ML20141H255
| ML20141H255 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 07/09/1997 |
| From: | Gordon Peterson DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M98326, TAC-M98327, NUDOCS 9707220147 | |
| Download: ML20141H255 (21) | |
Text
. II.
I Duke Ibwer Company (803)831 3000 Catawba NuclearStation 4800 Concord Road York, SC29745 -
DUKEPOWER
. July 9, 1997 U.S.
Nuclear Regulatory Commission Attention:
Document Control Desk Washington, D.C.
20555 subject:
Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Supplement to 1996 10 CFR 50.59 Report TAC Nos. M98326 and M98327
Reference:
Letter from W.R. McCollum, Jr. To NRC, dated March 31, 1997 Gentlemen:
Pursuant to 10 CFR.50.59, please find attached a supplement to the reference letter which includes a brief description of any' changes, tests, and experiments, which were completed under the provisions of 10 CFR 50.59 through December 31, 1996.
Following a review of the reference letter, it has been determined that a number of entries in the reference did not actually constitute an Unreviewed Safety Question (USQ) evaluation (i.e.,
they were " screened" from the 10 CFR 50.59 USQ evaluation process).
These entries should not have been included in the reference and should be deleted.
The
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remainder of the entries shown in the attachment to this k
letter should replace the corresponding entries in the reference letter in their entirety.
,L
/,[f This supplement is being submit t ed pursuant to a discussion with Mr.
P.S. Tam of your staff in June 24, 1997.
If you have any questions concerning this.information, please call L.J.
Rudy at (803) 831-3084.
9707220147 970709 PDR ADOCK 05000413
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' Document Control Desk Page 2 July'9, 1997.
Very trul yo s
t&E b?-
G.R.
Peterson, Vice President
' Catawba Nuclear Station GRP/ljr Attachment xc-(With attachment):
L.A. Reyes, Regional Administrator Region II R.J.
Freudenberger,-Senior Resident Inspector P.S. Tam, Senior Project Manager ONRR i
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i ENTRIES THAT SHOULD BE DELETED:
As a result of further review, the following entries, which were submitted in the reference letter, were determined to not actually constitute an Unreviewed Safety Question Evaluation.
Hence, these entries should not have been included in the reference letter, since they were screened from the 10 CFR 50.59 evaluation process:
4 Exempt Changes CE-3759, CE-4203, CE-4435, CE-4686, CE-4687, CE-4899, and CE-7040 EhTRIES THAT SHOULD BE REPLACED:
Exempt Change CE-1593 Description 4
This minor modification replaced flow elements OYTFE5120 and OYTFE5130, which are flow elements in the YT system (a 4
system associated with chemical treatment of the cooling towers).
The flow elements were replaced with Brooks Instruments flow elements, Model Number lll4CJ34CQDAA.
The original instruments were pegged high due to a new type of J
dispersant used in the system, which has a different specific gravity and viscosity from the solution used previously.
I Evaluation 1
This modification did not affect the intended function of the system.
This modification allows the associated equipment to be operated as designed.
The modification did not involve any unreviewed safety questions and did not adversely affect any structure, system, or component important to safe plant operation.
4 Exempt Change CE-4373 Description 4
This exempt change documents the installation of plugs in the Unit 1 steam generators during the end-of-cycle 7 refueling outage.
Although the actual field work took place during the above outage, the completion of the documentation (i.e.,
revision to steam generator tube sheet maps) occurred on 4/30/96.
1 Evaluation The installation of the tube plugs and sleeves did not degrade primary system pressure boundary integrity; rather, Page1
- - -. - ~.
4 i
4
- it served to maintain it This ensured compliance with Technical Specifications for. reactor coolant system leakage.
All materials complied with reactor coolant and interconnected secondary system requirements as specified in j
the Final Safety Analysis Report.
There were no unreviewed safety. questions associated with this exempt change.
Exempt Change CE-4382 l
Description This minor modification replaced the dual coil electrical trip solenoid valve-(ETSV) in the turbine control system j'
with the new single coil ETSV which has been developed by General Electric.
Experience over the past ten years has
{'
shown the old type ETSV has been known to fail on an average of six times per year.
The main mode of failure has been the coils either opening or shorting, which was due j
primarily to the high. operating temperature.
Evaluation j
i The function of the ETSV has not changed.
