IR 05000324/2003008

IR 05000325-03-008, IR 05000324-03-008, on 08/11-15/2003 and 08/25-29/2003, Brunswick Steam Electric Plant, Units 1 and 2; Safety System Design and Performance Capability

ML033240610
Person / Time
Site:	Brunswick
Issue date:	10/09/2003
From:	Ogle C NRC/RGN-II/DRS/EB
To:	Keenan J Carolina Power & Light Co
References
-RFPFR IR-03-008
Download: ML033240610 (45)

Text

ber 9, 2003

SUBJECT:

BRUNSWICK SEAM ELECTRIC PLANT - NRC SAFETY SYSTEM DESIGN AND PERFORMANCE CAPABILITY INSPECTION - REPORT NOS. 05000325/2003008and 05000324/2003008

Dear Mr. Keenan:

This refers to the safety system design and performance capability team inspection conducted on August 11 -1 5 and August 2549,2003, at the Brunswick facility. The enclosed inspection report documents the inspection findings, which were discussed on August 29, 2003, with Mr. C. J. Gannon and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license. The team reviewed selected procedures and records, observed activities, and interviewed personnel.

Based on the results of this inspection, one finding of very low safety significance (Green) was identified. This issue was determined to involve a violation of NRC requirements. This finding has very low safety significance and has been entered into your corrective action program. However, the NflC is withholding the treatment of this issue as a non-cited violation as provided by Section VI.A.4 of the NRC's Enforcement Policy, pending our review of your corrective actions related to restoration of compliance. lf you contest this finding, you should provide a response with the basis for your concern, within 40 days of the date of this inspection report to the Nuclear flegulatory Commission, ATTN: Document Control Desk, Washington, BC 20555- *1001 ~ with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001 ; and the NRC Resident Inspector at the Brunswick faciiity.

In accordance with 10CFR 2.790 of the NRC's "Rules of Practice,"

a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system ATTACHMENT

CP&L 2 (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

~ Enaineerina Bran Division of iieactor Safety Docket N O S.: 50-325,50-324 License Nos.: DPR-71, DPR-62

Enclosure:

NRC Inspection Report w/Attachment:

Supplemental Information

REGION 11 50-325,50-324 DPW-71, BPW-62 05000325/2003008 and 05000324/2003008 Carolina Power and Light Brunswick Steam Electric Plant, Units I and 2 8470 River Road SE Southport, NC 28461 August 11-15, 2003 August 25-29,2003 J. Moorrnan, Senior Reactor Inspector (Lead Inspector)

N. Merriweather, Senior Reactor Inspector R. Schin, Senior Reactor Inspector (Week 1 only) M. Thomas, Senior Reactor Inspector M. Mayrni, Reactor Inspector (Week 2 only) N. Staples, Reactor Inspector Charles R. Ogle, Chief Engineering Branch 1 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

bR 05000325/2003-008, 05000324/2003-008; 08/11-15/2003 and 08/25-29/2003;

Brunswick Steam Electric Plant, Units 1 and 2; safety system design and performance capability.

This inspection was conducted by a team of inspectors from the Region II office. The team identified 1 Green unresolved item. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 0609, "Significance Determination Process" (SBP). Findings for which the SBP does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated Juiy 2000.

A. NRC-Identified

and Self-Revealina Findinas

Cornerstone: Mitigating

Systems m. The team identified a violation of 10 CFR 50, Appendix B, Criterion Ill, Qesign Control requirements.

The Technical Specification (TS) allowable value for the Condensate Storage Tank (CST) Level - Low function, for automatic high pressure coolant injection (HPCI) pump suction transfer to the suppression pool, was not adequately supported by design calculations.

The calcuIations did not adequately address the potential for air entrainment in the HPCI process flow due to vortexing.

This finding is in the licensee's corrective action program as Action Request 102456. This finding is unresolved pending further NRC review of the requirements for the CST Level - Low function and of the corrective actions related to restoration of compliance with 10 CFR 50, Appendix B, Criterion 111, Design Control requirements.

The finding is greater than minor because it affects the design control attribute of the mitigating systems cornerstone objective.

It is of very low safety significance (Green) because the finding is a design deficiency that will not result in loss of the HPCl function per BL 91- 18 (Rev. I) and the likelihood of having a low level in the CST that would challenge the CST level - low automatic HPCI suction transfer function is very low. In addition, alternate core cooling methods would normally be available, including reactor core isolation cooling (RCIC) as well as automatic depressurization system and low pressure coolant injection. (Section 1821.1 1. b)

B. Licensee-Identified Violations

None

REPORT DETAILS

REACTOR SAFETY

Cornerstones:

Initiating Events and Mitigating Systems 1821 Safety Svstem Desian and Performance Casabilitv (71 11 1.21) This team inspection reviewed selected components and operator actions that would be used to prevent or mitigate the consequences of a loss of direct current power event. Components in the high pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), and 125E5.0 volt

(v) direct current

(dc) electrical systems were included.

This inspection also covered supporting equipment, equipment which provides power to these components, and the associated instrumentation and controls.

The loss of dc power event is a risk-significant event as determined by the licensee's probabilistic risk assessment. .I .I 1 a. b. Svstem Needs Process Medium Inspection Scowe The team reviewed the licensee's installed configuration and calculations for water volume in the condensate storage tank (CST) and for net positive suction head for the HPCI pump. This included reviews of system drawings and walkdown inspection of installed equipment to compare arrangements and dimensions to those used in the calculations.

The team also reviewed the licensee's calculations supporting the Technical Specification (TS) setpoint for the CST level instrumentation which initiates an automatic transfer of the HPCB pump suction from the CST to the suppression pool. This included checking the adequacy of the calculations and comparing calculated values to values in the TS and in the instrument calibration procedures.

Findines introduction:

An unresolved item of very low safety significance (Green) was identified for inadequate design control of the HPCI suction source from the CST. The calculations which determined the CST low level setpoint for automatic HPCl system suction transfer from the CST to the suppression pool did not adequately account for air entrainment in the process flow due to vortexing.

This finding involved a violation of NRC requirements.

However, it is unresolved pending further NRC review of the requirements for the CST bevel - bow function and corrective actions related to restoration of compliance.

Description:

Vortexing in pump suction sources is a well known phenomenon.

It is discussed in typical textbooks on centrifugal pumps. NRC Regulatory Guide I.8z5 "Sumps for Emergency Core Cooling and Containment Spray Systems,"

dated June 1974, discussed the need to preventing vortexing.

Regulatory Guide 1.82, Rev. 1, dated November 1985, and Rev. 2, dated May 1996, included specific guidance on how to prevent air ingestion due to vortexing in containment heat removal systems. That 2 guidance included limiting the Froude number (Fr) to less than 0.8 for BWW suppression pool suctions

[where Fr is equal to the inlet pipe velocity (U) in feet per second divided by the square root of (the suction pipe centerline submergence below the water level (S) in feet times gravity

(9) in feet per second squared}].

NRC NUREG / CR-2772, "Hydraulic Performance of Pump Suction Inlet for Emergency Core Cooling Systems in Boiling Water Reactors?" dated June 1982, included experiments on suctions from tanks and showed almost no air entrainment with a Fr of 0.8. The experiments also showed that air entrainment increased dramatically when Fr reached 1.0. The BWR Owners' Group Emergency Procedure Guidelines included guidance on preventing vortexing in emergency core cooling system pump suctions from the suppression pool. This guidance included a vortex limit curve based on maintaining Fr less than 0.8. All of the above references addressed suction pipes that extended into a LanWsump.

A more recent research paper published in 2001 by ASME titled "Air Entrainment in a Partially Filled Horizontal Pump Suction Line" described tests on air entrainment.

The tests were conducted at various flowrates, in a horizontal suction pipe that did not extend into the a tank; a configuration similar to the HPCl suction from the CST at Brunswick. The paper's conclusions about vortexing and air entrainment at high flow rates were similar to those of the previous references where a suction pipe extended into a tank. Brunswick Units 1 and 2 TS Table 3.3.5.1-1 stated that the allowable value for the HPCl system automatic suction transfer from the CST to the suppression pool was a low CST level of 2 23 feet 4 inches above mean sea level. (NQTE:

That value represented 3 feet 4 inches above the bottom of the CST.) Once initiated, the HPCI suction transfer involved first opening the suppression pool suction valves (E41-FO41 and F042) and then closing the CST suction valve (E41-FOO4). The Updated Final Safety Analysis Report (UFSAR) stated that for each unit's CST: "...the HPCl and RCIC pumps take suction through a 16-inch line connected to the tank with a nozzle centerline 2 feet above the tank bottom. Level instruments will initiate an automatic transfer of the pumps' suction path to the suppression pool suction if level approaches this connection. For HPCl the setpoint is above the 3.3-foot TS limit and below the 3.5-foot calibration maximum allowed value.

To allow time for the suction transfer to take place, this setpoint provides a margin of approximately 10,000 gallons in the tank after the setpoint is reached and before air will be entrained in the process flow." The calculation of record that supported the TS allowable value was Calculation OE41- 1001, "High Pressure Coolant Injection System Condensate Storage Tank Level Low Uncertainty and Scaling Calculation

[E41 -LSL-N002(3)

Loops]," Rev. 1, dated March 29, 1999. The team noted that Calculation OE41-1001 stated that its objective was to determine the allowable value and setpoint for the CST low water level trip function for the HPCl system. However, the calculation did not include a hydraulic analysis to determine the allowable value.

Instead, it relied on a design basis input from Engineering Service Request (ESR) 97-00026, Action Item 2, for the allowable value.

3 ESR 97-00026, Action Item 2, stated its objective:

"... the analytical limit for the HPCI and RClC CST low level transfer function is 23 feet 4 inches. Provide a basis for this analytical limit. The basis should address air voids ..." It also stated: "This ESR action item will show that using the TS limit as the analytical limit is acceptable."

The ESW included Condition Report (CR) 97-02379 Task 2 (approved August, 27,1997) as an attachment.

The team noted that the ESW relied entirely on CR 97-02379 Task 2 for concluding that using the TS limit as the analytical limit was acceptable.

However, the ESR also stated: "This CR review was not conducted as a design basis input with formal testing and design verification."

CR 97-02379 Task 2 stated that its objective was to determine if a vortexing problem existed in the CST when running the HPCO pump. Task 2 further stated that it was responding to an operating experience event where a nuclear plant had identified that they had failed to account for unusable volume In their CST due to vortexing concerns.

It described a scale model test that had been performed by another nuclear plant to conclude that no vortexing would occur in their CST. However, the CR noted reasons why this test could not be relied upon as a design input. The CR also contained results from an informal test performed by the licensee.

The CR concluded that, based on the results of the informal testing and engineering judgement, air ingestion may briefly occur during the transfer process; however, the air ingestion would be of such limited duration and such a small percentage that there was no concern for damage to the HPCI pumps. The team noted that the informal test used a small scale model without determination that the results would be applicable to the installed CST and HPCl suction, the test was performed without calibrated instruments, and the test was not independently verified.

The team considered that the informal test was not suitable for use as an input to a design basis calculation.

Subsequently, action request (AR) 00005402 documented an engineering audit concern with relying on ESR 97-80026 as a design basis input to a calculation.

ESW 01-00322 was then written to respond to AR 00005402.

ESR 01-08322 stated that its purpose was to document the technical resolution of the CST intake vortex formation issue and to insert appropriate references into design documents.

ESR 01 -00322 included an extensive review of reference documents on vortexing.

It included references to LERs and INPO Event Reports on vortexing issues at other nuclear plants; NUREWCR-2772; and several research papers on vortexing.

