ML20079B032

1993 Annual Rept Per 10CFR50.59(b)(2)

ML20079B032
Person / Time
Site:	Brunswick
Issue date:	12/30/1993
From:	CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20079B025	List: BSEP-94-0517, Transmits 1993 Annual Rept Per 10CFR50.59(b)(2), for Brunswick Nuclear Plant ML20079B032
References
NUDOCS 9501040248
Download: ML20079B032 (42)

Text

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I BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 i

DOCKET NOS. 50-325 & 50-324/ LICENSE NOS. DPR-71 & DPR-62 l

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l 1993 ANNUAL REPORT IN ACCORDANCE WITH 10 CFR 50.59(b)(2) l

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9501040248 941229 DR ADOCK 05000324 PDR

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TABLE OF CONTENTS TITLE PAGE NO, Time Delay Relay Direct Replacement for Core Spray, ADS and RHR Systems 1

Unit 2 - Installation of Sight Flow Indicator in RHR Service Water to RHR Cross-Tie Drain Line 2

Decommissioning of DC Panels in the Nitrogen & Off Gas Building 3

Meteorological Program Changes 4

Unit 2 - Evaluation of Separation Criteria and Cable Divisional Crossover at Cable Tray Intersections 5

Drywell Liner Qualification 6

T,tandby Gas Treatment System Flow Rate 7

Removal of Operators from Valves on Abandoned Lines 8

Overpressurization of Unit 1 and 2 MG Set Rooms 9

Revision to Reflect Decommissioned SBGTS Temperature Switch / Alarm 10 Unit 1 - Deletion of Multi-Point Area Radiation Monitors from Panel XU-4 in Main Control Room (MCR) 11 Decommissioning and in-Place Abandonment of Spent Resin Pump Pressure Switch 12 Unit 1 - Temporary Modification Converting Reactor Building Sprinkler System from Primed Pre-action Design to Wet Design 13 Unit 1 - Temporary Makeup and Backup to the Spent Fuel Pool 14 Unit 2 - Spent Fuel Pool Makeup Provisions During RHR B Loop Outage 15 Replacement of K-600 Circuit Breakers with K-800S Circuit Breakers 16 Organization Changes 17 Core Shroud Modification 1B Steam Leak Detection System Upgrade for HPCI, RCIC, RWCU, and RHR 19 Diesel Generator Building and Reactor Building Internal Wall Tornado Venting 20 Containment Atmosphere Control (CAC) Subsystem Design Compliance 21 l

Condenser Waterbox Air Removal Common Vacuum System 22 Turbine Building Oily Drains to Radwaste 23 Primary Containment Sample Probe Removal from Drywell Ducting 24 Standby Gas Treatment System Standby Mode of Operation Deletion 25 HPCI Vent Valve Addition 26 i

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y TABLE OF CONTENTS TITLE PAGE NO.

Addition of Hardened Wetwell Vent (HWV) Subsystem to Containment Atmosphere Control (CAC) System 27 Turbine Buildirig East & West Moisture Separator Drain Tanks Area Radiation Monitors 28 Digital Feedwater Control System Upgrade 29 Reactor Building Roof Vent Radiation Monitor Upgrade 30 Modificatien of Containment Cooling Logic for Cooling Valves 31 Plant Process Computer System Replacement 32 Reactor Vessel Water Level Reference Leg Continuous Backfill 33 Post-Accident Sampling System Small Volume Sample Valve 34 Feedwater Venturi Rubidium Nitrate Calibration Test 35 Unit 2 - Reactor Recirculation System Operating Procedure 36 Emergency Fuel Pool Makeup 37 Feedwater Venturi Sodium-24 Calibration Test 38 l

Containment Atmosphere Control System 39 Fuel Pool Cooling Special Operating Procedure 40 CRD Flow Maximization 41 1

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-* 4-CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Time Delay Relay Direct Replacement for Core Spray, ADS and RHR Systems BRIEF DESCRIPTION:

The evaluation addresses a time delay relay (TDR) direct replacement for the Core Spray, ADS, and RHR systems per the recommendation of General Electric (GE) Service Information Letter (SIL) No. 230. TDRs are set and calibrated to 180 seconds, but were at one time set at one minute.

SUMMARY

OF THE SAFETY EVALUATION:

GE Design Basis Documentation (DBD)-17 requires the LPCI valve open signals sent to the RHR Heat Exchanger Bypass Valve shall be maintained for a sufficient time to assure the valve fully opens, assuming it is fully closed when the LPCI initiatior signal is received.

The evaluation addresses increase in time delay from 60 second; to 180 seconds per GE SIL No. 230. This setpoint conforms to the original design basis. UFSAR accident probability is unchanged. The established time delay setpoint has no effect on fission product barriers. The increased delay time ensures the bypass valve strokes full open when LPCI is initiated following a LOCA. This assurance mitigates damage to the RHR heat exchanger, thus preventing an increase in the probability of RHR heat exchanger malfunction. Increasing the bypass valve opening time delay initiates no new type of accident than those previously evaluated. No new single failures are created that would exceed previously identified LOCA boundaries.

Plant

Reference:

EWR No. 81-450VR GE SIL No. 230 1

W CHANGE TO FACILITY AS DESCRIBED IN THE FSAR i

TITLE:

Unit 2 - Installation of Sight Flow Indicator in RHR Service Water to RHR Cross-Tie Drain Line BRIEF DESCRIPTION:

The evaluation addresses installation of a sight flow indicator in the RHR Service Water to RHR Cross-Tie drain line and evaluation of the addition of pipe supports to strengthen the drain line.

SUMMARY

OF THE SAFETY EVALUATION:

The drain line is non-Q and non-seismic. The sight flow indicator meets or exceeds the pressure and temperature ratings of the piping in which it is installed. Installation of the sight flow indicator will not affect the function of the drain line. It will only provide monitoring capability. Installation of additional pipe supports will enhance the line by providing greater structural integrity.

Plant

Reference:

EER 91-0124 l

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CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Decommissioning of DC Panels in the Nitrogen & Off Gas Building BRIEF DESCRIPTION:

The engineering design package addresses disconnection and in-place sparing of 125V DC Augmented Off Gas (AOG) distribution panels AOG-1 and AOG-2, cables GJ1-HBO and GMS-XN2, and breakers GJ1 and GM5 from 125/250V DC distribution switchboards 1 A and 28.