This modification l
merely intended to improve system reliability.
This modification did not increase the probability or consequences of any accident.
The probability or consequences of a malfunction of equipment important to f
safety was not increased.
This modification did not reduce the margin of safety of any Technical Specification.
No USQ was generated as a result of this modification.
It was subsequently determined during testing of this modification that numerous problems occurred with the new single coil ETSV.
Catawba eventually reinstalled the old dual coil ETSV.
Exempt Change CE-4706 Description This minor modification corrected the loss of flow alarm for 1 and 2 EMF 35 and 38.
The EMF system is the radiation monitoring system.
EMF 35 is the unit vent particulate monitor and EMF 38 is the containment particulate monitor.
This modification added a relay to the vacuum pump motor control circuit so that when the vacuum pump is off, the 1
loss of flow alarm will be on.
Previously, when the inlet valve to the vacuum pump was closed, the loss of flow alarm would come in due to high vacuum.
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' Evaluation This change did not increase the probability, consequences, or create the possibilicy of an accident evaluated in the SAR.
This change did not increase the probability, 1
consequences, or the possibility of a malfunction of I
equipment important to safety evaluated in the SAR.
The change did not reduce the margin of safety as defined in the basis for any Technical Specification.
4 Exempt Change CE-5017 Description This minor modification increased the Standby Nuclear 1
Service Water Pond (SNSWP) normal elevation from 571.0 feet to 572.0 feet by installing a %-inch stainless steel plate 1
across the SNSWP outlet discharge pipe headwall.
The increase in normal SNSWP level provides a larger volume of water and a larger pond surface area to aid in_ dissipation j
l of waste heat during a Loss of Coolant Accident (LOCA) plus 1
unit shutdown events.
This minor modification also increased the existing intercept basin weir invert elevation from 572.58 feet to 573.09 feet.
This increase in weir
}
height will reduce the potential that the SNSWP will back up into Secondary Containment Sump #3 via Intercept Basin #7.
1 1
Evaluation The described changes are conservative in that they enhance the volume and heat dissipation capability of the SNSWP and j
reduce the potential that the SNSWP will back up into the l
described secondary containment sump.
No USQs were created oy this minor modification.
No Technical Specification changes were required.
I Exempt Change CE-7218
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Description i
This modification changed the frequency of turbine l
stop/ control / intercept valve testing as currently required in Selected Licensee Commitment (SLC) 16.7-5.
Catawba previously tested the turbine stop and intercept valves on a weekly basis.
The turbine control valves were tested on a monthly basis.
This modification also changed the recommended test interval for the off-line test of the mechanical overspeed trip device on the main turbine.
General Electric (GE) has reviewed the operating experience of the mechanical overspeed trip device on nuclear steam turbines and has concluded that the intervals between tests can be extended to 18 to 24 months.
This will also align 4
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'the UE recommendation with the insurance requirement of once per fuel cycle.
Evaluation In Amendments 131/125 for Units 1 and 2, respectively, Catawba removed the prior Technical Specification for the turbine overspeed protection system and bases and relocated it to the SLC manual, which is Chapter 16 of the Final l
Safety Evaluation Report (FSAR).
The Safety Evaluation (SE) for these amendments acknowledged that the turbine overspeed i
control is not part of an initial condition of a det.ign basis accident or transient that either assumes the failure j
of, or presents a challenge to, the integrity of a fission product. barrier.
Placing the surveillance requirements in the SLC allows optimization of testing and inspection frequencies without prior NRC approval, as no Technical Specification change will be required.
The SE acknowledged that Catawba planned on making changes to inspection J
frequencies via SLC changes under 10 CFR 50.59, which was done via this modification.
SLC 16.7-5 was changed as
)
described above.
This change did not involve any USQ.
No a
Technical Specification changes were required.
Exempt Change CE-7930 t
l Description This editorial modification revised the Final Safety Analysis Report (FSAR) and the component cooling system i
Design Basis Document (DBD) to reflect current flow rates and heat load information as contained in calculations.
This will make the FSTR and DBD agree with the design basis
]
information contained in the calculations.
)
i Evaluation There were no USQs associated with this modification.
- Also, no Technical Specification changes were required.
The revised data reflects the correct flow rates and heat loads derived from the design calculations.
i Exempt Change CE-7998 l
Description This modification changed the documentation for the hydrogen ignitor system to reflect the fact that the hydrogen ignitor 3
part number has changed from 7G to 12G.