The team noted that ESR 01-00322 did not reference NRC Regulatory Guide 1.82. ESR 81-00322 agreed with the conclusions of CR 97-02379 and ESR 97-00026 that the TS allowable value of 23 feet 4 inches was adequate.

It concluded that the potential for a significant air ingestion event was of sufficiently low probability to be considered non- credible.

The team noted that this conclusion was based primarily on the CR 97-02379 informal test and on a research paper by A. Daemi of the Water Research Center in Tehran, Iran, that had been presented to the American Society of Civil Engineers in 1998. The research paper tested the effect of an intake pipe protruding various distances into a reservoir and found that a pipe that did not protrude into the reservoir showed some vortexing but no air entrainment while a pipe that did protrude into the reservoir would have significant vortexing and air entrainment into the pipe. ESR 01- 00322 considered that, since the NUREG/CR-2272 tests used a configuration where the 4 suction pipe protruded into the tank and the licensee's HPCl suction pipe did not protrude into the CST, the NUREG/CR-2272 conclusions were not applicable to the Brunswick design. The NRC team noted that the research paper by A. Baemi was significantly flawed for applicability to Brunswick in that it did not state what flowrates were used in its tests and apparently used gravity flow. Regulatory Guide 1.82 and NUREG/CR-2272 indicate that flow velocity is one of the most important factors in vortex formation.

A suction pipe that would have little or no vortexing at low flow velocities (e.g., gravity flow) could have significant vortexing at higher flow velocities (e.g., a HPCI pump at 4300 gprn). The team considered that both sources of information on which the conclusions of E§R 01-00322 were based were not suitable for use as inputs to safety-related design calculation OE41-1001.

The HPCl pump was designed to automatically start and establish a flowrate of 4300 gpm. Licensee procedures did not contain guidance to reduce that flowrate when the CST level approached the low level switchover setpoint.

Using the NUREG/CR-2272 methodology, the team calculated that, at a HPCI pump flowrate of 4300 gpm, an Fr of 0.8 would be reached at a CST level of 5.0 feet and an Fr of 1 .O would be reached at a CST level of 3.9 feet. Considering the automatic suction transfer actuation setpoint and the valve stroke times, the HPCB pump suction pipe could be exposed to a suction Fr in excess of 0.8 (some air entrainment)for about 8.9 minutes and over 1 .O (over 2% air entrainment)for about 5.0 minutes. Calculations that used the 2001 ASME research paper equations provided different results: air entrainment in the process flow would start at a tank level of 3.2 feet and would exceed 2% at tank levels below 3.0 feet. This would represent a HPCI pump suction pipe exposure to some air entrainment in the process flow for about 1.8 minutes and to over 2% air entrainment for about 1.1 minutes. The team concluded that the plant design was not consistent with the UFSAR in that the TS allowable value for the HPCl automatic suction transfer would not prevent air from becoming entrained in the HPCl process flow. During this inspection, team and licensee measurements of the installed CST configuration revealed non

-conservative errors of about 1.5 inches in the actual heights of the Units 1 and 2 CST level switches above the HPCl suction pipes. These would result in additional non-conservative errors in the HPCI automatic suction transfer setpoints.

The licensee entered this issue into their corrective action program as AR 102456. This AR included an operability determination and planned corrective actions that were reviewed by the team. The operability determination concluded that the CST Level - Low instrument was operable with the existing TS allowable value and related setpoint and no compensatory measures were needed. This conclusion was based on the following:

1) HPCl operation during design or licensing basis events would not challenge the CST Level bow instrument; and 2) Operator actions consistent with plant procedures would not result in 4300 gpm HPCl flow for the full duration of the suction transfer.

The operability determination did not include an analysis which assured that the instrument's allowable value was adequate to prevent significant air entrainment during the full duration of a CST bevel - Low setpoint initiated suction transfer while the HPCl pump was operating at its maximum flowrats of 4300 gpm.

However, the team's interpretation of licensing basis documents indicated that the CST 5 Level - Low function was required to be able to protect the HPCl pump from damage from any suction hazard that could occur. This inciuded air entrainment in the process flow due to vortexing that would result if the CST level became low while the HPCI pump was operating at about 4300 gpm, even if this could only occur outside of a design basis event. The licensee's corrective actions for this issue were in AR 102456. This AB included only two planned corrective actions.

The first corrective action was: "Issue a UFSAR change package to correct the description of HPCB air entrainment potential during suction swap." Phis was described in more detail in the AB under Section 3, Inappropriate Acts, item 4: "Error 4 was a simple text error by BNP engineering where the concept was understood (no significant air at the pump) but was not translated into specific detailed words." The second corrective action was: "Issue an evaluation to update the HPCI CST level switch design basis information to reflect the evaluation provided in the operability review portion of this AW." The operability determination portion of the AR concluded that the CST Level - Low automatic HPCl suction transfer function would not be challenged during design basis events and consequently the TS allowable value was adequate. The documented corrective actions in AR 102456 did not appear to be sufficiently comprehensive to restore compliance with 10 CFR 50, Appendix B, Criterion 111, Design Control. The licensee's planned corrective actions did not Specifically include revising the design calculation, OE41-1001.

In addition, they did not include assuring that the CST Level Low suction transfer function will protect the flPCl pump if it is operating at its maximum flowrate during the transfer.

The planned corrective actions identified in the AR did not include obtaining a certification from the pump vendor that the pump can withstand a certain amount of air in the process flow for a certain amount of time without pump damage. [This was subsequently done by the licensee.]

The planned corrective actions identified in the AR also did not include submitting a license amendment request to the NKC to revise the TS allowable value, remove the CST Level - Low function from TS, or add an operator action to throttle HPCl pump flow at low CST levels so that the existing setpoint will be able to protect the pump. This issue will remain unresolved pending further NRC review of the design basis and operability requirements for the CST Level - Low suction transfer function.

Specifically, the NRC will review whether the CST Level - Low function is required to be able to protect the HPCI pump from damage only during design basis events; or if it is required to be able to protect the HPCI pump from damage due to air entrainment if the level is the CSB becomes low with the HPCI pump operating at a flowrate of about 4300 gpm, even if this could only occur outside of a design basis event. Analvsis:

Design Calculation OE41-1001, for the CST Level - Low setpoint and TS aliowable value was inadequate.

The finding is greater than minor because it affects the design control attribute of the mitigating systems cornerstone objective.

It is of very low safety significance (Green) because the finding is a design deficiency that will not result in loss of the HPCl function per GL 91-18 (Rev. 1) and the likelihood of having a low level in the CST that would challenge the CST bevel - Low automatic HPCI suction transfer function is very low. In addition, alternate core cooling methods would normally 6 be available, including RCIC as well as automatic depressurization system and low pressure cooiant injection.

Enforcement:

10 CFR 50, Appendix B, Criterion Ill( Design Control, requires in part, that design control measures shall include provisions to assure that appropriate quality standards are specified and included in design documents.

Contrary to the above requirements, the NRC identified during this inspection that, from 1999 to August 2003, licensee Calculation OE41-1001 and associated design documents did not adequately consider air entrainment in the HPCl pump process flow due to vortexing in the CST for the current TS value for the CST Level bow setpoint for automatic transfer of the HPCl pump suction from the CST to the suppression pool. This finding was entered into the licensee's corrective action program as Action Request 102456 and is unresolved pending further NRC review of the requirements for the CST Level - Low function and of the licensee's corrective actions related to restoration of compliance with Criterion Ill of 18 CFW 50, Appendix E. This finding is identified as UBI 05000325, 324/2003008-01, Failure to Adequately Consider Vortexing in the Calculation for CST Level for Automatic Transfer of the HPCI Pump Suction. .I2 Enerav Sources a. lnsoection Scow The team reviewed appropriate test and design documents to verify that the 12.9250 vdc power source fur HPCl system valves and controls would be available and adequate in accordance with design basis documents.

Specifically, the team reviewed the 125'250 vdc battery lead study, 125 vdc battery charger sizing calculation, and 125/250 vdc system voltage drop study, and battery surveillance test results, to verify that the dc batteries and chargers had adequate capacity for the loading conditions which would be encountered during various operating scenarios.

The team reviewed a sample of HPCl motor operated valves (MOVs) to verify the adequacy of available motor output torque, stroke times, thermal overload heater sizing, and valve performance at reduced voltages.

The team also reviewed portions of a voltage study to verify adequacy of voltage for HPCl solenoid valves l-E41-F025 and -F026 under worst case voltage conditions.

A list of related documents reviewed are included in the attachment.

The team reviewed design basis descriptions and drawings and walked down the HPCl and RClC systems to verify that a steam supply would be available for pump operation during a loss of station dc power event. This included review of the steam supply drain systems and review of a recent modification to the HPCI steam supply drain system. The team reviewed the HPCl steam supply drain pot flow orifice inspections; the drain pot level switch logic and calibration records, and the drain pot drain line isolation valves modification to verify that the HPCl steam supply would be available if needed. The team reviewed functional valve testing fur the HBCl and RClC turbine exhaust vacuum breaker check valves to verify adequacy of acceptance criteria and to verify that vacuum breaker functionality was being maintained.

7 b. Findinas No findings of significance were identified. .I 3 Instrumentation and Controls

a. Inspection Scope

The team reviewed electrical elementary and logic diagrams depicting the WPCI pump start and stop logic, permissives, and interlocks to ensure that they were consistent with the system operational requirements described in the UFSAR. The team reviewed the HPCI auto-actuation and isolation functional surveillance procedures and completed test rscords to verify that the control system would be functional and provide desired control during accident and event conditions in accordance with design. The team reviewed the calibration test records for the CST low water level instrument channels to verify that the instruments were calibrated in accordance with setpoint documents. The team also reviewed the records demonstrating the calibration and functional testing of the HPCI suppression pool high level instrument channels to determine the operability of the high level interlock functions of HPCI. b. Findinas No findings of significance were identified. .I4 Operator Actions a. Inspection Scone The team assessed the plant and the operators' response to a Unit 1 initiating event involving a loss of station battery 18-2. The team focused on the installed equipment and operator actions that could initiate the event or would be used to mitigate the event. The team reviewed portions of emergency operating procedures (EOPs), abnormal operating procedures (AOPs), annunciator panel procedures (APPs), and operating procedures (OPs) to verify that the operators could perform the necessary actions to respond to a loss of dc power event. The team also observed simulation of a loss of dc power event on the plant simulator and walked down portions of Procedure OAOP-39, "Loss of DC Power." The simulator observations and procedure reviews focused on plant response and on verifying that operators had adequate instrumentation and procedures to respond to the event. The team reviewed operator training records (lesson plans, completed job performance measures, etc.) to verify that operators had received training related to a loss of dc power event.

b. Findinas No findings of significance were identified.

8 .I5 Heat Removal

a. Inspection Scope

The team reviewed historical temperature data for the Unit 2 battery rooms to verify that the minimum and maximum room temperatures were within the allowable temperature limits specified for the batteries. The team reviewed heat load and heat removal calculations for the HPCl and RClC rooms. The team also reviewed the calculated peak temperature and pressure responses during high energy line break and loss of coolant accidents for these rooms. The team reviewed service water temperature and flow requirement calculations for the HPCl and RClC rooms and fan coolers. These reviews were conducted to verify the adequacy of design for the room coolers, and to verify that heat will be adequately removed during a loss of dc power event. The team also reviewed HPCI and RClC room cooler thermostat calibrations, inspection and cleaning records, and corrective maintenance history to verify room coolers were properly maintained and would be available if called upon. b. Findinas No findings of significance were identified.

