
TN/2/A/0330/00/22A Original
+
Nuclear Station Modification CN-20330, Rov, o will modify the control circuit Wiring for 2CA007A, 98, 11A, 2FWO27A, 55B, 2KC051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 30B, 4 3 A, 2RN2'iOA, 110B, 2SM074B, 75A, 76B, J 77A, 2SV025B, 26B, 27A, and 29A to provide " limit actuated" torque switch bypass contacts which can be adjunted independently of indica-tions or interlooks and provide data to complete Mov testing of these j valves. The MOV testing information included in the NSM will super-l sede the old torque switch setting values end replace them with thrust i values. The purpose of the thrust valuos in to ensure that torque switch settings are selected, net, and maintained correctly to accom-modate the maximum differential prensure expected on the valve during both normal and abnormal cNents within the design basis. The new l thrust va3ues ensure the valve vill operato during normal and-abnormal 1
events by setting limitations cn Total Thrust, Differential Pressure Thrust, and Packing Loan. Thlu procedure will control work being
! performed on valve 2SV027Am IA2 will perform all work at the valve.
1 IAE will rewire the switch compartment, setup the cwitch rotors, and 2 perform MOV tecting on the vnive. Valve 2SV027A is the Main Steam 2A power operated re.lis f isolation valve. Prior to returning the valve to service, a functional verification will be performed to verify j valve operability.
This procedure may be inpioniented with Unit 2 in any mode. A USQ does not exist.
i TN/2/A/0330/00/23A Original Nuclear Station Modification CN-20330, Rev. O will modify the control
-circuit wiring for 2CA007A, 98,EA1A, 2FWO27A, 55B, 2KC051A, 54B, 2ND028A, 2NS001B, 3B, 20A, 38B, 43A, 2RN250A, 310B, 2SM074B, 75A, 76B, 77A, 2SV025B, 26B, 27A, and 28A to provide " limit actuated" torque-
. switch bypass contacts which can be adjusted independently of indica-tions-or interlocks and provide data to complete MOV testing of these valves. The MOV testing information included in the NSM will: super-sede the old torque switch setting values and replace them with thrust values. The purpose of the thrust values is to ensure that torque i switch settings are selected, set,.and maintained correctly to accom-modate the maximum differential pressure expected on the valve during.
both normal and abnormal events within the design basis.- The new -
thrust values ensure the valve will operate during normal and abnormal 45 l
1 1
events by setting limitations on Total Thrust, Differential Pressure Thrust, and Packing Load. This procedure will control work being performed on valve 2SV024A. IAE will perform all work at the valve.
IAE will rowire the switch compartment, setup the switch rotors, and perform MOV testing on the valve. Valve 2SV028A is the Main Steam 2B power operated relief iso.'ation valve. Prior to returning the valve to service, a functional varification will be performed to verify valve operability.
This procedure may be implemented with Unit 2 in any mode. A USQ does not exist.
TN/2/A/0396/01/01A Original NSM CN-20396, Rev. 1 modifies various piping system math models with the objective of reducing the number of mechanical snubbers required.
This work unit covers math model CAD. This math model includes supports is the Auxiliary Feedwater (CA) and Main Feedwater (CF) systems. This procedure provides guidance for the removal of snubbers deleted from math model CAD as well as the modification of certain other support / restraints.
These supports will either be deleted from the system or revised to a different configuration.
There are n2 system isolations required to implement this procedure.
The only concern is the seismic qualification of the affected systems' piping during implementation of this procedure. The Math Model has been qualified for the prosent support / restraint configuration. It has also been qualified for the support / restraint configuration which will be in place after this procedure has been implemented. However, the interim configuration (with some deleted snubbers removed and some
- still in place) has not been analyzed because the many possible combinations of Support / Restraint (S/R) configurations would require l numerous analyses. For this reason, Design has determined this work may be done while the affected system (s) are operable provided all the I support modifications for the entire math model are completed within the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed by the technical specification for snubbers. In order to avoid exceeding the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limit and declaring the affected system (s) inoperable, Design Engineering has performed analysos that indicate the affected piping could be qualified under any combination of snubbers removed provided the modifications to the existing sup-ports have been completed. This procedure is written to complete all modifications before deleting any supports. However, if the support modifications cannot be completed in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit, Design Engineering will be contacted to perform an analysis of the affected piping to determine operability. Design Engineering has stated they could justify operable status of any configuration of modified sup-ports provided the affected piping is adequately supported. Station maintenance procedures provide guidance to ensure piping is adequately supported. This procedure will provide the necessary controls to ensure compliance with all technical specification requirements.
46
Based on this discussion, there are no unreviewod safety questions associated with the implementation of this procedure.
TN/2/A/0396/01/02A Original NSM CN-20396, Rev. 1 modifies various piping system math models with the objective of reducing the number of mechanical snubbers required.
This work unit covers math model CAF. This math model includes supports is the CA and CF systems. This proceduro providos guidance for the removal of snubbers deleted from math model CAF as well as the modification of cortsin other support / restraints.
These supports will either bo deleted from the system or revised to a l different configuration.
There are n2 system isolations required to implomont this procedure.
The only concern is the soismic qualification of the affected systems' piping during implomontation of this proceduro. The Math Model has boon qualified for the present support / restraint configuration. It has also boon qualified for the support / restraint configuration which will be in place after this procedure has boon implomonted. However, the interim configuration (with some doloted snubbers removed and some still in place) has not been analyzed because the many possible combinations of Support / Restraint (S/R) configurations would require numerous analyses. For this reason, Design has determined this work may be done while the affected systom(s) are operable provided all the support modifications for the entiro math model are completed within the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed by the technical specification for snubbers. In order to avoid excooding tho 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> linit and declaring the affected systom(s) inoperable, Design Engineering has performed analyses that indicate the affected piping could be qualified under any combination of snubbers removed provided the modifications to the existing sup-ports have been completed. This procedure is written to complete all modifications before doloting any supports. However, if the support modifications cannot be completed in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> timo limit, Design Engineering will be contacted to perform an analysis of the affected piping to datormine operability. Design Engincoring has stated they could justify operable status of any configuration of modified sup-ports provided the affected piping is adoquately supported. Station maintenance proceduros provide guidance to ensure piping is adequately supported. This procedure will provide the necessary controls to ensure compliance with all technical specification requirements.
Based on this discussion, there are no unreviewod safety questions associated with the implomontation of this procedure.
TN/2/A/0396/01/03A Original NSM CN-20396, Rev. 1 modifies various piping system math models with the objective of reducing the number of mechanical snubbers required.
This work unit covers math model NDE. This math model includes supports in the Residual Heat Removal (ND), Containment Spray (NS) and 47
i Refueling Water (FW) systems. This proceduro provides guidanco for the removal of snubbers doloted from math model NDE as well as the modification of cortain other support / restraints.
These supports will either be doloted from the system or revised to a different configuration.
There are ng system isolations required to implomont this procedure.
The only concern is the seismic qualification of the affected systems' piping during implomontation of this procedure. The Math Model has boon qualified for the present support / restraint configuration. It has also boon qualified for the support /rostraint configuration which will bo in place after this proceduro has boon implemented. However, the interim configuration (with some doloted snubbers removed and some still in place) has not been analyzed because the many possible combinations of Support / Restraint (S/R) configurations would require numerous analyses. For this reason, Design has determined this work may be done while the affected system (s) are operable provided all the support modifications for the entiro math model are completed within the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed by the technical specification for snubbers. In order to avoid exceeding the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limit and declaring the affected system (s) inoperable, Design Engineering has performed analyses that indicate the affected piping could be qualified under any combination of snubbers removed provided the modifications to the existing sup-ports have boon completed. This procedure is written to complete all modifications before deleting any supports. however, if the support modifications cannot be completed in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> timo limit, Design Engineering will be contacted to perform an analysis of the affected piping to determino operability. Design Engineering has stated they could justify operablo status of any configuration of modified sup-ports provided the affected piping is adequately supported. Station maintenance procedures provide guidance to ensure piping is adequately supported. This procedure will provido the necessary controls to j onsure compliance with all technical specification requirements.
Based on this discussion, there are no unroviewed safety questions associated with the implomontation of this procedure.
TN/2/A/0396/01/04A Original NSM CN-20396, Rev. 1 modifies various piping system math models with the objective of reducing the number of mechanical snubbers required.
This work unit covers math models CAB and TAE. This math model includes supports in the Auxiliary Feodwater (CA), Main Feedwater Pump l Turbino Exhaust (TE) , Main Steam to Auxiliary Equipment (SA), and Main
! Feedwater (CF) systems. This procedure provides guidanco for the removal of snubbers deleted from math models CAE and TAE as well as the modification of certain other support / restraints.
Those supports will either be deleted from the system or revised to a different configuration.
48
There are no system isolations required to implement this procedure.
The only concern is the seismic qualification of the affected systems' piping during implementation of this procedure. The Math Model has been qualified for the present support / restraint configuration. It has also been qualified for the support / restraint configuration which will be in place after this procedure has been implemented. !kwever, the interim configuration (with some deleted snubbers removed and some still in place) has not been analyzed because the many possible combinations of Support / Restraint (S/R) configurations would require numerous analyses. For this reason, Design has datormined this work may be done while the affected system (n) are operable provided all the support modifications for the entire math model are completed within the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed by the technical specification for snubbers. In order to avoid exceeding tho 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limit and declaring the affected system (s) inoperable, Design Engineering has performed analyses that indicate the affected piping could be qualified under any combination of snubbers removed provided the modifications to the existing sup-ports have been completed. This procedure is written to complete all modifications before deleting any supports. However, if the support modifications cannot be completed in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit, Design Engineering will be contacted to perform an analysis of the affected piping to determine operability. Design Engineering has stated they could justify operable status of any configuration of modified sup-ports provided the affected piping is adequately supported. Station maintenance procedures provide guidance to ennure piping is adequately supported. This procedure will provide the necessary controls to ensure compliance with all technical specification requirements.
Eased on F.>is discussion, thero are no unreviewed safety questions associated with the implementation of this procedure.
TN/2/A/0396/01/05A Original NSM CN-20396, Rev. 1 modifies various piping system math models with the objectivo of reducing the number of mechanical snubbers required.
This work unit covers math model NDA. This math model includes supports in the ND system. This procedure provides guidance for the removal of snubbers deleted from math model NDA as well as the modifi-cation of certain other support /rostraints.
These supports will either be doloted from the system or revised to a different configuration.
There are D2 system isolations required to implement this procedure.
The only concern is the coismic qualification of the affected systems' piping during implementation of this procedure. The Math Model has been qualified for the present support / restraint configuration. It has also been qualified for the support / restraint configuration which will be in place after this procedure has been implemented. However, the interim configuration (with some deleted snubbers removed and some still in place) has not been analyzed because the many possible combinations of Support / Restraint (S/R) configurations would require
. numerous analyses. For this reason, Design has determined this work l
49 1
.~ _ ,- ., - - . , .
-.n --
I I
l may be done while the af fected system (s) are operable provided all the support modifications for the entiro math model are completed within the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed by the technical specification for snubbers. In ;
order to avoid exceeding the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limit and declaring the affected system (s) inoperable, Design Engineering has performed analysos that indicate the affected piping could be qualified under any combination of snubbers removed provided the modifications to the existing sup-ports have been completed. This procedure is written to complete all modifications beforo deleting any supports. However, if the support modifications cannot be completed in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit, Design Engineering will be contacted to perform an analysis of the affected piping to determine operability. Design Engineering has stated they could justify operable status of any configuration of modified sup-ports provided the affected piping is adequately supported. Station maintenance procedures provide guidance to ensure piping is adequately supported. This procedure will provide the necessary controls to ensure compliance with all technical specification requirements.
Based on this discussion, there are no unreviewed safety questions aesociated with the implementation of this procedure.
TN/2/A/0396/01/06A Original NSM CH-20396, Rev. 1 modifies various piping system math models with the objective of reducing the number of mechanical snubbers required.
This work unit covers math model NDP. This math model includes supports in the ND and NS systums. This proceduro provides guidance for the removal of snubbers deleted from math model NDP as well as the modification of certain other support / restraints.
These supports will either be deleted from the system or revised to a different configuration.
There are n2 system isolations required to implement this procedure.
The only concern is the seismic qualification of the affected systems' piping during implementation of thic procedure. The Math Model has been qualified for the present support / restraint configuration. It has also been qualified for the nupport/ restraint configuration which will be in place after this procedure has been implemented. However, the interim configuration (with some deleted snubbers removed and some still in place) has not been analyzed because the many possible combinations of Support /Rostraiht (S/R) configurations would require numerous analyses. For this reason, Design has determined this work may be done while the affected system (s) are operable provided all the support modifications for the entire math model are completed within the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed by the technicel specification for snubbers. In order to avoid exceeding the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limit and declaring the affected systea(s) inoperable, Design Engineering has performed analyses that indicate the affected piping could be qualified under any combination of snubbers removed provided the modifications to the existing sup-ports have been completed. Thi.s procedure is written to complete all modifications beforo deleting any supports. However, if the support modifications cannot be completed in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit, Design 50
l Engineering will be contacted to perform an analysis of the affected piping to determine operability. Design Engineering has stated they could justify operable status of any configuration of modified sup-ports provided the affected piping is adequately supported. Station ;
maintenance procedures provide guidance to ensure piping is adequately !
l supported. This procedure will provide the necessary controls to I
ensure complian.:o with all technical specification requirements.
Based on this djocussion, there are no unroviewed safety questions associated with the implementation of this procedure.
TN/2/A/0396/01/07A Original NSM CH-20396, Rev. 1 modifies various piping system math models with the objective of reducing the number of mechanical snubbers required.
This work unit covers math model SVA. This math model includes supports in the Main Steam Vent to Atmosphere system. This procedure provides guidance for the removal of snubbers deleted from math model SVA as well as the modification of certain other support / restraints.
These supports will either be deleted from the system or revised to a different configuration.
There are D2 system isolations required to implement this procedure.
The only concern is the seismic qualification of the affected systems' piping during implementation of this procedure. The Math Model has been qualified for the present support / restraint configuration. It has also been qualified for the support / restraint configuration which will be in place after this proceduro has been implemented. However, the interim configuration (with some deleted snubbers removed and some still ir, place) has not been analyzed because the many possible combinations of Support / Restraint (S/R) configurations would require numerous analyses. For this reason, Design has determined this work may be done while the affected system (s) are operable provided all the support modifications for the entire math model are completed within the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed by the technical specification for snubbers. In order to avoid exceeding the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limit and declaring the affected system (s) inoperable, Design Engineering has performed analyses that indicate the affected piping could be qualified under any combination of snubbers removed provided the modifications to the existing sup-ports have been completed. This procedure is written to complete all modifications before deleting any supports. However, if the support modifications cannot be completed in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time limit, Design Engineering will be contacted to perform an analysis of the affected piping to determine operability. Design Engineering has stated they could justify operable status of any configuration of modified sup-l ports provided the affected piping is adequately supported. Station l maintenance procedures provide guidance to ensure piping is adequately supported. This procedure will provide the necessary controls to ensure compliance with all technical specification requirements.
i Based on this discussion, there are no unreviewed cafety questions associated with the implementation of this procedure.
l 51
TN/2/A/0451/00/01A Original This procedure provides implementation guidelines for Nuclear Station Modification (HSM) CN-20451, Rev. O. This NSM will provide a manual bypass of the p-14 safety signal (Steam Generator Hi-Hi Level) to prevent nuisance feedwater isolations during Modes 4, 5, and 6. This procedure will provide guidelines for pulling cables, performing wiring changes, and installing switches on Control Board 2MC3. The only tie in that will be made in this procedure is in the Demultiplexer Cabinet, 2ELCC0011. A functional of the relay affected will be performed in this procedure. The functional and retest for the intent of the modification will be performed under procedures TN/2/A/0451/00/02A and TN/2/A/0451/00/03A.
No work will begin on this procedure until Unit 2 is in Modes 5, _6, or No Mode. No systems will be prevented from performing any function important to safety while this work is being performed. During the implementation of this procedure, as a result of the isolations, various indicating lights, status lights, annunciators, and events recorder points will be out of service. None of these indications are required to be operable during the modes that we are implementing the procedure. As a result of the isolations, the Refueling Water Storage Tank Heater groups A and B will bo inoperable. This will not be a problem since we will be implementing this procedure in June, July and August. No USQ exists.
TN/2/A/0451/00/02A Original This procedure provides implementation guidelines for NSM CN-20451, Rev. O. This NSM will provide a manual bypass of the P-14 safety signal (Steam Generator Hi-Hi Level) to prevent nuisance feedwater isolations during Modes 4, 5, and 6. This procedure will provide guidelines for complating the Train A wiring and functional testing of this modification.
No work will be performed on this procedure until Unit 2 is in Modes 5, 6, or No Mode. No aystems will be prevented from performing any function important to safety while this work is being performed.
During the implementation of this procedure, as a result of the isolations, the indicating lights for "EXRAUST FAN 2A1 (2A2) ON (AND OFF)" and "VF TRN A FILTERED (BYPASSED)" will be inoperable. The loss of this indication does not render Spent Fuel Pool Ventilation Train A inoperable. Operations will be notified of these indications being removed from service prior to the isolations being made.
This procedure also removes power from Unit 2 Solid State Protection System (SSPS) Train A. However, the SSPS is not required to be l
l operable in Modes 5, 6, or No Mode. The SSPS is required for Mode 4 l and will be returned to service prior to Unit 2 entering Mode 4.
l Thus, no USQ is involved with this procedure.
l l
TN/2/A/0451/00/03A Original 52
- This procedure provides implementation guidelines for NSM CN-20451, Rev. O. This NSM will provide a manual bypass of the P-14 safety signal (Steam Generator Hi-Hi . Level) to prevent nuisance feedwater isolations during Modes 4, 5, and 6. This procedure will provide guidelines for completing the Train B wiring and functional testing of this modification.
No work will begin on this procedure until Unit 2 is in Modes 5, 6, or No Mode. No system will be prevented from performing any function important to safety while this work is being performed.
During the implementation of steps 8.1.2 through 8.1.5, Operations will not have the ability to place Spent Fuel Pool Ventilation (VF)
Train B in the " BYPASS" mode. As a result, VF Train B will be inoper-able. This precedure will require VF Train A to be operable during the implementation of those steps.
This procedure also doenergizes Solid State Protection 6ystem (SSPS)
Train B. However, the SSPS is not required to be operable in Modes 5, 6, or No Mode. The SSPS will be returned to service prior to Unit 2 entering Mode 4. Based on tho above discussion, this procedure does not create a USQ.
TH/2/A/0566/00/01A Original This procedure provides instructions for implementation of Nuclear Station Modification (NSM) CN-20566, Rev. O Work Unit 01. This modification will replace Steam Generator 2D Blowdown (PB) Isolation valve 2BB008A with a new gate valve having item #06H-210. This is being performed due to numerous maintenance problems and marginally sized operators. Also, one additional cable and new overload heaters will bo installed. Upon completion of electrical work, a Leak Rate Test will be performed on 2 PENT 0115 to ensure its pressure boundary is maintained. Testing of the new valve will consist of performing Motor Operated Valve (MOV) testing and verifying all remote position and status light indications. Stroke time tests will be performed prior to Mode 4 and again in Mode 3. A differential pressure test will also be performed on valve 2BB008A. Hydrostatic and appropriate Non-De-structive Examination (NDE) tests will be performed by station and Quality Assurance (QA) procedures.
The flow path from Steam Generator 2D will be out of service during the replacement of valve 2BB008A. The containment isolation valves downstream of 2BB008A will be utilized to satisfy. Tech. Spec. require-ments for control of penetrations which have direct acccc: to outside atmosphere for containment integrity / closure during core alterations, fuel movement, and Reactor Coolant (NC) System mid-loop conditions.
Additionally, the work associated with 2 PENT 0115 will not break the pressure boundary of the penetration. Breakers F04B in 2EMXS and F04B in 2EMXK will be opened and Red Tagged to ensure valve 2BB008A, and plug #7 in 2 PENT 0115 are de-energized for electrical work. Breaker F04B in 2EMXK also supplies power to the Electric Motor Operator '(EMO) for containment isolation valve 2NM190A, and Operations has 53
. ~
responsibility for this valvo to ensure containment integrity / closure.
These isolatione do not present a concern for the safo operation of the BB system because the DB system is not required operablo in modes 5, 6, or No Modo. Rod Tags will be removed and breakers F04B in 2EMXS and F04B in 2EMXK will be closed for valvo not up, verification of remoto position indication, MOV, and stroke time topting. Thus, no USQ is created by this procedure.
TN/2/A/0566/00/02A Original This procedure provides instructions for implementation of NSM CN-20566, Rev. O Work Unit 02. This modification will replace Steam Generator 2B blowdown Isolation valve 2BB019A with a new gate valve having item #06H-210. Thio is being performed due to numerous mainte-nance problems and marginally sized operators. Also, one additional cable and now overload heators will be installed. Upon completion of electrical work, a Loak Rate Tout will be performed on 2 PENT 0115 to ensure its pressure boundary is maintained. Testing of the new valve will consist of performing MOV testing and verifying all remote position and status light indications. Stroke time tests will be performed prior to Mode 4 and again in Mode 3. A differential pres-sure test will also be performed on Ysivo 2BB019A. Hydrostatic and appropriate NDE examinations will be performed by station and QA procedures.