The form, fit, function, and material remained the same.
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' Evaluation This modification did not affect the form, fit, function, or material of the subject system.
The design parameters of 4
the ignitors were not affected.
No USQs were generated as a result of this modification.
Exempt Change CE-8245 1
l Description
{
This evaluation was performed to determine if a USQ existed j
4 i
due to the operable but degraded condition on the Unit 1 and i
2 standby makeup pump (SMP).
The SMP sizing calculation 4
contained errors which did not account for elevated spent l
fuel pool temperatures and higher SMP flow rates, both of which could lead to inadequate available suction head for
. the pump.
4 Evaluation The evaluation determined that in order to consider the SMP l
and the standby shutdown system (SSS) operable, the following conditions must be met:
- 1) Spent fuel pool.
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temperature needs to be maintained at or below 125F as read i
on the control room gauge.
- 2) The fuel pool level needs to l
be kept at or above the 596 feet elevation.
- 3) The suction damper charge pressure needs to be kept between 5 and 10 psig relative to a 70F charge temperature.
These conditions j
are. met based on adherence to current procedures and j
Technical Specifications.
There were no USQs associated with this evaluation.
The SMP and SSS are technically operable from an equipment / system standpoint, but are considered to be operable but degraded due to the unconservative borated water volume stated in Technical Specification 4.7.13.3 (a) (2).
A Technical Specification change was submitted on May 27, 1997 to delete the required volume for the SMP.
Exempt Change CE-61162 Description This was an elective modification, editorial in nature, which allowed the installation of block ice cable suspension system parts as a basis for loading large diameter block ice in the ice condenser.
No physical work was performed under this minor modification.
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' Evaluation This modification only created an additional option as to j
the parts that can'be installed in ice baskets under outage work requests.
The new basket parts described in this modification were designed to be functional in the ice condenser in all possible containment conditions while performing their intended benefit of easing ice basket 3
maintenance activities and reducing plant operations and maintenance costs.
A complete analysis was performed to ensure that modified baskets and the ice bed will perform in a manner consistent with the original requirements for the ice condenser when modified using the cable suspension system and block ice.
No USQs were generated as a result of this modification.
Exempt Change CE-61214 Description This modification changed the non-safety related digital feedwater control system main feedwater pump turbine circuit.
The prev'ous design had two contacts in series for each redundant train parallel to each other; the current design has all four contacts in series.
The previous design
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provided non-safety related protection even if one train of the reactor trip system failed, due to the parallel circuit design.
This design, however, was not conducive to reliable plant operation should spurious conditions occur in the circuit.
Evaluation This modification did not affect accident initiation I
frequency or consequences.
Consequences of accidents and l
malfunctions of equipment important to safety were not increased.
The modification did not affect the fuel, cladding, reactor coolant pressure boundary, or containment.
No USQ was generated as a result of this modification.
No Technical Specification changes were required.
Exempt Change CE-61234 Description This modification lowered the 2C cold leg accumulator low level alarm setpoint, in order to reduce makeup frequency i
and associated dose due to sampling of the accumulator.
This is the accumulator for which a procedure had been developed for obtaining a sample through an alternate flow path inside containment due to the normal sample flowpath not being available.
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Evaluati,n.
All Teck_lcal Specifications are being met and the new j!
accumulator low level setpoint contains adequate margin for j
instrument uncertainty and operator action.
Therefore, the accumulator will perform as designed in the event of an accident and dose terms are unaffected by the different level alarm setpoint.
No Technical Specification changes
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were required and no USQ was identified, i
i FSAR 6.4.4.2 - Control Room Chlorine Analysis f
Description j
Problem Investigation Process report PIP 0-C95-1158 was written because FSAR Chapter 6.4 and Table 6-100 indicated that Catawba complied with Regulatory Guide (RG) 1.95 while
'certain requirements of the RG seemed unresolved.
Through j
an NRC questionnaire (a survey from the Region to resident inspectors regarding an event at another plant where a SO2 gassing of the control room occurred from an adjacent fossil j
unit; the survey led to the resident inspector evaluating I
the Catawba control room habitability system for compliance j
with RG 1.95), it was discovered that Catawba was not I
I' meeting the requirements of the RG with respect to response time and single failure as stated in the FSAR.