2 System Condition and CaDability 21 Installed Confiauration

a. Inspection Scope

The team visually inspected the 125/250 vdc batteries and battery chargers, dc distribution panels, dc switchgear, and dc ground detection systems in both units to verify that the dc system was in good material condition with no alarms or abnormal conditions present and to verify that alignments were consistent with the actions needed to mitigate a loss of dc power event. The batteries were inspected for signs of degradation such as corrosion, cell discoloration, plate buckling, grid cracks, and excessive plate growth.

The team waiked down the HPCI and RCIC systems and the CST to verify that the installed configuration was consistent with design basis information and would support system function during a loss of dc power event. The team walked down portions of the HPCI system to verify that it was aligned so that it would be available for operators to mitigate a loss of dc power event. During this walkdown, the team compared valve positions with those specified in the HPCI system operating procedure lineup, and observed the material condition of the plant to verify that it would be adequate to support operator actions to mitigate a loss of dc power 9 event. This also included reviewing completed surveillance tests which verified selected breaker positions and alignments. b. Findines No findings of significance were identified.

22 Desian Calculations a. Inspection ScoDe The team reviewed the thermal overload sizing calculations for a sample of Unit 1 HPCI MOVs to verify adequacy of the installed overload relay heaters. The team also reviewed calculations that assessed the stroke times and motor torque produced at reduced voltage to verify that they would exceed or meet minimum specified requirements. The valves and calculations reviewed are listed in the attachment.

The team reviewed design basis documents, probabilistic risk assessment system notebooks, UFSAR, selected piping and instrumentation diagrams, selected TSs, system reviews, ARs, and the corrective maintenance history for HPCl and RClC systems to assess the implementation and maintenance of the HPCI and RCIC design basis. b. Findinas No findings of significance were identified.

.23 Testing and InsDection

a. The team reviewed the 125/250 vdc battery surveillance test records, including performance and service test results, to verify that the batteries were capable of meeting design basis load requirements. The team reviewed functional and valve operability testing (stroke times), and corrective maintenance records for HPCl and RClC selected valves, including the minimum flow bypass valves, and steam admission valve. This review was conducted to verify the availability of the selected valves, adequacy of surveillance testing acceptance criteria, and monitoring of selected valves for degradation. The team reviewed HPCI and RCIC system operability tests to verify the adequacy of acceptance criteria, pump performance under accident conditions, and monitoring of system components for degradation.

b. Findinas No findings of significance were identified.

.3 31

a. b. 32 a. b.

.33 a. Selected Components

Component Dearadation InsDection Scope The team reviewed in-service trending data for selected components, including the HPC! and RClC pumps, to verify that the components were continuing to perform within the limits specified by the test. The team reviewed the maintenance history of the 125/250 vdc batteries, 125 vdc battery chargers, and selected 41 60 v alternating current

(ac) and 480 vac breakers to assess the licensee's actions to verify and maintain the safety function, reliability, and availability of the components in the system. The team also reviewed the preventive maintenance performed on selected 41 60 vac and 480 vac breakers to verify that preventive maintenance was being performed in accordance with maintenance procedures and vendor recommendations.

The specific work orders and other related documents reviewed are listed in the attachment.

Findinas No findings of significance were identified.

Eauipment/Environmental Qualification Inspection Scope The team conducted in-plant walkdowns to verify that the observable portion of selected mechanical components and electrical connections to those components were suitable for the environment expected under all conditions, including high energy line breaks. Findinos No findings of significance were identified.

Eauipment Protection inspection Scope The team conducted in-plant walkdowns to verify that there was no observable damage to installations designed to protect selected components from potential effects of high winds, flooding, and high or low outdoor temperatures.

The team walked down the HPCI and RClC systems and the CST to verify that they were adequately protected against external events and a high energy line break.

11 Findinas No findings of significance were identified.

Oueratinq Experience lnsuection Scope The team reviewed the licensee's dispositions of operating experience reports applicable to the loss of de power event to verify that applicable insights from those reports had been applied to the appropriate components.

Findinos No findings of significance were identified.

Identification and Resolution of Problems lnsuection Scose The team reviewed corrective maintenance work orders on batteries, battery chargers, and ac breakers to evaluate failure trends. The team also reviewed Action Requests involving battery problems, battery charger problems, and charger output breaker problems to verify that appropriate corrective action had been taken to resolve the problem. The specific Action Requests reviewed are listed in the attachment.

The team reviewed selected system health reports, maintenance records, surveillance test records, calibration test records, and action requests to verify that design problems were identified and entered into the corrective action program. Findinus No findings of significance were identified.

Other Activities b.

.34 a.

b.

.4 a.

b. 4. 40A6 Meetinos.

lncludina Exit The lead inspector presented the inspection results to Mr. C. J. Gannon, and other members of the licensee staff, at an exit meeting on August 29, 2003. The inspectors confirmed that proprietary information was not provided or examined during this inspection.

SUPPLEMENTAL INFORMATION

KEY PQINTS OF CONTACT Licensee b. Beller, Supervisor, Licensing

E. Browne, Engineer, Probabilistic

Safety Assessment

8. Cowan, Engineer 6. Elberfeld, Lead Engineer

P. Flados, HPCB System Engineer

N. Gannon, Director, Site Operations

M. Grantham, Design

C. Hester, Operations

Support

D. Hinds, Manager, Engineering

G. Johnson, NAS Supervisor

W. Leonard, Engineer

T. Mascareno, Operations

Support

J. Parchman, Shift Technical

Advisor, Operatiofls

C. Schacker, Engineer 6. Stackhouse, Systems

H. Wall, Manager, Maintenance

K. Ward, Technical

Services _D NRC (attended

exit meeting)

E. DiPaoio, Senior flesident

Jnspector

J. Austin, Resident Inspector

LIST OF ITEMS OPENED, CLOSED AND DISCUSSED

Opened 0500032~,324/2003008-~~

UBI Failure to Adequately

Consider Vortexing

in the Calculation

for CST Level for Automatic

Transfer of the HPCI Pump Suction (Section 7 R21.17. b) Attachment

LISP OF DOCUMENTS

REVIEWED Procedures

OAI-115, 125/250 VPC System Ground Correction

Guidelines, Rev. 6 OAOP-36.1, boss of Any 41 60V Buses or 48OV E-Buses, Rev. 25 OAOP-39.0, Loss of DC Power, Rev.

001-01.02, Shift Routines and Operating Practices, Rev. 31 001-50, 125i250 VDC Electrical

Load List, Rev.

OOP-50.1, Diesel Generator

Emergency

Power System Operating Procedure, Rev. 55 OPM-ACU500, Inspection and

Cleaning of the RHWCore Spray Room Aerofin Cooler Air Filters

1APP-,445, Annunciator Procedure for Panel A

-05, Rev. 46 IAPP-UA-23, Annunciator Procedure for

Panel UA-23, Rev. 45 1 EOP-01 -RSP, Reactor Scram Procedure, Rev. 8 f OP-19, High Pressure

Coolant Injection

System Operating

Procedure, Rev. 58 16P-50, Plant Electrical System Operating Procedure, Rev. 64 1OP-51, DC Electrical System Operating Procedure, Rev. $0 2APP-A-01, Annunciator Procedure for Panel

A-81, Rev. 44 OPIC-TMRQ02, Calibration

of Agastat 7020 Series Time Delay Off Relays, Rev. 18 OPM-BKR001 , ITE 4KV-line Breaker and compartment checkout, Rev

OPM-BKR002A, IT* K-line Circuit Breakers, Rev

OPM-TRB518, HPCI & WClC Steam Inlet

Brain Pot Flow Orifices Inspection, Rev. 3 Drawinqs 1-FP-60085, High Pressure Coolant Injection System Unit

1, Rev. J Contract No. 71-2162, Dwg. No. 1, General Plan for Condensate Storage Tanks by Brown

& D-02523, High Pressure Coolant Injection

System Unit 2, Sh. 1 & 2, Rev. 52 & 45 8-02529, Reactor Core Isolation Cooling System

Unit 2, Sh. 1 & 2, Rev. 52 & 36 8-25023, Sheet 2, Unit 1 High Pressure Coolant Injection System Piping Diagram, Rev. 45 D-25023, Sheet

I Unit 1 High Pressure Coolant Injection System Piping Diagram, Rev. 54 F-03044, Units 1 & 2 480 Volt System Key Qne Line Diagram, Rev. 38 LL-7044, Instrument Installation

Details Units 1 & 2, Sh. 15, Rev. 10 Calculations

OE41-1001;

High Pressure Coolant Injection System - Condensate Storage Tank Level - Low 9527-8-E41-06-F;

NPSH Requirements - HPCI and RCIC; dated March 26, 1987

BNP-E-6.033, AC/DC MOV Thermal Overload Sizing Calculations, Rev. 3 BNP-E-6.062, 125i250 Volt DC System Voltage

Drop Study, Rev.

BNP-E-6.074, 125i.250 Volt DC Battery Load Study, Rev. 2

BNP-E-6.079, 125 Volt DC Battery Charger Sizing Calculation, Revision

BNP-E-6.109, Unit 1 Stroke and Motor Torque Calculations for

250VDC Safety-Related MOVs, BNP-E-8.013, Motor Torque Analysis for

AC MQVs, Rev. 4 and Coolers, Rev. 7' Root, lnc; Rev. C Uncertainty and Scaling Calculation (E41 -bSL-N002(3)

Loops), Rev. I, dated March

29, 1999 Rev. 5

BMP-EQ-4.001, Temperature

Response in RHR and HPCl Rooms Following

LBCA with BNP-MECH-E4I -F002, Mechanical

Analysis Report to Verify Minimum Torque Availability, BNP-MECH-RBER-001, Reactor Building Environmental

Report, Rev. OA WAC Flow Rates, Rev. 0 M-89-0021; HPCllRCIC

NPSH with Suction from the CST; Rev. 0, dated November 27, 1989 PCN-G0050A, RHR Room Cooler Allowable

Service Water Inlet Temperature, Rev. 2 Desian Basis Bocuments

DBD-19, High Pressure Coolant Injection

System, Rev. f 1 DBD-51, DC Electrical

System, Rev. 5 Enaineerina

Service Requests ESR 97-0026; Provide a Basis for the Analytical

Limit for the HPCl and RCIC CST bow bevel ESR 98-00067;

HPCI/RCIC

Reserve Capacity in CST; Rev. 1, dated February 17, 1998 *SI? 99-00404; #PCI/WCIC

Drain Pot Piping Boundary Changes; dated February 25,2000 ESR 01-00322;

Document the Technical

Resolution

of the CST Intake Vortex Formation

Issue; ESR 99-00405, HPCl Design Conversion

To Fail Open for E-41-F028/29, Rev. 0 Updated Final Safetv Analvsis Reuort UFSAR Section 54.6, Reactor Core Isolation

Cooling System UFSAR Section 6.3, Identification

of Safety Related Systems - Emergency

Core Cooling UFSAR Section 7.1.1.2, Emergency

Core Cooling Systems UFSAR Section 8.3.2, BC Power Systems UFSAR Section 9.2.6, Condensate

Storage Facilities

Improved Technical

Soecifications

Section 3.5.1, ECCS - Operating

Section 3.5.3, RCIC System Section 3.8.4, DC Sources - Operating

Section 3.8.6, Battery Cell Parameters

Section 3.8.7, Electrical

Distribution

Systems s Operating

TS Bases Section 3.5; Emergency

Core Cooling Systems and Reactor Core Isolation

Cooling Reduced Rev. 3 Transfer Function;

dated November 24, 1997 dated September

25,2001 Systems System List of Valves lnsoected

1-E41-F0011

HPCl Steam Supply Valve l-E41-F006, HPCI Main Pump Discharge Valve

1-E41-F007, HPCl Main Pump Discharge

Valve ?-E41+008, HPCI Test Bypass to CST Valve

1-*41-F011, WPCl Redundant

Shutoff to CST Valve 1-E41-F012, HPCl Test Line Miniflow Valve 1-E41-F04lI

HPCI Suppression

Pool Suction Valve 1-E41-F042, HPCE Pump Suction Valve Completed

Maintenance

and Tests OPT-09.2, HPCI System Operability

Test, completed

06/29/03, 04/03/03, 01/10/03, 08/20/03, OPT-20.10, Testing of Valves E4l-FO96, E44 -FO99, *51 -F063, E51 -F064, completed