The flow path from Steam Generator 2B will bo out of service during the replacement of valve 2BB019A. The containment isolation valves downstream of 2BB019A will be utilized to satisfy Tech. Spec. require-monts for control of penetrations which have direct access to outsido atmosphere for containment integrity / closure during core alterations, fuel movement, and NC System mid-loop conditions. Additionally, the work associated with 2 PENT 0115 will not break the pressure boundary of the penetration. Breakers F04C in 2EMXS and F04B in 2EMXK will be opened and Red Tagged to ensure valvo 2BB019A, and plug #7 in 2 PENT 0115 are do-energized for electrical work. Breaker F04B in 2EMXK also supplies power to the EMO for containment isolation valve 2NM190A, and Operations has responsibility for this valve to ensure containment integrity / closure. These isolations do not present a concern for the safe operation of the BB system because the BB system is not required operable in modos $, 6, or No Modo. Red Tags will be removed and breakers F04C in 2EMXS and F04B in 2EMXK will be closed for valve set up, verification of remoto position indication, MOV, and stroke time testing. Thus, no USQ is creatc6 by this procedure.
TN/2/A/0566/00/03A original This procedure provides instructions for implementation of NSM CN-20566, Rev. O Work Unit 03. This modification will replace Steam Gonorator 2A blowdown Isolation valvo 2BB056A with a new gate valve having item #06H-210. This is being performed due to numerous mainte-nance problems and marginally sized operators. Also, one additional cabic and new overload heaters will bo installed. Upon completion of 54
electrical work, a Leak Rate Test will be performed on 2 PENT 0115 to ensure its pressure boundary is maintained. Testing of the new valve will consist of performing MOV testing and verifying all remote position and status light indications. Stroke time tests will be performed prior to Mode 4 and again in Mode 3. A differential pres-sure test will also be performed on valve 2BB056A. Hydrostatic and appropriate NDE examinations will be performed by station and QA procedures.
The flow path from steam Generator 2A will be out of service during the replacement of valve 2BB056A. The containment isolation valves downstream of 2BB056A will be utilized to satisfy Tech. Spec. require-ments for control of penetrations which have direct access to outside atmosphere for containment integrity / closure during core alterations, fuel movement, and NC System mid-loop conditions. Additionally, the work associated with 2 PENT 0115 will not break the pressure boundary of the penetration. Breakers FOSA in 2EMXS and F04B in 2EMXK will be opened and Red Tagged to ensure valve 2BB056A, and plug #7 in 2 PENT 0115 are de-energized for electrical work. Breaker F04B in 2EMXK .
also supplies power to the EMO for containment isolation valve l 2NM190A, and Operations has responsibility for this valve to ensure ,
containment integrity / closure. These isolations do not present a l concern for the safe operation of the BB system because the BB system (
is not required operable in modes 5, 6, or No Mode. Red Tags will be i removed and breakers F05A in 2EMXS and F04B in 2EMXK will be closed for valve set up, verification of remote position indication, MOV, and stroke time testing. Thus, no USQ is created by this procedure.
TN/2/A/0566/00/04A original This procedure provides instructions for implementation of NSM CN-20566, Rev. O Work Unit 04. This modification will replace Steam Generator 2C blowdown Isolation valve 2BB060A with a new gate valve having item /06H-210. This is being performed due to numerous mainte-nance problems and marginally sized operatorn. Also, one additional cable and new overload heaters will be installed. Upon completion of electrical work, a Leak Rate Test will be performed on 2 PENT 0115 to ensure its pressure boundary is maintained. Testing of the new valve will consist of performing MOV testing and verifying all remote position and status light indications. Stroke time tests will be performed prior to Mode 4 and again in Mode 3. A differential pres-sure test will also be performed on valve 2BB060A. Hydrostatic and appropriate NDE examinations will be performed by station and QA procedures.
The flow path from Steam Generator 2C will be out of service during l
the replacement of valve 2BB060A. The containment isolation valves l downstream of 2BB060A will be utilize 6 to satisfy Tech. Spec. require-ments for control of penetrations which have direct access to outside atmosphere for containment integrity / closure during core alterations, fuel movement, and NC System mid-loop conditions. Additionally, the work associated with 2 PENT 0115 will not break-the pressure boundary of the penetration. Breakers F05B in 2EMXS and F04B in 2EMXK will be 55
, l opened and Red Tagged to ensure valve 2BB060A, and plug #7 in 2 PENT 0115 are de-energized for electrical work. Breaker F04B in 2EMXK also supplies power to the EMO for containment isolation valve 2NM190A, and Operations has responsibility for this valve to ensure i containment integrity / closure. These isolations do not present a concern for the safe operation of the BB system because the BB system ;
is not required operable in modes 5, 6, or No Mode. Red Tags will be removed and breakers F05B in 2EMXS and F04B in 2EMXK will oe closed for valve set up, verification of remote position indication, MOV, and stroke time testing. Thus, no USQ is created by this procedure.
TN/2/A/0567/00/01A Original This procedure provides implementation guidelines for Nuclear Station Modification (NSM) CN-20567, Rev. O, Work Unit 01. This modification will replace Auxiliary Feedwater (CA) Discharge Isolation Valve 2CA066B with a new gate valve, item # 06J-339. This is being done to assure that there is sufficient torque available to meet design requirements. Also, new valve 2CA288, item #06J-500, is being added between the bonnet for valve 2CA066B and the upstream process-piping.
This will prevent bonnet overpressurization on 2CA066B. This proce-dure provides guidance for the removal / reinstallation of valve 2CA066B, and the installation of new valve 2CA288. Testing of the new valve will consist of performing Motor Operated Valve (MOV) testing and verifying all remote position indications. Stroke time tests will be performed prior to Mode 3, and again in Mode 3. A differential pressure test will not be performed on valve 2CA066B due to it being l one of the discharge isolation valves for the turbine driven auxiliary feedwater pump, and sufficient differential pressure cannot be ob-tained. Hydrostatic and appropriate Non-Destructive Examination (NDE) tests will be performed by station and Quality Assurance (QA) proce-dure s .-
A blank is being installed in 2 CAFE 5090, by procedure-TN/2/A/0567/00/02A, to ensure that the Auxiliary Feedwater flow path to Steam Generator 2A is isolated-during the replacement of valve 2CA66B. This blank will also be utilized to maintain containment integrity / closure during core alterations, fuel movement, and Reactor Coolant (NC) System mid-loop conditions. Breaker F08A in 2EMXL will be opened and Rcd Tagged under Block Tagout CAlVLV to ensure valve 2CA66B-is de-energized for electrical work. This does not present a concern for safe operation of the CA system because the CA system is l - not required to be operable'in modes 4, 5, 6, or No Mode. The Red Tag j for breaker F08A will be lifted from block tagout'CA1VLV for valve set up, verification of remote position indication, MOV, and stroke time testing. Thus, no USQ is created by this procedure.
TN/2/A/0567/00/02A Original This procedure provides. implementation guidelines for NSM CN-20567, Rev. O, Work Unit 02. This modification Will replace Auxiliary -
Feedwater Discharge Isolation Valve 2CA062A with a'new gate valve, -
item / 06J-339. This is being done to assure that there is sufficient torque available to mest design requirements. Also, new-valve 2CA287, j 56
,r,,ew,,<v w+~. ,
y e w , yr- - - + ,,,, g gw v- s-, ,g,-+,-wm,. s. --,m , s e ,m - s e , y m e --m ,+ ,_,ww--gw--r
l 1
item #06J-500, is being added between the bonnet for valve 2CA062A and the upstream process piping. This will prevent bonnet '
overpressurization on 2CA062A. This procedure provides guidance for the removal /reinst'.11ation of valve 2CA062A, and the installation of i new valve 2CA287. Testing of the new valve will consist of performing i MOV testing and verifying all remote position indications. Stroke time tests will be performed prior to Mode 3, and again in Mode 3. A differential pressure test will also be performed on valve 2CA062A. ,
Hydrostatic and appropriate NDE tests will be performed by station and j QA procedures. ,
+
A blank is being installed in 2 CAFE 5090 by this procedure to ensure that the Auxiliary Feedwater flow path to Steam Generator 2A is isolated during the replacement of valve 2CA62A. This blank will also be utilized to maintain containment integrity / closure during core alterations, fuel movement, and NC System mid-loop conditions.
Breaker FOBA in 2EMXK will be opened and Red Tagged under Block Tagout CAlVLV to ensure valve 2CA62A is de-energized for electrical work.
This does not present a concern for safe operation of the CA system ,
because the CA system is not required to be operable in modes 4, 5, 6, '
or No Mode. The Red Tag for breaker FOBA will be lifted from block tagout CA1VLV for valve set up, verification of remote position indication, MOV, and stroke time testing. Additionally, an Out of Service sticker will be placed on control room indicator 2 cap 5090 to ensure awareness of out of service instruments.
Based upon the above discussion, a USQ is not created by this proce-dure.
TN/2/A/0567/00/03A priginal This procedure provides implementation guidelines for NSM CN-20567, Rev. O, Work Unit 03. This modification will replace Auxiliary Feedwater Discharge Isolation Valve 2CA058A with a new' gate valve, item / 06J-339. This is being done to assure that there is sufficient torque available to meet design requirements. Also, new valve 2CA286, item #06J-500, is being added between the bonnet for valve 2CA058A and the upstream process piping. .This will prevent bonnet overpressurization on 2CA05BA.- This procedure provides guidance for the removal / reinstallation of valve 2CA058A, and the installation of new valve 2CA286._ Testing of the new valve will consist of performing MOV testing and verifying all remote position indications. Stroke time tests will be performed prior to Mode 3, and again in Mode 3. A differential pressure test will also be performed on valve 2CA058A.
Hydrostatic and appropriate NDE tests will be performed by station and QA procedures.
A blank is being installed in 2 CAFE 5100 by this procedure to ensure that the Auxiliary Feedwater flow path to Steam Generator 2B is isolated during the replacement of valve 2CA58A. This blank will'also l
be utilized to maintain containment integrity / closure during core l alterations, fuel movement, and NC System.mid-loop conditions.
Breaker R03C in 2EMXI will be opened and Red Tagged under Block Tagout
-57
i CA2VLV to ensure valve 2CA58A is de-energized for electrical work.
This does not present a concern for safe operation of the CA system i
because the CA system is not required to be operable in modes 4, 5, 6, or No Mode. The Red Tag for breaker R03C will be lifted from block tegout CA2VLV for valve set up, verification of remote position <
l indication, MOV, and stroke time testing. Additionally, an out of Service sticker will be placed on control room indicator 2 CAP 5100 to ensure awareness of out of service instruments.
Based upon the above discussion, a USQ is not created by this proce-dure.
TN/2/A/0567/00/04A original This procedure provides implementation guidelines for NSM CN-20567, Rev. O, Work Unit 04. This modification will replace Auxiliary Feedwater Diccharge Isolation Valvo 2CA054B with a new gate vaF 1, item # 06J-339. This is being done to assure that there is sufricient torque available to meet design requiremaints. Also, new valve 2CA285, item #06J-500, is being added between the bonnet for valve 2CA054B and the upstream procesa piping. This will prevent bonnet overpressurization on 2CA054B. This procedure provides guidance for the removal / reinstallation of valve 2CA0540, and the installation of new valve 2CA285. Testing of the new valve will consist of performing MOV testing and verifying all remote position indications. Stroke time tests will be performed prior to Mode 3, and again in Mode 3. A differential pressure test will not be be performed on valve 2CA054B due to it being one of the discharge isolation valves for the turbine driven auxiliary feedwater pump, and sufficient differential pressure cannot be obtained. Hydrostatic and appropriate NDE tests will be performed by station and QA procedures.
A blank is being installed in 2 CAFE 5100, by procedure TN/2/A/OS67/00/03A, to ensure that the Auxiliary Feedwater flow path to Steam Generator 2B is isolated during the replacement of valve 2CA548. This blank will also be utilized to maintain containment integrity / closure during core alterations, fuel movement, and NC System mid-loop conditions. Breaker R03C in 2EMXB will be opened and Red Tagged under Block Tagout CA2VLV to ensure valve 2CAS4B is de-en-orgized for electrical work.- This does not present a concern for safe operation of the CA system because_the CA system is not_ required to be operable in modes 4, 5, 6, or No Mode. The Red Tag for breaker R03C will be lifted from block tagout CA2VLV for valve set up, verification of remote position indication, MOV, and stroke time testing.
Based on the above, this procedure does not involve a USQ.
TN/2/A/0567/00/05A Original This procedure provides-implementation guid lines for NSM CN-20567, Rev. O, Work Unit 05. .This-modification will replace Auxiliary Feedwater Discharge Isolation Valve 2CA050A with a new gate valve, item # 06J-339. This is being'done to assule that there is sufficient 58
i torque available to meet design requirements. Also, new valve 2CA284, item #06J-500, is being added between the bonnet for valve 2CA050A and the upstream process piping. This will prevent bonnet overpressurization on 2CA050A. This procedure provides guidance for the removal / reinstallation of valve 2CA050A, and the installation of new valve 2CA284. Testing of the new valve will consist of performing MOV testing and verifying all remote position indications. Stroke time tests will be performed prior to Mode 3, and again in Mode 3. A differential pressure test will not be be performed on valve 2CA050A due to it being one of the discharge isolation valves for the turbine driven auxiliary feedwater pump, and sufficient differential pressure cannot be obtained. Hydrostatic and appropriate NDE tests will be performed by station and QA procedures.
A blank is being installed in 2 CAFE 5110 by this procedure to ensure that the Auxiliary Feedwater flow path to Steam Generator 2C la isolated during the replacement of valvo 2CA50A. This blank will also be utilized to maintain containment integrity / closure during core alterations, fuel movement, and NC System mid-loop conditions.
Breaker F04A in 2EMXS will be opened and Red Tagged under Block Tagout CA3VLV to ensure valve 2CA50A is de-energized for electrical work.
This does not present a concern ror safe operation of the CA system because the CA system is not required to be operable in modes 4, 5, 6, or No Mode. The Red Tag for breaker F04A will be lifted from block tagout CA3VLV for valvu set up, verification of remote position indication, MOV, and stroke time testing. Additionally, an Out of Service sticker will be placed on control room indicator 2 CAP 5110 to ensure awareness of out of service instruments.
Based on the above, this procedure does not involve a USQ.
TN/2/A/0567/00/06A original This procedure provides implementation guidelines for NSM CN-20567, Rev. O, Work Unit 06. This modification will replace Auxiliary Feodwater Discharge Isolation Valve 2CA046B with a new gate valve, item / 06J-339. This is being done to assure that there is sufficient torque available to meet design requirements. Also, new valve 2CA283, item #06J-500, is being added between the bonnet for valve 2CA046B and the upstream process piping. This will prevent bonnet overpressurization on 2CA046B. This procedure provides guidance for the removal /roinstallation of valvo 2CA046B, and the installation of new valve 2CA283. Testing of the new valve will consist of performing MOV testing and verifying all remote position indications. Otroke time tests will be performed prior to Mode 3, and again in Mode 3. A differential pressure test will be performed on valve 2CA046B.
Hydrostatic and appropriate NDE tests will be performed by station and QA procedures.
A blank is being installed in 2 CAFE 5110, by procedure
, TN/2/A/0567/00/05A, to ensure that the Auxiliary Feedwater flow path to Steam Generator 2C is isolated during the replacement of valve 2CA46B. This blank will also be utilized to maintain containment 59
integrity / closure during core alterations, fuel movement, and NC System mid-loop conditions. Breaker R03A in 2EMXB will be opened and Red Tagged under Block Tagout CA3VLV to ensure valve 2CA46B is de-on-orgized for electrical work. This does not present a concern for safe operation of the CA system because the CA system is not required to be operable in modos 4, 5, 6, or No Mode. The Red Tag for breaker R03A will be lifted from block tagout CA3VLV for valve set up, verification of remote position indication, MOV, and stroke time testing.
Based on the above, this proceduro does not create a USQ.
TH/2/A/0567/00/07A original This procedure provides impicmontation guidelines for NSM CN-20567, Rev. O, Work Unit 07. This modification will replace Auxiliary Feedwater Discharge Isolation Valvo 2CA042B with a now gate valve, item # 06J-339. This is being done to assure that there is sufficient torque available to meet design requirements. Also, new valve 2CA282, item #06J-500, is being added betwoon the bonnet for valvo 2CA042B and the upstream process piping. This will provent bonnet overprossurization on 2CA0428. This proceduro provides guidance for the removal / reinstallation of valvo 2CA042B, and the installation of new valve 2CA282. Testing of the new valve will consist of performing MOV testing and verifying all remote position indications. Stroke timo tests will be performed prior to Mode 3, and again in Mode 3. A differential pressure test will be performed on valve 2CA0428.
Hydrostatic and appropriate NDE tests will be performed by station and QA procedures.
A blank is being installed in 2 CAFE 5120, by procedure TN/2/A/0567/00/08A, to ensure that the Auxiliary Feedwater flow path to Steam Generator 2D is isolated during the replacement of valve 2CA42B. This blank will also be utilized to maintain containment integrity / closure during core alterations, fuel movement, and NC System mid-loop conditions. Breaker F08B in 2EMXL will be opened and Red Tagged under Block Tagout CA4VLV to ensure valve 2CA42B is de-en-orgized for electrical work. This does not present a concern for safe operation of the CA system because the CA system is not required to be operable in modes 4, 5, 6, or No Mode. The Rod Tag for breaker F08B will be lifted from block tagout CA4VLV for valve set up, verification of remote position indication, MOV, and stroke time testing.
Based on the above, this procedure does not create a USQ.
TN/2/A/0567/00/08A original This procedure provides implementation guidelines for NSM CN-20567, Rev. O, Work Unit 08. This modification will replace Auxiliary Feedwater Discharge Isolation Valve 2CA038A with a new gate valve, i
item / 06J-339. This is being done to assure that there is sufficient torque available to meet design requirements. Also, new valve 2CA281, item #06J-500, is being added betwoon the bonnet for valve 2CA038A and 60
l the upstream process piping. This will prevent bonnet overpressurization on 2CA038A. This procedure provides guidance for the removal /roinstallation of valve 2CA038A, and the installation of new valve 2CA281. Testing of the new valve will consist of performing MOV testing and verifying all remote position indications. Stroke l time tests will be performed prior to Mode 3, and again in Mode 3. A differential pressure test will not be performed on valve 2CA038A, due to it being one of the diucharge isolation valves for the turbine driven auxiliary feedwater pump and sufficient differential pressure cannot be obtained. Hydrostatic and appropriate NDE tests will be performed by station and QA procedures.
A blank is being installed in 2 CAFE 5120 by this procedure to ensure that the Auxiliary Feedwater flow path to Steam Generator 2D is isolated during the replacement of valve 2CA38A. This blank will also be utilized to maintain containment integrity / closure during core alterations, fuel ~~vement, and NC System mid-loop conditions.
Breaker F08B in 22 = will be opened and Red Tagged under Block Tagout CA4VLV to ensure yh.ve 2CA38A is de-energized for electrical work.
This does not present a concern for safe operation of the CA system because the CA system is not required to be operable in modes 4, 5, 6, or No Mode. The Red Tag for breaker F08B will be lifted from block tagout CA4VLV for valve set up, verification of remote position indication, MOV, and stroke time testing. Additionally, an Out of Service sticker will be placed on control room indicator 2 CAP 5120 to ensure awareness of out of service instruments.
Based on the above, this a USQ je not created by this procedure.
i
! TN/2/A/0572/00/01A original Nuclear Station Modification (NSM) CN-20572, Rev. O relocates flow restricting orifices 2CFFE6330 and 2CFFE6340, installed in the main feedwater (CF) loops 2C and 2D respectively, from the horizontal to 1 l
the vertical position. Single hole split orifices will be installed at these locations to reduce the potential for fouling while providing the necessary pressure differential to divert flow to the Auxiliary Feedwater (CA) nozzles. Additionally, check valve 2CF168 will be relocated to reduce flow turbulence, and larger diameter piping will be installed downstream of 2CF96 for stress analysis reasons. This procedure provides guidance for mechanical activities which will be performed during the Unit 2 End of Cycle 3 refueling outage. This l
procedure will modify the CF Loop 2D piping to relocate flow orifice 2CFFE6340 to the vertical position and install the single hole split orifice plate. After completion of the modification work, the piping will be inspected for leaks at system temperature and pressure to verify piping integrity. CA and CF flow and pressure parameters will be verified to ensure the orifices provide the necessary pressure differential to divert the required flow to the CA nozzle, and.CF Loop
- D pressure is consistent with CF loops A, B, and C. Also, the check valves downstream of flow orifice 2CFFE6340 (2CF058 and 2CF169) will be acoustically monitored to ensure flow through the orifice does not induce fluttering of the check valvo disks.