SUMMARY

OF THE SAFETY EVALUATION:

Sparing of these panels, cables and breakers does not impact system operation. No potential impact to plant safety is of concern after these cables are physically disconnected, since all possible electrical pathways for problems such as ground faults, induced voltage, short circuiting with separate divisions or sneak circuits are removed for credible failure modes. No safety margin was reduced by performance of this activity.

Performance of this activity does not increase the probability or consequences of any accident previously evaluated in the UFSAR since the identified equipment is being disconnected from the power source. Nor willit create the possibility of an accident of a different type for the system because the number of feeder cables that previously connected panels and circuit breakers are being reduced. There is no possibility of a malfunction to other equipment that is important to safety resulting from this activity because of the elimination of the branch circuitry from the 125V DC switchboards.

Plant

Reference:

EDP 92-030 1

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. 6-4 CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

. Meteorological Program Changes BRIEF DESCRIPTION:

Brunswick Nuclear Plant meteorological program changes include removal of the Westinghouse sensor system and Esterline Angus strip chart recorders and incorporation of the Yokogawa Hybrid Recorder. This represents a change of the historical data base input and primary data acquisition system from the Westinghouse sensor system to the ADAC sensor system, with the hybrid recorder serving as the backup sensor system.

SUMMARY

OF THE SAFETY EVALUATION:

The purpose of the meteorological program is to measure and display meteorological data during normal and accident conditions. The system has no effect on the design, operation, and maintenance of systems important to safety. No changes have been made to the meteorological program that would decrease data recovery rates or data availability.

Changes to the program have increased data availabilities and recovery rates because of the increased serviceability of the newer technology used in the hybrid recorder. Data recovery rates between the ADAC sensor system and the Westinghouse sensor system are comparable and rate within Regulatory Guide 1.23 recommendations. No significant differences exist in the values produced by the two systems; consequently, there is no change in the safety margin resulting from these meteorological program changes.

Plant

Reference:

10CFR 50.59 Safety Analysis " Meteorological Program" i

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CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Unit 2 - Evaluation of Separation Criteria and Cable Divisional Crossover at Cable Tray Intersections BRIEF DESCRIPTION:

This EER analyzes cable tray divisionti separation for isolated incidences that cannot conform due to unusual conditions and circumstances.

SUMMARY

OF THE SAFETY EVALUATION:

This engineering evaluation applied IEEE Standard 379, "lEEE Standard Application of the Single-Failure Criterion to Nuclear Power Generating Station Class 1E Systems" to acceptability of three incidences that cannot conform to cable tray divisional separation criteria for Class 1E electric systems. The analysis concludes that none of the conditions would contribute to or cause a plant condition which would adversely affect safety. These conditions will not increase the probability of occurrence of an accident previously evaluated in the UFSAR. The consequences of an accident or equipment failure remain unchanged. Based on this engineering evaluation, the conditions will not introduce any factors which, under credible accident scenarios, degrade the performance of or increase the challenges to any safety systems. These conditions do not introduce a new initiating factor for an accident or equipment failure. These conditions do not result in an addition or deletion to previously evaluated plant configuration. Therefore, the possibility of an accident of a different type than previously evaluated does not exist. The subject conditions do not adversely affect any specific Technical Specification requirements and therefore will not reduce the margin of safety.

Plant

Reference:

EER 93-0446 5

CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TG 5:

Drywell Liner Qualification i

BRIEF DESCRIPTION:

The engineering evaluation analyzes localized corrosion of the drywell liner at the interface between the liner and the concrete floor slab at Elevation 4 ft. 6 in. resulting in less than the 5/16 in. thickness specified on the liner drawings. The liner was repaired in five local areas to meet a calculated minimum thickness for local areas.

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SUMMARY

OF THE SAFETY EVALUATION:

i Based on industry accepted approach and experimental evidence, thicknesses less than 5/16 in are acceptable in local areas at Elevation 4 ft. 6 in. Since the acceptance limit of 1.18 is not exceeded and adequate margin to preclude liner tearing exists, the liner will maintain leak-tight integrity and a progressive failure of the anchorage system will not e c..t er. At the repair locations, liner thickness was increased to meet minimum thickness ter !" al areas. Other areas were evaluated as acceptable. Drywell liner is not an initiator e:

. accident. Reduction in drywell liner thickness does not create credible accident scenarios.

Plant

Reference:

EER 93-0173, Rev. 3 i

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CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Standby Gas Treatment System Flow Rate j

BRIEF DESCRIPTION:

The evaluation analyzes operation of the Standby Gas Treatment (SBGT) System under maximum flows and determines that those operating values are acceptable. The SBGT

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system reactor building inlet valves lock full open on automatic initiation of the system, resulting in insufficient throttling to keep the system flow within technical specification surveillance test flow values. Actual baseline flow (unthrottled) is 4200 cfm at 130 degrees F.

SUMMARY

OF THE SAFETY EVALUATION:

The equipment will be operated in the same manner for which it has been designed, and the increases in flow rates have been shown to be within the acceptable design limits of the components and systems. Conduct of an unreviewed safety question determination regarding this issue indicated that operation of this system as analyzed by this EER does not reduce the margin of safety or result in an unreviewed safety question. Increases in radiation dose remain well within the acceptable limit.

Plant

Reference:

EER 93-037 7

i CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Removal of Operators from Valves on Abandoned Lines BRIEF DESCRIPTION:

The evaluation addresses removal of operators from valves on abandoned lit.ed so the lines f

and associated supports can be seismically qualified. Removal of operators will allow the lines and associated supports to be qualified "as is."

SUMMARY

OF THE SAFETY EVALUATION:

The lines are no longer a pressure boundary and are classified as Class B-01, Seismic 1, and non-ISI. This engineering evaluation report (EER) is classified as a permanent repair and enhances the abandoned lines by removing weight which will allow the lines and associated piping to be seismically qualified. No new accident scenarios are introduced and there is negligible effect on any existing accidents, consequences or events as previously evaluated in the UFSAR. No unreviewed safety questions exists and the margin of safety is not reduced as defined in the basis of any technical Specification.

Plant

Reference:

EER 93-0147 r

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l CHANGE TO FACILITY AS DESCRIBED IN THE FSAR

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TITLE:

Overpressurization of Unit 1 and 2 MG Set Rooms BRIEF DESCRIPTION:

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The evaluation addresses disconnecting and removing a duct spool piece on the MG Set i

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Room return / exhaust air ductwork located outside the Turbine Building on the 70' elevation. This change was made in order to equalize return and supply airflow in an effort to remove the overpressurization problem in the MG Set Room.