04/24/02, OPT-10.1 1, RClC System Operability

Test, completed

06/06/03, 03/14/03, 12/20/82, 07/31/03, OPT-09.3, HPCl System ~ I65 Psig Flow Test, completed

04/20/03, 03/26/01, 03/29/02, OPT-09.7, HPCl System Valve Operability

Test, completed

09/25/03, 05/02/03, 02/07/03, 05/01/03, 04/01/03 OPT-10.1 .El, RClC System Valve Operability

Test, completed

09/04/03, 0411 0103, 07/03/03, 04/09/030PT-10.1.3, RClC System Operability

Test - Flow Rates at 150 Psig, completed

0311 8/QO, 03/29/02, 03/23/01, 04/02/03 05/29/03,04/04/03

03/08/02, 0311 0/03,04/22/02

05/08/03, 04/03/03 03/23/00 Completed

Work Orders (WOs) and Work Requests (WRs) WO 49443-01, HPCl Turbine Restricting

Orifices Inspection, completed

0311 3/01 WO 49442-01, RClC Turbine Restricting

Orifices Inspection, completed

03/15/01 WQ 45998-01, HPCl Turbine Supply Steam Drain Pot Hi Level Switch Calibration (Unit 2), WQ 192543-01, HPCl Steam Supply Valve 2-E41-F001

Repairs due to Leakage Past the Seat, WO 4581941. HPCl Turbine Sugnlv Steam Drain Pot Hi bevel Switch Calibration (Unit I), completed

2/06/01 completed

03/31/03 .. ~ completed

i/25/Oi WO 46107-01, Calibration

of RHR Room Cooler Thermostats, completed

11/09/80 WO 53172-01;

Inspection

& Cleaning of iqe RHR Roorrl Cooler, cotnpleted

03/05/02 WO 50171-01, Inspectioil

R Cleartiny

of the HI-iR Room Cooler, completed

03/05/02 WR AFQO 001, HPCI Turbine Supply Stem Drain Pct Hi Level Switch Calibration (Uqit 2), WR AlTl 001, HPCI Turui!ie Supply Steam Drain Po! Hi Level Switch Caliwation (Unit 1). WR ABPD 063, Calibration

of PCIR Room Cooler Thetmostars, completed

09/13/00 WR ABPD 002. Caiibratiori

of HHH Room Cooler Thermosta!s, completed

08/25/97 WR AGEB 002, Calibratiop

of HHH Room Cooler Thsrmosats, comple;ed

08/21/97 WR AlWK 004, Inspectian

& Cleaning of the HI-IH Rocm Cooler, completed

C3/09/02 WWJO ANRROOl, 1A-1 Ba:teries, 125 VDC, Perfcrmacice

Capaci!y Test WW:O ANTKGOI, 1A-2 Bat:er:es, 'I25 VUC, Performarice

Capacity Test WWLO ANSN001, 1 B-1 Batteries, 125 VDC, Performarm?

Capacity Test WR/;O ANSTOOl, 10-2 Batteries, 125 VDC, Performance

Capacity Test WO 0004C;46SOI, 28-1 Batteries, 125 VDC, Performance

Capacity Test WO 0004546C3:, 28-2 Batteiies, 125 VDC, Pertormance

Capacity Test completed

06/07/96 cmpieted 08/03/95

WO 0004546301,2A-I

Batteries, 125 VDC, Performance

Capacity Test WO 0004546601,2A-2

Batteries, 125 VBC, Performance

Capacity Test WO 0004635001, 18-2 Batteries, 125 VDC, Service Capacity Test WO 0004635101, 1A-1 Batteries, 125 VDC, Service Capacity Test WO 0004634901, 1 B-1 Batteries, 125 VDC, Service Capacity Test WO 0004634801, 1 B-2 Batteries, 125 VDC, Service Capacity Test WO 0017812801, 2B-2 Batteries, 125 VDC, 28-2 Service Capacity Test WO 0017569601, 28-1 Batteries, 125 VDC, 2B-1 Service Capacity Test WB 8019450581,2A-l

Batteries, 625 VDC, 2A-1 Service Capacity Test WO 0017414101,2A-2

Batteries, 625 VDC, 28-2 Service Capacity Test WO 0040923401,OMST-BAW11

W, 525 VDC, Weekly Test WO 5040495901, OMST-BATTI 1 W, 125 VDC, Weekly Test WO 0040496001,OMST-BAW11

W, I25 VDC, Weekly Test WO 0040734401, OMST-BATTI 1 W, 125 VDC, Weekly Test WO 003991 4901, 15-1 & 18-2 OMST-BATTI 1 Q Quarterly

MI0 0031256501, 18-1 & 1 B-2 OMST-BATTI 1 Q Quarterly

WB 80309501 01,15-1 & 1 B-2 QMST-BATTI

Q Quarterly

MI0 0028265501, SB-1 & 1 B-2 OMST-BATTl

IQ Quarterly

WO 0038119301, ?A-1 & 1A-2 OMST-BATTIIQ

Quarterly

WO 0031639601, SA-1 & 18-2 OMST-BATTI IQ Quarterly

WO 0031256401,lA-1

& 1A-2 OMST-BATTIlQ

Quarterly

WO 0028260601, 1A-1 & 18-2 OMST-BATTI 3Q Quarterly

WB 0030391 401.2A-1 & 2A-2 OMST-BATTI 1 Q Quarterly

WO 0530391 501,2B-1 & 28-2 OMST-BATTI 1 Q Quarterly

WO 0031256201,2A-l

& 2A-2 OMST-BATTI 1Q Quarterly

WO 0531256301,2A-I

& 28-2 OMST-BATTI 16 Quarterly

WO 0031256601,2!3-1

& 28-2 OMST-BATTI t Q Quarterly

WO 0031256701,2B-I

& 28-2 OMST-BAW1

Q Quarterly

WO 0004680801, HPCl Auto-Actuation

and Isolation

Logic System Functional

Test WO 0067956801, HPCl Auto-Actuation

and Isolation

Logic System Functional

Test WB 003971 1701, 1 MST-HPCi27Q

and RCIC CST Low Water bevel Instrument

Catibration

WB 0031316101, 1 MST-HPC1270

and RClC CST Low Water Level Instrument

Calibration

WO 0539317801,2MST-HPC127Q

and RClC CST Low Water Level Instrument

Calibration

WO 0031323101,2MST-HPC127Q

and RClC CST Low Water bevel Instrument

Calibration

WO 0038679201, HPCI Suppression

Pool High Level Instrument

Channel Calibration

WO 0031264601, HPCl Suppression

Pool High Level Instrument

Channel Calibration

WO 0038677301

I HPCl Suppression

Pool High Level Instrument

Channel Calibration

WO 0004589001, Calibrate

14541 -FSHL-NO06

in accordance

with OPIC-DP-SO01

WO 0007165106, Replace HPCl pump discharge

line flow switch WO 00431 63606, Perform single cell charging on 1-1 A-2 Cell #43 IAW BSPP-BAT010 WO 0043161306, Perform single cell charging on 1-18-1 Cell #13 IAW BSPP-BAT010

WO 0042888401, 1-1 B-1 125 VBC Battery Cell # 13 has a low voltage reading WO 0044659406, Perform single cell charging on 1A-2 Battery Cell # 1 WO 0037821401, 18-2 Battery Cell ?# 53 has a cell voltage of 2.124, minimum voltage is 2.1 3 WO 0033286001, 1-1 8-2 Battery corrosion

found on positive terminal of battery cell # 52 WO 0033285401

~ I-1A-1 Battery corrosion

found WO 0033285301, l-IAP-125VDC-BAT.

Replace Cell # 4 on Battery 1A-2 WO 001 6351401, Equalize 1-1 8-2-1 25VBC-BAT

IAW OPM-BAT004

WO 0014092401, 1-152 Cell # I needs to be replaced due to low specific gravity reading WO 0006930901, Using ESR 00-00345 and WO Task knstructions, Replace Cell # 54 in I-1B- WO WRiJO 99-ADIK1, Troubleshoot

and assist operations

in ground hunting for 18 Battery WO 0043131301, 1-1A-2-125VDC-CHRGW

investigate

breaker tripkharger

voltage card WO WWJO 99-AFEC1, Replace floatlequalize

toggle switch on I-$A-1-125VBC-CHWGR

WO WWJO 99-AFED1, Replace floaffequalize toggie switch

on 1 -lA-2-125VQC-CHRGR WO WWJO 99-AFEEI Replace floatlequalize

toggle switch on 1-1 B-1-125VDC-CHRGR

WO WWJO 99-AFEE2, Place 1-1 B-I-125VDC-BAT on equalize WO WWJO 99-AGKAI, Investigate

problem with 1-18-2-125VDC-CHRGR WO WWJO 99-AGKA2, Troubleshoot

ground on 1-1B-2 Battery Charger during Unit 1 outage WO WWJO 99-AFEF1, Replace floatlequalize

toggle switch on 1-1 8-2-125VDC-CHRGR

WO WWJO 98-ACNW 1, Troubleshoot

and Repair 1-1 B-2-125VDC-CHRGR

WO 0033286301, Perform OMST-BAWI

SQ to remove corrosion

from battery terminals

WO 0033286201, Perform OMST-BATTI 1Q to remove corrosion

WO 0027849301, 2-2A-1-125VDC-BAT, Petform DLRO measurements

WO 0027849201,2-28-1

-125VDC-BAT, Perform DLRO measurements

WQ 0016331601, 2-2B-I-125VDC-CHRGR

has no output voltage please investigate

and repair WO 001 3345101, The corrected

specific gravity was less than the required 1.205 tolerance

WO WWJO 99-ADMLI, Place 125 VDC Battery Banks 2A-1,2A-2,2B-II

2B-2 on equalize WO WWJO 00-ADJS1, Replace Cell # 27 in 2-2A-2-125VDC-BAT

WO WWJO 00-ADEEf , Clean off electrolyte

on cell #27 of 2-28-2 Battery WQ WWJO 99-AAGJI, 2-28-2-125VDC-BAT

individual

ceil voltage out of tolerance

WO WWJQ 00-AARJ1, Troubleshoot

2-28 battery bus ground WO WWJO 99-ACRSI , Replace floatlequalize toggle switch

on 2-2A-2-125VDC-CHRGR WO WR/JO 99-ACSWI, Replace floatlequalize toggle switch

on 2-2A-1-125VDC-CHRGR WO 001 11 66201, Replace floaffequalize

toggle switch on 2-28-1-125VBC-CHRGR

WO 0017170101, Specific gravity on Cell #56 of battery 1B-2 out of tolerance

WO WWJO 99-AAGEd.