61
_ _. _ _ .___ _ _ _.._ _ __ _ ___.__ _ _ _ _m.__ ___ _ _ _ _ _ .
CF Loop 2D must be isolated and drained for this work. All modifica-tion work under this procedure will be performed with Unit 2 in Mode 5 1 or below. CF Loop 2D is not required to be in service during this i period. Also, all modification work is outside of containment, so containment integrity is not a factor. All piping modifications will be complete prior to mode 4 of Unit 2 startup. This will allow CF Loop 2D to be filled to support power escalation.
All post-modification testing will be done in Mode 1 with CF in normal alignments. No USQ is created by this procedure.
TN/2/A/0572/00/02A Original NSM CN-20572, Rev. O relocates flow restricting orifices 2CFFE6330 and 2CFFE6340, installed in the main feedwater (CF) loops 2C and 2D respectively, from the horizontal to the vertical position. Single i hole split orifices will be installed at these locations to reduce the potential for fouling while providing the necessary pressure differen-tial to divert flow to the Auxiliary Feedwater (CA) nozzles. Addi-tionally, check valve 2CP168 will be relocated to reduce flow turbu-lence, and larger diameter piping will be installed downstream of 2CF96 for stress analysis reasons. This procedure provides guidance for mechanical activities which will be performed during the Unit 2 End of Cycle 3 refueling outage. - This procedure will modify the CF Loop 2C piping to relocate flow orifice 2CFFE6330 to the vertical position and install the single hole split orifice plate. Also, check valve 2CF168 will be relocated under this procedure to further down-stream in the piping run to reduce flow turbulence.- After completion of the modification work, the piping will be inspected for leaks at system temperature and pressure to verify piping integrity. CA and CF flow and pressure paramotors will be verified to ensure the orifices provide the necessary pressure differential to divert the required flow to the CA nozzle, and CF Loop C pressure is consistent with CP loops A, B, and D. Also, the check valves downstream of flow orifice 2CFFE6330 (2CF049 and 2CF168) will be acoustically monitored to ensure flow through the orifice does not induce fluttering of the check valve disks.
CF Loop 2C must be isolated and drained for this work. All-modifica-tion work on this procedure will be performed with Unit 2 in Mode 5 or below. CF Loop 2C is not required to be in service during this period. Also, all modification work is outside of containment, so containment integrity is not a factor. All piping modifications will-l be complete prior to mode 4 of Unit 2 startup. This will allow CF Loop 2C to be filled to support power escalation..
1 Certain welds within the modification boundary require hydrostatic testing in accordance with ASME Section XI. All hydrostatic testing will be performed under TN/2/A/0572/00/03A and completed prior to Mode 4.
All post-modification testing.will be done in. Mode 1 with CF in norma) l alignments. No USQ is created by this procedure. 1 62 a
l TN/2/A/0572/00/03A Original NSM CN-20572, Rev. O relocates flow restricting orifices 2CFFE6330 and 2CFFE6340, installed in the main feedwater (CF) loops 2C and 2D respectively, from the horizontal to the vertical position. Single hole split orifices will be installed at these locations to reduce the potential for fouling while providing the necessary pressure differen-tial to divert flow to the Auxiliary Feedwater (CA) nozzles. Addi-tionally, check valve 2CF168 will be relocated to reduce flow turbu- '
lence, and larger diameter piping will be installed downstream of 2CF96 for stress analysis reasons. This procedure provides guidance for mechanical activities which will be performed during the Unit 2 End of cycle 3 refueling outage. This procedure will install larger diameter piping downstream of 2CF96 and perform the required hydrostatic tests for the modification work performed on CF loop 20.
After completion of the modification work, the piping will be inspect-ed for leaks at system temperature and pressure to verify piping integrity. Chemical Addition System to-CF flow through 2CF96 Will be verified to ensure no obstruction exists in the piping run.
In order to perform this work, Unit 2 must be in No Mode with CF Loop 2C isolated and drained. This modification 580rk is within the con-l tainment closure boundary. Howaver, since Unit 2 is required to be in No Mode, containment closure is not required, and CF Loop 20 is not required to be in service. This procedure requires the piping modifi-cations be conpleted prior to Mode 6. This will-allow containment closure requirements to be met as required by Tech. Specs.
This procedure requires all hydrostatic tests be complete prior _to Mode 4. This will allow CF Loop 2C to be returned to service to support Unit 2 power escalation. The remaining post-modification testing will be performed in Mode 1 using normal CF system alignments.
No USQ is created by this procedure.
63
Catawba Nuclear Station Summary of Procedure Changes, Tests, and Experiments l
Completed from 11/1/89 to 9/30/90 -- Volume 4 TN/2/A/0579/00/01A Original This procedure provides implementation instructions for Nuclear Station Modification (NSM) CN-20579, Rev. O. Thin NSM replaces the r
BIF butterfly vu3vns currently installed for 2RN232A and 2RN292B.
These valves have not proven to be reliable for isolation. These valves serve as the Diesel Generator Engine Jacket Water Cooler (KD HX) Nuclear Service Water (RN) inlet isolation valves. Seat leakage past these valves affects the ability to keep the lube oil warm for fluidity during engine standby. Posi Seal valves, which are more reliable and easier to maintain, will be installed at these applica-tions. This procedure provides guidelines for replacing 2RN232A with valve item number 05B-477 and repositioning the Limitorque operator for correct operation.
Replacing 2RN232A requires Diesel Generator (D/G) 2A to be taken out of service. The redundant D/G will be available to supply emergency power. Shared valves normally powered from the 2A D/G are provided with a manual switchover to the ethnr unit's diesel of the corre-sponding channel. Therefore, any one D/G can be down for maintenance and the RN system can still shut the plant down. The loss of one train will not prevent safety systems from performing their-intended functions. Based on these considerations, no unreviewed safety questions (USQ) are associated with implementation of this procedure.
TN/2/A/0579/00/02A Original This procedure provides implementation instructions for NSM CN-20579, Rev. O. This NSM replaces the BIF butterfly valves currently in-stalled for 2RN232A and . 2RN292B. These valves have not proven to be reliable for isolation. These valves serve as the Diesel Generator Engine Jacket Water Cooler (KD HX) RN inlet isolation valves. Seat leakage past these valves affects the ability to keep the lube oil warm for fluidity during engine standby. Posi Seal valves, which are more reliable and easier to maintain, will be installed at these applications. This procedure provides guidelines for replacing 2RN292B with valve item number 05B-477 and repositioning the Limitorque operator for correct operation.
2 Replacing 2RN232A requires Diesel Generator (D/G) 2A to bJ taken out of service. The redundant D/G will be available to supply emergency power.- Shared valves normally powered from the 2A D/G are provided with a manual switchover to the other unit's dissel of the corre-sponding channel. Therefore, any one D/c can be down for maintenance and the RN. system can still shut the plant down. The loss of one train will not prevent safety systems from performing their-intended 1-
- ~ - - _ _ , - _
l I
l functions. Based on these considerations, no unreviewed safety questions (USQ) are associated with impleinentation of this procedure.
TN/2/A/0594/00/01A Original This procedure provides instructions for deletion of the Unit 2 Annulus Ventilation (VE) System Train A lower ring header duct work under NSM CN-20594, Rev.0. The removal of this duct is necessary to provide access to inspect and repair / maintain the containment vessel.
A Post-Modification Test (for negative pressure distribution) is required to verify that the duct in the ring header may be permanent-ly deleted. This test may be performed as soon as the lower ring header is disconnected from the duct work above by the removal of two short vertical pieces of duct.
The test can be most offectively performed while containment is still sealed at the beginning of the Unit 2 End of Cycle 3 Outage. This means this ductwork should be removed as the unit goes from mode 1 to mode 4.
Design Engineering has documented that the remaining ductwork will be seismically adequate. VE train A will remain operable during and after the lower ring header deletion. Since the function of the VE system will not be affected by the work on the duct, no USQ is creat-ed by implementation of this procedure.
TN/2/A/2039/CE/01A Re-type #1 This procedure provides for imple.montation on Exempt Change CE-2039 Work Unit 01. CE-2039 provides an access hole in the main steam (SM) system piping at steam Generator (S/G) 2A. "his hole will be used to access the inside of the-piping in order
- arform a radiographic examination of the weld between the piping 1.id the steam generator.
This inspection is being done in order to comply with the ASME Code Section XI Inservice Inspection requirements. After the examination I is complete, the hole will be plugged and seal welded.
l The purpose of this procedure is to provide guidance for drilling the hole and installing a half coupling and plug. Testing for CE-2039 will be performed in accordance with the Post Modification Testing program at the station. Since the hole is one inch in diameter, a i pressure test is not required per ASME Section XI exemptions; howe \-
er, a visual inspection for 3eaks will be performed in Mode 3 with the affected main steam piping at normal system temperature and pressure. This inspection will not affect the function or operation of the main steam system.
The Operations group will coordinate the isolations necessary to implement this procedure. S/G 2A and its associated SM line will be out of service during this procedure. This drilling of the hole and installation of the plug will be performed in Modes 6, 1, and No mode. The Operations group will be notified to ensure containment 2
I
integrity is maintained while the affected SM line is open to con-tainment. Dased on this discussion, no USQ is involved with the implementation of this procedure.
TN/2/B/0073/01/01A Original I This procedure provides guidelines for installing portions of the Post- Accident Liquid Sampling (PALS) panel piping used to return l samples taken from the Reactor Coolant (NC) and Residual Heat Removal (ND) System after an accident to the Volume control Tank (VCT). This i system will consist of a positive displacement sump pump and the associated piping and valves. These activities are part of NSM CN-20073, Rev. 1. This specific procedure provides guidance for mechanical activities which will be performed during the Unit 2 End of Cycle 3 refueling outage. It will install portions of the PALS return piping, along with a manual isolation valves and two check valvas and hangers, and makes the required tie-ine to the VCT sample recirculation piping. After installation, the piping and valves will be visually inspected at system temperature and pressure t- rerify piping <ntegrity.
In order to perform this procedure, the VCT, along with the VCT sample recirculation piping, must be isolated and drained. This procedure will be perforved during No mode while the NC system is out of service and the VCT is not required. After the completion of this procedure, the VCT sample recirculation piping pressure boundary will be restored and the VCT can be returned to service. Based on the above, no USQ is involved with the implementation of thia procedure.
MP/0/A/7600/10 Retype, Changes 0 to 7 Incorporated This procedure provide thod of disassembly, reassembly, and correctire maintenance 1. 11 Westinghouse Swing check Valves.
Inntruction Manual CNM-1105.00-322 provided the technical information used in this procedure. Specific information relating to Catawba
' Nuclear Station is also included in this procedure. The following is a summary of the changes made to this procedure as a result of the procedure upgrade process. A step was added to instruct the mechanic to contact Maintenance Engineering Services (MES) if any check valve part does not pass inspection. Due to the heightened awareness of check failures in the nuclear industry, it is necessary for the responsible check valve engineer to stay abreast of any wear and/or degradation of critical cneck valve parts. Steps were added to the procedure giving guidance to the mechanic on how to properly de-ten-l sion the bonnet studs / nuts, as stud to body galling has been a repet-itive problem for this valve design. A step was added to verify the integrity of the disc anti-rotation pin, as failure of this pin could result in disc spinning, and subsequent disc stud fatigue / failure.
l These additions will decrease the chance of equipment failure and
- unnecessary maintenance, and consequently, increase equipment reli-ability. For this reason, the probability of an accident or malfunc-tion will not be increased. No USQ is created.
3
MP/0/A/7150/67 Retype, Changes 0 to 11 Incorporated This procedure is for performing latching and unlatching operations on the Control Rod Drive Mechanism drive shafts. Instruction Manuals CNM-12C1.00-0030 and CNM-1201.13-039 provided technical information for the development of this procedure. Plant specific information was al o included in the development of this procedure. This retype incl Wew the following changes. Step 11.7.1 was added to provide sign C:f for attaching safety sling to tool. In step 11.11.3, tool operating pressure was revised to 110-125 psig, based on actual operating experience. Sign offs for elevation marks were added to Step 11.12 and 31.14.3. Section 11.13 was revised for clarification.
The sequence % to lock drive shafts after unlatching was removed, per Wcstinghouse technical justification to Design Engineering. Section 11.14 wae revised sur clarification. The sequences performed if the drive sbtit was in tha Aciked position prior to latching were re-moved. bactions 11.15 anu 11.16 were revised for clarification of 9 Jence steps. No technical information changed. Enclosures 13.1 13.2 were revised per changes made in the procedure body. In
. 4ure 13.1, Section H was deleted and moved to Section F. Enclo-
- m. '3.4 was rovisad to add building and vessel location orienta-
' on- No USQ was deemed to exist.
M1 / A/7150/ 4, Ret /pe, Changes O to 9 Incorporated Thia procedure is for performing removal and replacement activities required on the Reactor Vessel Head during refueling outages. The following changes were made during the procedure review following the Unit. 1 End of Cycle 4 refueling outage. These changes were required to clarify steps for better understanding and to add manufacturer's information on new equipment.
Step Chanqc 2.1.6 Added CNM-1144.28-0037-001, Instruction Manual i fcr L. V. Nozzle Inspection Hatch Covers Changed eyebolt size to 1-1/4 "
1.1.5 .
Changed tubing size to 3/8 "
1.3.17 1.3.1J Changed rope size to 3/8 ".
9.1 Revised elongation readings to .002 +.006 per CE VN-2532.
101 3 Added note for head shielding installation (moved from 10.7) 11.3.2 Added SWR # 5343 SW1 and 6672 SWR 11.3.6 Added 5292 SW1 4
d l
1 l
11.3.8 to Deleted due to new mechanical seals. ]
11.3.8.2 1;.4.4 Add note #6 11.4.33 Changed word install to inspect.
11.4.33 thru Revised to add installation instruction for mechanical sealing nozzle covers per 11.4.37.1 manufacturer's manual.
11.4.45 Added visual inspection requirements for Maintenance Representative r
11.4.56 Added visual inspection requirements for shroud area 11.5.15 Revised for clarity 11.5.20 Changed rope and tubing size 11.5.23 Added caution 11.5.58 Revised elongation requirements 11.5.60.2 Revised elongation requirements 11.5.61.15 Revised elongatien requirements l 11.5.68 thru Revised to add removal instructions for nozzle 1 11.5.72 cc c 1
._ In Section 13.0, enci fares 13.6 and 13.7 were added. The Data I sheets were revised to reflect the abovefchanges. This procedure
! will be used to maintain the Reactor Vessel Head in its original-design requirements and specifications. No USQ is involved.
OP/1/A/6700/01 Change #181 This change replaces-all 4 pages of Figure 1.1 with new Axial Flux Difference (AFD)' Targets obtained per PT/1/A/4150/08, Target Flux Difference Calculation. The target AFD is changed to keep control Bank D at approximately 215 steps withdrawn at 100% full power while allowing for changes in the natural axial power. shape.that occur with burnup. The targets are-set by procedure to be within the operating bounds of Tech._ Spec. 3/4.2.1. The. targets _are an operating guide-line only to aid the control room operators in maintaining AFD within the limits of1P.S. 3/4.2.1. .The targets serve no other purpose.
They do not feed any trip function or serve any safety relatea func-tions. The limits that must be observed in T.S. 3/4.2.1 are set by cycle specific analysis. All accidents analyzed in the FSAR chapter 15 have as one of the initial conditions that AFD is within the-5
l l
limits of T.S. 3/4.2.1. As such, the targets have no effect on the ;
accident analysis. No USQ is involved.
MP/0/A/7150/16A Change /3 This change added two steps which were omitted during the procedure j upgrade. The correction made by this procedure change has been reviewed against approved vendor manuals, design documents, and station procedures to ensure that the corrective maintenance con- I trolled by these procedure will return the pump to as-built /as-designed conditions. These actions will ensure the pump's compliance with FSAR accident analysis. Therefore, no USQ exists.
PT/1/A/4200/09 Change #88 This restricted change adds steps to retest componente which were not verified to operate properly during Engineered Safety Features (ESP) testing. It will document retests of Auxiliary Feedwater Valves 1CA149 and 1CA152 after completion of Work Requests 7453PRF and 7454PRF, verify the proper operation of Incore Instrument Room Vent Unit 1A, and verify proper operation of Control Rod Drive vent fan
- 2. It also documents calibration and time delay for load group 3 after calibration following failure during B train LOCA test. The rotests involve only the components being retested and will not place the equipment in any abnormal alignment. The change also deleted Containment Purge Valve 1VP6B from the enclosures where response times are recorded. IVP6B exceeded its allowable stroke time; howev-er, T.S. 3.6.1.9 requires that the valve be sealed closed during modes 1 through 4. As a result, power is removed from this valve with the valve in its safety position (closed) during the modes in which is required to be able to close in a specified time. No sys-tems are made inoperable 7r degraded by these tests. The possibility
- of an accident different than any already evaluated in the FSAR will l not be increased by this change. Neither the probability nor the consequences of a malfunction of equipment important to safety will be created by these changes. Thus, no USQ is deemed to exist.
t PT/0/A/4400/08 Change # 46 This restricted change is being made to allow the 1A Nuclear Service Water (RN) pump house balance to be performed with an operating B train RN pump. The two trains of RN will be separated, with A train initially supplying the Unit i non-essential header. If insufficient flow path is available to achieve the required train A header pres-
! sure, steps have been added to align Unit 2 non-essential header to A train also. The operating B train RN pump will supply the B train essential header, and RN pump 1A will supply the A train essential header, noth trains of RN will continue to be supplied during the test. All valves will remain operable, with the exception of 1RN291 and 2RN291 (Component Cooling heat exchanger outlet valves) which 6
1 l
will be f ailed open (open is their safety position) . All equipment will continue to receive RN during the test. No USQ exists, l
PT/0/A/4400/08 Change #45 This restricted change is being made so that the RN pump house bal-ance may be performed on RN pump 1B without shutting down RN pump 1A.
The two truins of RN will be separated with B train supplying the non-essential headers. RN pump 1A will supply the A train essential header, and RN Pump 1B will supply the B train essential headers.
Both trains of RN will continue to be supplied during the test. All equipment will continue to receive RN during the test. All valves will remain operable, with the exception of 2RN351 (Component Cool-ing heat exchanger 2B outlet valve) which will be failed open (safety position). Both Containment Spray heat exchangers will be used. The Diesel Generator Engir. Cooling Water heat exchangers will be used if necessary to get the required flow. No USQ is deemed to exist.-
MP/0/B/7400/43 Re-type This is a previously approved procedure that was upgraded to the new procedure format. Tne changes are a result of the new format, are
! minor in nature, do not create an unreviewed safety question, and have no effect on the identified sections. The scope of the proce-dure has increased to embrace the complete asserably rather than just the pump itself. The title has been changed to include the pump's gear reducer and motor. There are increased-references to include asbestos controls. Manufacturer references were added for the gear reducer and motor. In the body of the procedure, steps are delineat-I ed for the removal and replacement of~the associated gear reducers and motors (Section 11.4 -- Gear Reducer Removal,- Section 11.5 --
Gear Reducer Installation, Section 11.6 Motor Removal, and Section 11.7 -- Motor Installation). While coupling alignment is addressed, additional information was added to-complete coupling installation and alignment at the motor / gear reducer, gear reducer / pump interface.
The incorporation of the gear reducer and motor into the procedure.is to organize and consolidate closely interrelated component sequences under one procedure. They are recommendations and reqairements of the manufacturer's instruction manuals. The purpose of these addi-tions is to perform deliberate and controlled steps'in the removal
, and replacement of these components. This will insure continued operailonal performance and maintenance of material condition excel-lence. The changes do not adversely effect.any FSAR section in a l significant manner. They do not adversely effect equipment operabil-l ity.
B&WFC FO-019 Initial Issue This procedure provides information and instructions concerning the removal, disassembly, and packaging of the B&W Fuel Company's-(B&W 7
1 l
FC) Remote Upper-internals Dimensional Inspector (RUDI) from the refueling cavity.
The RUDI system was designed to inspect the reactor upper internals for wear caused from vibration of the control rods in the upper internals. The area of concern is the individual guide cards which support tSe control rods. The RUDI system will travel Cown into the upper intarnals, where a profiling transducer will align with a guide card and perform profiling measurements (non-destructive). The B&W ,
inspection probe has the ability to collapse into a configuration of less than 2.3 inches in diameter (which allows the profile probe to travel through the Control Rod Drive Shaft hole at the top of each guide tube assembly). The system will measure the slot and holes in the guide cards of the upper internals guide tubes. The continuous section in the lower portion of the guide tube will not be inspected.
This program is part of the Rod Control Cluster Assembly (RCCA) demonstration program which will determine the effects of using hardened control rods in the reactor. A total of three B&WFC control rod assemblies, which have Maen coated, will be placed in the reac-tor. Two assemblies will :e Armoloy plated 304 stainless steel clad, and one will be chromium carbide coated Inconel 625 clad. A fourth control location will be supplied with a standard Westinghouse con-trol rod assembly. Over several operating cycles, these control rod assemblies and the corresponding upper internals will be inspected for wear. The wear measurements over several operating cycles vill establish the rate of wear on the control rods and upper internals.