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SUMMARY

OF THE SAFETY EVALUATION:

I This change reduces the potential impact on Control Room habitability because it results in j

minimizing the difference between supply and return airflow of the MG Set Room l

ventilation system. This change also assures that the pressure in the MG Set Room i

remains positive in respect to the Turbine Building pressure. Both the MG Set Room i

ventilation system and the MG set itself are non-safety related and nonseismic. The Control Room contains equipment / systems which are Quality Class A and are seismically qualified as well, but the safe operation of these systems and components are not l

impacted by this activity. Ongoing testing and monitoring on MG Set Room and MG Set ~

Motor bearing / winding temperatures will be performed using approved plant testing 1

procedures.

u Plant P.eference: EER 93-0280 i

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t CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Revision to Reflect Decommissioned SBGTS Temperature Switch / Alarm BRIEF DESCRIPTION:

This equipment decommissioning package (EDP) decommissions two Standby Gas Treatment System (SBGTS) temperature switches and associated wiring / cabling. It also decommissions the corresponding Main Control Room (MCR) annunciators that provide high humidity indication when a low temperature is detected by the subject temperature j

switches.

SUMMARY

OF THE SAFETY EVALUATION:

The subject temperature switches sense a temperature change in the last stage HEPA SBGT filter compartment that may be due to moisture in the filters. The loop was designed to alert operators of potential problems with SBGT heater. The loops are no longer necessary because inlet / outlet temperatures vary throughout the year and therefore a meaningful setpoint cannot be established. Outlet temperature is not an acceptable method for monitoring humidity. Delta T indication which can be determined from MCR temperature indicators is more appropriate for such determination as relative humidity decreases with increasing temperature.

This change does not affect any UFSAR Chapter 15 analysis previously evaluated nor create the probability of any new accident not already evaluated in the UFSAR. The proposed activity does not change the way the system is intended to operate and does not l

alter the original design intent. The margin of safety defined in the Technical Specifications remains unaffected by this change.

Plant

Reference:

EDP 92-020, Rev.1 i

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b CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Unit 1 - Deletion of Multi-Point Area Radiation Monitors from Panel XU-4 in Main Control Room (MCR)

BRIEF DESCRIPTION:

This equipment decommissioning package decommissions two multipoint area radiation recorders in the Main Control Room. Removal of these recorders from Panel XU-4 in the Main Control Room requires personnel to obtain area radiation levels from indicators located on the area radiation cabinet located in the electronic equipment room area which is adjacent to the Main Control Room.

SUMMARY

OF THE SAFETY EVALUATION:

Removal of the subject recorders has no effect on the performance of the Area Radiation Monitoring (ARM) system. The recorders are used for data recording only. Accident evaluation in the UFSAR and radiological consequences are not affected by removal of these recorders.

Plant

Reference:

EDP 93-005, Rev. O 11

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.i CHANGE TO FACILITY AS DESCRIBED IN THE FSAR l

TITLE:

Decommissioning and in-Place Abandonment of Spent Resin Pump Pressure Switch BRIEF DESCRIPTION:

The engineering design package (EDP) addresses in-place decommissioning of the spent I

resin pump low suction pressure switch. The spent resin pump low suction pressure trip function has been jumpered out for several years. The purpose of this trip is to prevent the pump from operating dry resulting in damage. There is a pump trip on low tank level.

This trip provides sufficient pump protection. Although changing the setpoint would provide protection from running the pump dry, decommissioning the switch is more viable because it eliminates calibration and repair of the switch resulting in reduced personnel I

radiation exposure.

SUMMARY

OF THE SAFETY EVALUATION:

In-place decommissioning of this Quality Class D switch per this EDP will not affect any accident description in Chapter 15 of the UFSAR. This change does not create the probability of any new accident not already evaluated in the UFSAR. The proposed activity does not change the way the system is intended to operate and does not alter the original design intent. The margin of safety defined in the Technical Specifications remains unaffected by this change.

Plant

Reference:

EDP 92-005, Rev.1 i

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CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Unit 1 - Temporary Modification Converting Reactor Building Sprinkler System from Primed Pre-action Design to Wet Design l

BRIEF DESCRIPTION:

A recurring problem existed with the fire protection system deluge valves associated with the automatic fire suppression sprinkler systems. This engineering evaluation report (EER) evaluated a temporary repair involving conversion of the Unit 1 Reactor Building sprinkler system from a primed preaction design to a wet-pipe design by closing the opening in the valve body with a new plate and gasket.

SUMMARY

OF THE SAFETY EVALUATION:

This temporary change does not affect the ability of the fire protection sprinkler system in l

the Unit 1 Reactor Building to perform its design function of controlling and suppressing postulated fires. The conversion of this valve to a wet-pipe design eliminates the interaction between the detection system and the valve. In the event of a pipe rupture downstream of the valve, the wet-pipe design allows water to flow and the full flow from the fire pumps will be discharged through the break. The previous primed preaction design of the valve would have resulted in the valve not opening in such an instance with a I

substantiallimitation of the flow of water through the break to about 25 gpm.

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L The fire suppression system is not an accident initiating system, and does not interface j

with fission products nor safety systems. This activity does not result in non-credible events becoming credible, nor change any bounding conditions. No new single failures are introduced. The possibility of an accident of a different type than any evaluated previously in the UFSAR is not created. There are no reductions to any margin of safety as defined in the bases of the Technical Specifications.

Plant

Reference:

EER 93-0312, Rev.1 i

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CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Unit 1 - Temporary Makeup and Backup to the Spent Fuel Pool BRIEF DESCRIPTION:

The evaluation addresses the installation of temporary seismic safety-related supply lines from the Core Spray A Loop and Service Water Conventional Header to the spent fuel pool for use in the event that the Fuel Pool Cooling and Makeup System becomes inoperable and alternate means become necessary to ensure makeup and backup capability.