I-lB-2-125VDC-BAT

Cell #37 voltage low WWJQ ASLEOOI ,I -E6-AV4-52, 5175 480 VAC Distribution

System, Substation

Breaker PM WWJO ADUEQOl ,l-Es-AU9-52, 5175 480 VAC Distribution

System, Substation

Breaker PM WWJO ADKC007 ,1 -EB-AXI-52,5175

480 VAC Distribution

System, Substation

Breaker PM WWJO 99-ACPTI ,2-2CB-C56, 5175 480 VAC Distribution

System, Substation

Breaker WR/JO 00-ABHD2,1-1CA-C05, 5175 480 VAC Distribution

System, Substation

Breaker WWJO 00-ABDH1 ,1 -1 CAC05, 5175 480 VAC Distribution

System, Substation

Breaker WWJO ACDUOO-i, 2-2A-GKO-72, 5240 125 VDC Battery Charger System, Circuit Breaker WWJO ACDXOOI, 2-2A-GK3-72,5240

25 VDC Battery Charger System, Circuit Breaker WR/J0 AAKOOOI, 2-2CB-656-52, 5240 125 VDC Battery Charger System, Circuit Breaker WO 0005034401, PM on 1 -E2-A#1 WO 0017871402, In-situ Test of Mag Latch for 1-E6-AV4-52

25VDC-BAT

while batteries

remain on line BUS IAW OAl-I 15 and IOP-51 replacement

Maintenance

Maintenance

Maintenance

Functional

Test Functional

Test Maintenance

WB 0030223001, Overload Relay Setting Change WO 0019871802, In-situ Test on 143-AV4-52

WO 0029973501, Circuit Breaker Tie Between Unit Substation

E5&E6 WO 0017868201, in-situ Test of Mag Latch of E5E6 Tie Breaker WO 0005033201, PM on I-E2-AH1 WO 0012789501, Breaker Operator Replacement

WO 0005030701

PM on Breaker 1 -dB-GMI -72 WO 5005009301, PM on Breaker 1-1B-GM4-72

WO 0029610701

I PM O R Breaker 2-25-GM1-72

WO 0029609301, PM on Breaker 2-25-GM4-72 WO 0013432712, Test/Replace

Breaker 2B-l-125VDC-Charger

AC CKT Comcdeted

Surveillance

Procedures.

Preventive

Maintenance (PM). and Test Records OPT-12.6, Breaker Alignment

Surveillance, Rev. 42, Completed

8/2/03, 8/9/03, 8/16/03, 8/23/03 Action Reauests (ARs. 087358, Deficiencies

related with valve 2-E41-F001

CR 97-02379;

Determine

if Vortexing

Problem Exists in the CST When Running the HPCl AB 00005402;

Vortexing

in CST Needs More Formal Analysis than CR 97-02379;

dated AR 00098654,125

VDC 1A-2 Battery Charger Main Supply Breaker Trip AR 00047078, 1 B-2 Cell # 56 Failed Specific Gravity AR 00091O76, Positive Plate Discoloration

and Expansion

AR 00071079, 16-2 Battery cells have positive piate discoloration

and expansion

AR 00058078, Battery $A-2 has low voltage cells AR 00053109, Visual signs of degradation

on 213-1 battery AR 00083997,2A-I

Battery Cell #31 cracked cell top AR 00085750, 1B-2 Battery Cell #53 has a low voltage AB 00044684, 15-2 Batteries

are A(1) under new Maintenance

Rule criteria AI? 00052618, BC MOV Thermal Overload Heater Sizing AI? 00076440, BESS Caiculatiofls

Self Assessment

50952 Action Reauests Written Due to this lnsnection

101924, Update periodic maintenance

program to add periodic replacement

of diaphram in Pump; dated August 27, 1997. December 30,1998. valve E41-PCV-152, dated 08/14/03 102321, Valve E41-FC42, reduced voltage strike time calculation

basis, dated 08/14/03 102456, CST Vortexing

Documentation

Discrepancies;

dated 08/20/03 103005, Note in OPT-09.2 Referring

to Auto Closure of HPCl Steam Line Brains (F029 and F028) should have been removed by ESR 99-00405, dated 08/26/04

103106, Correct procedure

inconsistencies

in preventative

maintenance

Procedure

OQM-EfKR001, ITE 4KV Breaker and Compartment

Checkout, dated 08/27/03 103252, Procedure

Enhancement

to OPT-09.3, Rev. 50, HPCl System - 165 Psig Flow Test. Add Procedural

Guidance to Ensure that HPCl Minimum Flow isolation

Valve E41-FO12 Goes Closed After Proper Flow Setpoint is Reached, dated 08/28/03 103256, Procedure

Enhancement

to OPT-09.2, Rev. 1 11, HPCl System Operability

Test. Add Procedural

Guidance to Ensure that HPCl Minimum Flow Isolation

Valve E41-FO12 Goes Closed After Proper Flow Setpoint is Reached, dated 08/28/03 103299, Provide procedural

guidance as io when a Shift Technical

Advisor should activate their post, dated 08/28/03 Lesson Plans/Job

Performance

Measures (JPM) Lesson Plan CLS-LP-51, BC Distribution, Rev. 0 Lesson Plan CkS-LP-402-G, Electrical

Failure Related AOPs (AQP-32.0, AOP-22.0, AOP-36.1, AOT-OJP-JP-O51-AOI, DC Ground Isolation

for P, N, and P/N, Rev. 1 AOT-OJT-JP-302-GO1, Loss of BC Power - Transfer of DC Control Power, Rev. 2 Miscellaneous

Documents:

Brunswick

Nuclear Plant Probabilistic

Safety Assessment

RSC 98-24, Reactor Core Isolation

Cooling System Notebook, Rev. 0 RSC 98-23, HPCl System Notebook, Rev. O HPCI System Periodic Review, dated 02/20/03 RClC System Periodic Review, dated 02/20/03 Maintenance

Rule §coping and Performance

Criteria, System 1001, ECCS Suction Strainer Vendor Manual FP-3808, Battery Charger, Rev. G Specification

137-002, 125 Volt Battery Chargers, Rev. 9 Engineering

Evaluation

BNP-DC-03, Overload Heater Resizing for Valves 1-E41-F00II

FOQ6, and AQP-39.0).

Rev. 0 FOOT, and FOO8, Rev. 0

BCT-09-2083

W3:41 PPl BRUNSWICK

REG BFF 9104573014

P. 16 AII 106230-10

Operability

Review Page 1 of 20 AR 102,456 was written to address documentation

discrqsancies

with respect to pottntkl air entrainment

in the con,ndensate

storage tank (CST) ~~pply line due to vortex a1 the suction nozzle prior to completion

of the HE1 pump suction auto transfer on low CST level. An initia? operability

evduation

concluded

that the low CST WCI level insbmmentathn

ia still operable.

Due to additional

questions

and concerns, a more detailed operability

evaluation

was desired. 'This evaluation

provides additional

detail. When more detail was added tQ the review, some unneeded conservatism

were no longer applied and the end results actudly improved, The issue in question, foe both Units 1 and 2, is whether the setpoint for the Technical

Specification (TS) Table 3.3.5.1-1 Function 3.d. HPCI Condensate

Srmge Tank Level -Low insmentation

is appropriate.

This instrumentstion

is required when the plant is in MODE 1 and ah when in MODES 2 and 3 with reactor stem dome pressure water than 150 pig. TS Bases B 33.5.1 discusem the PIPGI Condensate

Storage Tank Level-Low function:

LOOW level in the CST indicates

the unavairability

of an tldequste

supply of makeup water from this normal source. Normally 6he suction valves between HpeI and the CST are open and, upon receiving

a HPCI initiation

signal, water for KPCI injection

wouldbt taken from the CS

T. However, if the water level in the CST falls below a psesclecteci

level, fimt the 8UppSdOn pol suction valves automatically

open, and then the CST suction valve automatically

cio&es. This ensures that an adequate supply of makeup water is available

to the MlpcI pump. To prevent losing suction to the pump, the suction valves are interlwked

sion pool suction valves m~~t bc open before the CST suction valve automatically

chses. The Function is implicitly

assumed in the accident and transient

analyses (which take credit for HPCI) since the analyses assume that the HPCI suction sow is the suppression

pool. The Condensate

Storage Tank Level-Low signal is initiated

from two level switches.

The lo& ie arranged slack that either level switch cxn cause the suppression

pool suction valves to open and the CST suction valve to close. The Condensate

Storage Tank Level--Low

FURC~~DII

Allowable

Value is high enough to ensure adequate pump suction head while water is being takrn faom the CST. Two channels of the Condensate

Storage Tank Level-Low Function are nquired to be OPERABLE only When HPCI is required to be OPERABLE to en8uTe that no single insmmenr failure can preclude HPCi swap to suppression

pool source. H41-ULNW and Mi-LSL-NOQS

are TS required instrumentation

and are designated

8s Q Clslla A (safety related).

Elquipmcnt

datnbase (H>B) describes

the active function as P~wv&% a signal to the WPCI logic when the condensate

storage tank level is low. This opens valves E41- FM1 and E41-FQ42 to dlow WPCl pump suction from the suppnssion

p~o!." This review was performed

in accordance

with EGR-NGGC-0019, Engineering

Operability

Assessment, and makes dime reference

to NRC Inspection

Manual, Part 9900: Technical

Guidance STS1Oo.TG

and STS IOOPSTS. It supports the determination

that the deficiencies

are. dacumentation

problems only and that no oprability

coneem exists. ATTACHMENT

AR 106230-10 Operability

Review Page 2 of 20 The definition

ofOPERABLBO?ERAB~LITY

is contained

in Chapter 1 of the plant's Technical

Specifications

which states: A system, subsystem, division, component, or device shall be O?ERABLB OT have OPmAI4ILITY

when it is capable of perfoming

its specified

safety funCtion(s)

and when dl necessary attendant instrumentation, controls, normal or emergency

elect13cd

per, cooling and seal water, lubrication, and other auxiliary

equipment

that are required for the system, ~ubsystern, division, component, or device to perfom its specified

safety function(@

ate also capable of pefloming

their related support function(s).

For the HE1 CST Level-Low

instmmenratioa

to be OPERABLE, the chawlaels

must be in calibration

and the CST Level-Low

Function Allowable

Value must bc high enough Io ensm an sdquate 8upply of water is available

for all MPCI system specified

functions.

The preaence of vwtexing in the CST wm not initially

factored into the setpoint development.

This evalunlticm

demonstrates

that the current TS Allowable

Value for the instmentation

setpaint ie appropriate

for all HPC1 system specified

fUnCtiQn9

with the effects of the CST suction vortexing

phenomenon

considered.

As stared in MC Inspection

Manual, Part 9900: Technicai

Guidance, STSlOOP.Sri'S, 3.3 Specified

Function(s):

%e definition

of operability

refers to capability

to perfom the " specified

functione," The SpeciEied

bclim(s) of the system. subsystem, train, component, or device (hereafter

refed to a!? system) is that specified

safety function(8)

in the cumnt licensing

basis for the facility.

In addition to providing

the specified

safety function, a system is expected to perform a designed, test&, and maintained.

When system capabiiity

is de to a point where it cannot periWm with reasonable

assurance

of reliability, the system ahould be judged inopefable, even if at this instantaneous

pint in time the system could provide the specified

safety function.