The RUDI probe and housing will be maneuvered by the monorail hoist on the Reactor Building Manipulator crane and positioned on the top hat of each guide tube, in turn, to be inspected. The total weight of this system is 560 lbs. This will not challenge the ability of the hoist, since the hoist is rated at 1.5 tons. Even when the RUDI housing is setting on the guide tube top hats, the system will still be attached to the monorail hoist to preclude the possibility of the housing from causing undesirable eccentricity forces on the guide tube. The Civil Engineering Section of Design Engineering has per-
! formed an analysis concerning this. In addition, B&W has designed this system to be non-detrimental to the guide tubes. Any guide tube problems resulting from this activity will manifest themselves during the control rod latching procedures and/or IP/0/B/3220/01, Control Rod Drop Timing Test (i.e., before unit goes critical). This activi-ty may be performed under water (i.e., in contact with the Reactor l Coolant Water). All materials used by this system meet the require-ments for contacting the Reactor Coolant. One item, the alignment bladder (ethylene propylene diene monomer with a polyethylene mesh)
I required analysis by Duke Corporate Chemistry. The resulting Materi-al Data Sheet (DP No. 10029) approved this substance as a Category I item based on the leachable results.
l This activity will be performed during No Mode with the water level l approximately at the 594 foot elevation. At elevation the water level over the vessel / internals would be sufficient to attenuate the majority of the gamma radiation; ALARA sect 4.on concurs with this.
i 8
This piecedure creates no accident scenarios that are not already analyzed. This procedure does not require off-normal operation of safety equipment. The margin of safety will be enhanced by the use of this procedure since the flow induced vibratory wear problem will become better understood and actions can be taken to determine a definitive solution to this industry-wide problem. In addition, the Catawba Unit 2 (CNS-2) upper internals guide tube assemblies can be replaced if the RUDI study indicates and unacceptable amount of wear on the guide cards. CNS-2 Upper Internals design includes a set of 8 part length guide tubes which are identical to the full length guide tubes. Since the part length control rods are not utilized at CNS-2,
! these guide tubes can be implemented as replacements if necessary.
l No USQ is involved.
B&WFC FO-018 Initial Issue This procedure provides information and instructions concerning the operation of the B&W Fuel Company's (B&WFC) Remoca Upper-internals Dimensional Inspector (RUDI).
The RUDI system was designed to inspect the reactor upper internals for wear caused from vibration of the control rods in the uppor internals. The area of concern is the individual guide cards which support the control rods. The RUDI system will travel down into the upper internals where a profiling transducer will align with a guide card and perform profiling measurements (non-destructive).
The B&W inspection probe has the ability to collapse into a configu-ration of less than 2.3 inches in diameter (which allows the profile probe to travel through the Control Rod Drive Shaft hole at the top of each guide tube assembly). The system will measure the slot and holes in the guide cards of the upper internals guide tubes. The continuous section in the lower portion of the guide tube will not be inspected.
This program is part of the RCCA demonstration program which will determine the effects of using hardened control rods in the reactor.
A total of three B&WFC control rod assemblies, which have been coat-ed, will be placed in the reactor. Two assemblies will be Arnoloy plated 304 stainless steel clad, and one will be chromium carbide l coated Inconel 625 clad. A fourth control location will be supplied l with a standard Westinghouse control rod assembly. Over several operating cycles, these control rod assemblies and the corresponding l upper internals will be inspected for wear. The wear measurements over several operating cycles will establish the rate of wear on the i
control rods and upper internals.
l The RUDI probe and housing will be maneuvered by the monorail hoist on the Reactor Building Manipulator crane and positioned on the top hat of each guide tube in turn to be inspected. The total weight of this system is 560 lbs. This will not challenge the ability of the hoict since the hoist is rated at 1.5 tons.
9
I l
Even when the RUDI housing is aetting (.: the guide tube top hats, the !
system will still be attached to the merarail hoist to preclude the !'
possibility of the housing from causing undesirable eccentricity forces on the guide tube. The Civil Engineering Section of Design i Engineering has performed an analysis concerning this. In addition, o B&W has designed this system to be non-detrimental to-the guide
, tubes. Any guide tube problems resulting from this activity will r manifest themselves during the control rod latching procedures and/or IP/0/B/3220/01, Control Rod Drop Timing Test-(i.e., before unit goes critical). This activity may be performed under water (i.e., in contact with the Reactor Coolant Water). All materials used by this system meet the requirements for contacting the Reactor Coolant.
One item, the alignment bladder (ethylene propylene diene monomer with a polyethylene mosh) required-analysis by Duke Corporate Chemis-try. The resulting' Material Data Sheet (DP No. 10029) approved this-substance as a Category I item based on the'leachable results.
This activity will be performed during No Mode with the water level approximately at the 594 foot elevation. At elevation the water level over the vessel / internals would be sufficient to actenuate the majority of the gamma radiation; ALARA section concurs with this.
This procedure creates no accident scenarios that are not already analyzed. This procedure.does not require off-normal operation of safety equipment. The margin of safety will be enhanced by the use of this procedure.since the flow induced vibratory wear problem will become better understcod and actions can be taken to determine a definitive solution to this industry-wide problem. In addition, the CNS-2 upper internals guide tube assembl'ies can be replaced if the RUDI study indicates and unacceptable. amount of wear on the guide cards. CNS-2 Upper Internals design includes a set of 8 part length guide tubes which are identical to the full length guide tubes. Since the part length control' rods are not utilized at CNS-2, these guide tubes can be implemented as replacements if necessary.
PT/2/A/4150/31 Initial Issue This procedure provides assembly and installation instructions for the B&W Fuel Coifany's (S&WFC) Remote Upper-internals Dimensional Inspector.(RUDI). It provides. operating instructions-for the B&WFC RUDI to inspect the Unit 2 Reactor Upper Internals for wear caused from vibration of the RCCAs in the Upper Internal guide cards. It provides removal, disassembly, and packaging instructions for the B&WFC RUDI.
The RUDI system was designed to inspect the reactor upper internals ;
for wear caused from vibration of the control rods in the upper -)
internals. The area of concern is the individual guide cards which l i
support the control rods. The RUDI system will travel down into the upper internals where a profiling transducer will align with a guide card and perform profiling measurements (non-destructive). This program is part of the RCCA demonstration program which will 10 l
l l
l I
l determine the effects of using hardened control rods in the reactor. l A total of three B&WFC control rod assemblies, which have been coat- i ed, were placed in the reactor.
The B&W inspection probe has the ability to collapse into a configu-ration of loss than 2.3 inches in diameter (which allows the profile probe to travel through the Control Rod Drive Shaft hole at the top of each guide tube assembly). The system will measure the slot and holes in the guide cards of the upper internals guide tubes. The continuous section in the lower portion of the guide tube will not be inspected.
The RUDI probe and housing will be maneuvered by the monorail hoist on the Reactor Building Manipulator crane and positioned on the top hat of each guide tube in turn to be inspected. The total weight of this system is 560 lbs. This will not challenge the ability of the hoist since the hoist is rated at 1.5 tons.
Even when the RUDI housing is setting on the guide tube top hats, the system will still be attached to the monorail-hoist to preclude the possibility of the housing from causing undesirable eccentricity forces on the guide tube. The Civil Engineering Section of Design Engineering has performed an analysis concerning this. In addition, B&W has designed this system to be non-detrimental to the guide tubes. Any guide tube problems resulting from this activity will manifest themselves during the control rod latching procedures and/or IP/0/B/3220/01, Control Rod Drop Timing Test (i.e., before-unit goes critical).
This activity may be performed under water (i.e., in contact with the Peactor Coolant Water). All materials used by this system meet the requirements for contacting the Reactor Coolant. One item, the alignment bladder (ethylene propylene diene monomer with a polyethyl-ene mesh) required analysis by Duke Corporate Chemistry. The result-ing Material Data Sheet (DP No. 10029) approved this substance as a Category I item based on the leachable results. This activity will be l performed during No Mode with the water level approximately at the l
594 foot elevation. At elevation the water level over the ves-sel/ internals would be sufficient to attenuate the majority of the I gamma radiation; ALARA section concurs with this.
This procedure creates no accident scenarios that are not already analyzed. This procedure does not-require off-normal operation of safety equipment. The margin of safety will be enhanced by the use of this procedure since the flow induced vibratory wear problem will _
become better understood and actions can be taken to determine a
! definitive solution to this industry-wide problem. In addition, the )
CNS-2 upper internals guide tube assemblies can be replaced if the RUDI study indicates and unacceptable amount of wear on the guide cards. CNS-2 Upper Internals design jncludes a set of 8 part l length guide tubes which are identical to the full length guide ,
tubes. Since the part length control rods are not utilized at CNS-2, these guide tubes can be implemented as replacements if necessary.
11
l' PT/0/A/4150/18 Initial Issue This procedure is designed to provide a safe transfer of fuel assem-bly inserts from the previous cycle fuel assemblies to their next cycle fuel assemblies. Only inserts are moved in this procedure, not fuel assemblies. The general procedure is for the fuel handling personnel to move the appropriate insert handling tool over the specified insert, remove the insert, and place it in its final fuel assembly. Independent Verification is required both for the initial position of the insert and the final position of the insert. Fuel assembly Region Reference Numbers and insert ids are not checked by this procedure. It is designed to move inserts based only on the positions in the spent Fuel Pool. However, PT/0/A/4550/09, Fuel Assembly / Insert Verification, is required to be performed after the shuffle is complete. This test will verify that each fuel assembly that is going back into the core contains the correct insert for the next cycle and is in the proper Spent Fuel Pool location. In addi-tion, PT/0/A/4550/03C, Core Verification, will be performed after fuel loading to verify that the fuel assemblies are in the proper core location for the next cycle, and that all fuel assemblies that are supposed to contain Rod Control Cluster Assemblies (RCCAs) actu-ally contain RCCAs. The activities described by this procedure are bounded by the accident analysis performed in FSAR Chapter 15.4.7, Inadvertent Loading and operation of a Fuel Assembly in an Improper Location. This chapter also analyzed operation and loading of a fuel assembly with an incorrect insert. This procedure requires movement of inserts and fuel handling tools over the. spent fuel in the Spent Fuel Pool, but the inserts and handling tools weigh less than a fuel assembly. Therefore, the accident analysic.in FSAR Chapter 15.7.4, Fuel Handling Accidents, is bounding. In addition, precautions are taken in the procedure to ensure that-objects are not dropped into_
the Spent Fuel Pool. The latest philosophy on thimble' plugs is that if the thimble plug rods are bent more than 1 inch, then the thimble plug should not be bent back into alignment and reinserted into a Fuel Assembly going back into the core, due to the possibility of the rod breaking off during operation. This could possibly cause-a loose part and allow bypass flow through the F/A guide thimble. As a response to this latest concern, this procedure will not reinsert any bent thimble plugs. This procedure does not require off-normal operation of safety equipment. No USQs are created by this proce-dure.
PT/0/A/4150/17 Initial Issue This procedure is designed to unload the core in a safe and orderly manner. 'This test is not described in the FSAR. The closest-de-scription is found in FSAR section 9.1.4 which describes the process used in an incore fuel assembly / insert shuffle. FSAR section 9.1.4 includes a description of how the fuel is off-loaded from the core to the Spent Fuel Pool. This procedure complies with the off-load description found there. It is different in that it completely unloads the core, while FSAR section 9.1.4 leaves fuel assemblies in the core at all times.
12
This procedure includes statements requiring the Fuel Handling Senior Reactor Operator to be present in the Reactor Building whenever fuel is moved in the Reactor Vessel by non-licensed personnel. This is to comply with 10CFR55.13.b. The only loads being moved are fuel assem-blies. The accident discussed by FSAR section 15.7.4, Fuel Handling ,
Accidents in the Containment and Spent Fuel Storage Buildings, is i bounding. No USQ is created by this procedure.
l TN/2/A/0299/01/02A Initial Issue Nuclear Station Modification CN-20299, Rev. 1, Will complete all the electrical work remaining with the Upper Head Injection (UHI) System deletion. The system was mechanically and electrically isolated by CN-20299, Rev. O. CN-20299, Rev. 1, will delete the controls and cables from various cabinets throughout the plant. The purpose of this procedure is to control the removal of wiring which is associut-ed with the removal of the UHI system. This procedure deals with the wiring in the Auxiliary Building. The isolations required by this procedure are needed so that deleted cables may be removed from the cabinets. No equipment will be isolated until the plant is in a condition in which the equipment is not required to be operable.
Prior to completion of the procedure, functionals will be performed to ensure that equipment affected by this procedure has not been adversely affected. Thus, an unroviewed safety question will not be created by this procedure.
PT/0/A/4550/03C Initial Issue This procedure is designed to verify that all fuel assemblies that are loaded into the core for the next cycle are in the proper loca-tions and contain the proper type inserts. To do this, each core location is inspected, the fuel assembly Region Reference Number is recorded, and the insert type is checked. Then, this record is independently compared to a prepared map for the next cycle. If any discrepancies are found, they are resolved before completion of the procedure. This procedure is designed to prevent the accident ana-lyzed in FSAR Chapter 15.4.7, Inadvertent Loading and Operation of a Fuel t;sembly in an Improper Location. This activity involves using an underwater camera to read the fuel assembly Region Reference Numbers. This activity does not constitute Core Alterations. No USQ is created by this procedure.
PT/2/A/4600/10 Retype, Changes 0 to 1 Incorporated This reissue principally incorporates the latest revision of the vendor procedure under which Incore Thimble Eddy Current Testing is conducted. One major change is the addition of the requirement that
! the Reactor Coolant (NC) System pressure be 5 5 psig as a Prerequi-site System Condition. This is due to the fact that the system will not be vented to atmosphere at the time of this testing. A nitrogen overpressure of no more than 5 psig is placed on the system to facil-itate the draining. The latest revision of the Vendor Procedure 13
(Revision 3) is unchanged from a methodology standpoint. The only changes in this document involve updating of revision numbers of reference documents. Eddy current testing will still be conducted by the passage of a test probe (which is no larger in diameter than an actual Incore Detector) the length of each incore thimble for the purpose of assessing the magnitude of the vibration induced thimble tube wall thinning.
Based on the evaluation of the eddy current results, thimbl%s with excessive through wall wear shall either be isolated at the seal table or repositioned to move the region of thimble wear out of the area where wear occurs. This testing program is being undertaken to anticipate and prevent potential thimble failures. This procedure differs from the Unit 1 procedure only in that Thimble cleaning is not required before testing, as long as the thimbles have been shown to be unobstructed. A new step was created to allow extent (in inches) of thimble repositioning to be recorded for the purpose of future reference. Provisions were added l'or recording the procedure under which positioning is performed in the event that a special procedure is created for this. FSAR Section 15.6.5, Loss of Coolant Accidents, has been evaluated from the standpoint of the small break LOCA induced by the eddy current test probe causing a failure of.a thimble with excessive through wall thinning. This scer.ario is bounded by the small break LOCA analysis, since it would occur in Mode 5 with the NC system depressarized to less than 5 psig. The isolation of a leaking thimble is addressed specifically in FSAR Section 7.7.1.9. Repositioning degraded Incore Thimbles will ensure the operability of the Incore Detector System. There are no unreviewed safety questions involved with the implementation of this procedure change.
PT/2/A/4250/14 Initial Issue The purpose of this procedure is to response time test the Auxiliary l Feedwater Pump Turbine (CAPT) #2 governor and any equipment which could directly or indirectly affect its performance. CAPT #2 will be
' run in recirculation to the Upper Surge Tank. In this test align-ment, CAPT #2 is not available for emergency actuations without local
- operator actions. This condition exists for Sections 12.1 and 12.2, l in which flow-is set up to a desired value to the Upper Surge Tank.
l In Sections 12.3 and 12.4, the pump is run in miniflow to the Upper l Surge Tank, and is thus still capable of providing flow to the. Steam l Generators in.the event of a CA autostart. The operations shift Unit Senior Reactor Operator will verify that proper CA system operability concerns are addressed before-allowing testing to begin. The elec-trical circuit modifications needed to perform this test only affect CAPT #2 and do not affect CA pumps 2A and 2B, These modifications have the appropriate independently verified placement and removal steps. During Sections 12.1 had 12.3, CAPT #2 is started by using a test box switch to deenergize solenoid valves for Main Steam to Auxiliary Equipment (SA) Valves 2SA2 and 2SAS from 2AFWPTCP by open-ing the solenoid valves power circuits directly. When this test box switch is in the closed position, the Control Room Operators will 14
still have full use of the CAPT /2 START /STOP switch. When the test box switch is opened, CAPT #2 will start, and the Control Room Opera-tors will not be able to stop CAPT #2 using the CAPT /2 START /STOP switch unless the test box switch is closed, which in itself will stop the pump. At any time, the Control Room Operators will be able to stop CAPT #2 by closing 2SA145, CAPT #2 Trip and Throttle Valve.
Procedural CAUTION and NOTE statements have been included to inform the control Room Operators of these CAPT #2 control function details.
The required number of CA pumps will remain operable throughout the test. CAPT #2 will not be operated outside of design parameters. No USQ is created by the performance of this procedure.
TN/2/A/0606/00/01A Initial Issue Nuclear Station Modification (NSM) CN-20606, Rev. O, is to install a bypass from the Nuclear Service Water (RN) system to the. Condenser Circulating Unter system for the purpose of flushing the RN piping that branches off to supply the Auxiliary Feedwater (CA) pumps and Auxiliary Shutdown Panel Supply Units. This procedure is to provide guidance for the installation of the piping, supports, and components located between the 'A' Train tie-in and the first qualified hanger past tne isolation valve. This implementation procedure provides guidance for the installation of piping, valves, supports, a Y-strainer, and miscellaneous components in the RN System. The new piping installed in this procedure will tie into the RN system ap-proximately one foot upstream of valve 2RN250A. The work associated with this procedure is to be performed with Unit 2 in Modes 5, 6, or
! no mode. The 'A' Train portion of the RN system to the component Cooling System (KC) Heat Exchanger will be isolated during implemen-tation of this procedure. The CA portion of this procedure only involves some CA supports and a few feet of pipe that will not be tied into the CA system under this procedure. The isolations re-quired to perform this modification will be scheduled with other i outage activities such that the components isolated do not impact the I operability of any systems required to be operable during implementa-tion of this procedure. This procedure, along with any other work control procedures, will adequately govern the return to service of all components / systems affected by this modification. The isolation, implementation, and testing performed in this procedure will not increase the probability, consequences, or possibility of new or previously evaluated accidents. Thus, it is concluded that there are no unreviewed safety questions associated with the implementation of this procedure.
l TN/2/A/0606/00/02A Initial Issue NSM CN-20606, Rev. O, is to install a bypass from the RN system to the Condenser Circulating Water system for the purpose of flushing the RN piping that branches off to supply the CA pumps and Auxiliary Shutdown Panel Supply Units. This procedure is to provide guidance for the installation of the piping, supports, and components located (
between the 'B' Train tie-in and the piping installed under 15 !
l l
TN/2/A/0606/00/01A. This implementation procedure provides guidance for the installation of piping, valves, supports, and miscellaneous components in the RN System. The new piping installed in this proce-dure will tie into the RN system approximately one foot upstream of valve 2RN310B. The work associated with this procedere is to be performed with Unit 2 in Modes 5, 6, or no mode. The 'B' Train portion of the RN system to the KC Heat Exchanger will be isolated during implementation of this procedure. The isolations required to perform this modification will be scheduled with other outage activi-ties such that the components isolated do not impact the operability of any systems required to be operable during implementation of this procedure. This procedure, along with any other work control proce-dures, will adequately govern the return to service of all compo-nents/ systems affected by this modification. The isolations, imple-mentation, and testing performed in this procedure will not increase the probability, consequences, or possibility of new or previously evaluated accidents. Thus, it is concluded that there are no unreviewed safety questions associated with the implementation of this procedure.
TN/2/A/0606/00/03A Initial Issue NSM CN-20606, Rev. O, is to install a bypass from the RH system to the Condenser Circulating Water (RC) system for the purpose of flush-ing the RN piping that branches off to supply the CA pumps and Auxil-iary Shutdown Panel Supply Units. This procedure is to provide guidance for the installation of the piping, supports, and components that tie into the RC system side of CA. Also, the piping and hanger work not completed under Work Units 1 and 2 for this NSM will be finished in this procedure. This implementation procedura provides guidance for the installation of piping, valves, supports, and mis-cellaneous components in the RN and CA Systems. The new piping installed in this procedure will tie into the RN system piping in-stalled under TN/2/A/0606/00/01A and also tie into the CA system
(
piping en the RC supply line. The work associated with this-proce-dure is to be performed with Unit 2 in Modes 4, 5, 6, or no mode.