SUMMARY

OF THE SAFETY EVALUATION:

The purpose of the engineering evaluation report (EER) is to provide compensatory actions for the unavailability of the Residual Heat Removal System "B" loop to provide for fuel pool makeup. The change evaluated by this EER only adds mechanical seismic connections to existing taps on the Core Spray System. There are no electrical components added; therefore, these changes can not increase the load on the onsite

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electrical distribution system. This change does not affect any UFSAR Chapter 15 analysis previously evaluated nor create the probability of any new accident not already evaluated in the UFSAR because use of this EER is restricted to Operational Mode 4 or 5 in order to prevent demands on the CSS and Service Water System during power operations. The proposed activity does not change the way the system is intended to operate and does not alter the original design intent. The only likely malfunction that could result from the temporary lines and connections is a break in the hose after it has been installed. The connections to the Core Spray System (CSS) or Conventional header (if needed) will cause no failure of a different type because each is prevented by the seismic qualification. The margin of safety defined in the Technical Specifications remains unaffected by this change.

Plant

Reference:

EER-92-0336 BNP Procedure 1-SP-93-008 i

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i CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Unit 2 - Spent Fuel Pool Makeup Provisions During RHR B Loop Outage BRIEF DESCRIPTION:

The evaluation addresses the measures taken to supply makeup water to the Fuel Pool during residual heat removal (RHR) Loop B inoperability resulting from RHR maintenance work.

SUMMARY

OF THE SAFETY EVALUATION:

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During the RHR Loop B work, the B loop was not in the normal configuration that would allow valving activities to establish the lineup necessary to achieve the Brunswick Nuclear Plant method of compliance with the regulatory requirement for a Seismic Category I makeup water supply. The evaluation documents that the time required from the ic,ss of fuel pool cooling until reaching the minimum Technical Specification allowable level of the fuel pool is adequate to allow RHR restoration to the FSAR described configuration. The evaluation establishes backup provisions that do not rely on RHR piping in order to provide additional assurance of makeup capability. The evaluation also documents that the RHR B Loop work does not affect the ability to comply with the FSAR analysis conditions associated with Fuel l'ool cooling.

The evaluation does not affect systems in a way that could increase the probability of an accident evaluated previously in the UFSAR. The affected permanent equipment is controlled such that there will be no significant reduction in the probability that the l

equipment will be available to perform the assumed mitigation for any accident. The af,fected permanent equipment is restored to a functional condition prior to use. The EER does not degrade the affected permanent equipment in its ability to mitigate the consequences of an event. The preferred configuration and operation of the plant is not significantly changed per this EER. Temporary configurations are established by this EER if the preferred condition cannot be used. The temporary configuration does not create any new accidents that are not similar to and bounded by existing accidents. The preferred configuration and operation of the plant is not changed. No equipment capability is affected or Technical Specification limit is changed.

Plant

Reference:

EER-91-0137 BNP Procedure 2-SP-93-009 i

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CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Replacement of K-600 Circuit Breakers with K-800S Circuit Breakers i

l BRIEF DESCRIPTION:

This direct replacement (DR) addresses refurbishment and upgrade of existing K-600 mechanical type circuit breakers with solid state 800 amp frame K-Line breakers in 480 I

VAC Busses E5, E6, E7, and E8.

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SUMMARY

OF THE SAFETY EVALUATION:

The solid state trip devices which are used as replacement items have been proven to be suitable replacements which serve exactly the same function within the same or possibly even more conservative trip time frame. Replacement of the mechanical trip devices in the 480 V Emergency Bus feeder breakers does not affect the safety evaluation of any accident scenario as previously evaluated in the UFSAR. This replacement will have no effect on the margin of safety as defined in the bases to the Technical Specification.

Plant

Reference:

DR 92-0088, Rev. O F

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CHANGE TO FACILITY AS DESCRIBED IN THE FSAR '

I TITLE:

Organization Changes BRIEF DESCRIPTION:

l This change incorporates corporate and plant organizationa! changes which resulted from the CP&L Corporate improvement initiatives into the UFSAR.

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SUMMARY

OF THE SAFETY EVALUATION:

These changes represent administrative changes which do not impact accidents which have been evaluated in the UFSAR. Overall system performance of any plant systems have not been altered by this update. Radiological consequences of any accidents evaluated in the UFSAR remain unchanged as a result of this revision. This administrative f

change does not create the possibility of any type of credible accident, affect the failure mode of any type of equipment, nor reduce the margin of safety as defined in the Technical Specifications.

Plant

Reference:

10CFR 50.59 Safety Review for BNP Updated FSAR Interim Amendment No.11A CP&L Letter (Serial GLS-93-216) from Mr. R. E. Rogan to US NRC Document Control Desk, entitled " Updated Organization Structure."

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j CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Core Shroud Modification BRIEF DESCRIPTION:

This plant modification installed mechanical clamps on the core shroud at the H2 and H3 welds. The clamps are designed to provide structural integrity across the H2/H3 top guide j

support ring interface and thereby eliminate reliance on the H2 and H3 welds.

SUMMARY

OF THE SAFETY EVALUATION:

Installation of the mechanical clamps on the core shroud results in maintenance of its design basis safety-related requirements. Changes made by this modification did not introduce new effects or factors which prevent any other equipment or system from meeting desigt: basis function. Additionally, installation processes did not create safety concerns.

i The design, materials and installation requirements for the this modification met the original design basis function for the core shroud. This modification ensures the shroud will meet its design function of maintaining the core geometry and providing a floodable volume and that the Core Spray piping and spargers will perform their intended safety function. Insignificant effects on peak cladding temperature for the design basis LOCA result from the clamp leakage flow paths, with no impact on the release from the analyzed accident. Thus, the modification did not increase the consequences of any accident.

Leakage effects on jet pump operation were evaluated and shown to be acceptable.

t Therefore the probability of a malfunction of equipment important to safety remains i

unchanged. Leakage flow at the bolt holes were analyzed to have negligible effect on system operation, thus there is no different accident that can be postulated as occurring as a result of the modification. No reduction in a safety margin of any kind results from the modification.

Plant

Reference:

PM 93-038 and PM 94-007 r

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CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Steam Leak Detection System Upgrade for HPCl, RCIC, RWCU, and RHR BRIEF DESCRIPTION:

i The plant modification implemented steam leak detection system configuration changes encompassing replacement of Riley hardware and associated components with General Electric (GE) nuclear measurement analysis and control (NUMAC) microprocessor units, replacement of high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) Fenwat switches with thermocouples that will connect to NUMAC monitoring / trip channels, removal of General Electric measurement analysis and control (GEMAC) and Agastat components used for reactor water cleanup (RWCU) differential flow monitoring, i

and replacement of the 125 V DC power supply with 120 V AC emergency power.