A B stated in NRC h6pction Mwual, Pan 9900: Technical

Guidance, STSIOOP.STS, 2.1 Cmnt Licensing

Bassis: Cunent licensing

basis (CLB) is the set of NRC requirements

applicable

to a spific plant, and a licensee's

written commitments

for =wring compliance

with and operation

within applicable

NRC requirements

and the plant-specific design basis (including

all modifications

and additions

to such commitments

over the life of the license) that an? docketed and in effect. The CLB includes the NRC ngulations

contained

in IO Cm Parts 2,19.2D, 21,30,40,50, SI, 55,?2,73,100

and appendices

thereto; orden: license conditions;

exemptions, and Technical

Specifications (TS). It also includes the plant- specific design basis infomation

defined in 10 CFR 50.2 a5 documented

in the rnmt mnt Find Safety Analysis Repon (FSAR) as required by 10 CFR S0.71 mad the licmsm's comiome~ts

remaining

in effect that were made in hketed licensing

c~mspondence

such 88 licensee respanscs

to NRC bullctins, generic Ictcers, and enforcement

Bctions. BS well as licensee eomrnitnaents

documented

in NRC safety evaluations

or licensee event repone. P. 17

OCT-89-2003 03:42 PM BRUNSUICK

RE4 eFF 9184553814

P. 1B AR 106230-10 Operability

Review Page 3 of 20 A5 stated in NRC Inspection Manual, Part 9908: Technical

Guidance, STS100.'Ki, ScctiOn 1.0, C.S. Principal

Criteria, the following are the principal criteria

for technical

speGification

operability

rquirem~ts:

a, The system oprability

requirements

should ke consistent

with the safety ana)ySiS Of b. The system operability

quirernemts, including related

regulato~

requirements, my be c. Design-basis events are plant specific

and regulatory

requirements

may have plant- d. The system opesability

quiremen&

that are based on safety analysjs of spcific desip- specific desipbases

events and regulatory requirements.

waived BI~ a consequence

of swified action statements.

spedflc considerations

related to technical specification

operability.

bmis events fer one mode or condition

of operation

may not be the same for ail modes 0% conditions

of operation.

e. The system qxrability

requirements

extend to necess~sy

support systems regardless

of the existence

or absence ~fsttpp~n

system quiroments.

f. lphe operability

of necessary

support systems includes regulatory

requimnentli.

It doca not include consideration

of the Dccumnce of multiple (simultaneous)

&sign buls events. Also applicable

to this discussion

is NRC Inspection Manual, Part 990: Technic& Guidme, STSlO(9.TG.

Section 1.0, D. Conclusion:

Many systems and components

perform dual-function roles with ?egard to midart mitigation

and Foe events for which safe plant shutdown

is required.

The cotrcct application

of operability

quirenuents

for them systems and components

requins additiond

reliance on a knowlededge

of design bssis events. Thus, it is essential

for the proper application

of technical

specification

operability

requirements, to know the applicable

design-basis events for the facility.

. OCT--BS-2883

83:42 PW BRUNSWICK

REG FIFF 9104573014

P. 19 AR 196230-10

Ojknrbility

Review Page 4 of 20 The specified

functions

for the IfpcI spstem for the purposes of this operability

evduatim are as follows: F-B: HPCI LoeA Licensing Basis Function

The Oriri$inal

mI design and limnsing basis requirements

were established

such th$K HecI was a part of the integrated

ECCS group of systems that provide a LOCA response capability

consistent

with the requirements

of 1QCFw50.46.

O R March 29 1989, CP&E submitted

an evaluation

to the NRC for revised LEA licensing

basis rand to update the demonstration

of conformance

to the ceiteria provided in iOCPR50.46, a6 modified by SECY-83-472, Emergency Core Coolant System Analysis

Methods. This evduati~n, Brunswick

Stem Electric Plant, Units 1 & 2, SAFEWGESTR-LOCA

bnas-of- Coolant Accident Anfdysie, NEDC31624P, assumed less performance

from ECCS systems to allow for relaxation

of some selected requirements, On May 17,19&9,6P&L

submitted

a written response to 0 verbal NRC request for additional

information.

I"XC Question 2 was given

as: Relative to relaxations

of input values (Table AI), what ate all of the nlaxatims

between the new analysis and the analysis of record (Le., the current analysis).

The respnse to Quwtim 2 grovided a tiable which included the following:

rnM ANALYSIS OFRECORD NEW ANALYSIS HPCI hump Minimum Flew 4250 gpm 0 gPm On June I, 1989, the NRC iaswd a Safety Evaluation

for the CP&L submittal.

This SER included "tsstly the staff notes that significant system

or component

assumptions

included no offsite pawet, RO high pressu~ coolant injection

system, two SRVIADS valves out

of servkc and a SRV setpint tolerance

of 3% The assumptions

are acceptable." It also pviddthe fdowing " On this basis. the analysis contsined

in the GE report can be Used to @rdde B nvkd LOCA licmnsing

basis for both Brunswlck

units, and can be referenced

in futuro submittals." The HK.1 pufomce requirements

were discussed

more recently in NEDG-33039P, The Safety Andysis Report for Brunswick

Units 1 and 2 Extanded Power Uprate (pUsAI6), that WBB part of the 08M/01 120% power uprate submittal.

The report included the fdowing "Ori@inally, the HITI system was primarily

for the mitigation

of small break ILEA8 where the depressurization

function [Automatic

Depressurization

System (ADS) I SRVa] WW assumed TO fail. Fw BSEPP, the depressurization

function is Fully redundant, and no accidenr mitigation

credit is taken for the HPCI system." On the bmis of the 1989 NRC SER, the cutrent safety related

LNA licensing

basis prrformance

criteria for KPCI at BSEP is 0 gpm. Given the above, the potential

for air enrPainmnt

the CST suction nozzle during

HpcI operation

is not a concern with respect to the ECCS rcquircments

of 1OCFR50.46

and no further discus5bn

of this function will be prOVi&.

OCT-E9-20E3

03:42 PM BRUNSWICK

REG FlFF 91R4373014

._ AR 106238-10

Operability

Review Page 5 of 20 Fm: Piefed Response to a 1" Line Break Function Although not Wuired for the BSEB JAXA licensing

basis as discwssed

in Function 1 above, BSEP dws consider HPCI operation

to be the preferred

method of responding

to very srnd line breaks. VFSAR 6.3.1.2 and 6.3.3.5 have the following

statements

which go along with this fundon: One high pressure cooling system is provided, which is capable of maintaining (he water level above the top

of the core and preventing

ABS actuation

for small b~aks. and For the HPCI, a criterion

was used (in addition to the criterion that it

depxc~s~~

pprly in conjunction with

the low pmsure systems) which prevents cfaddlng headng far haks less than a 1-in. pipe when functioning

alone, This wm done to ensum maincen@rmce

of level at rated vessei pressure for the more probable leaks

thst might occur QVCT plant life. Since I-in. lines predominate, this provided a good basis for such a criterion.

This flow io also orders of magnitude

in excess of leakage that would occur for cracb approaching

critical size in large pipes. The abve IJFS.4.R 8tatetnCntS

provided the basis for the following

portion of the PWSAR described

WPCI funnctim: "me primary remaining

purpose of the FECI system is to maintain reactor level above the top of the active fuel (TAR and prevent ADS actuatim for line breake up tQ I" in dim*." ESR 99-0062 evaluated

the ability of WI to meet the above

requirements

in response t0 response the testing concerns.

This ESR documented

that less than l@lO gpm of makeup flow was required in response to a 1" line break, Bad on the above this is an explicit function associeted

with :he BNP specific HPCI Licensing

his. Function 2 88 described above

does not inherently exclude the

possibility

of HPCl suction transfer m !OW CST level. Evaluation

of the potential

for air entrainment

at the CST suction noule duhg HPCI Qperaaion

for this function will be evaluated

a8 Case 1

. OCT--89--2803

243 Bbl BRUNSWICK

REG eFF - - -~ 9104573814

P.21 AR 104230-10

Operability

Review Page 6 of 20 Function 3: Backup to RCIC Function WPCH also ha a design requirement

that it be capable of providing

a backup to the non safety related RCE fuwtiOR for loss of feedwater

and vessel isolation

events. Technical

Specifications

require that RGIC be able to inject water to the vessei at 400 gpm over the same mge of vessel pressme as is specified

for WCI. The RCIC functional

nquiwnents

specified

in UPSAR 5.4.6 include: The RCIC system operates automatically

to maintain sufficient

coolant in the reactor veswl to prevent overhesting

of the reactor fuel, in the event of reactor isolation

accompanied

by loss of feedwater

flow. The system functions

in a timeiy manner so that integrity

of the rgxtioactive

material bamer is not compromised.

This is a transient

response function and is not a Safety Related function.

Technical

Specification

aquirements

have been maintained

because of the contribution

to the reduction

of overall plant risk provided by RCI

C. After the 105% Power Uprate, analysis showed that the original RCIC performslace

quhmenbs (4W gpm starting 30 seconds after initiation)

would result iIl a lowest level Inside rtme shmud of no less than 5.4 ft above the top of active fuel. Even with relared perfomnce

requirements

of 360 gpm starting 66) seconds after initiation, the lowest level Insick the shroud would be no less than 4.7 ft above the top of active fuel. Either nspon8e ia aeccptable.

RCIC operetion

can prevent the need for ABS biowdown and low preressupe

ECCS injection

following

a loss of feedwater.

Transient

rcsponse graphs in NEF1Bc-30106-P (the GE basis for changing the MSIV isolation

setpoint from LL2 to LId that provrded LTSAR Figure 15.2.6-3)

and GE-NE-187-26-1292 (Power Upate Transient

Analysis for Bmnswick Steam Electric Plant) indicate water level may drop far enough to cwe LL3 actuation (level olttside the shroud between 33.3' and 35.3' above vessel zero). For thie event, operators

would inhibit ADS a5 directed in EBPs due to the large margin between the LJ3 setpoint and top of active fuel, the lack of LQCA indications

and the slow fate of level decrease.

A slow downward trend would follow as the mass of steam flew for decay heat removal via SRV actuations

initially

exceeds the RCXC makeup flow. At 15 to 20 minutes into the event, the level trend would stabilize

and then later start to increase a8 the RCIC makeup matches and then exceeds the steam flew for decay heat removal. The above UFSAR statements

are consistent

with the following

portion of the PSAR dessnbed HPCI function:

"'Kc HPCI system also serve6 as a backup to the Reactor Core Isolation

Cooling (RCIC) system to provide makeup water in the event of a loss of feedwater

flow transient.

For the loss of feedwater

flow transient, which assumes closure of the Mslin steam halation ValVeP (MSrVs), the currentty

specified

WCI system minimum injection

rate of 3825 gpm would pvide sufficient

makeup water to maintain the level inside the shroud well above TAP. DMwg tfiis transient

event, the SRVs would open, then cycle, and the WCI system would quickly retwm the reactor water level to P~WIIIR~, or to the reactor high water level trip (i.e., kvel 8 shutoffh" Note that the 3825 gpm vaiue used above is 90% of the original design Row and is the value that BE would have specified

for HPCI in the SAIFEWGESTR-LQCA

evaluation

had KKI operatton

bn credited.

A high HPCI flow rate is appropriate

only fer the ATWS function not

. OCT--Y9-2003

53:43 PPI BRUNSWICK

RE6 QFF 9104573814

P.22 AR 106230-10

Operability

Review Page 7 of ZQ the backup to RCIC function.

A flow rate of 400 gpm is the ticensing

basis flow rat0 requirement

for the HPCI Backup to RCIC Function.

Based on the above,this

HETI function is an expiicit fUIICtiOR

associated

with ?-he BNP specific IIPCURCIC

Licensing

basis. Function 3 as described

above does not inherently

exciude the possibility

ofml SWtia transfer on low C§T level. Evaluation

of the potential

for air entrainment

at the CST suction nozzle during NPCI operation

for this function will be evaluated

as Case 2. Case 3 and Case 4. Function 4: SB6 Function Although not pan of the original HPCI design basis, the HPCI system has been credited fW providing

makeup water during B postulated

Station Blackout (SBO) event. The most recent SBO evdu~tion

required HPCI to deliver approximately

86,080 gallons of CST water to the Reactor in a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> time mod. This is an average flow wte of only 3.58 gpm. The peak flow requirement

for this event can be estimated

as the decay heat removal plow rate nonndy provided by RCIC at 4QO gppm combined with an assumed 61 gpm win: pump seal leak or 461 kpm. Although the WSAR did not explicitly

describe the above "CI function, this function waa an essential

pan of the SBO evaluation

th&t was described

at the summary level in the PWSAR. Bd on the above this WCI function is an implied function associated

with the BNP specific SBQ Licensing

basis. Since RHB operation

is not assumed for the initial SBO response, significant

Suppression

Pwl heating is anticipated.