Evaluation i
Upon further detailed review, it was determined that the i
control room ventilation system is capable of maintaining I
control room habitability following receipt of a chlorine detection alarm without the use of the current breathing apparatus.
No USQs were created by this FSAR change.
No l
Technical Specification changes were required.
i i
FSAR Table 6 Potential Bypass Leak Paths Through Containment Isolation Valves 1
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Description j
This FSAR change resulted from Nuclear Station Modifications (NSMs) CN-10911 and CN-20300.
The original 10 CFR 50.59 evaluation for these NSMs, which was conducted in 1990, failed to determine that a Technical Specification change was required prior to implementation of the NSMs.
These j
NSMs created an additional secondary containment bypass leakage pathway and should have required a Technical L
Specification change to add this pathway to Table 3.6-1.
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' Evaluation Table 3.6-1 will be relocated to the FSAR as part of the Improved Technical Specifications conversion submittal, which was made on May 27, 1997.
Changes to this table will be made under 10 CFR 50.59 following relocation to the FSAR.
Catawba is presently in compliance with all existing Technical Specification requirements.
NSM CN-11375 Description This modification changed the design temperature for portions of the auxiliary feedwater (CA) system piping located between the steam generator isolation check valves and the CA punps' flow control valves.
The design temperature increased from 160F to 250F.
These lines included two paths for each of two motor-driven CA pumps and four paths for the turbine-driven CA pump for a total of eight lines.
This NSM also added local pressure indication (one high-range gauge per pump) in the affected lines at a location that is upstream of the steam generator check valves and downstream of the pump discharge check valves 1CA20, 27, and 32.
These pressurr. gauges will provide assurance that when certain temperature conditions exist due to check valve backleakage, the means exist to determine if saturated /subcooled conditions are present.
Evaluation The function of the CA system remained unchanged.
The likelihood of having a water hammer condition in the CA system has not increased with the increase in design temperature.
Temperatures will be monitored to assure j
saturated conditions are not present in the CA piping.
Predetermined alarm setpoints will assure appropriate action is taken to preclude the formation of steam voiding in the piping.
No adverse effects were created with respect to pipe rupture analysis as a result of this modification.
The I
response of the CA system to any accident which requires its use has not been affected.
No USQ was created as a result of this NSM.
No Technical Specification changes were I
required.
NSM CN-21367 Description This NSM reduced the full power operating temperature of Unit 2's reactor coolant system hot leg temperature (T-hot) by 3.0F.
This temperature reduction will extend the life of 4
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'the steam generators' Inconel-600 alloy tubes by slowing the rate of primary water stress corrosion cracking (PWSCC).
Evaluation This modification did not require any physical changes to the plant.
The change was to how the existing plant is operated; therefore, procedures reflect the temperature reduction.
Equipment operating sequences were not affected and no control room instrumentation changes were needed.
This modification did not involve an unreviewed safety question or safety concern.
Other than the changes to the K6 penalty coefficient definition submitted on September 13, 1995 and approved in Amendments 137/131, no additional Technical Specification changes were required.
NSM CN-50441 Description This NSM modified the control area chilled water system by replacing the ITT hydromotors (electro-hydraulic valve actuators) on twelve valves in the system with manual valve actuators.
These hydromotors had frequent reliability concerns and high maintenance costs.
Evaluation This modification did not degrade the ability of the YC system to provide a sufficiently cool environment for the operation of equipment in the control room area.
No control changes were made to the rest of the subject system.
The system will be made more reliable and less likely to malfunction and challenge safety systems.
No USQs were created by this modification and no Technical Specification changes were required.
NSM CN-50443 Description This modification replaced carbon steel pipe and valves in the nuclear service water strainer backwash lines with stainless steel piping and valves.
The modification added a flow orificu and a local gauge to assist in system flow determinations and balancing.
A set of bolted flanges was added for removing a section of pipe for accessibility to backwash strainer internals for periodic cleaning.
Valves 1 and 2 RNE39 and 1 and 2 RNE47 were added by this modification.
The modification changed %-inch flanges and approximately 6-inch long nipples from the high and low Page 9
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' pressure connections instrumentation and controls at the backwash strainer to stainless steel.