SUMMARY

OF THE SAFETY EVALUATION:

i This change does not affect any UFSAR Chapter 15 analysis previously evaluated nor 1

create the probability of any new accident not already evaluated in the UFSAR. The proposed activity does not change the way the system is intended to operate and does not alter the original design intent. The margin of safety defined in the Technical Specifications remains unaffected by this change.

Plant

Reference:

PM 91-038 and PM 91-039 l

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CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Diesel Generator Building and Reactor Building Internal Wall Tornado Venting BRIEF DESCRIPTION:

The modification addresses venting of the Diesel Generator Building and the Reactor Building during a design basis tornado.

SUMMARY

OF THE SAFETY EVALUATION:

These changes protect the walls of the diesel generator building and the reactor building against a design basis tornado without diminishing the fire protection and seismic design criteria of the walls. These activities do not increase the probability of occurrence of any accident previously evaluated in the UFSAR. The subject structures, equipment, and components are not initiators of accidents evaluated in the UFSAR. The modification increases the integrity of structures supporting safety systems and thus reduces the possibility of failure of such systems. Potential consequences of accidents previously evaluated in the UFSAR are not increased due to the modification. The modification increases the integrity of structures supporting safety-related equipment, therefore, probability of failure is reduced. The only safety-related devices installed as a result of the modification that can failin a manner such that their safety functions are compromised are the fire dampers. These dampers are included in a surveillance program that inspects the fusible links that support the damper curtains for possible deterioration.

Plant

Reference:

PM 92-069 20

CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Containment Atmosphere Control (CAC) Subsystem Design Compliance i

BRIEF DESCRIPTION:

The modification provides divisional separation between cabling for containment atmosphere control valves in control room panels. In addition, the modification changes the power source for one CAC valve from a Division ll panel to a Division Q panel.

SUMMARY

OF THE SAFETY EVALUATION:

Physical separation of the cables satisfies commitments to IEEE 279-1971. The changing of the power source for the CAC valve will enhance operation of the CAC venting system in the event of an accident. This change does not affect any UFSAR Chapter 15 analysis previously evaluated nor create the probability of any new accident not already evaluated in the UFSAR. The proposed activity does not change the way the system is intended to operate and does not alter the original design intent. The margin of safety defined in the Technical Specifications remains unaffected by this change.

Plant

Reference:

PM 90-065 and PM 90-066 1

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9 CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Condenser Waterbox Air Removal Common Vacuum System BRIEF DESCRIPTION:

The modification changed the existing plant to add a waterbox air removal / priming system consisting of two mechanical vacuum pumps and a priming tank.

SUMMARY

OF THE SAFETY EVALUATION:

The modification has no effect on the probability of occurrence of any accidents previously evaluated in the UFSAR. The consequences of any accident previously evaluated in the UFSAR are not increased as a result of these changes. The probability of occurrence of a malfunction of equipment important to safety evaluated previously in the UFSAR is not increased as the modification affects balance of plant equipment only. The consequences of a malfunction of equipment important to safety evaluated previously in the UFSAR are not increased. The modification does not create new failure modes and thus cannot create the possibility of an accident of a different type than any evaluated previously in the UFSAR. The modification does not create the possibility for a malfunction of equipment important to safety of a different type than any evaluated previously in the UFSAR. There is no reduction in system qualification or performance which results from these changes, therefore there is no possibility that any current margin of safety is reduced. Waterbox air removal enhances the performance of the circulating water system and increases margin between operation and trip with respect to condenser vacuum.

Plant

Reference:

PM 91-065 and PM 91-066 22

CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Turbine Building Oily Drains to Radwaste BRIEF DESCRIPTION:

The modification installs a new vendor-supplied oil / water separator skid for the Oily Drain tank.

SUMMARY

OF THE SAFETY EVALUATION:

The system is not an initiator of nor a contributor to any previously analyzed accident in the UFSAR. The probability of occurren'ce of or consequences of any accident previously evaluated in the UFSAR is not increased because the oil processing system is not.

addressed in the UFSAR accident evaluation. The oil / water separator skid system is located inside the Radwaste Building and has no effect on the operability of any safety related equipment. The probability of occurrence of or consequences of an accident affecting equipment important to safety evaluated previously in the UFSAR or of a different type than previously evaluated in the UFSAR is not created, since the system does not interface with safety-related equipment, and because no safety-related equipment is located in the proximity of the nil separation skid system. The possibility of an accident of a different type than any evaluated previously in the UFSAR also is not created, since no safety-related equipment is located in the proximity of the oil separation skid system.

The margin of safety as defined in the basis of the Technical Specifications is not reduced.

Plant

Reference:

PM 92-007

(

23

CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Primary Containment Sample Probe Removal from Drywell Ducting -

BRIEF DESCRIPTION:

The modification removed the sample supply tubing from inside the drywell for containment atmospheric monitors (which are located outside the drywell) on the 20 foot elevation in order to meet the requirements of Regulatory Guide 1.45.

r

SUMMARY

OF THE SAFETY EVALUATION:

Seismically qualifying these items represents and upgrade of the system. This upgrade does not increase challenges to existing safety systems. This change does not affect any UFSAR Chapter 15 analysis previously evaluated nor create the probability of any new accident not already evaluated in the UFSAR. The proposed activity does not change the way the system is intended to operate and does not alter the original design intent. The margin of safety defined in the Technical Specifications remains unaffected by this change.

Plant

Reference:

PM 93-009 and PM 93-013 6

24

I CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Standby Gas Tr'eatment System Standby Mode of Operation Deletion j

BRIEF DESCRIPTION:

f The modification changes the configuration of the standby gas treatment system (SBGTS) heaters from a delta configuration to a star configuration and deletes the automatic t

temperature control design feature from the standby mode of operation of the SBGTS.

l

SUMMARY

OF THE SAFETY EVALUATION:

The modification does not increase the probability of any accident evaluated in the UFSAR because the changes to the SBGTS heaters will not change the design functions of the system. The higher power output of the heaters increases the margin of safety from the calculated power output needed from the heaters. Consequences and probability of occurrence of accidents previously evaluated in the UFSAR are not increased. This activity does not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the UFSAR. The proposed activity will not increase the consequence of a malfunction of equipment important to safety as previously evaluated in the UFSAR. The possibility of an accident of a different type than previously evaluated in the UFSAR is not created by this plant modification.