Due to HPCI system process fluid temperacue

limitations, the event explicitly

excludes allowing CST depletion.

This requirement

establishes

a limit on the highest allowed actuation

of the low CST level HLPCL instruments.

Function 5: Appendix R Function Although not pant of the original FPCX design basis, the I;IpcI sysfem has been credited for providing

makeup water during a postulated

Appendix R event. Appendix R evaluations

squired WI to deliver CST water to the Reactor for decay heat removal when manually started after a number of other manual operator actions are completed.

RCIC has a similar Appendix R function.

The use of RCIC for the similar Appendix R event was found to quire a peak flow rate of 500 gpm. Although the MJSAR did not explicitly

describe the above WI function, this funCtiOn i5 essential

for Appendix R compliance.

Appendix R compliance

a uprated conditions

is descrjbctl

at the summary level in the PUSAR. Based on &e above this HKX function is an implied function apsociatd

with the BNP specific Appendix R Licendng basis. SirrPiliv

to the SBO event, the Appendix R event is evaluated

over a specific time penOd. The mal required makeup inventory

for this event will not exceed the required makeup for thc SBO event. Suppression

pool temperature

is expected to exceed the allowed temperaturn

for #pcI operation, CST depletion

is not a required assumption

for this evenr.

" OCT-B9-288%

83:44 PPI BRUNSWIGK

REG BFF ~- 9104573614

P. 2s AR 106230-10

Operabillry

Review Page 8 of 20 pction 6: HPCB Rod Drop Function I-PCI may be used for vessel inventory

makeup following

a rod drop accident.

Although a 03/IY02 Extended Power Uprate RAI response documented

that neither HBCI nw RCIC operation

is required for a rod drop event, HPCl usage would be expected if RCIC is not available.

The nquired makeup during this event is based on decay heat alone where either HPCI or RCIC operation

would be sufficient.

This function is essentially

the game as the Backup to RCIC function that is addressed

in the Cw 2, Case 3 and Case 4 cvdulaiione.

Function 7: HPCI ATWS Function When the 120% power uprate site specific ATWS evaluation

was performed, KPCI operation

WBS assumed. The operation

of HPCI during iin ATWS is based entirely on manual operator actions including

inhibiting

the auto start at Low Level 2, manually allowing WCX to start just prior to reaching the desired level, and then promptly adjusting

the flow controller

secpolnt a8 ne%clled to control level in B nmw band. Although the FUSAR did not explicitly

describe the above HPCI function, this hn~tim was an essential

pdin afthe ATWS evaluation

&hat was described

at the susnmtppy

level in the PUSAR. Baaed on the above this MPGI function is an implied function associatd

with the BNP specific ATWS Licensing

basis. This event is also an event where Suppmssion

Pod temperatun%

are expected to exceed the limit for w?cI operation.

ASSKIW~ WCI operation

for an ATWS response will be for a relatively

short duration and the event does not amme CST depletion.

OCT-WS-~~W~

m:44 PM BRUNSWICK

R E G AFF - 9184573814

P. 24 AR 106230-10 Operability

Review Page 9 Qf 20 The hw CST level setpoint

does not need to provide any pmtection

for LOCA even&. It do= provide yrotectios

when either an operator action in accordance with

existing procedures, suppflsiiwr

pool level reduction

is credited, or when early MSIV Closure is Rat assumed. For all LOCA response wen&, operator actions

to drain the suppression

pa01 or to jumper the high suppsion pool level FPCf instntments

would not be allowed by proceduns.

The "CI sactian transfee occurs based on high suppression

pool level and the CST inventory

is new fully depleted.

No air ~xhes the HPCI pump and all HPCI performance

is consistent

with UFSAR descriptions.

The Tech Spec hstrumen!

function is however required for HPCI when it is pmviding the

backup to RCIC function.

This funstton can requin extended NPCI operation, either at a reduced flow

rate or intermittently.

The potentid fw an acceptabte

operator action in reccordence

with existing procedmo (educing suppression

pa01 level) could result in pump damage if the stpoint is not adequate.

Additionally, if early MSIV closure does

not occur, a loss of feedwater

event may result in CST depletionc

For this backup to RCK function, opcrarer actians for mnudIycmtrolling

vesseS level late in the event are appropriate.

Etthtr the WCI flow rata would be reduced acceptably

or HFC6 would be operated at full flow for only 60 seeonds. For dhe full faow caw, no air would Each the pump during the last injection

with CST suction and the WCI suction swap would then

be completed

prior to the next HPCI injection.

This proVides the Protechicpn

that is nm&d to prevent continued "Cl operation

with the suction lined

up to a depleted CST. 1 TS Table 3.3.5.1-1.

Function 3.e. #pcI Suppression

Chamber ~vel-High

Instrumentation

Channels are operable (otherwise, WCI pump suction would

be aligned to the suppfession

PI). NP@I auto transfer on high suppression

pool kvel starts at the - 24 inch Tech Spec limit. 2 Cofhmak Stomp Tank level is being maintained

at a minimum of 10' in accordance

wiKh UPSAR 9.262 requirements.

See Attachment

for CST volumes at variom Icveb. 3 WCI auto transfer on low CST level start5 at the 23' 4 Tech Spec limit. 4 Supssion pl Ievel is assumed to start at the -31 inch Tech Spec low level limiL 5 w"cI suction valves operate with maximum stroke times allowed during sUndat9Ce

testing. 6 The HPCI system will respond to automatic

signals at Tech Spec specified

serpoints, and OpMatora will operate the plant in accordance

with existfng design basis, training

and prOCC&*S.

If NPCf actuates automatically (Le,, due to low reactor water

kwl) RCIC will also actutatc if available.

CRlp is nDt taking suction from the CST as the bottom of the suction nozzle

supplying

CRD is more than 9' above the bottom of the tank. 9 Ne sources ~ke ndding waiet to the GST and no actions are taken to refill the CS

T. 10 The plant is at noma full power, 2923 MWt.

I OCT-09-2083

83:45 PM BRUNSUICK

REG AFF 9104Ei73014

AR 106230-10

Operability

Review Page $0 of20 e IIpeI is providing

the Pafemd Response to 1" Line Break function 0 Operators

may or may not manually control vessel level

Requind manual operation

of RHK is assumed in accordance

with proccdurcs

FOF Case 1, HPCI and RClC will inject QR low wactor water !eve1 (LL2, 105"). If not manually secured due to the standad post trip 170" to 200" level control band

procedure

requiremenl, WBCI and RCIC will trip when level reaches the high feactor level trip setpoint

at 206". Level will then continue to cycle between 105" and 286" if RO operator actctrons

are assumed 01: 190" and 200" if operatom RE performing

normal event response actions. Level control assumptions

do nor affect the outcome of this case. Since this

event involves a small break LWA, significant

drywell heating and pssurizatim

would mur. Operatom would place at least one loop of WI-IR in suppres8ion

pool cooling at 18 minutes consisknslt

witlh existing BSEP Licensing basis assumptions (ref. UFSAR 6.2.2.3).

RWR would also be used for containment

spray if drywell pressure approaches

or exceeds 11 pBig, but containment

spray operation would be terminated prior

to resetting

the Group 2 isolation

instrumentation

that actutltes

at 2 psig. With drywell pres5ure above 2 psig, no flow path is available

for reducing suppression

pool ievel due to the isolation

of Ell-FW md Ell-FW9. With RHR in auppmsioil

pool cooling and the reactor not depressurized via

SRVs, suppnssion

peol ternpeRlture8

would not increw to a value where overriding

the HPCI high suppreselon

pol level transfer inemmentation

is allowed. Continued

operation

of KPCI and/or RCJC rends to depressurize

the vessel 8s it nmoves steam from the reactor and 8s it injects low temperature

wster into the vessel. Although it

is possible that cmtinued HFCI operation

could reduce vessel

pressure to below the "CI isolation

8etpdnt prim to my automatic

auction transfer for larger small breaks, this is not expected for the 1" line break king considend

here. The HPCI suction transfer will stm after 94,330 gallons of water Is injected based on high suppression

pwl level, not low CST level (see Attachment

for supporting

&tds). The CST lswl would k at least 8.0 inches above the top of &e CST suction nozzle after the transfer k complete.

A recent industry paper, JBOC200UPWR-190010, presents the best published

information

applicable

to this appIication

that BNP has been able to find. Although the plant review indicates

that the nominal equation provides

a conservative

estimate for our CST, the "bounding"

eqUQtiOn for 0% air from JPGC2001/PWR-1$010, Equation 10, was used in this case for conservatism:

Sa% I 1.363*FrA0.261

where Fr = V1(32.2*(d/12)"0.5

and S = (d+Lll/d d Pipam now Velocity Fr S-0% L14% I5 1.23 470 8.53 1.345 1.473 7.09 (in) (frA22) (gem) Wet) (in) This shows that no airentrainment

at the CST nozzle will occur far CrrSR I. P.25

OCT--89-2003

245 PPl BRUNSUICK

REG QFF 9184573814

P. 26 AR 106230-10

Operability

Review Page1lofaO

HPCI is providing

the Backup to RCIC function 0 hpt MSIV closure occurs e No cperstor actions assumed other than the required

initiation

of suppression

pool cooling For Cwe 2, WBCI operation

alone will bc considered

as RCiC unavailability

is part of the CBBC definition.

Wl will auto start on low reactor water level (LU, l05"). "(3 will Wip Wh level reaches the high reactor level trip setpoint at 206". Level will then continue to cycle between 105" and 286". This event

does not involve a

small break LOCA, but it may involve a loss of drywell cooling. Drywell heating and pressurization

to above 2 peig may or may not occur. Operatma would place & feast one loop ofM in suppression.

pool cooling a1 10 minutes. With RHR in suppression

pool cooiing and the reactor nut depressurized

vie SRVs, suppssion

pool temperatures

would not increase fo a value whea overriding the

HPCI high suppression

pl level transfer instrumentation

is allowed. Note that if RHR suppression

pool cooling is not 5tute5, "CI would eventually

be operating

with the suction lined up to the suppression

and the supppessim

pool water remperanurc

above the value allowed for Hp@I operation.

Conhued ophn of IipCI tends to depressurize

the veasel as it removes stem fmm the reactor and 8s it iaajecte low ternpalure

water into the vessel. Although it is possible that Continued

mI operation

could reduce vessel pressure to below the HPCI isolation

setpoint prior tn any automatic

suction transfer for small breaks, this is not expected for the case being considerect

here. With MSIV closure, all coolant removed from the vessel will be discharged

tD the mpp,ssion

pl via SRVs and the HPCK turbine exhaust. For this case, the suction transfer Will start after only 43,160 gallons of water is injected to the vessel based on high suppression

pi level. The volume would be less than for Case 1 as the lower

elevation

of the drywell does not collect my water. Also the qqulnd submergence

would be less than for Case 1 since only HPCI operation

is assumed. The margin for avoiding air

entrainment

is therefore increased and

the event would be acceptable.

___ ~- - P. 2s OCT--89-2BBS

03:46 PM BRUNSWICK

REG AFF 9164573Ef14

AR 106230-10 Operability

Review Page 12 of 20 #pcI is providing the

Baekekup to RCIC function m Prompt MSIV closure occuls Qpmtors initiate suppression

pool cooling e Operators

perform suppression

pooi level contml in accordanhe

with proceduns

e Operators eventually perform vessel level

control in accordance

with procedures

WCI operation alone

will be considered

as RCIC unavailability is

part uf the case definition.