' The RC to CA portion of the CA system will be isolated during imple-montation of this procedure. The isolations required to perform this modification will be scheduled with other outage activities such that the components isolated do not impact the operability of any systems required to be operable during implementation of this procedure. The fire barrier block wall for the CAPT Control Panel room will be i
penetrated during installation. Since fire barriers are required to l be operable at all times, the requirements of Technical Specification 3/4.7.11 will be adhered to until the fire barrier is returned to an i operable condition. The Post Modification Testing to be performed in this procedure does not involve any system or component alignments that cannot be quickly re-aligned to their normal position if re-quired. The testing will take place before Mode 3;'therefore, the operability of the CA system will not be jeopardized. This proce-dure, along with any other work control procedures, will adequately govern the return to service of all components / systems affected by this modification. The isolations, implementation, and testing 16
l performed in this procedure sill not increase the probability, conse-quences, or possibility of new or previously evaluated accidents.
Thus, it is concluded that there are no unreviewed safety questions associated with the implementation of_this procedure.
I l
l PT/2/A/4200/09 Re-issue, Changes 0 to 47 Incorporated In addition to completely changing the format of the procedure,.
several additional changes havo been made on this reissue of the j
procedure which were not previously approved.
- 1) Enclosure 13.21 was added to provide additional guidance on the placement of the jumpers on the Blackout load centers during sections 12.3 and 12.6. This-change does not affect this evaluation since it does not change the way'the proce-dure is performed.
- 2) Power no longer needs to be removed from Safety Injection (NI) valves 1NI188A, 1NI252A, 1NI189B, and 1NI253B. A modification during the Unit 1 End-of-Cycle 4 outage changed the power supply on the Volume Control Tank (VCT) level instrumentation that causes these valves to swap on low VCT leve , Previously, these circuits lost power during the UI.it 2 Engineered Safety Features (ESF) tost, but they no longer will.
l l
- 3) A CAUTION statement was added to ensure that the test does not start aftdr 2345 hours0.0271 days <br />0.651 hours <br />0.00388 weeks <br />8.922725e-4 months <br />. The ResponsefTime Test (RTT) program on the OAC will not function properly if the test extends through 0000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. This change does not affect any systems or components. The change was made to ensure that the test does not have to be repected due.to incorrect l response times.
l l
- 4) Steps have been added_to check the circuitry for the hydro-gen recombiners in the Blackout /LOCA sections of the test.
i This is to ensure that the recombiners get the permissive
! to start on Load Group-13. The requirement for closing the supply breakers in the system alignment was deleted since
~
the recorbiners will not be started; only the permissive will be verified. This change has no'effect on-operation of any equipment other than the previous requirement to close the supply breakers.-
- 5) Steps have been added-to remove power from_ valves 2NI9A (sections 12.1 and 12.2) and 2 nil 0B (sections 12.4 and '
12.5) if the. opposite train. Chemical and Volume-control _
l System (NV) pump is running._ These steps were added to L prevent injection into the Reactor Coolant system.
- 6) The alignment for the Spent Fuel Cooling (KF) system is now specified so that the IGP pump on'the train under test may be started and left running for-the duration of_the test.
17
1 Previously, the pump could not always be started due to lack of a flow path. This will provide for a better test
, and a more accurate indication of the Diesel Generator load during an accident.
l
- 7) The breaker for the Unit 1 Nuclear Service Water Pump Structure Ventilation (V2) Fans will be taken to the OFF position in the Blackout section so that the Unit 2 fans will start during the test. This change is being made so that all VZ Fans will be verified to start during the Unit 1 and Unit 2 ESF tests. The VZ fans will only be verified in the Blackout sections. They will not be verified in the other sections since the same relay actuates them.
- 8) In sections 12.9 and 12.10, steps have been added to verify the trip circuitry on NV Pump #2. The change is being made so that all components on relays K648 and K649 will be tested and PT/2/A/4200/09A, Auxiliary Safeguards Test Cabinet Periodic Test, will not have to be performed to verify the trip circuitry. No systems or components will be affected by this change since NV Pump #1 will be racked out in Modes 5 and 6.
- 9) In sections 12.9 and 12.10, valves 2KF101B and 2KF103A will be verified to reposition during the test. This will require that 2KF101B initially be open during ssetion 12.10 and that 2KF103A be open during section 12.9. These valves are normally closed. This will not affect the KF system since the other isolation valve in the line will remain closed by procedure. The change is being made so that all components on relays K648 and K649 will be tested, and PT/2/A/4200/09A, Auxiliary Safeguards Test Cabinet Periodic Test, will not have to be performed to verify the valves will reposition.
- 10) Thc ateam Generator Power Operated Relief Valve (PORV) bloc % valves were added to the lineups so that opening t' 2 POP,Ve for the test would have no effect.
- 11) Main Feedwater (CF) Valves 2CF21 and 2CF25 were added to enclosures 13.18A and 13.19A. These valves have been closed by Operations in the past to prevent-the Steam Generators from filling during the test. They.have beJn added to the procedure to document that they are closed and that they are returned to their "As Found" position.
- 12) Enclosures were added to sections 12.3 and 12.6 to aid Operations in recovering from the blackout sections of the test. The enclosures are to assist in returning systems (components) to their pretest aljgnment.
- 13) Steps have been added to verify status lights and annunciators that are expected during the test. These steps were added so that the test coordinator would be 18
aware of any problems that may arise during the setup in l the process cabinets, and so that the control room opera- ;
tors would know to expect the alarms. l Neither the probability nor the consequences of a malfunction of equipment important to safety will be increased by the performance of ,
this test. Pumps which receive a start signal are aligned to assure that a minimum flow path is available. The procedure requires imme- l diate verification of flow upon actuation. Testing is performed in accordance w4th the FSAR and Technical Specifications. The bases to Technical Specifications were reviewed and thu margin of safety is not reduced by the performance of the Engineered Safety Features Actuation Periodic Test.
B&WFC FO-017 Initial Issue This procedure provides information and instructions concerning the assembly and installation of the B&W Fuel Company's (B&WFC) Remote Upper-internals Dimensional Inspector (RUDI).
The RUDI system was designed to inspect the reactor upper internals for wear caused from vibration of the control rods in the upper internals. The area of concern is the individual guido cards which support the control rods. The RUDI system will travel down into the upper internals where a profiling transducer will align with a guide card and perform profiling measurements (non-destt;ative). The B&W inspection probe has the ability to collapse into a configuration of less than 2.3 inches in diamoter (which allows the profile probe to travel through the Control Rod Drive Shaft hole at the top of each guide tube assembly). The system will measure the slot and holes in the guido cards of the upper internals guide tubes. The continuous section in the lower portion of the guide tube vill not be inspected.
The RUDI probe and housing will be maneuvered by the monorail hoist on the Reactor Building Manipulator crane and positioned on the top hat of each guide tube in turn to be inspected. The total weight of this rystem is 560 lbs. This will not challenge the ability of the hoist sinco the hoist is rated at 1.5 tons. Eve.n when the RUDI housing is sitting on the guide tube top hats, the system will still be attacbed to the monorail hoist to preclude the possibility of the housing from causing undesirable eccentricity forces on the guide tube. The Civil Engineering Section of Design Engineering has per-formed an analysis concerning this. In addition, B&W has designed this system to be non-detrimental to the guide tubes. Any-guide tube problems resulting from this activity will manifest themselves during the control rod latching procedures and/or IP/0/B/3220/01, Control Rod Drop Timing Test (i.e., before unit goes critical). This activi-ty may be performed under water (i.e., in contact with the Reactor Coolant Water). All materials used by this system meet the require-ments for contacting the Reactor Coolant. One item, the alignment bladder (ethylene propylene diene monomer with a polyethylene mesh) required analysis by Duke Corporate Chamistry. The resulting l
19
i l
l Material Data Sheet (DP No. 10029) approved this substance as a i Category I item based on the leachable results.
This activity will be performed during No Mode with the water level approximately at the 594 foot elevation. At elevation the water level over the vessel / internals would be sufficient to attenuate the majority of the gamma radiation; ALARA section concurs with this.
This procedure creates no accident scenarios that are not already analyzed. This procedure does not require off-normal operation of safety equipment. The margin of safety will be enhanced by the use of this proc ure since the flow induced vibratory wear problem will become betto anderstood and actions can be taken to determins a definitive solution to this industry-wide problem. In addit;on, the Catawba Unit 2 (CNS-2) upper internals guide tube assemblies can be replaced if the RUDI study indicates an unacceptable amount of wear on the guide cards. CNS-2 Upper Internals design includes a set of 8 part length guide tubes which are identical to the full length guide tubes. Since the part length control rods are not utilized at CNS-2, these guide tubes can be implemented as replacements if necessary.
PT/0/A/4150/22 Initial Issue This procedure is designed to load the core in a safe and orderly
( manner. This procedure is not described in the FSAR. The closest I description is found in FSAR Section 9.1.4 which describes the pro-cess used in an incore fuel assembly / insert shuffle. FSAR Section 9.1.4 includes a description of how the fuel is loaded from the Spent l
Fuel Pool to the core. This procedure complies with the loading description found there with the exception of Part 9.1.4.2.2t on pages 9.1-20 and 9.1-21, in which the FSAR should be revised to reflect the current Tech. Spec. requirements of 3/4.9.1. It is also different in that the core is completely unloaded at the beginning, while FSAR Section 9.1.4 leaves fuel assemblies in the core at all times.
FSAR Table 14.2.12-2 (page 2), Initial Fuel Loading Abstract, de-scribes the initial fuel loading procedure used.at Catawba. This procedure complies with the abstract except for the use of temporary detectors. In addition, this procedure allows the use of the Boron Dilution Mitigation System (BDMS) detectors (excore fission chambers) instead of the Source Range detectors. The BDMS detectors are just as accurate and were determined to be functional for subcriticality monitoring during the Catawba Unit 1 End-of-Cycle 4 refueling outage.
Use of BDMS allows N-31 and N-32 Containment evacuation setpoint to b reset without stopping the loading. But more importantly, the BDMS detector will initiate automatic boration if a dilution or subcritical multiplication accident should occur.
This procedure is in compliance with the Duke Power Response to NRC Bulletin 89-03, Potential Loss of Required Shutdown Margin During Refueling Operations. Instructions are included to avoid formation of unanalyzed clusters of fuel assemblies. Part of the attempt to 20
avoid unanalyzed clusters of fuel assemblies is to avoid using new fuel assemblies in intermediato locations to " box in" other assem-blies. The preferred method to load difficult fuel assemblies is with the use of a fuel assembly shoehorn. A 10CFR50.59 Evaluation has been developed to use a fuel assembly shochorn.
This procedure includes the requirement that tho Fuel Handling Senior Reactor Operator is to be present in the Reactor Building whenever fuel is moved in the Reactor Vessel by non-licensed personnel. This is to comply with the latest version of 10CFR55.13b.
No increase in the probability or consequences of an accident evalu-ated in the FSAR is created by this procedure. This procedure cre-ates no accident scenarios that are not already analyzed. The only loads being moved are fuel assemblies. The accident discussed by FSAR Section 15.7.4, Fuel Handling Accidents in the Containment and Spent Fuel Storage Buildings, is bounding for fuel movement. The accident discussed in FSAR Chapter 15.4.7, Inadvertent Loading and Operation of a Fuel Assembly in an Improper Location, is also bound-ing for this procedure. To prevent this accident, the fuel assem-blies are identified as they are put into the core, and PT/0/A/4550/03C, Core Verification, is performed after all fuel assemblies are loaded. PT/0/A/4550/03C also verifies that fuel assemblies designated for Rod Cluster Control Assemblies (RCCAs) have a RCCA as their core component.
Evaluation of Fuel Assembly Shoehorn When a fuel assembly is bowed to an extent that the bottom nozzle is kicked out past the tapered ends of the lower core plate guide pins, the only way to seat the assembly is by use of a box or a wedge to push the bowed acaembly into position.
One positioning aid is a wedge shaped guide or a part length " shoe-horn". Daden Products Corporation has manufactured a fuel assembly shoehorn which is positioned adjacent to the slot into which the distorted assembly is to be positioned. The shoehorn is 8 to 12 i
inches high and tapered to-allow the bottom nozzle of the fuel assem-l bly to slide down the taper and engage on the lower core plate guide pins.
Since this shochorn is positioned manually with use of a rope (or cable) or a poll arm (standard Daden AIR-GRIPS handling system), it does not require manipulation by the refueling bridge handling mast.
Since it does not make contact with the refueling mast, it cannot jeopardize the mechanical interface of the fuel assembly with the refueling mast.
Daden Products has developed this shoehorn with the aid of experi-enced fuel handlers and has taken advantage of previously used wedge I designs. The Daden shoehorn has four large, tapered pins which are located in the flow holes; a 2.5 inch high collar separates the bottom of the shoehorn from the lower core plate, thus providing 1/2 inch clearance from the bottom of the shoehorn to the top of the 21
_. - - - -_ . . -. . . . - - ~ -, .-
guide pins. The shoehorn does not require use of lower core plate guide pins in order to position the shoehorn (preventing risk of damage to the guide pins). Instead, the shoehorn pins use the core plate flow holes.
The shoehorn is constructed of 1/4 inch stainless steel plate (50 pounds dry weight), which'makes it heavy enough to hold itself into position; because the four tapered pins have a one inch long engage-ment with the lower core plate, it cannot be tipped over if a fuel assembly is rested against it. The outside dimensions of the shoe-horn are slightly smaller than the fuel assembly dimensions; there-fore, the positioned shoehorn would not touch an already seated fuel assembly.
Once the shoehorn is in position, the fuel handlers lower the dis-torted fuel assembly using a standard open water movement until the assembly is approximately 12 inches from the lower core plate. The fuel assembly is then moved 1/2 pitch or more (even on indexed) toward the final position and then lowered. As the fuel assenbly is lowered, the shoehorn wedges the bottom of the assembly 11nto posi-tion. The 1/4 inch Stainless Steel plate has a' glass beaded finish (slick when wet) which helps eliminate any possibility of a fuel assembly snagging on the wedge.
The fuel assembly will be suspended from the refueling bridge mast and will not be using the shoehorn as a rest. No significantly
, higher loads are imparted on the fuel assembly legs as a result of l using this shoehorn as a guide, even if the assembly-is twisted'and only one leg contacts the shoehorn.-
The height of the shoehorn is made so that.it remains inLeontact with the bottom nozzle of the assembly until the lower core plate guide pins are aligned with the assembly's bottom' nozzle S-holes. The guide pins are aligned before the top of the bottom nozzle reaches the front face of the shoehorn. Thus, the fuel assembly will be lined up with its indexed position prior to the fuel rods (assumed to be on the bottom nozzle) reaching the front face of the shoehorn.
Use of this shoehorn will help reduce grid-to-grid. interfaces between fuel assemblies, because the distorted assemblies can be lowered off-indexed into open water until about 12 inches'above the lower core plate. The shoehorn will also help eliminate the use.of tempo-rary locations of non-distorted assemblies-in order to build the
" boxes," thus helping to ensure Tech. Spec. 3.9.1 is not' violated.
The initial conditions for the analysis for damaged-fueliin Section 15.7.4 of.the FSAR have not been changed.' Use of the fuel-assembly shoehorn will not cause the effects and consequences of a damaged fuel accident to exceed the calculated values of offsite dose-on Table 15.7.4-2 and 15.7.4-3. Also, as long as the assembly-is prop-l erly seated in the core, structural integrity and coolable geometry is assured. No USQ is created by either this procedure re-issue cn:
the use-of a shoehorn.
22
PT/2/A/4200/04E Initial Issue This test is specified to be performed in Mode 5 (with Reactor Vessel '
i Head installed) or No mode, at which time the containment Spray (NS) system is not required to be operable. During these modes, NS pumps are tagged out, and the test lineup oncures that the containment spray initiating valves are isolated with perer removed for the spray ring being tested. This ensures that an inadvertent spray down of i containment will not occur. The specified condition for this test I also ensures that any debris created by this test will not reach the ,
fuel by ensuring that either the reactor vessel head is installed or l all fuel is removed. Prior to initiation of air flow through the nozzles, the piping is drained to ensure all trapped water is removed
( from the piping.
The test procedure also ensures that the Instrument Air (VI) header pressure is maintained above 85 psig while using VI to pressurize the spray header to verify flow from the spray nozzles. This will ensure that all equipment supported by VI will have adequate air pressure for proper operation. Adequate VI compressors with backup from Station Air are available to ensure the VI header pressure is main-tained at the normal level. This test is required by Tech. Spec.
4.6.2.d. No USQ is created by this procedure.
PT/1/A/4400/03F Initial Issue The purpose of this test is to verify that the head curves for Compo-nent Cooling (KC) Pumps 1A1, 1A2, 1B1, and 1B2 meet design criterias For each pump, KC flow is adjusted to various rates by throttling its corresponding KC to Residual Heat Removal (ND) heat exchanger isola-tion valve and KC pump motor cooler outlet isolation valve. Thus, the total developed head will be determined at the different flow rates and compared to the design point.
The KC system is designed for operation during all phases of plant operation and shutdown, supplying cooling water to the safety-related heat exchangers of the ND system and various other equipment. Only one train of KC equipment is necessary to supply normal operating plant loads, as well as minimum engineered safety requirements.
i During this test, each pump will be individually tested within the action time allowed by Tech. Spec. 3/4.7.3. The opposite train will remain in service, supplying cooling water to the essential and non-essential components. Thus, the margin of safety defined in the bases of Tech. Spec. 3/4.7.3 is not degraded, and a USQ does not exist.
PT/1/A/4400/03E Initial Issue The purpose of this test is to assure that the automatic operation of the Component Cooling (KC) miniflow valves, 1KCC37A and 1KCC40B conforms to design criteria and to balance the flow through the miniflow line for each train of KC pumps. The miniflow valves will 23
i be adjusted to provide adequate flow to prevent dead-heading the two running pumps when alternate flow paths are isolated. It will fur-ther be verified that the throttled positions of the miniflow valves provide adequate back-pressure to prevent run-out by a single operat-ing pump.
The KC system is designed for eperation during all phases of plant operation and shutdown, supplying cooling water to the safety-related heat exchangers of the ND system and to various other components.
Only one train of KC is necessary to supply normal operating plant loads, as well as minimum engineered safety requirements. During this test, only one train of KC will be balanced at a time within the action time allowed by Tech. Spec. 3/4.7.3. The opposite train of KC will remain in service. In the unlikely event that the train of KC being tested is needed, the miniflow valve can be manually returned to the closed position and the train returned to service. Having the miniflow valves properly set up will provide protection against KC pumps dead-heading, so that the probability or consequences of an equipment malfunction will not be increased, and the margin of safety as defined in the bases to Tech. Spec. 3/4.7.3 will not be reduced.
No USQ is involved with the implementation of this procedure.
PT/2/A/4400/01 Re-issue, Changes 0 to 5 Incorporated This procedure has been rewritten and dirfers to some degree from the I procedure it is replacing. Previously approved changes have been incorporated in this rewrite as well as general procedure upgrades to enhance procedure execution and provide better plant configuration control. Thus, this procedure was considered entirely new for review purposes.
This procedure is required to be performed in "No Mode", with no fuel in the core. No immediate accident scenarios are affected by its performance. No portion of the Emergency Core Cooling System (ECCS) is required to be operable during this time, and Reactor Coolant System (NC) boron concentration control is not critical during the test. Since all equipment manipulated by this procedure will be operated within the appropriate design limitations, no equipment malfunctions are expected to occur. With the Vessel Head removed, there is no danger of overpressurizing the NC system by an inventory increase. This procedure assumes full control of the ECCS alignment '
and ensures that the pre- and post-test system alignments are identi-cal, facilitating an error free configuration control transfer back to Operations upon test completion. All non Tech. Spec. flow accep-tance criteria have been provided by either Westinghouse or Design Engineering. These acceptance criteria have been analyzed by these groups to ensure proper delivery of ECCS water inventory in all phases of the various Design Basis Accident scenarios. All accep-tance criteria that had not already been error corrected by these groups have been corrected by station personnel to account for test instrument inaccuracies. Thus, no accident scenarios are created.
The Tech. Spec. margin of safety is not reduced. No USQ is created by this procedure.
l 24 {
l
l TN/2/A/0608/00/01A Initial Issue Nuclear Station Modification (NSM) CN-t0608, Rev. O, provides an access hole in Steam Generator (S/G) .a to be used for Foreign Object Search and Retrieval (FOSAR). This procedure provides guidance for activities associated with machining the access hole in S/G 2A and securing the access hole to restore the S/G 2A secondary pressure boundary. Upon completion of this procedure, the S/G 2A access hole will be secured, and an access hole will be available for future S/G 2A inspections.
This procedure requires that the secondary side of S/G 2A be isolated and drained. This will be performed with Unit 2 in Mode 5 or below.
Isolating S/G 2A will remove it from service. Therefore, it will not be available for use as a heat nink. Tech. Spec. allows S/Gs to be removed from service for maintenance and inspections. Tech. Spec.
3.4.1.4.1 requires at least one Residual Heat Removal (ND) loop be operable, or secondary side water level of at least two S/Gs be greater than 12%. This procedure specifies that during implementa-tion of this procedure, S/G 2A cannot be used to meet the Require-ments of Tech. Spec. 3 . 4 .1. 4 .1. It also requires that an additional ND loop be operable, or that the secondary side water level of two-S/Gs, other than 2A, be greater than 12% when Unit 2 is in Mode 5 with the Reactor Coolant System loops filled. This will ensure provisions are available for decay heat removal as required by Tech.-
Specs.