Evaluation This modification did not increase any accident probabilities or consequences; nor did it create the possibility for any new types of accidents or increase the probability of malfunctions of equipment-important to safety.
It did not create the possibility for any new types of malfunctions of equipment.
No USQs were created by this modification and no Technical Specification changes were required.
PIP 0-C96-1824 - Operable But Degraded Evaluation of the Unit 1 and 2 Standby Makeup Pump Description PIP 0-C96-1824 identified errors in the calculation for Standby Makeup Pump Sizing.
Because of these errors, operability of the Unit 1 and Unit 2 Standby Makeup Pump (SMP) was in question.
This 10 CFR 50.59 evaluation discussed the operability concerns associated with the Unit 1 and 2 SMP and the standby shutdown system (SSS).
Evaluation This evaluation was performed to determine if a USQ existed due to the operable but degraded condition of the Unit 1 and 2 SMP.
The SMP sizing calculation contained' errors which did not account for elevated spent fuel pool temperatures and higher SMP flow rates, both of which could lead to inadequate available suction head for the pump.
In order to consider the SMP and SSS as operable, the following conditions must be met:
- 1) Spent fuel pool temperatures need to be maintained at or below 125F as read on the control room gauge, 2) The fuel pool level needs te be kept at or above the 596 feet elevation, 3) The suction camper charge pressure needs to be kept between 5 and 10 psig relative to a 70F charge temperature.
These conditions were met based on adherence to current procedures and Technical Specifications.
There were no USQs associated with this operable but degraded evaluation.
Also, no FSAR changes were required.
The SMP and SSS were technically operable from an equipment / system standpoint, but were considered to be operable but degraded due to the unconservative borated water volume stated in Technical Specification
- 4. 7.13. 3 (a) (2 ).
On May 27, 1997, a Technical Specification change request was submitted to the NRC to delete the subject water volume.
Other existing Technical Page 10
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' Specifications ensure at least 23 feet of water are maintained over the top'of the fuel' assemblies at all times.
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7/1/94 - 1RF859 Fire Protection (RF) System Description
' A review of the Unit 1 containment pipe chase fire j
protection sprinkler system identified that this system is a normally manual system with dry' piping located within containment.
Indications of a fire inside containment would have to be recognized by control room operators prior to
- opening valve 1RF447B and charging the containment RF header with' water.
Based on this review, the following compensatory actions were considered to be adequate:
- 1) Utilize the following EFA. zones, which are located within the Unit 1 containment pipe chase area, to satisfy the continuous fire watch requirement.
If any of these EFA zones becomes inoperable,. Selected Licensee Commitment Section 16.9-6, Remedial Action for inoperable detection zones within containment,.will be initiated (EFA zone 131, 132, 133, 134, 136, and 137).
2)-The fire hose' stations' located within the Unit 1 reactor building will not be impaired by the RF isolation required q
to perform maintenance on 1RF859.
These fire hose stations will serve as backup fire supression equipment.
Evaluation-Per this discussion, it was considered that the Unit 1 reactor building fire protection compensatory actions used during maintenance on 1RF859 did not adversely affect any structure, system, or component important to the safe operation of Catawba Nuclear Station.
These fire protection compensatory measures did not increase the probability or
. consequences of any accidents evaluated in the FSAR.
PT/2/A/4200/13H, Change 10 - Safety Injection (NI)/ Chemical
'and volume Control (NV) Check valve Test Description The purpose of PT/2/A/4200/13H, NI and NV Check Valve Test, is to comply with Catawba Inservice Testing for Valve (IWV) program requirements for operability for those valves listed in'the procedure.
The change.to this procedure involved fourteen separate changes.
The most significant changes included adding valve 2NI118A, Safety Injection Pump 2A Cold Leg Injection Isolation, to the procedure to allow for static and differential pressure testing of this valve, deleting certain steps pertaining to differential pressure Page11
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' testing-of other valves, and ensuring adequate miniflow protection for-the centrifugal charging pumps.
Evaluation During the performance of this test, the reactor vessel is open with no fuel inside the core.
Performance of this test i
provides assurance that adequate emergency core cooling flows will be delivered to the reactor coolant system in the event of a loss of coolant accident.
No USQ was created as a result of these procedure changes.