Plant

Reference:

PM 92-105 and PM 92-106 25

i.

P CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

HPCI Vent Valve Addition BRIEF DESCRIPTION:

The modification provides for the installation of a tee, vent valve, and section of vent line between the levelinstruments and the first isolation valve on the gaseous side of each instrument on the high pressure coolant injection (HPCI) system.

SUMMARY

OF THE SAFETY EVALUATION:

Adding the vent lines to the suppression pool level instrument loops does not increase the probability or consequences of a previously analyzed accident or of equipment malfunction. The vent addition does not create the possibility of an accident of a different type than previously analyzed in the UFSAR or reduce the margin of safety as defined in the basis of any Technical Specification.

Plant

Reference:

PM 90 056 and 90-057 I

26 i

l

t CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Addition of Hardened Wetwell Vent (HWV) Subsystem to Containment Atmosphere Control (CAC) System 1

BRIEF DESCRIPTION:

The modification installed a hardened vent path, with associated isolation, controls, and radiation monitoring equipment, from the torus to a suitable release point on the Reactor Building roof.

(

SUMMARY

OF THE SAFETY EVALUATION:

The modification does not increase the probability of an accident previously evaluated in the UFSAR. The proposed modification does not increase the consequences of an accident as described in the UFSAR. The new equipment and systems meet the design basis requirements for design, materials, and installation for the systems in which they are installed, and therefore the new equipment does not increase the probability of an occurrence of a malfunction of equipment. Installation of a rupture disc downstream of the redundant primary containment isolation valves decreases the consequences of the l

malfunction of the PCIVs. The piping system associated with the modification is supported Seismic 1, and therefore does not allow the new piping to create the possibility of an accident by impacting other equipment. The margins of safety described in the

- Technical Specifications are not reduced by the modification as the design closing time for the new PCIS outboard valve is consistent with the requirement for that penetration and the total acceptable leakage for the unit will not be increased as a result of this modification.

Plant

Reference:

PM 91-001 and PM 92-073 t

h 27 1

0

,v-,.

-a.

CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Turbine Building East & West Moisture Separator Drain Tanks Area Radiation Monitors BRIEF DESCRIPTION:

The modification changes the upscale trip (alarm) setpoint associated with the Turbine Building East and West Moisture Separator Drain Tanks area radiation monitors to account for the increase in the background radiation levels resulting from the location of these tanks in a high radiation area as defined by the Technical Specifications and 10CFR 20.23.

SUMMARY

OF THE SAFETY EVALUATION:

These monitors perform no safety related function. Minor adjustments to compensate for changes in background radiation will not increase the probability of occurrence or consequences of an accident previously evaluated in the UFSAR. The modification does not change the design basis for functional capabilities of any equipment important to safety. The radiation monitors perform an alarm funcdon only and are not interlocked with any safety related control function. This change does not affect any UFSAR Chapter 15 analysis previously evaluated nor create the probability of any new accident not already evaluated in the UFSAR. The proposed activity does not change the way the system is intended to operate and does not alter the original design intent. The margin of safety defined in the Technica! Specifications remains unaffected by this change.

Plant

Reference:

PM 92-109 and PM 92-110 i

28

~.

s

~5 o

CHANGE TO FACILITY AS DESCRIBED IN THE FSAR I

TITLE:

Digital Feedwater Control System Upgrade BRIEF DESCRIPTION:

- The modification installed a microprocessor based digital feedwater control system to improve feedwater control maintainability and reduce SCRAM potential.

t i

.i

SUMMARY

OF THE SAFETY EVALUATION:

The installation of the digital feedwater control system (DFCS) does not increase the probability of other components and systems, or of the feedwater control system to initiate accidents evaluated in the UFSAR. The DFCS is not a mitigating system for the radiological consequences in the UFSAR. The probability of a malfunction of the DFCS equipment is not increased and does not impact that equipment which is important to safety per the UFSAR. The consequences of a malfunction of equipment important to safety cannot increase as a result of the DFCS. The possibility of an accident of a _

different type than any evaluated in the UFSAR is not created. The feedwater control system is not safety related and is not required to mitigate any accident or ensure

- operability of any equipment important to safety. The possibility of a malfunction of equipment important to safety of a different type than any evaluated in the UFSAR is not created. The margin of safety as defined in the basis of the Technical Specifications is not reduced by the modification.

Plant

Reference:

PM 89-001 and 89-002 i

l 29 i

9

.O.

l CHANGE TO FACILITY AS DESCRIBED IN THE FSAR j

i TITLE:

Reactor Building Roof Vent Radiation Monitor Upgrade i

BRIEF DESCRIPTION:

This change modified the Reactor Building Roof Vent Radiation Monitor to resolve its -

l temporary condition, made additional changes to the as-built radiation monitor skid, and completed electronic changes in the control room.

SUMMARY

OF THE SAFETY EVALUATION:

The radiation monitoring system is non-safety related. The system determines radioactive effluent from the Reactor Building and has no control functions in avoiding or mitigating the consequences of an accident. The radiation monitor itself is not discussed in-Chapter 15 of the UFSAR. The probability of occurrence of a malfunction of equipment-important to safety previously evaluated in the UFSAR is not increased by the modification because the radiation monitor has no interfaces with plant equipment physically or logically except for power, air sample tap, control room annunciation, ratemeters, and supports.

The modification removed the ratemeters and control room annunciation and will not degrade any other of the items. The consequences of a malfunction of equipment important to safety evaluated previously in the UFSAR is not increased. The modification -

does not create the possibility of an accident different than any evaluated previously in the UFSAR, nor create the possibility of malfunction of equipment important to safety of a different typt; than any evaluated previously in the UFSAR. The margin of safety as defined in the basis of any Technical Specification is not reduced as a result of the modification.

Plant

Reference:

PM 92-016 and PM 92-017

)

L t

30

4 SHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Modification of Containment Cooling Logic for Cooling Valvos BRIEF DESCRIPTION:

The modification removes the high drywell pressure permissive from the control logic of suppression pool cooling valves 1-E11-F024A/B and 1-E11-F028A/B in order to allow suppression pool cooling if a LOCA signalis present.