During implementation of this procedure, installation of the access hole in S/G 2A could affect Unit 2 Containment Integrity. This procedure requires that the isolations used during implementation of the procedure for S/G 2A meet the requirements for maintaining Con-tainment Integrity as specified by Tech. Specs.
This procedure requires that S/G 2A be returned to service prior to entry of Unit 2 into Mode 4. Restoring the S/G 2A secondary _ pressure boundary and performing a Post-Modification operability review will l
return S/G 2A to service and meet Tech. Spec. requirements. This will ensure S/G 2A is available to be used as a heat sink to support Unit 2 reactor operation.
Therefore, temporarily removing S/G 2A from service does not increase-the probability or consequences of an accident or equipment malfunc-tion previously evaluated in the FSAR. No new failure modes or operating characteristics are introduced by this procedure. All systems affected by this procedure wil_ be able to perform their intended functions. No USQs are created by this procedure.
TN/2/A/2568/CE/01A Initial-Issue This procedure will provide guidelines for replacing the time delay relays associated with Component Cooling (KC)' valves 2KC394A and 2KC413B. This time delay allows transients to occur (e.g. swapping 100 trains, start 3ng an additional KC pump) without causing valves to 25
l 1
close. This procedure will be implemented with Unit 2 in mode 5, 6, l or no mode with the reactor coolant (NC) pumps off. During the implementation of this procedure, power will be removed from valves 2KC394A and 2KC4138. These valves are the NC pump 2A and 2D thermal barrier outlet isolation valves, respectively. During the time this procedure is implemented, outlet flow from the thermal barrier is not essential because the NC pumps will not be running and the NC system will not be hot.
A complete function test of the affected circuits will be performed prior to returning the valves to service, including a train swap on KC, to verify the valves do not close. No USQ is created by this procedure.
MP/0/A/7600/54 Change #3 This procedure change adds steps to the procedure to provide instruc-tion for proper installation of the valve actuator. The change is editorial in nature and will not significantly affect the valve or system performance. Post-Maintenance Testing will be performed to verify equipment performance prior to return to service. The mainte-nance procedure was developed from the manufacturer's instruction manual, and providos guidelines for maintaining the valve within their original design specifications. Thus, there are no USQs aris-ing from this procedure change.
PT/0/A/4150/26 Change #1 The main changes to the procedure are 1) allow this activity to be done in parallel with PT/0/A/4150/18, Fuel Assembly / Insert Shuffle, and 2) incorporate the added concerns of the Justification for Con-tinued Operation / Operability Determination Evaluation for Auxiliary Building Ventilation, Control Area Ventilation, Annulus Ventilation, and Spent Fuel Pool Ventilation (VF) Systems. The 100 deg F maximum limit for VF is conservative relative to the design maximum tempera-ture of 110 deg F as stated in FSAR 9.4.2.1. For both items, no safety system will be placed in an off normal configuration. The margin of safety as defined in Tech. Specs, is enhanced. No unreviewed safety question is involved.
PT/2/A/4350/15A Change #7 This change revises the test method for the verification of the l Diesel Generator (D/G) overspeed trip. The test method is being
- changed so that the Unit 1 and Unit 2 procedures will be identical in the testing of the overspeed trip. The same equipment is being tested as before, but in a different manner. Jumpers are being placed across the same contacts as before, but in a different order.
The D/G will be inoperable during this test; however, D/G 2B will be operable. The D/G will no longer be taken to an actual overspeed condition to test the overspeed trip. The combination of this test with the Instrument and Electrical group testing (calibrations) 26
performed ensures that the overspeed trip circuitry works properly l without actually overspeeding the D/G. The margin of safety as I defined in the Tech. Spec. bases will not be reduced. No USQ is created by this change.
l PT/1/A/4206/06 Change /17 l
l This procedure change will delete the requirement to inspect rooms for leakage where the expected dose to personnel will exceed 100mR/hr. This administrative control will ensure that personnel dose remains hLARA. The probability and consequences for accidents evaluated in the FSAR are not increased since this test does not inspect for gross failure of equipment. Large leaks will be quickly identified by wster through the door or floor drain leak detection systems. There are no new accidents or malfunctions created by this change. Therefore, no USQ exists.
PT/1/A/4201/01 Change #25 A number of editorial changes were made. In addition, the Limit and Precaution concerning Safety Injection (NI) pump flow was deleted because there is no indication available to monitor NI pump flow I during this test. However, a miniflow path to and from the Refueling Water Storage Tank (FWST) is maintained and ensures that adequate NI pump flow is maintained at all times during the test. The Required Station Status Section 7.3 has been deleted. This section was in-tended to assure that FWST level remained above 37% during the test.
This is redundant, because Section 7.1 requires Reactor Coolant System pressure to be greater than 1900 psig during the test. In order to meet this requirement, this test can only be performed when in modes 1 through 4. Technical Specifications require FWST level i remain.above 92.3% when in modes 1 through 4. These two changes do I not involve any modifications to the system alignments or any other aspect of the test. Thus, they do not decrease the margin of safety.
Another change adds a new requirement to notify maintenance of any indication of boric acid corrosion identified during this test. This is a new reporting requirement, and does not affect the performance of the test. A)so, an administrative clarification was made to the acceptance criteria section to reflect that only one section of the procedure may have been completed. In addition, the existing Enclo-sure 13.1 footnote (*) was deleted. This footnote is being deleted because is does not apply to any valve in the enclosure. These three changes are of an administrative nature only, and do not involve any safety concerns.
The final change adds valve 1NI-441 to the Enclosure 13.1 list of valves to be placed in the open position prior to initiation of the -
test. Valve 1NI-441 was previously isolated during the performance of this test. This change will allow for testing of valve 1NI-98 for external system leakage. Valve 1NI-98 has been added to enclosure 13.4, NI system leak test worksheet. NI flow to the Waste Evaporator 27
Feed Tank Sump A and to the sample line downstream of valves 1NI-441 I and 1NI-98 remains isolated by valves 1NI-186 and 1NI-187, which are verified to be in the closed position per Enclosure 13.1. These i changes provide additional assurance that external NI leakage icvols l are as low as possible. They do not involve any accident initiation mechanism different than any already evaluated in the FSAR. They do not result in any new or different malfunctions of equipmenc impor-tant to safety. Thus, they do not involve a USQ. l l
PT/2/A/4200/08A Change #4 This procedure change expands the available modes in which the peri-odic test may be performed to include No Mode. No Mode full flow testing of the Safety Injection check valves will allow for earlier detection of any discrepancies and allow earlier corroution, if necessary. A review of Tech. Specs, and the FSAR have revealed no reason that parformance of the test in No Mode would hour any effect on nuclear safety. The only concerns for the operatica of the pump with no fuel in the core were strictly for adequate suction head for the pump. This is adequately addressed in OP/2/A/6203/0/. 9esidual Heat Removal System procedure, but was reinforced in tne F4. There-fore, no USQs are created by this change.
PT/0/A/4400/08 Change #51 This restricted change is being made so that the pump house balance may be performed on Nuclear Service Water (RN) pumps 1A and 2A with RN to the Containment Spray (NS) heat exchanger 2A isolated.- RN pump 1A or 2A will supply the flow requirements of the RN system during the test. This change allows 1RN291 and 2RN291 to be failed open during the test (open is their safety position). Both trains of RN will continue to be supplied during the test. All valves will remain operable, with the exception of 1RN291 and 2RN291. Thus, no unreviewed safety question is created by this test.
PT/0/A/4400/08 Change #52 Je s ricted change is being made so that the pump house balance e .rformed on RN pump 2A with RN to the NS heat exchanger 2A
.. RN pump 2A will supply the flow requirements cf the RN tw during the test. This change allows 1RN291 and 2RN291 to be open during the test (open is their safety position). Both
.. of RN will continue to be supplied during the test. All valveu will remain operable, with the exception of 1RN291 and 2RN291.
Thus, no unreviewed safety question is created by this test.
TN/2/B/0420/00/02A Chapoe #1 This change adds isolation steps necessary to isolate electrical penetration 2 PENT 0245, Plug 2. The new isolation will isolate sample 28
. _ __ __ . _ _ _ _ _ _ . _ _ ~ _ . , _ _ _ _ . .
lines for Radiation Monitors (EMF) 2 EMF 38, 39, and 40. This will make these EMPs inoperable, and will suspend all containment purge (VP) operations. Because of this, these isolations will be performed when VP is out of service. The isolations will be restored per this procedure and a voltage check will be_ performed to ensure the EMPs are fully operable again. As a result, this procedure will not increase the consequences of an accident as stated in the FSAR.
Accordingly, this procedure change will not create a USQ.
PT/2/A/4200/13A Change #13 This change adds steps to the enclosure which tests the Safety _Injec-tion (NI) pump Hot Leg isolation valves. The steps specify to close the NI Pump Cold Leg isolation valve for the train under test-prior to opening the hot leg valve. This change will prevent the possibil-ity of affecting the operability of the opposite train by eliminating the cross-train flow path to the Hot Leg which could-invalidate the Emergency Core Cooling System (ECCS) flow balance. This change alsc '
clarifies Limit and Precaution 6.2 which states that the train under test is rendered inoperable during the test. Technical Specifica-tions allow one train of ECCS to be inoperable for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Thus, this change does not increase the consequences of an accident as described in the FSAR. This change does not affect the NI pump's minimum flow path. The margin of safety in the bases to the Tech.
Spec. will not be reduced since the train under test will be inopera-ble and the Limiting Condition for operationLfor ECCS will be adhered to. Ne USQ exists.
HP/0/B/1000/10 Re-issue, Changes 0 to 12 Incorporated The Waste Monitor Tank (WMT) Building radiation monitors (EMF) are not addressed in the FSAR. The WMT Building Ventilation system is not a normal gaseous > effluent release pathway. The ventilation noble gas monitor was installed to acce,unt for the'small atonnts of en-trained noble gases that may be celeased during-liquid waste process-ing. This change provides a revision-to the WMT noble gas monitor l setpoint. This change is in accordance with' Tech. Spec. Section 3/4.3 and is intended only to more. accurately reflect t'Te setpoint.
Thus, operation of EMF 58 with the ner setpoint will not increa.me the l consequence of accidents previously evaluated in the FSAR or create i
the possibility of an accident. different than those previously evalu-ated, i
The setpoint concentration for EMF 58 (UMTB Ventilation Noble Gas l
Monitor) was recalculated. _The previoas setpoint was calculated using Unit Vent flow rate, providing excessively conservative values.
The new setpoint was corrected for actual ftow rate of the WMTB ventilation exhaust. This change still ensures Tech. Spec, limits for gaseous releases are not exceeded, but provides for a more rea-sonable setpoint. Steps were added to ensute personnel performing l setpoint adjustment noto current output module reading prior to changing setpoints and verify that the module returns to the current 29
reading following adjustment. The affect of this change is te pro-vide increased assurance that output modules are functional following setpoint adjustment. An example of sensitivity determination was added to Enclosure 5.4oto assist personnel in making operability determinctions. This example illustrates instructions previously incorporated in the procedure.
PT/0/A/4150/11A Re-issue This procedure measures the reactivity worth of RCCA banks using the boration/ dilution method. This method consists of initiating a Reactor Coolant (HC) boration/ dilution and moving the test bank in discrete steps to compensate for the reactivity effects due to the changes in boron concentration. This process is continued until.the entire bank is measured. This procedure is written to accommodate the testing on any RCCA bank. A Tech. Spec, special test exception allows the test to be performed on shutdown banks which are normally required to be fully withdrawn while critical. This same exception allows the test to be performed on Control Bank C which is normally-limited to an insertion limit of 47 steps at 0% full power. The surveillance requirements for this exception are that the reactor power must be limited to less than 5% full power, and T-avg be above 541 degrees F. In addition, Intermediate Range and' Power Range trip setpointo must be set to less than 25% full power.
The significant areas of safety concern are as follows:
- 1. Shutdown banks not all fully withdrawn. The Tech. Spec, special test exception provides protection by its limits on reactor trip setpoints.
- 2. Control Bank D below insertion limits. The Tech. Spec. special test exception provides protection by its limits on reactor trip setpoints.
- 3. Loss of Shutdown (S/D) margin. This test is performed under the direction of the Control Rod Worth Measurement by Rod Swap Test (PT/0/A/4150/11D). This test specifies the rod configuration and ensures that S/D margin is not violated.- During all physica testing, the requirements in Tech. Specs, concerning maintaining-shutdown margin are met.
- 4. Excessive rates of reactivity insertion. This test limits the rates to less than 40 pcm per rod movemenc and less than 400 pcm l per hour for boration and dilution. The state of core criti-cality is constantly monitored during this test by a reactivity computer. In addition, the flux level is monitored-and limited in this test to a small band just below the point of adding nuclear heat. This test' band is set in PT/1(2)/4150/21, Post Refueling Controlling Procedure for Startup Testing. These procedural limits, plus the available reactor trips (Intermedi-ate and Power Range high flux trips and Power Range high flux rate trips) prevent reactivity insertion accidents. The rate of 30
._. ._. __ _ .~ - - - - .
reactivity insertion due to rod withdrawal or 1-acrtion is inns than that assukod in the FSAR anolysis for the o.opped RCCA bank and uncontrolled bank withdrawal accidents. The rate of reac-tivity change due to the boron dilution is loss than that au-sumed in the FSAR for boron dilution accidents.
- 5. Adverse corn power distribution. Since this test involves rod mcyomonts, the power distribution in the core is affocted.
However, since this test is performed at zero power, this poses no safety concern. The reactor trip sotpoint reductions provide adequate protection.
This test is bounded by the uncontrolled rod withdrawal, dropped bank, and dilution accidents. (FSAR sections IS.4.1, 15.4.3, and 15.4.6). Thus, an unreviewed safety question does not 3xist.
TN/2/A/0396/01/08A Initial Issue Nuclear Station Modification (NSM) CN-20396, Rev. 1 modifies various piping system analyson with the objective of reducing the number of mechanical snubbers. This work unit provides guidance for the remov-al of pipe attachments and structural attachmants for snubbers delet-ed from a rigorous analysis model on the Auxiliary Feedwater (CA) system. (It covers math model CAD.) The snubborn were removed by tork unit 01. This work unit is merely a " clean-up" work unit. The work unit removes support steel, repairs abandoned concrete anchor holes, and touches up coatings for Support / Restraints (S/R) deleted by NSM CN-20396, Rev. 1. This work unit doco not affect or impact any equipment, system, or structure necessary to the safe operation of the plant. The S/R support steel is no longer physically attached to the CA and Main Feedwater (CF) cystems, and serves no function.
The pipe clamps are also removed by the work unit. The pipe clamps will have a negligible offect on the stress analysis of the CA and the CF systema, and it was not necessary to remove them under Work Unit 01. Therefore, removing the pipe clamps in this work unit has no affect on the CA and CF systems. For the above reasons o it is concluded that no USQ or safety concerns will arise as a result of the implementation of this work unit. TN/2/A/0396/01/09A Initial Issue NSM CN-20396, Rev. 1 modifica various piping system analyses with the i
objective of reducing the number of mechanical snubbers. This work l unit provides guidance for the removal of pipe attachments and-struc-tural attachments for snubberc deleted from a rigorous analysie m0 del on the CA system. (It covers math model CAF.) The snubbers ware removed by work unit 02. This work unit is merely a " clean-up" work unit. The work unit removes support steel, repairs abandoned con-crete anchor holes, and touches up coatings for Support / Restraints deleted by NSM CN-20396, Rev. 1. This work unit does not affect or impact any equipment, system, or structure necessary to the safe operation of the plant. The S/R support stecl is no longer physical-ly attached to the CA and CF systems, and serves no function. The pipe clamps are also removed by the work unit. The pipe clamps will 31
have a negligible effect on the stress analysis of the CA and the CF systems, and it was not necessary to remove them under Work Unit 02.
Therefore, removing the pipe clamps in this work unit has no effect on the CA and CF systems. For the above reasons, it is concluded !
that no USQ or safety concerns will arise as a result of the imple-mentation of this work unit.
i TN/2/A/0396/01/10A Initial Issue i NSM CN-20396, Rev. 1 nodifies various piping system analyses with the objective of reducing the number of mechanical snubbers. This work unit provides guidance for the removal of pipe attachments and struc-tural attachments for snubbers deleted from a rigorous analysis model on the Residual Heat Removal (ND) system. It covers math model NDE.) The snubbers were removed by work uni (t 03. This work unit is merely a " clean-up" work unit. The work unit removes support steel, I repairs abandoned concrete anchor holes, and touches up coatings for 1 Support / Restraints deleted by NSM CN-20396, Rev. 1. This work unit does not affect or impact any equipment, system, or structure neces-sary to the safe operation of the plant. The S/R support steel is no longer physically attached to the ND and Refueling Water (FW) cys-tems, and serves no function. The pipe clamps are also removed by the work unit. The pipe clamps will have a negligible effect on the stress analysis of the ND and the YW systems, and it was not neces-sary to remove them under Work Unit 03. Therefore, removing the pipe clamps in this work unit has no affect on the ND and FW systems. For-the above reasons, it is concluded that no USQ or safety concerns will arise as a result of the implementation of this work unit.
TN/2/A/0396/01/12A Initial Issue NSM CN-20396, Rev. 1 modifies various piping system analyses with the objective of reducing the number of mechanical snubbers. This work unit provides guidance for the removal of pipo attachments and struc-tural attachments for snubbers-deleted from a rigorous-analysis model on the ND system. (It covers math model NDA.) The snubbers were removed by work unit 05. This work unit is merely_a " clean-up" work .
unit. The work unit removes cupport steel, repairs abandoned con-crete anchor holes, and touches vn coatings for Support / Restraints deleted by NSM CN-20396, Rev. 1. This work unit does not affect or impact any equipment, system, or structure necessary to the safe operation of the plant. The S/R support steel is no longer physical-ly attached to the HD system, and serves no function. The pipe clamps are also removed by the work unit. The pipe clamps will have a negligible effect on the stress analysis of the ND system, and it was not necessary to remove them under Work Unit 05. Therefore, removing the pipe clamps in this work unit has no affect on the ND system. For the. above reasons, it is concluded that no USQ or safety concerns will artae as a result of the1 implementation of this work unit.
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TN/2/A/0396/01/13A Initial Issuo NSM CN-20396, Rev. 1 modifies various piping system analyses with the objective of reducing the number of mechanical snubbers. This work unit provides guidance for the removal of pipe attachments and struc-tural attachments for snubbers doloted from a rigorous analysis model on the Main Steam Vent to Atmosphoro (SV) system. (It covers math model SVA.) The snubbers were removed by work unit 07. This work unit is merely a " clean-up" work unit. The work unit removes support steel, rcpairs abandoned concrete anchor holes, and touches up coat-ings for Support / Restraints delettd by NSM CN-20396, Rev. 1. This work unit does not affect or impact any equipment, system, or struc-ture necessary to the safe operation of the plant. The S/h support steel is no longer physically attached to the SV system, and serves no function. The pipo clamps are also removed by the work unit. The pipo clamps will have a negligibic offect on the stress analysis of the 0 - system, and it was not necessary to remove them under Work Unit 07. Thorofore, removing the pipo clamps in this work unit has no affect on the SV system. For the above reasons, it is concluded 1 that no USQ or safety concerns will arico as a result of the imple- l montation of this work unit.
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i Catawba Nuclear Station Summary of Procedure Changes, Tests, and Experiments i, Completed from 11/1/89 to 9/30/90 -- Volume 5 TH/2/A/0618/00/01A Initial Issue 4 This procedure provides implementation instructions for Nuclear Station Modification (NSM) CN-20618, Rev. O, Work Unit 1. NSM CN-20618 will install a bypass check valve around Safety Injection l (NI) Valve 2NIO95A for overpressurization protection of Penetration H322. The check valve will afford leakoff when the penetration i
isolation valves close in the event of a LOCA, thereby preventing 1
overpressurization due to thermal expansion. The purpose of this j procedure is to provide guidance for installing valve 2NI471 with i valve item number 9J-356 and associated piping. It will also assure
- Post Modification Testing is completed in order to maintain system J integrity. An unreviewed safety question (USQ) is not created by the implementation of this procedure.
T7/2/A/9200/59 Initial Issue This procedure involves a test not addressed in the FSAR. This test requires that the Chemical and Volume control (NV) System is operating in normal letdown and charging mode. During normal operation, valves 2NV294 and 2NV309 are approximately 50% open to maintain stable NV and Reactor Coolant (NC) system operation. This test instructs the operatora to fully open those valves. (These valves are " fail open" valves.) Fully opening 2NV294 and 2NV309 will result in an increase in charging flow with some chan9a in NC pump seal flow rates. The increase in charging flow will result in increasing pressurizer level
- and decreasing Volume Control Tank (VCT) level. These transients can j be mitigated by opening both the 45 and the 75 gpm Ictdown orificas if the Control Room Operator (CRO) so chooses, per the guidance in this procedure and in accordance with the NV operating procedure. The subject procedure provides limits and precautions to the operator.