No Technical Specification changes were required.
l PT/2/A/4200/01N, Change 31 - Reactor Coolant System Pressure Boundary Valve Leak Rate Test Description The change to this procedure involved starting safety injection UNI) pump 2B and establishing flow through valve 2 nil 69 in order to flush any debris that may have collected on the seat, preventing the valve from seating properly.
Evaluation Starting an NI pump would not increase accident probabilities.
The path opened to the test line sample hood would be automatically isolated in the event of an actual safety injection.
This change did not adversely affect the system's ability to perform its accident mitigation function.
None of the equipment put in service as a result of this change was operated outside the bounds of what would normally be required.
No new failure modes were created.
The margin of safety for system operation was not reduced.
PT/2/A/4400/01, Change 22 - ECCS Flow Procedure Description This restricted change allowed the safety injection CNI) cold leg balance for NI pump 2A to be performed.
This change was necessary so the NI pump 2A balance could be performed without performing NI pump 2B cold leg injection flow balance.
Evaluation The NI pump 2B cold leg balance has been performed in accordance with Technical Specification surveillance requirement 4.5.5h; therefore, this activity did not involve a USQ.
No Technical Specification changes were required.
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'AM/1/A/5100/07, Change 0, Retype 0 - Changing Positioner Action to Operate 1V010 Description The subject procedure provides guidance for Maintenance personnel to open containment air release and addition (VQ) system valve IVQ10 during a loss of offsite power (LOOP) event.
Tf containment pressure must be reduced during a Loop event and power is not available for IVQ10, the Fisher Model 3582 positioner for valve IVQ10 will be changed from direct to reverse acting to open the valve.
This was the initial issue for this procedure.
Evaluation Placing IVQ10 in an open position will not increase the probability of an accident because the valve is not an accident initiator.
Accident consequences will not be increased because this valve is not required to mitigate the consequences of an accident.
This procedure will not affect any safety related equipment.
This valve does not interface with the VQ system containment isolation valves.
Opening this valve does not affect any safety related equipment necessary to mitigate an accident.
No safety margins were reduced as a result of use of this procedure.
AP/0/A/5500/39, Change 0, Retype 0 - Control Room High Temperature Description This procedure was written to mitigato the consequences of a total loss of control room cooling.
This procedure uses the auxiliary building ventilation (VA) system to pull air from the service building through the control room.
The two front doors of the control room leading to the service building will be opened, as well as the back doors leading to the auxiliary building.
L.
- his configuration, air will be drawn from the service building, past the protection cabinets, and then out into the auxiliary building.
If a total loss of control room cooling event were to occur, both units would enter Technical Specification 3.0.3 and would commence a shutdown if one train could not be restored operable within the allowable time limitation.
Opening the control room doors under these circumstances would aid in keeping control room equipment within acceptable operating temperature limits.
This was the initial issue of this procedure.
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' Evaluation This procedure would be used as a defense-in-depth measure l'
to provide additional control room cooling under extreme conditions.
No accident probabilities will be increased.
i No accident consequences will be increased.
Operating with l
4 the control room doors open will not adversely impact any j
l, other systems that could potentially cause an accident.
No j
increase in probability of a malfunction of safety equipment l
4 will occur.
Neither will such consequences be increased.
No possibility for a malfunction of a new type will be created.
No safety margins will be adversely impacted.
In addition, this procedure contains appropriate security i
requirements for having the control room doors open.
1 i
d CP/0/B/8500/30,- Change 3 - Chemistry Procedure for the Determination of Clam-Trol-(CT-1) i
{
Description l
The subject procedure was the analytical procedure for utilizing Clam-Trol for control of asiatic clams at Catawba.
The subject procedure was deleted, since it was determined j.
that use of this particular biocide was unsuitable due to its toxicity and its detrimental effects on Lake Wylie, t
Catawba could not meet its National Pollution Discharge i
. Elimination System (NPDES) limits while utilizing Clam-Trol.
t -
Evaluation j
In lieu of using Clam-Trol, clam control will be maintained i
by a flushing program which is used on affected piping.
j Catawba's flushing program is administered under procedure i
OP/0/A/6400/06F, " Nuclear Service Water System Flush Procedure".
Deletion of the subject procedure did not i
result in any USQs being created.
No Technical Specification changes were required.
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IP/0/A/3200/12, Change 3 - Procedure for Blocking P-14 l
and/or Placing Both Trains of the Solid State Protection l
System (SSPS) in Test Description i
l This change added a new step to the procedure which verifies i
that no output relays are latched prior to placing SSPS trains in test.