SUMMARY

OF THE SAFETY EVALUATION:

The probability of occurrence, and consequences of any accident previously evaluated in the UFSAR are not increased as a result of this modification. The probability of occurrence and consequences of malfunction of equipment important to safety previously evaluated in the UFSAR also are not increased. The modification does not create the possibility of an accident of a different type than already evaluated in the UFSAR. The possibility of a malfunction of equipment important to safety of a different type than evaluated previously in the UFSAR is not created by the modification, nor is the margin of safety as defined in the Technical Specifications reduced.

Plant

Reference:

PM 92-046 9

31

^'..

a:

s F

CHANGE TO FACILITY AS DESCRIBED IN THE FSAR TITLE:

Plant Process Computer System Replacement

_ BRIEF DESCRIPTION:

j L

' The modification replaced the Brunswick Nuclear Plant (BNP) Plant Process Computer l

(PPC) System with a new system which utilizes Digital Equipment Corporation's VAX

'4000 computers and associated peripherals. The new PPC has greater hardwar'e/ software.

capability, expansion capability, reliability, and maintainability; as well as upgraded functions to perform more advanced computing and monitoring using the latest computer I

graphic capabilities.

SUMMARY

OF THE SAFETY EVALUATION:

The PPC is used to aid Control Room personnel and other technical staff in assessing i

whether existing plant conditions warrant corrective action. It is not intended as a i

substitute for any safety related equipment or instrumentation, but rather as a supplement to such equipment. The PPC is not essential to the safe operation of the plant; nor is it essential to the prevention of events potentially harmful to the public health and safety.

The PPC is not essential to the mitigation of the consequences of an accident.

The PPC is discussed in UFSAR Section 7. " Control Systems HQ.1 Required for Safety,"

which states "this section discusses the instrumentation and controls of systems whose -

functions are not essential for the safety of the plant. No credit is taken for the operability of these control systems in the plant accident analysis. Their failure does not impair the capability of the protection system in any significant manner nor cause plant conditions to be more severe than those for which the safety systems are designed." The modification does not change the intent of Section 7 of the UFSAR. The modification does not represent an unreviewed safety question.

Plant

Reference:

PM 90-004 and PM 90-005 f

u i

32 y

T g

vv-y tm1

• -m,---MD.

t e

m+

i

• ~

i CHANGE TO FACILITY AS DESCRIBED IN THE FSAR l

t TITLE:

Reactor Vessel Water Level Reference Leg Continuous Backfill BRIEF DESCRIPTION:

The modification installed a continuous water backfillinstallation to the cold reference legs associated with the Reactor Water Levelinstrumentation.

SUMMARY

OF THE SAFETY EVALUATION:

Installation of the modification assures the accuracy of levelindication is within acceptable limits which are assumed in the accident analysis for all operating conditions, including those involving reactor vessel depressurization. The modification does not introduce any i

new accident initiators and thus will not increase the probability of occurrence of an accident evaluated in the UFSAR. The consequences of an accident evaluated previously in the UFSAR are not increased since the backfillinstallation is connected to the non-Q i

portion of the ORD system. The modification does not increase the probability of occurrence of, or consequences of, a malfunction of equipment important to safety evaluated previously in the UFSAR. Analysis of the modification design has determined that the possibility of an accident of a different type than any previously evaluated in the UFSAR is not created. The modification does not create the possibility of a malfunction of equipment impodant to safety of a different type than any previously evaluated in the UFSAR. The margin of safety as defined in the basis of the Technical Specifications is not-reduced by the modification.

Plant

Reference:

PM 93-031 and PM 93-032 33

m CHANGE TO FACluTY AS DESCRIBED IN THE FSAR

)

TITLE:

Post-Accident Sampling System Small Volume Sample Valve

~

BRIEF DESCRIPTION:

The modification replaced the Post-Accident Sampling System (PASS) Small Volume Sample Valve with a Rheodyne Sample injection Valve which has an external volume loop.

i

SUMMARY

OF THE SAFETY EVALUATION:

The PASS is functionally classified as a non-safety-related system. It is not needed to prevent or mitigate the consequences of any accident evaluated in the UFSAR. The modification does not increase the probability or consequences of an analyzed accident.

Neither the probability nor consequences of equipment malfunction is increased. No new accident or failure modes are created. No Technical Specification safety margin is reduced as a result of the modification.

Plant

Reference:

PM 92-113 and PM 92-114 i

34

~

i i

CHANGE TO PROCEDURES AS DESCRISED IN THE FSAR TITLE:

Feedwater Venturi Rubidium Nitrate Calibration Test f

BRIEF DESCRIPTION:

This procedure provides for a more accurate means to determine the feedwater flow rate by using chemical dilution techniques using Rubidium Nitrate (RbNO ) as the tracer.

j 3

SUMMARY

OF THE SAFETY EVALUATION:

This procedure involves the injection of RbNO into the feedwater system to obtain i

3 accurate flow rate measurement. This procedure does not require, alter, or affect the operation of any equipment important to safety. Therefore, no failure of such equipment can be postulated by performing the procedure. The injected RbNO, in a very diluted 3

form, will contact important-to-safety equipment, including control rods, cladding, and other reactor vesselinternals. Likewise, the RbNO will be removed by the RWCU filters.

3 Therefore, injection of RbNO into the feedwater flow stream will not increase the 3

probability or severity of an accident. RbNO was specifically chosen for use as the tracer 3

due to its low impact upon plant systems. Injection of RbNO into the feedwater flow 3

stream will not increase the probability or severity of an accident. No Technical Specification margins are impacted by performing this procedure.

l Plant

Reference:

- BNP Procedure 2-SP-92-001, m

35

c, e

l CHANGE TO PROCEDURES AS DESCRIBED IN THE FSAR TITLE:

Unit 2 - Reactor Recirculation System Operating Procedure BRIEF DESCRIPTION:

The procedure revision adds a section to the operating procedure for the Reactor Recirculation System addressing manual operation of the Unit 2 Reactor Recirculation Pump speed control at the Bailey positioner. The UFSAR was revised to indicate that Section 15.2.6.3 discusses the current analysis for the loss of feedwater transient analysis for both units.

i

SUMMARY

OF THE SAFETY EVALUATION:

The Unit 2 loss of feedwater transient analysis assumes that a Recirculation Pump runback occurs. Manual operation of the scoop tube would disable automatic runbacks to speed limiters Number 1 and 2, and would therefore seem to prohibit any operation of the i

Reactor Recirculation System from the local scoop tube positioner since automatic runback is defeated in this mode. This safety analysis determines that Unit 2 operation in manual scoop tube control may be completely bounded by the Unit 1 analysis, as reported in UFSAR Section 15.OA.6.3, Event 23, " Loss of Feedwater Flow."