Pressurizer and VCT minimum and maximum level, pressurizer temperature l rates of change, NC system temperature rates of change, and NC pump minimum and maximum seal injection limits are outlined in the proce-dure. If any of these limits are exceeded, the procedure instructs the Test Coordinator to abort the test and instruct the CRO to manipu-late the above mentioned vslves to achieve desired system conditions.
These limits are within alurm setpoints for pressurizer and VCT levels and within administrative limits for temperature rates of change for i the pressurizer and NC system. The allowable seal flows are within normal operational limits for the individual NC pumps and ensure that the Tech. Spec. total seal flow rate criterion is not exceeded. In the required mode for this test (mode 3), all pressurizer Power Operated Relief Valves (PORVs) are required to be operable in accor-dance with Tech, bpocs. No_ Emergency Core Cooling System (ECCS) subsystem will be rendered inoperable as a result of the test. Since the charging' pump till essentially be operating in recirculation
(
I 1
through the NC system with no boron dilution occurring (no water from the Reactor Makeup Water Storage Tank being added), and since adequate shutdown margin in ensured at mode 3 boration levels at 0 Cycle Burnup (EFPD) to below mode 6 NC temperatures, any shutdown margin concerns are precluded. The initial conditions for a design basis accident are not being altered by this test. Since this test does not impact accident initiation nachanisms, the probability or consequences of an accident are not increased. This test does not involve any hardware modifications or operation outside normal limits, and the possibilits of an accident which is different than any already evaluated in the FSAR is not created. This test does not adversely affect any equip-ment important to safety. No new malfunction of equipment important ;
4 to safety different than any evaluated in the FSAR is created. The margin of safety as defined by the Tech. Specs, is not being affected by the performance of the test. Based on the above discussion, it is determined that this test does not involve a USQ.
TN/1/A/0067/01/02A Initial Issue This procedure provides implemantation instructior.: for NSM CN-10067, Rev. 1 (Part 2) work unit 2. NSM CN-10067, Rev. 1 provides a closed systam for the Unit 1 Post Accident Liquid Sample (PALS) panel effluent in the event of an accident. This system will consist of a positive displacement sump pump and the associated piping and valves required to return the effluent to the Volume control Tank (VCT).
This modification will be implemented in parts, with portions of the piping being installed during Unit 1 End of Cycle 4 Refueling Outage, and the remainder of the piping, along with the pump and the associat-ed electrical activities installed non-outage. This procedure pro-vides guidance for mechanical and electrical activities (Part 2) which will be performed non-outage. This procedure will install the sump pump, associated valves and supports, electrical modifications, and make tie-ins to the piping installed under Part M1.
In order to perform this procedure, the Unit 1 PALS panel sump pump must be isolated, and the Unit 1 PALS panel removed from service. The PALS system is part of the Nuclear Sampling (NM) system.which provides various system sampling capabilities. The PALS system allows sampling of the Reactor Coolant (NC) and Residual Heat Removal (ND) systems and the containment sump during accident conditions for laboratory analysis. With the PALS panel out of service, the NC and ND system could be sampled through tL9 NM system. Therefore, in the event of an accident during implementation of _ this procedure, samplir.g capabili-ties are available.
This procedure can be performed during any Mode of Unit 1 operation.
The PALS sump pump discharge piping will be tied into the VCT sample recirculation piping to provide a closed system for the PALS effluent. An isolation valve is available at the tie-in point.
4 Therefore, the VCT does not have to be drained to allow implementation of this procedure. Only the PALS panel sump pump and the PALS panel must be removed from service. The remaining portions of the NM system '
2 l
l will not be affected by this procedure. No other system must be removed from service to allow implementation of this procedure.
After installation, the piping will be visually inspected at system temperature and pressure to verify piping integrity. Also, sump pump flow to Waste Evaporator Feed Tank (WEFT) Sump A and the VCT will be verified to ensure the modification performs as designed. Continuity checks will be performed on sll new and revised circuits to ensure operability. After completion of this procedure, the PALS sump pump discharge piping pressure boundary will be restored and the PALS panel returned to service.
Since sampling capabilities are available during implementation of this procedure, the margin of safety as defined in bases of the Tech.
Specs will not be reduced. Based on the considerations above, a USQ is not created by this procedure.
TN/2/A/0073/01/02A Initial Issue This procedure provides implementation instructions for NSM CN-20073, Rev. 1 (Part 2) work unit 2. NSM CN-20073, Rev. 1 provides a closed system for the Unit 2 Post Accident Liquid Sample (PALS) panel effluent in the event of an accident. This system will consist of a positive displacement sump pump and the associated piping and valves required to return the effluent to the Volunie Control Tank (VCT).
This modification will be implemented in parts, with portions of the piping being installed during Unit 2 End of Cycle 3 Refueling Outage, and the remainder of the piping, along with the pump and the associat-ed electrical activities, installed non-outage. This procedure provides guidance for mechanical and electrical activities (Part 2) which will be performed non-outage. This procedure will install the sump pump, associated valves and supports, electrical modifications, ,
and make tie-ins to the piping installed under Part M1.
In order to perform this procedure, the Unit 2 PALS panel cump pump l
must be isolated and the Unit 2 PALS panel removed from service. The PALS system is part of the Nuclear Sampling (NM) system which provides various system campling capabilities. The PALS system allows sampling of the Reactor Coolant (NC) and Residual Heat Removal (ND) systems and the containment sump during accident conditions for laboratory analy-sin. With the PALS panel out of service, the NC and ND system could be sampled through the NM system. Therefore, in the event of an accident during implementation of this procedure, sampling capabili-ties are available.
This procedure can be performed during any Mode of Unit 2 Operation.
The PALS sump pump discharge piping will be tied into the VCT sample recirculation piping to provide a closed system for the PALS effluent.
An isolation valve is available at the tie-in point. Therefore, the VCT does not have to be drained to allow implementation of this procedure. Only the PALS panel sump pump and the PALS panel must be removed from service. The remaining portions of the NM system will not be affected by this procedure. This procedure also requires 3
certain firestops to be opened to allow rerouting of electrical cable.
These firestops will be opened and closed in accordance with station Tech. Specs. to ensure firestop integrity.
After installation, the piping will be visually inspected at system temperature and pressure to verify piping integrity. Also, sump pump flow to Waste Evaporator Feed Tank (WEFT) Sump B and the VCT will be verified to ensure the modification performs as designed. Continuity checks will be performed on all new and revised circuits to ensure operability. After completion of this procedure, the PALS sump pump discharge piping pressure boundary will be restored and the PALS panel returned to service.
Since sampling capabilities are available during implementation of this procedure, the mergin of safety as defined in babss of the Tech.
Specs, will not be reduced. Based on the consideration 2 above, a USQ is not created by this procedure.
TN/2/A/2765/CE/01A Initial Issue This procedure provides for implementation of Exempt Change CE-2765.
This Exempt Change will reposition the operating elevation of the fully withdrawn Rod Control cluster Assemblics (RCCAs) from 230 steps to 225 steps. In addition, the bank overlap will be changed to 112 steps. This is being dono due to a wear problem on the RCCAs. This will be accomplished by changing the thumbwheel settings in the Rod control System logic cabinets. No isolations are required by this procedure and the changes will be made while Unit 2 is in Modes 3, 4, l 5, 6, or No Mode. Functional verification will be done by both Instrument and Electrical (IAE) and Performance. The functional verification will consist of calibrations, rod drop time testing, overlap check, and insertion limit checks. This testing has been determined to adequately test the nodification. A revision to the Core Operating Limits Report (COLR) will be made and issued by Design Engineering. Implementing this procedure will have no effect on Technical Specifications. An unreviewed safety question will not be created during the implementation of this procedure.
OP/1/A/6700/01 Change #187 This change replaces page 1 of Table 2.2 to incorporate the Power Range Nuclear Instrumentation System full power calibration currents generated per PT/1/A/4600/05A. This table is used to record the 100%
full power calibration currents (at axial offsets of +20%, 0%, and
-20%) and the M Factors for each of the power range excore detectors.
Data is obtained for this table only by approved procedure. The data recorded on this table is used by IAE to adjust the Axial Flux Differ-ence (AFD) calculating circuitry and Operator Aid Computer (OAC) programs. It may also be used to manually calculate AFD and Quadrant Power Tilt Ratic- (QPTR) if the OAC is inoperable.
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Since AFD is used to dynamically adjust both the overtemperature-Differential Temperature (OTDT) and Overpower Differential Temperature (OPDT) Setpoints, the data herein is safety-related. This change affects no other equipment than the Power Range Nuclear Instrumenta-tion System (NIS). This change does not constitute a USQ.
OP/1/A/6700/01 Change #188 This change replaces all 4 pages of Figure 1.1 to incorporate the new Axial Flux Difference Target Values obtained per PT/1/A/4150/08, Target Flux Difference Calculation. The target AFD is-changed to keep control Bank D at approximately 215 steps withdrawn at 100% full power while allowing for changes in the notural axial power _ shape that occur with burnup. The targets are set by procedure to-be within-the-operating bounds of Tech. Spec. 3/4.2.1. The. targets are an operating guideline only to aid the control room operators in maintaining the 4 AFD with the limits of Tech. Spec. 3/4.2.1. The targets serve no other purpose. They do not feed any trip function or serve any safety related functions. The limits that must be observed in Tech. Spec.
3/4.2.1 are set by cycle specific analysis.
AFD is an input to the OTDT Trip Setpoint. Targets may be. set so that the AFD input function to the OTOT will impose a penalty on OTDT.
However, this penalty will automatically be imposed by the 7300 system per=the formulation of Tech. Spec. 2.2.1. AFD and AFD Targets are I monitored the the OAC NUCLEAR 06 (AFD ALARM NONITOR) program. AFD Targets on the OAC are changed by the above mentioned procedure as well.
All accidents analyzed in FSAR Chapter 15 have as one of the initial conditions that the AFD is within the limits of Tech. Spec. 3/4.2.1. i As such, the targets do not have any effect on the accident analysis.
OP/1/A/6100/09A Change #20 This change places a step in the Immediate Action Section which reads:
- 1. Fuel 011 Filter maximum allowable DP is 25 paid.' Ensure supplementary actions.are completed before reaching 25 psid.
This change gives the operators the-maximum allowable differential pressure value for the Diesel Generator Engine Fuel 011 (FD)-filter so that the equipment safety determination for a Diesel Generator (D/G) with the High Differential Procsure (DP)-Fuel Filter alarm is more readily determined. CNM 1301.99-237 Volume III, TransAmerica DeLaval (TDI) Diesel Engine Manual, references the-maximum allowable-fuel' oil filter-differential pressure as 25 psid. .The high DP Fuel Filter alarm is conservatively set below that value at 20 paid. The Mainte-nance Engineering Services (MES) Engineer for the Diesel Generators, in agreement with the TDI engineers and.the:above mentioned D/G Manual,. approved the operability of the D/G with a fuel filter DP-above the alarm point, as-long as the 25 psid maximum was adhered to.
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i The operation of the D/G with the alarm actuated does not create any new safoty connidoratior.c since the alarm was designed to actuate at a lower than maximum DP, and time was allowed for the operator to take -
corroctive uction before reaching a critical differential pressure.
(TDI gavw the valve of 35 paid.) Therefore, no unreviewed safety
- question is created by this-procedure change.
OP/1/J./6100/09B Change #19 This change places a step in the Immediate Action Section which reads:-
- 1. Fuel Oil Filter maximum allowable DP is 25 paid. Ensure supplementary actions are completed before reaching 25 paid.
This change give: the operators the maximum allowable differential pressure value for the FD filter so that the equipment safety determi- >
nation for a D/G with the High DP Fuel Filter alarm is more readily determined. CNM 1301.99-237 Volume III, TransAmerica DeLaval Diesel Engine Manual, references the maximum allowable fuel oil filter differential pressure as 25 psid. The high DP Fuel Filter alarm is conservatively sat below that value at 20 psid. The MES-Engineer for the Diesel Generators, in agreement with the TDI engineers and the above mentioned D/G Manual, approved the operability of the D/G with a fuel filter DP above the alarm point, as long as the 25 paid maximum .
was adhered to. The operation of the D/G with the alarm actuated does not create any new safety considerations since the alarm was designed to actuate at a lower than maximum DP, and time was allowed for the operator to take corrective action before reaching a critical differential pressure. (TDI gave the value of 35 paid.) Therefore, no unreviewed safety question is created by this procedure change.
OP/2/A/6100/09A Change #15 This change places a step in the Immediate Action Section which reads:
- 1. Fuel Oil Filter maximum allowable DP is 25;psid. Ensure
- supplementary actions are completed before reaching 25 paid.
This change gives the op rators the maximum allowable differential pressure value for the FD filter so that the equipment safety _determi-nation for a D/G with the High DP Fuel Filter alarm is more readily determined. CNM 1301.99-237-Volume III, TransAmerica DeLaval Diesel Engine-Manual, references the maximum allowable-fuel oil filter differential pressure as 25 paid. The high.DP Fuel Filter alarm-is conservatively set below that value at 20 paid. The-MES Engineer for the Diesel Generators, in agreement with the'TDI engineers and'the above mentioned D/G Manual, approved the operability of the D/G-with a l fuel-filter DP above the alarm point, as long as the 25 psid maximum was adhered to. The operation of the D/Glwith-the alarm actuated does not create any new safety considerations since the alarm was L
' designed to actuate at a lower than maximum DP,-and time was-allowed far the operator to take corrective action- before reaching a critical 6
- - - . _ _ _ _ _ _ _ ..~. _. __ _ .. a _ _ _ _ . , _ . _.-._a.., . . _ _ . ~ _ , _
differential pressure. (TD2 gave the value of 35 paid.) Therefore, no unreviewed safety question in created by this procedure change.
i i OP/2/A/6100/09B Change #14 This change places a step in the Immediate Action Section which reads:
i
- 1. Fuel Oil Filter maximum allowable DP is 25 psid. Ensure supplementary actione are completed before reaching 25 paid. ]
1
! This change givus the operators the maximum allowable differential-
)
pressure value for the FD filter so that the equipment safety determi-nation for a D/G with the High DP Fuel-Filter alarm is more readily determined. CNM 1301.99-237 Volume III, TransAmerica DeLaval Diesel Engine Manual, references the maximum allowable fuel oil filter j differential pressure as 25 paid. The high DP Fuel Filter alarm is conservatively set below that value at 20 psid. The MES Enginear for the Diesel Generators, in agreement with the TDI engineers and the above mentioned D/G Manual, approved the operability of the D/G with a fuel filter DP above the alarm point, as long as the 25 paid maximum was adhered to. The operation of the D/G with the alarm actuated does not create any new safety considerations since the alarm was designed to actuate at a lower than maximum DP, and time was allowed
! for the operator to take corrective action before reaching a critical differential pressure. (TDI gave the value of 35 paid.) Therefore, i no unreviewed safety question is created by this procedure change.
PT/1/A/4200/28A Change #13 i
This change adds Fire Protection (RF) valve 1RF457B to the lineup for stroking valve IRF447B. The lineup calls for 1RF457B to be closed.
This isolates RF flow to the annulus. RF flow to the annulus is isolated before closing 1RF457B by closing 1RF859. Therefore, closing i 1RF457B is of no consequence as.long as it is opened after testing is complete. This change adds steps to record the "As Found" position of 1RF457B and to return the valve to the "As Found" position. The time-that the header must be isolated is decreased by this change because drain time is decreased. There are no USQs resulting from this change.
l l PT/2/A/4200/28A Change #6 i
This change adds Fire Protection (RF) valve 2RF457B to the lineup for stroking valve 2RF447B. The lineup calls for 2RF457B to be closed.
This isolates RF flow to the annulus. RF flow to-the annulus is-isolated before closing 2RF457B by closing 2RF859. Therefore, closing 2RF457B is of no consequence as long as it is opened after testing-is complete. This change adds steps to record the "As Found" position of '
2RF457B and to return the valve to the "As Found" position. The time that the header must be isolated is decreased by this change because 7
drain time is decreased. There are no USQs resulting from this change.
i PT/2/A/4200/13E Change #24 This restricted change is being made to allow for the retest for Nuclear Station Modification (NSM) CN-20567. This change will allow stroke time testing of Auxiliary Feedwater (CA) valves 2CA58A, 2CA62A, 2CA42B, and 2CA46B under full differential pressure. These valves were replaced during the NSM. All sections of this change require that Unit 2 be in modes 4, 5, 6, or No Mode. Since the auxiliary feedwater system is required only in Modes 1, 2, and 3, no accident scenarios are impacted. Also, the margin of safety as defined in the Design Bases is not decreased. Water will be injected into the Steam Generators from the Auxiliary Feedwater system, which will have no effect on plant safety. Precautions are given in this change to ensure that the Steam Generator levels stay below 70% narrow range.
PT/2/A/4200/34 Change #8 This change revises the IWV stroke time value for containment Purge (VP) valves to 10 seconds to eliminate the need to maintain the valves with an unnecessary limiting value of 5 seconds. IWV-3410 requires valves to be cycled and stroke time tested quarterly. Technical Specification 3.6.1.9 places severe restrictions on the alignment and operation of the VP system. All VP valves are administratively closed (power removed) during Modes 1 through 4. The valves are never opened I
in modes 1 through 4, and, therefore, are not timed to close except in Modes 5, 6, or No Mode. During cold shutdown, the VP valves mue'. be full stroke exnrcised, stroke timed, and fail safe operation verified.
l Because Tech. Spec. 3.6.1.9 does not apply during cold shutdown, the 5 l second stroke time criteria also does not apply. During Moden 5, 6, and No Mode, the only requirements for these valves are for contain-ment closure during mid-loop operation and Tech. Spec. 4.6.3.2 for automatic closure on high radiation and high relative humidity. The isolation time required for Tech. Spec. 3.6.1.9 does not apply to j
these testing requirements. Table 3.6.2B times are limiting values that ensure that the release of radioactive material to the environ-ment will be consistent with the assumptions used in the analysis for a LOCA. Containment closure requires isolation within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and Tech. Spec. 4.6.3.2 does not specify a closure time. Therefore, to detect valve degradation, the IWV stroke time will be measured with an acceptance criteria of 10 seconds. The average stroke time for the VP valves is 3.2 seconds. The limiting value of full stroke time will be set by multiplying the average stroke time (reference) by 2.25 and rounding up to the nearest 5 seconds. Therefore, the limiting value of full stroke time for the VP valves will be set at 10 seconds for cold shutdown testing. This value will cover all of the VP valves under normal conditions. The required action for stroke time increases is also changed. Trending will be evaluated on a case-by-case basis.
The evaluation will be performed by the responsible engineer. The VP system is only operated in Mode 5 or lower, and timing the valves to l 8 I
l close within 10 seconds will ensure containment closure for high radiation and high relative humidity.
PT/2/A/4450/03C Change #9 This restricted change was written to allow performance of the annulus vacuum decay time test, in response to the HVAC Review Committee's findings, pIlt 2-C90-0072, and incorrect error adjustment to the acceptance criteria. The vacuum decay time test measures the time that it takes the annulus pressure to decay from -3.5 inches of water to -0.5 inches of water with Annulus Ventilation (VE) shutdown. The purpose of this test is to ensure less than 2000 scfm air in-leakage into the ar.nulus during post-LOCA operation of VE. The 2000 scfm in-leakage is the assumed value used in CANVENT and ACTOS to calculate dose for the post-LOCA conditions. This test is not required directly by Tech. Specs., but in-leakage outside the above mentioned valuo may increase the dose above the appropriate 10CFR/FSAR limits.
s Section 12.9 was re-written so that the vacuum decay time test may Le repeated for varioua alignments of the backdraft dampers, unit vent isolation dampers, and access doors on filter units for each train of VE. A340, Spent Fuel pool Ventilation (VF) on Unit 2 will be shutdown during tno test, and A Train of Auxiliary Building Ventilation (Vh) may be aligned into the post-LOCA mode of operation while B Trair of VA is shutdown. Aligning the backdraft dampers and filter unit doors will be accomplished by physically tying the component open. The unit vent isolation dampers will be aligned by placing the switch in the Control Room into the desired position. One momentary jumper is placed in 2DGLSA-2 that initiates the signal to the A Train VA, This will align both Unit 1 and Unit 2 VA A trains in the post-LOCA align-ment. If an accident occur: auring this period on Unit 1, VA will function as designed because the LOCA alignment will already be established for VA (A Train). VF is only required to be operable during fuel movement or with any overhead load abovl the fuel pool. A prerequisite to the test ensures that there is no work in p7 ogress or planned during the performance of this test that may requi'te the operability of Unit 2 VF. Another prerequisite to the ter,t requires that Unit 2 be in Modes 5 or 6. Therefore, VA and Unit L VF may be aligned as necessary for this test without increasing the consequences of an accident. The above will test the pressure boundary of the annulus for all possible post-LOCA VE operating alignments (assuming a single failure) and determining if normal routine work performed on the filter units affects system operability.