If any are latched, Operations is to be l'
notified and required blocks and/or resets are to be i
reestablished.
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' Evaluation This change did not impact any accident probabilities or ccnsequences.
No new accident possibilities were created.
No increase in probability or consequences of equipment malfunctions was created.
No USQ was generated and no Technical Specification changes were required.
OP/1/B/6200/22, Change 0 - Unit 1 Operating Procedure for Composite Crud Sampling Description This procedure was revised in order to insert a statement to I
notify Radwaste Chemistry of intention to sample for composite crud.
This will allow them to more effectively 4
track inputs to their systems and thus reduce the time spent looking for unidentified inputs.
Also, steps were added to direct technicians to flush sample sinks by operating makeup i
demineralized water (YM) supply valves.
This will provide
{
more consistency in sampling routines.
~.
Evaluation i
No accident probabilities or consequences were impacted.
The possibility for creation of different accident types was not created.
No increase in probability or consequences of equipment malfunctions was created.
No safety margins were reduced by this change.
No USQ was created.
J OP/1/A/6200/28, Change 8 - Operating Procedure for the 4
Addition of Chemicals to the Reactor Coolant (NC) System i
i Description This procedure was revised in order to allow for multiple chemical additions and to provide a mechanism for the i
suspension of the procedure at a certain point to facilitate the addition of chemicals to the chemical mixing tank at one l
time and the injection of chemicals to the NC system at a later time.
Several steps were being repeated and additional signoffs were added as needed.
Evaluation No accident probabilities or consequences were impacted by this change.
The possibility for different accident types was not created.
No increase in probability or consequences of equipment malfunctions was created.
No safety margins j
were reduced.
No USQ was created by this procedure change.
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'OP/0/A/6200/26, Change 2 - Operating Procedure for the Radiation Monitoring (EMF) System Troubleshooting Description This procedure was revised in order to delete two enclosures which pertain to the flushing of EMFs on both units.
A different type of EMF, which does not utilize a crud trap, j
is being used.
Also, the note pertaining to temperature was changed to read less than 140F to make it consistent with other Chemistry procedures.
Evaluation No accident probabilities or consequences were impacted by this change.
The possibility for different accident types was not created.
No increase in probability or consequences of equipment malfunctions was created.
No safety margins were reduced.
No USQ was created by this procedure change.
OP/0/B/6500/34, Change 8 - Operating Procedure for Liquid j
Waste Recycle System Laundry and Hot Shower Description This revision deleted requirements for placing the Laundry, Hot Shower Tank (LHST) in recirculation prior to processing
)
to a monitor tank.
The requirement of the LHST being in j
recirculation has been evaluated and determined that the intent of moving water from the LHST to a monitor tank can be accomplished without the extra steps of placing the LHST in recirculation.
Evaluation No accident probabilities or consequences were impacted by this change.
The possibility for different accident types was not created.
No increase in probability or consequences of equipment malfunctions was created.
No safety margins were reduced.
No ilSQ was created by this procedure change.
PT/1/A/4250/02B, Changes 13-15, Retype 11 - Monthly Main Turbine Valve Movement Description This change changed the frequency of turbine stop/ control / intercept valve testing as previously required in Selected Licensee Commitment (SLC) 16.7-5.
Catawba previously tested the turbine stop and intercept valves on a weekly basis.
The turbine control valves were tested on a monthly basis.
This change also changed the recommended test interval for the offline test of the mechanical Page 16
'overspeed trip device on the main turbine.
General Electric (GE) reviewed the operating experience of the mechanical overspeed trip device on nuclear steam turbines and concluded that the interval between tests can be extended to 18 to 24 months.
This aligned the GE recommendation with the insurance requirement of once per fuel cycle.
Evaluation SLC 16.7-5 was changed as described above per Exempt Change CE-7218.
No increase in any accident probabilities or consequences was generated as a result of this change.
No new accident types were created.
No increase in probabilities or consequences of any equipment failure was created.
There were no USQs associated with this' change.
PT/2/A/4400/09, Change 18 - Cooling Water Flow Monitoring for Asiatic Clams and Mussels Quarterly Test Description This PT measures the flow and pressure drop across the component cooling water, diesel generator engine jacket cooling water, and containment spray heat exchangers.