The addition of procedural guidance for manual operation of the Recirculation Pump speed control does not affect the probability of any previously analyzed accident or transient.

Manual operation of the Recirculation Pump speed control places no more liability on the operation of the plant than operation with the scoop tube locked as far as the plant response to transients is concerned. Operation with the Bailey positioner in manual does not affect the reliability of any required safety equipment. This change does not increase the consequence of a malfunction of equipment important to safety evaluated previously in j

the UFSAR. No accident initiators are affected by operation of the Reactor Recirculation pump speed controller in manual at the Bailey controller.

Plant

Reference:

EWR 09064 BNP Procedure 2-OP-02, Rev. 77 i

36

I CHANGE TO PROCEDURES AS DESCRIBED IN THE FSAR TITLE:

Emergency Fuel Pool Makeup i

BRIEF DESCRIPTION:

This new special procedure provides guidance for emergency fuel pool makeup when j

required as an alternate method of makeup.

SUMMARY

OF THE SAFETY EVALUATION:

This new special procedure is fully bounded by the safety analysis of EER 92-0336 " Unit 1 Temporary Makeup & Backup to the Spent Fuel Pool" and EER 93-0137 " Unit 2 Spent Fuel Pool Makeup Provisions During RHR B Loop Outage."

I Plant

Reference:

1 -SP-93-008 2-SP-93-009 EER 92-0336 EER 93-0137 e

i I

37

e i

e' CHANGE TO PROCEDURES AS DESCRIBED IN THE FSAR TITLE:

Feedwater Venturi Sodium 24 Calibration Test i

i BRIEF DESCRIPTION:

E This procedure provides instructions on performing a Feedwater Flow Venturi Calibration test on flow instruments. This tert is based on a radioactive isotope dilution technique, using Sodium-24 (Na-24) as the tracer. The test is accomplished by injecting a Na-24 solution of known concentration into a section of feedwater piping upstream of the venturi. Therefore, this provides a means to accurately measure the feedwater flow and l

to compare the flowrate to the instrument readings.

SUMMARY

OF THE SAFETY EVALUATION:

This procedure invoives the injection of Na 24 into the feedwater system to obtain accurate flow rate measurement. Performance of this test does not require operation of i

any equipment important to safety. Therefore, no failure of such equipment can be postulated by performing the procedure. Injection of the Na-24 will not pose any risk to the plant since Na-24 is sufficiently diluted in the feedwater flow stream such that no radiation monitors downstream of the venturi are expected to be tripped. In addition, the reactor coolant specific activity restrictions given in Technical Specification 3/4.4.5 will not be exceeded by performing this test. Therefore, injection of Na-24 into the feedwater flow stream will not increase the probability or severity of an accident and no technical I

specification margins are impacted.

l Plant

Reference:

BNP Procedure 2-SP-91-044 38

+

6 CHANGE TO PROCEDURES AS DESCRIBED IN THE FSAR I

TITLE:

Containment Atmosphere Control System BRIEF DESCRIPTION:

i Section 5.1 of this procedure was revised to inert the containment through the purge exhaust fans instead of the Standby Gas Treatment System.

I

SUMMARY

OF THE SAFETY EVALUATION:

I This revision brings the procedure into compliance with the requirements of the UFSAR.

The probability of occurrence of or consequences of an accident evaluated previously in the UFSAR is not increased by this procedure revision. The probability of occurrence of or consequences of a malfunction of equipment important to safety evaluated previously in i

the UFSAR is not increased as a result of this change. This new method of operation is the method that is required by the FSAR and the possibility of an accident of a different type, or of a malfunction of equipment of a different type, than any evaluated previously in the UFSAR is not created by this method. The margin of safety of any Technical Specification basis is not reduced by this method.

j Plant

Reference:

BNP Procedure 1-OP-24 I

l j

39

=..

CHANGE TO PROCEDURES AS DESCRIBED IN THE FSAR TITLE:

Fuel Pool Cooling Special Operating Procedure BRIEF DESCRIPTION:

The procedure contains the steps to implement the recommended compensatory actions for loss of fuel pool cooling on a single or common mode failure in the system.

SUMMARY

OF THE SAFETY EVALUATION:

The special procedure will not result in additional loss of fuel pool level beyond that which has already been analyzed in the UFSAR. The operation of the fuel pool cooling systems per this special procedure will not increase the consequences of an accident or increase the probability of occurrence of a malfunction of equipment important to safety evaluated previously in the UFSAR. The special procedure does not increase the probability of occurrence of a malfunction to the fuel pool cooling system nor create the possibility of a different type of accident. The procedure does not affect systems important to safety.

Use of this procedure does not result in a reduction of the fuel pool level below the levels required by Technical Specifications.

t i

Plant

Reference:

1 -SP-93-037 EER 93-0251 l

40

S..

TEST OR EXPERIMENT NOT DESCRIBED IN THE FSAR TITLE:

CRD Flow Maximization BRIEF DESCRIPTION:

This special procedure provides backup methods of decay heat removal.

SUMMARY

OF THE SAFETY EVALUATION:

This procedure adds a step to allow automatic CRD flow control below 95 F as an alternate method of decay heat removal. Loss of shutdown cooling has already been analyzed in the UFSAR. Use of this procedure will be in the post-accident phase of a Loss of Shutdown Cooling. This procedure only helps mitigate the consequences of a loss of shutdown cooling. Initial conditions for other accidents and potential malfunction of equipment important to safety evaluated previously in the UFSAR will not be affected.

CRD flow maximization is currently a method employed by the Emergency Operating Procedures to maintain reactor vessel water level. No new initiators for an accident or a malfunction of equipment of a different type than any evaluated previously in the UFSAR are created by use of this procedure. Margins of safety are unaffected by this procedure.

Plant

Reference:

BNP Procedure 2-SP-93-001 41

4s.,

List of Regulatory Commitments The following table identifies those actions committed to by Carolina Power & Light l

Company in this document. Any other actions discussed in the submittal represent intended or planned actions by Carolina Power & Light Company. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Manager-Regulatory Affairs at the Brunswick Nuclear Plant of any questions regarding this document or any associated regulatory commitments.

Committed Commitment date or outage 1.

NONE 2.

3.

i m

--