Section 12.10 was added as a system restoration section after the 3ast test run is performed under section 12.9. This section will ensure that all dampers, access doors, failed transmitters, installed instru-mentation and other ventilation systems are returned to " normal."
Acceptance criteria are also included in section 12.10. If any of the vacuum decay test runs should fail, Design Engineering will be con-tacted for an operability determination. Because the aligning of-the backdraft dampers may affect the flow, section 12.10 requires that flow on A and B train of VE be checked. This will ensure that the 9 l l
mm. .
flow for VE remains within the required limits after testing is complete. This section is required by this change to be performed after the last test run of section 12.9. Old section 12.10 was changed to section 12.11 and steps were renumbered.
The limits of the decay test were error adjusted incorrectly on past tests, and this change implements the correct error adjustment which will give more conservative results. Another part of this change reissued a previously approved change in order to perform the drawdown test using the correct values after implementation of Exempt Change CE-2467. The safety evaluation for this reissued change is still valid. Additional changes were made to include failing both the A and B Train transmitters so that either train may be used to pull the negative pressure on the annulus. Any alignment of VF will be conduct-ed under the appropriate Operations procedures. The VA system will be aligned using the Operations procedure except when the signal is simulated to establish the post-LOCA mode. If VA is aligned to the post-LOCA mode, it will be returned to normal operation by section 12.9 of this change.
Unit 2 is required by this change to be in Mode 5 or 6. VE is only required to be operable in Modos 1 to 4 by Tech. Spec. VE is not and docs not affect an accident initiator. All components are indepen-dently verified to be returned to the normal position after the last test run, and parameters that may have been affected are required to be checked. For these reasons, this procedure change does not create a USQ.
PT/0/A/4400/05 Change #5 The catch basins function to ensure sufficient flow off the power house yard in the event of Probable Maximum Precipitation. Two types of catch basin inlets are used. Type II inlets consist of steel grating covers over the catch basins, with an area approximately equal to the area of the pipe inlet. Type I inlets are protected by steel grating on the top and four sides. The area of the Type I inlets varies from 4 to 15 times the area of the pipe inlets, virtually eliminating the posnibility of complete blockage. Design analysis was done assuming 100% blockage of the type II inlets. Since only the type I inlets were considered, they are the only inlets that need to be inspected. Enclosure 13.1 has been revised to agree with table 2.4.2-2 in the FSAR.
The yard drainage system does not directly affect the safe shutdown of the plant. This action creates no new accident scenarios. This is only an inspection of the catch basin inlets. Also, the catch basin i inlets are not discussed in the Tech. Spec. Thus, there is no USQ l created by this procedure change.
PT/1/A/4400/06B Change /7 10
Step 8.9 in the procedure was changed to ensure that the FWST heater setpoints are adjusted to maintain a FWST temperature of 93 dog F.
This temperature will ensure that a sufficient temperature differen-tial exists between the Refueling Water Storage Tank (FWST) and Nuclear Service Water (RN) to perform the heat capacity test. The j setpoints will be returned to normal upon the completion of the test. .
Tech. Specs, specify that the temperatura must remain between 70 and !
100 deg F. Since the temperature will romain within the allowable limits as specified in the Tech. Spec., the margin of safety will not be reduced. Thus, no USQ is created by this change.
PT/2/A/4400/06B Change #4 Step 8.9 in the procedure is changed to ensure that the FWST heater setpoints are adjusted to maintain a FWST temperature of 93 deg F.
This temperature will ensure that a sufficient temperature differen-tial exists between the FWST and RN to perform the heat capacity test.
The setpoints will be returned to normal upon the completion of the test.
Tech. Specs, specify that the temperature must remain between 70 and 100 deg F. Since the temperature will remain within the allowable limits as specified in the Tech. Spec., the margin of safety will not be reduced. Thus, no USQ is created by this change.
PT/2/A/4350/15B Change #7 This change alters the test method for the verification of the overspeed trip. The test method is being changed so that the Unit 1 and Unit 2 procedures will be identical in the testing of the overspeed trip. Neither-the probability or the consequences of an accident will be increased by this change. The same equipment is being tested but in a different manner. Jumpers.are being placed across the same contacts as before, but in a different order. The Diesel Generator (D/G) will be inoperable during the test; however, the 2A D/G will be operable. The Diesel Generator will no longer be taken to an actual overspeed condition to test the overspeed trip.
l The combination of this test with Instrument and Electrical testing (calibrations)- performed ensures that the overspeed trip-circuitry works properly without actually.overspeeding the D/G. No USQ is created by this procedure change.
PT/2/A/4200/13E Change #25 i
This restricted procedure change allows for stroke time testing of Auxiliary Feedwater (CA) Valve 2CA46B under full-differential pres-sure. This test is required in order to fulfill the Retest require-ments of Nuclear Station Modification (NSM) CN-20567. All sections of this procedure change require that the unit be in Modes 4, 5, 6, or No Mode. Since the CA system is required only in modes 1, 2, and 3, no l accident scenarios are impacted. Water will be injected into-the 1
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,-_ _ _ - _ _ - ~- _ _.- __ _ _ _ . _._
steam generators from the CA system, which will have no effect on plant safety. Precautions are given in thic procedure change to ensure that the steam generator levels remain below 70% narrow range.
Thus, no USQ is created by this change.
PT/2/A/4350/02A Change #23 This restricted change deletes the step which requires that the Diesel Generator (D/G) be allowed to cool to less than 190 deg F and a minimum of 5 minutes before tripping the D/G. This change is being made to perform two tests simultaneously for D/G 2A. The ESF test, in order to meet Tech. Spec. 4.0.1.1.2 g(15), requires a full load run of the D/G for one hour. The D/G is then shut down, a five minute wait elapses, and then the D/G is restarted. This is to verify the hot restart ability of the engine. The normal, monthly Periodic Test (PT) runs the D/G for a one hour period. The D/G is then unloaded, allowed to cool to less than 190 deg F on the jacket water and lube oil systems, and then is chut down.
In order to save engine run time, it was decided to run both tests simultaneously. To accomplish the ESF test, the normal engineering practice of a cooldown period after a run to even out temperatures on the D/G, required by PT/2/A/4350/02A, must be waived. This require-ment will be reinstated before the next performance of the PT. All intents and purposes of running PT/2/A/4350/02A are met by conducting the PT simultaneously with the ESP testing. No increased potential for failure of, or damage to, equipment important to safety is created by this procedure change.
PT/2/A/4250/03D Change /18 There are several portions of the procedure affected by this change.
- 1) Section 5 has been changed to add test equipment which had been l left off in the past. This change ensures that the procedure can be followed accurately and with minimal problems.
- 2) Several steps in section 12.2 were changed for clarification.
This change ensures that the procedure can be followed accurately and with minimal problems.
- 3) Notes were added because the Auxiliary Feedwater (CA) Auto Start Defeat will not go into DEFEAT mode when the Solid State Protec-tion System (SSPS) is in TEST. This ensures that the affected procedure steps can be completed under varying system configura-tions (i.e., SSPS in TEST or NDRMAL.)
t
- 4) Steps were added to open the CA pump motor breakers. These steps are needed becabse the appropriate motor breakers close on a simulated motor driven pump autostart. These items ensure that I
the motor breakers are returned to the open position. Note that the~ motor breaker for the pump receiving the simulated autostart 12
is " racked to test," while the other pump is " racked out."
Alignments ensure that both motor breakers were returned to their "as found" positions following test completion.
Since the CA system is not required in Modes 4, 5, or 6, no accident scenarios are impacted. Thus, no USQ is created by this procedure change.
PT/0/A/4150/11B Initial Issue This procedure is used to verify that the reactivity worth of the reference bank, as determined through reactivity computer measurement data, is consistent with design predictions. (The reference-bank is the bank which has the highest predicted reactivity worth when insert-ed into an otherwise unrodded core.) This procedure also verifies that the reactivity worth of each control and shutdown bank (except the reference bank), as inferred from data following iso-reactivity interchange with the reference bank, is consistent with design predic-tions. A description of this procedure is included in the FSAR (Section 14.3.2.3). l The available reactivity insertion will remain with the assumptions described in FSAR section 15.0.5. The analyses in Section 15.4 1 (Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low Pv.er Startup Condition), Section 15.4.3 (Rod cluster Control Assembly Minoperation), and Section 15.4.8 (Spectrum of RCCA Ejection Acci-dents) are bounding and unaffected by this procedure. Manipulations of control rod banks are performed with rods in MANUAL (bank select) by a licensed operator using normal operating procedure.
The restrictions of Technical Specification 3.10.3 ensure that, although operations is outside Tech. Specs. 3.1.3.5 and 3.1.3.6, the margin of safety provided by these specifications is maintained.
In summary, this procedure does not create an unreviewed safety question because:
- 1. acceptable power distribution limits are maintained by limiting the test to zero power, t 2. the minimum shutdown margin is maintained, as confirmed by a calculation in PT/1(2)/A/4150/ 1, and
- 3. the potential effects of rod misalignment are limited by limiting the test to zero power and by moving banks of RCCAs, not individ-ual rods.
PT/0/A/4150/12A Initial Issue This test determines the isothermal temperature coefficient of reac-tivity, and is used to derive tha moderator temperature coefficient of reactivity from the isothermal data. FSAR section 14.3.3.2 describes this test'. The measurement _of the temperature coefficient of J 13
i reactivity of the core is performed at the beginning of each fuel cyclo during zero power physics testing. Reactor coolant (NC) tempora-ture is slowly decreased whiln reactivity is plotted against tempora-turo. Tho same process is repeated during subsequent heatup. The average slope for the test is the isothermal temperature coefficient.
From this value, the moderator temperature coefficient is calculated.
This value is compared to design predictions and is also used in the determination of temporary rod withdrawal limits. Only a few degrees of cooldown and heatup occur during this test. The minimum tempera-ture is within the limits of Toch. Spec. 3.10.3. The rate is procedur-ally limited to loss than 20 dog F, which is well with in the Tech.
Spoc. Limits for NC heatup/cooldown. The reactivity changes due to temperature changes are typically loss than 20 pcm. The procedure requires that flux be maintained within the zero power physics test band. Also, por Tech. Spec. 3.10.3, power is limited to less than 5%
full power. No NC boron concentration changes are permitted during the test. The offects of this test are bounded by the FSAR analysos for the increase and decrease in heat removal by the secondary system.
TT/2/A/9200/16 Retype, Changes 0 to 1 Incorporated The original procedure utilized simulated signals to generato a simultaneous automatic start of Auxiliary Foodwater (CA) pumps 2A and
- 28. This retype requires a simultaneous manuni start of both pumps.
The original proceduro REQUIRED data for motor driven pump speed and pressure switch contact status to be taken for further analysis. The rotype makes this OPi10NAL. As a result of this rotype, the CA system is aligned in Standby Readiness Mode, and only one of the three CA pumps (2B) must bo declared inoperable during testing. Making the pump spood and contact status OPTIONAL cnsures that cnly the appropri-ate data is taken, and that all the procedure steps can be successful-ly completed.
l This test can be performed when Unit 2 is in modos 1, 2, or 3 only.
l The CA System is required to be operable in those modos. As a result of this retype, only one CA Pump (2B) must be declared inoperable during the test. This is within the conditions allowed under the Toch. Spoc. action statement. CA flow still can be provided as l
assumed in the FSAR. The CA pumps retain auto start capability. The probability of a malfunction of safety related equipment is not increased even with the pump trip timer circuits defeated sinco a specific low suction pressure limit (4 psig) is established for the operators to trip the pumps. This limit has boon previously used in testing, and has boon found to be an acceptable mode of operation. No USQ is created by this procedure.
PT/2/A/4350/02E Re-type, Changes 0 to 33 Incorporated The description and evaluation of this procedure is broken into 5 parts.
- 1. Motor Driven Auxiliary 'eedwater (CA) Pump Auto-Starts 14
In these sections, each pump is aligned to recirculate to the Upper Surge Tank separately and is response time tested for various automatic start signals. Actuation of associated flow control valves is also tested in response to these signals.
Accident flowrate is set up in advance by Operations. These sections are tested in either Mode 4, 5, 6, or No Mode, when the CA system is not required to be operable. The pumps are operated well within mechanical design limits.
- 2. Turbine Driven Auxiliary Foodwater Pump Auto-Starts In these sections, CA pump #2 is alignod to the Upper Surge Tank and is response time tested for various automatic start signals.
Actuation of associated flow control valves is also tested in response to these signals. Accident flowrate is set up in advance by Operations. This test must be performed in Modes 1, 2, or 3 since Main Steam must be available to run the pump.
However, at least one motor driven CA pump will still be opera-ble, and both will be available, during this testing.
This pump will be run in recirculation modo and started by a jumper that affects no other components. The pump will not be operated outside of design parameters. No new equipment malfunc-tion possibilities are created. The required number of CA pumps are still available within the Tech. Spec. Action Statement times; the margin of safety is not reduced.
- 3. Main Feedwater Isolation These sections will test Main Feedwater (CF) isolation on Hi-Hi Doghouse Level and Hi-H1 Steam Generator Level. High Doghouse level will be sin. lated by placement of jumpers. High steam Generator level will simulated by manipulation of Process Control Cabinet logic cards and placing one train of the Solid State Protection System (SSPS) in test. This testing will be performed in Modes 5, 6, or No Mode.
Placing the jumpers for Hi-Hi Doghouse Level causes Feedwater Isolation of the valves pertinent to the affected doghouse.
These actuations fail these valves to their safety positions, and thus do not cause the inoperability of any components, regardless of mode. Process Control Cabinet modifications and placement of one train of SSPS in test are allowable in modes 5 or 6. The reactor will already be tripped, and the CA system is not re-quired in these modes. The only equipment operation is the normal stroking of valves. No new equipment malfunction possi-
! bilities are created.
- 4. Main Turbine Trip l The main turbine will be tripped on the following signals: trip
- of both main feedwater pumps, reactor trip, Hi-Hi steam generator level, and manual trip signal. The main turbine is not safety related, and since these trips will be performed with no steam 15
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passing through the turbine by having the main steam isolation valves closed, there will be no transient effect on the reactor coolant system and no interaction with any safety related system.
The turbine trip is not required in the modes that this test will l be performed.
- 5. ATWS/AMSAC Actuations The main turbine will be tripped and both motor driven CA pumps will be simulated to start on the following ATWS/AMSAC signalst 2 '
out of 2 main feedwater pumps tripped, and 3 out of 4 main feedwater flow paths isolated. These sect ions will be performed in Modes 4, 5, 6, or No Mode. Since the ATWS/AMSAC system is not a Technical Specifications required system, operability is not a concern, except that in the test procedure, some blocking of safety related trip signals and interlocks are necessary to ensure positively that the test actuation is caused by the ATWC/AMSAC system. The valves manipulated for this section are not required for containment closure, since the steam generators are assumed to be adequate barriers to the potential release of radioactivity. In any case,1the test may be aborted and all CF valves closed immediately, should this be required. None of the components are required in acdos 4, 5, 6, or No Mode.
PT/0/1/4400/08 Change #47 l
This restricted change was made so that the pumphouse balance might be performed on Nuclear Service Water (RN) pump 1A with RN-train 2A essential header isolated. RN pump 1A supplied the flow requirements of the RN system during the test. The change allowed 1RN291 and .
2RN351 to be failed open during the test, and allowed jumpers to be placed to allow RN flow to the Unit 1 and Unit 2 Diesel Generator (D/G) Engine Cooling Water (KD) heat exchangers. Failing open valves 1RN291 and 2RN351 increased cooling water flow to the Component Cooling System (KC) heat exchangers. Both trains of RN were suppled during the test. All valves remained operable, except 2RN351 and 1RN291 which were failed open. (Open is their safety position.) )
Opening 1RN232A and 2RN232A could cool the KD system down and make the '
A train D/Gs inoperable; however, both A train D/Gs were inoperable because the A train-RN pumps were inoperable. All equipment continued to receive RN during the test. The B train RN pumps were operable ,
during this test. The B D/Gs were available for-providing emergency !
power. No USQ was created by the performance of the test.
l pT/2/A/4200/07A Change #13 This' restricted change provides an alternate test flow path by requir-ing the normal charging flow path to be isolated and the Centrifugal H Charging Pump (CCP) to be aligned to the cold leg injection lines. l This change prevents leakage through-the Chemical and Volume Control !
System Cold Leg Injection Isolation Valves 2NI9 and 2 nil 0 into the reactor coolant system, which is expected if this test is performed 1
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during mode 5, and flow is aligned to the normal charging flow path.
This flow diversion to the cold leg injection lines could invalidate i test results unless properly quantified and accounted for. The proposed flow path is equivalent to the normal testing alignment, and it eliminates the flow diversion concern.
During the test, CCp 2A remains operable. The pump remains capable of injecting borated water into the reactor coolant system via the cold leg injection lines or the normal charging flow path. The pump is placed in recirculation-during portions of this test. This does not render the boration flow path inoperable because the operator can realign CCP-2A to inject into the reacter coolant system in the required time frame if needed during this test. The prerequisite system conditions have been revised to indicate that this te6t-is to be only performed in a conditions where seal injection and normal charging are not required. An operable emergency flow path will be available during performance of this test. The new test method does not involve a USQ.
MP/0/A/7300/03 Change #2 This change adds a new section to this procedure. This section provides a method-of procedural documentation and guidance for remov-al, replacement, and corrective maintenance of the dryer pneumatic control valves. This is_being_done to consolidate inspections and activities under one procedure for this component. The enclosures have been upgraded to document requirements for this new section.
This addition verifies or restores the pneumatic control valves back to their original material and operating condition. No USQ is creat-ed.
7N/2/A/1323/CS/01A Initial Issue This procedure.provides guidance for implementation of Exempt Change CE-1323. This procedure will replace 13 Valcor Solenoid Valves Model V70900-21 with-Model V70900-65. The-implementation will-involve removing the old valves, installing the new valve with a shim plata due to the size differenes, functionally testing the solenoid valve, and, as required, retesting the coresponding control valve.- The new nolenoid valve is a direct replacement for the old model valve. The nnw V70900-65 valves have a coil life of 40 years and an 0-ring life ut 5 years.
This procedure may be implemented with Unit 2 in any mode. The solenoid valves will be replaced with.all air and electrical power inolated. With air and electrical power isolated from the solenoid valve, the associated control valve will fail-to its safe position.-
1he time. required to complete the work on-an' individual valve has been determined to be within the allowable limits of its respective Tech.
spec. action statement. Thus, no USQ exists.
I 17
EP/2/A/5000/1C Re-type #6 Step 6 was reworded to correct a human factors deficiency. Stop 11 was reworded to clarify the intent of this step. Step 15 was reworded to say, "IF 2NI-;38 can be accessed prior to initiating Cold Leg Recirculation, THEN dispatch operator to close 2NI-208." The intent of this step is to close this valvo, if possible, to prevent high radiation level in the Auxiliary Building in the vicinity of the leakoff header being supplied by 2NI-200. However, during Cold Leo Recirculation, the radiation level in the proximity of 2NI-200 wj?
very high. Thus, if this valve cannot be closed orlor to the in.
tion of Cold Leg Recirculation, it would be bottor to leave the i
~
open rather than placing the operator in such a high dose enviros Substep 17.b was added to dispatch an operator to place the hydroge.
recombiners in service. This change is being made to ensura that the Energency Procedure (EP) properly reflects the guidance given in Emergency Procedure Guideline (EPG) E-1, High Energy Line Break Inside w Containment. '
The setpoint for verifying proper Annulus Ventilation (VE) system operation was modified. The new setpoint is the cand from -1.4 to
-1.8 inches Water Column (W.C.) This setpoint is mentioned in step 17 and Enclosures 1 and 4. This new setpoint band takes into accou't the tolerances associated with the VE System instruments. A stop was added in Enclosure 1, item H to monitor Annulus pressure between -1.4 to -1.8 inches W.C. at 30 minuto intervals. Enclosure 4 was modified to provide operator actions if Annulus pressure is more negative than
-1.8 inches W.C. These changes were made as a temporary measure to satisfy the requirements in PIR 0-C89-0283. 3 A step was added to Enclosure 1, item I, to ensure that the proper minimum flow requirements are met for the Residual Heat Removal (ND) pumps.
This information was added as a result of a re-evaluation of ND pump minimum flow requirements which was performed by Design Engineering and Ingersoll-Rand. Substep 5.c was added to Enclosure 3.
This provides additional guidance to the operator in case the previous steps failed to start the fans. Step 6 on the same Enclosure was reworded to say " Ensure NS pumps and Containment Air Return fans operate as containments pressure changes." The step previously said "VX fans." Only the Containment Air Return fans will start and stop as Containment pressure changes. The Hydrogen Skimmer fans will run continuously until secured.
The positions of dampers 2AVS-D-3 and 2AVS-D-8 listed on Enclosure 4 was changed from closed to open. If Annulus pressure is more positive than -1.4 inches W.C., these two dampers should be open.
Other changes to this procedure included correcting various typograph-ical and format errors to ensure compliance with the EP Writer's Guide. None of the above changes alters the intent of the procedure or affects.the analysis in the FSAR. Thus, No USQ is created.
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