The 1A and 2B component cooling water heat exchangers have recently had the original brass tube bundles replaced with j
stainless steel tube bundles.
Results c' the PT indicated that the stainless steel tubes foul mor apidly than did the brass tubes.
Prior to tube replace
'n t, these heat exchangers did not approach the PT's a vptance criteria.
'This change modified the component coo : 2g water resistance factor in the acceptance criteria sect n of the procedure.
Evaluation There were no changes to any design limit or setpoint.
No fission product barrier was affected.
The change did not modify the flow, temperature, or pressure of cooling water supplied to any component.
No control, instrument function, or performance of any structure, system, or component was degraded.
Therefore, the margin of safety as defined in the basis to any Technical Specification was not reduced.
There were no USQs associated with this change.
PT/1/A/4600/02A, Change 115 - Mode 1 Periodic Surveillance l
Items Description The purpose of this change was to modify the normal pre-accident operating temperature limits utilized in the 3
subject mode periodic surveillance procedure.
Calculations Page 17
'dete' mined new, normal pre-accident total loop uncertainties r
for the upper and lower containment temperature detectors.
The total loop uncertainties were then used to calculate a l
reduced error depending upon the number of operating air handling units and available temperature (re-istance temperature detector, or RTD) loops in service.
The RTDs and their associated alarms are utilized by operators to verify and maintain the bulk average containment Lir temperature below Technical Specification limits.
These containment temperature loops are non-safety related; however, they are utilized to verify the containment
-temperature parameter associated with Catawba's safety analyses.
1 i
Evaluation The allowable temperature limits for equipment located within containment was not impacted.
No Technical Specification Limiting Condition for Operation was affected.
No increase in any accident probabilities or consequences was generated.
No new accident scenarios were created as a result of'this change.
No impact on the probabilities or consequences of equipment malfunctions was created.
No USQ was genere ced as a result of this change.
No change to Technical Specifications was required.
TT/1/A/9200/88, Original Issue - Component Cooling (KC)
System One Pump Flow Verification for CN-11372 Description In order to obtain flow data to evaluate whether or not one KC pump can supply the necessary flows to all the normal operational loads, a test procedure was developed.
This TT has the KC system in its normal operational alignment.
Only one KC pump is running during the test while flow rates through all affected components are verified and adjusted as necessary.
Evaluation Sufficient precautions are addressed in the procedure to preclude degrading any system or component.
The normal operation of the KC system, or any system which interfaces with KC, is not affected by the performance of this test.
There were no USQs associated with this test procedure.
No FSAR or Technical Specification changes were required.
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1 l
'TT/2'/A/9200/88, Original Issue - Component Cooling (KC)
System One Pump Flow Verification for CN-21372 Description In order to obtain flow data to evaluate whether or not one KC pump can supply the necessary flows to all the normal operational loads, a test procedure was developed.
This TT has the KC system in its normal operational alignment.
Only one KC pump is running during the test while flow rates through all affected components are verified and adjusted as I
necessary.
Evaluation l
Sufficient precautions are addressed in the procedure to preclude degrading any system or component.
The normal operation of the KC system, or any system which interfaces with KC, is not affected by the performance of this test.
There were no USQs associated with this test procedure.
No FSAR or Technical Specification changes were required.
Simulate Revision 4 Description This smumary involved the 10 CFR 50.59 evaluation of the SIMULATE Revision 4 methodology for Catawba and McGuire Nuclear Stations.
The SIMULATE Revision 3 methodology has been approved by the NRC for use in safety related reload design calculations as described in topical report DPC-NE-1004A.
Calculations have been performed using Revision 4 to demonstrate the applicability of the conclusions of that report.
Specifically, the benchmarking and uncertainty calculations performed with Revision 3 to support the l
topical report have been repeated with similar or better
)
results using Revision 4.
Evaluation The upgrade from Revision 3 to Revision 4 of the CASMO-3/ SIMULATE-3P methodology (including changing from 12 to 18 axial levels) did not involve any USQ.
No changes to the i
FSAR or Technical Specifications were required.
The calculations performed to support this change demonstrated i
similar comparisons of predicted and measured reactivity i.
parameters to those provided in the topical report.
Explicit calculations demonstrated that the application of the uncertainties provided in the topical report remain conservative.
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