=I will auto start on low reactor water

level (LL2, lO5"). HPCI will trip when level

%aches the high reactor Ievd trip setpoint at 206". kvel may continue to cycle between 105" and 206' until such time that operators

have had time to assess plant conditions

and complctc any other mm important actions. Additional discussion

of manual actions to control level in the spified 170" to 2oh)" ievef contpol band will be pmvided below. This event does not involve R smaH bfeak LOCA, but it may involve a loss of drywell cooling. Drywell heating and pmsurization

to above 2 psig may or may not occur. opmttors would place at least one Imp of RHB in suppression

pool cooling consistent

with existing BSEP licensing

basis assumptions.

With RHA in suppression

pool cooling and the reactor not depnssutized

via SRVs, suppression

pool tempecnrtures

would no? increase to a value whm oveniding

the Hpcy hi& suppression

pool level transfer instrumentation

is allowed. The coolant removed from the vessel will be discharged

to the suppression

pool via SRVs and the HPCI turbine exhaust and the lower elevation

of the drywell will not fill with water. For chis case it will be assumed that prim to reaching the high suppression

pod Hg61 level instrument

Setpoint, dfpel1 pressure has been controlled

or restored such the manually

reducing suplprcssion

pol level is possible.

It wa8 estimated

that this would occur at between 0.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and 1.8 how into the event depending

on starting suppression

pod IeveI. For this case CST depletion

at some time after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of intermittent

HPCI operation needs

to be considered.

Prior to considering

the plant level response, it is appropriate

to take a close look at the cumnt BSEP design basis for the instrument

in question.

The original licensing

bssis for the switch did

not provide an explicit descripien

of the plant IeVd condtions

as&wiated

with actustion.

It simply indicated that the

switch would actuate on 10W CST level to onsure that an adequate supply of makeup water is available

to the HPCI pump. The original licensing basis

for the switch went with an original design basis that specified

the nominal trip setpoint be at a value that "corresponds

to 10,000 gallons capacity".

The documekd design basis did not specify a flow rate and it did not specify the

refmnce point foF the capacity.

The documented design

basis also did not link this setpoint to any stroke time limits on the WPCI suction valves. There

is no indication

that a margin for unccrtsdnties

such 86 temperature effects, suction vortexing, seismic concerns, e&. had to be Considercd.

Aftcr evaluating

OE item PS 5 109, BSEP changed the design basis

for the switches in 1997. The combination

of ESR 97-WO26 and ceiculation

0E41-1001

documented that setpoint

was acceptable

when continuous

HPCI plus RCIC oQeraticn

at 4700 gpm considered

This determination

WBS made based on engineering judgment.

The stroke time limits for the HPCi

9164973W14

OCT-%9-4003

83:46 PM BRUHSWICK

REG AFF P. 2% AR 106230-10 Operability

Review Page 13 ofu) suction valves were also updated and linked to the transfer function.

UncesOainties

were ewssed. Dudng an intarnal system review in

1999, it wa determined that

a more defendable

basis for the vottex aspect of setpint WEIS needed and AR 5402 was generated.

ESR 01-00322 was issued in 2001 88 a ckct mult of this AR. ESR 01-QO322 updated the switch

design as allowed by 1QCFR5Q.59

and was issued in accordance

with CB&L procedures

foe a design chge. The EX noted that the Hpcl system level functional

requirements

did not include actuation

of the switch at the flow rates pnviousty

consi&d. It documented that the highest

applicable

event respnsc flow rate requirement

far WCI was approximately

Io00 gpm. It noted that the HPCI operating procedure

instructs

operators

to adjust HPCI flow after stanup to mainfain stable rcactw vessel levd within the normal range. It established

that fer the HPCl system to be operating at

a high flow rate where significant

air entrainment

would occur due to the lack of adequate reactor level control

mmua! actions is conriderad

non credible.

AK greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into an event where E1 is pmviding the backup to RCIC function, it is apppriate

to Consider operam actions with respect to vessel level control. The following

guidance in the UBSAR is applicable

to this discussion:

UFSAR 5.4.6 inclwdes:

Following

any reactor shutdown, steam generation

continues

due to decay heat. hitidy, the rate of stem $enemtion

can be as much as six percent of rated flow. Thc stem normally flow8 to the main condenser

through the turbine

bypass a, if the emdenser is isolated, through the relief valves to the suppression

pool. The fluid removed from the reactor vmsel either can be furnished

entirely by the feedwater

pumps or can be partially

funti6ked

by the control rod drive (CRD) system, which is supplied by the CRD feed pumps. Lf makeup water is required to supplement

these sources of water, the RCIC turbine-pump unit either start?, automatically

upon receipt of a reactor vewi low water level signal (Bigurn 7.3.3-2) or is started by the operator from the Centhol Room by fernot~ mmud controls.

Re szme low level signal also energizes the

high prcssun coolant injection

system. The RCIC system delivers its design flow approximately

8&c after actuation.

WFSAR 6.3.2.8 System Operation

includes the following:

The ECCS have been designed to atart automatically

in the event of an accident that

threatens the adequacy

of core cooling. Manual operations

are required to Wntain long term cooling. The description

that follows details the

opedon of the systems needed to achieve initial

con mlhg followed by containment

cmling and then followed by extended cm cooiing for a long term plant shutdown

for the case of a non-opcrable

main feedwater

system. The manual operations

deseribcd

we generally

similar to those stquid in the event of a LOCA. The discussion

below also includes the

operation

ob the non-ECCS, non-safety relate$ RClC system. This system is designed to operate dueng loss Of feedwater

events, but is not relied upon to mitigate any accidents.

P.29 OCT-09-2003

03:46 PM BRUNSWICK

REG RFF 9104573B14

AR 186230-10

Operability

Review Page 14 of 20 Following

8. loss of feedwater

and reactor scram, a low reactor water

level signal (he1 2) will automatically initiate

a signei which places the HPCl and RCIC Systems into the reactor coolant makeup

injecrion

mode, These systems will inject water

into the Vemel until a high water level

signal automatically

trips the system. Following

a high reactor water level trip, the HPCI and RCIC Systems will automatically

ninitiate

when =tor water level agdn &creases to low water Level 2, Later in WSAR 6.3.2.8, the discussion

includes:

The aperator can manually initiate the "CI and RCIC systems fmm the ConrrOl Room befere the bel 2 automatic initiation

level is reached.

ahe OperW3' has the Option of manual control

or automatic

initiation

and can maintain

xactor water level

by throttling

system flow rates.

The applicable

operator actions asissodated

with reacror vessel level mtrol level for the non safety dated Backup to RCIC function iire the manual starting of HPCI, the adjusting

of the HBcl flow rate and the stopping of HPCI. The staning and stopping of WCI arc manual actions that also kave associated automatic actions.

1. pcI does not have pin automatic feature

to adjust the flow rate to control vessel level

within the procedurally

specified

170" to 200" range. NRC gddrmce wm reviewed with respect to Operator actions. As described

in MC IN 97-78, GL 41-18 rev. 1 states: "it is not appropriate

to take cndit for manual action in place of automatic

action fa protection

of safety limits to consider equipment operable.

This does not preclude opcpator action to put the plant in P safe condition, but operator action canna be a substitute

for automatic

safety limit protec~im."

It is notable that the OL text was specifically

far "automatic

safety limit protection"

and not "any automatic

WtkiR s@ecifid in tkc FSAR or Technical

Spccificatiorms".

Ttie text of IN 99-78 then goes on to quote the following

from ANSI-58.8: "Nuclear safety-related operator actaons or sequences

of actions may

be pcrfarmed

by an operator only whepe a single operator crror of one manipulation

does not Tesult in exceeding

the &sign requirements

for design basis events." Again the text rsfers to "safety-relaled operator actions" and not UFSAR described

actions for a non safety related function.

The text of Cy 97-76 then goes on to discuss that it is pctentid%ly

acceptable

to rely on operator actions, but that the requirements

of 1WFR50.59

eppiy, and @or NRC approval is applicable

when an Unreviewed Safety Question (WSQ) is involved.

A IoCpR50.59

review of the changes of the changes did not constitute

a WSQ. If it is desind to conservatively neglect the

manual actions associated

with starting and stowing HPCI due 10 the associated automatic features, then the ESR 01-00322 design basis for the switches yuire.8 that tRe manual action for adjusting

the HPCI flow controller

(&er flow in automatic

mc& or speed in manual mode) is assumed ro reduce flow such that significant

air entrainment

doe$ not occur. 01-00322 was performed

and it was identified

that

OCf--03--ZBB%

03:47 BM BRUNSWICK

REG FIFF 9104573014

P.30 AR 106230-10

Operability

Review Page IS of 20 Using JPGCXt01/PWR-19010

Equation 8, it was determined

that 2% air entminment

at cbe CST nozzle would be expected at 3000 gpm when LI reaches 2.6". With m assumed average HPCI flow of 3ooO gpm, the 2% entrainment

would start at 1 I7 seconds afta level switch actuation.

With a 45 second transpoet

time, "significant air

entrainment"

would not reach the HPCI pump bedm the lf4 seconds suction tmnsfer is complete.

With a flow rate requirement

that will be no mose than 400 gpm, it would be reasonable

to assume that the injection

flow rate would bc 3000 gpm or less for the last injection

from the CST. This assumption

is not contrary to any regulatory

guidance fer this non safety related function, is consistent

with WSAR descriptions

for sptem operetion

and is applicable given the switch

desigo basis. Regwdlcss

of whether 01 not the manual actions of starting and stopping HFCI am credit4 these actions very likely and need io be considered

for completeness.

Ef an operator decides that he d~ not want to adjust the HPCI flow rate, he can maintain the specified

vessel level by npeatedy starting I%pCI at 2 170" and then securing MPCI at 5 ZOO" whiIe leaving the flow controller

Bet for 4300 gprn. Operating

history was reviewed &J undemnd the plant

response to a full flow "cI[ injection.

Only one HWI injection was

found that was at full flow

for lag enough to determine the expected

plani response, As documented

in AR 102456-10

Atta&ment

5, JJ Unit 2 HPCI scram response injection

on 8/16/90 increased level

from 123" to 153" in just less than 60 seconds. This short response takes less time than would be first expected BB the increase in indicated

8evd is caused by both the inventory

mskeup md "level swell"

cwRlsed by the "CI steam flow induced

vessel pressure reduction.

Since level increased

30" in 6Q seconds, this is an amate duration fer assumed RCIC backup HPCI full flow injections

while opemtom arc maintaining vessel level between

170" and 200". Ah 4 horn, if 8.4300 gpm injection

were tu Stan witk CST level at just above slevatkm 23' 4", air entrainment

could stafl at L1= 5.3.7 inch based on JP(jc2QQ1/PWR-19010

Equation 6, (31 seconds into the injection, see Attachment

for details).

It would require 62 seconds of HK.1 injection

for air to travel the 228' to the pump, Since only 60 SWQ~~S of injection

is expscted, no air will reach the pump.

Any postulated

1. pCI full flow rate injection

for this case with CST level starting at just above elevation

23'4" will result in no air reaching the

pump during that speeific injection.

The Wpcl suction swap would then be completed

prior to the next HPCI injection.

This provides the protection

that is nw$ed to prevent continued

HPCI operation with

the suction lid up Ma depktsd CST.

OCT--D?-2005

03:47 PN BRUHSWICK

REG F1FF 9104573614

8.31 AR 106230-10

Operability

Review Page 16 of 20 * HgCI is providing the Backup

to RCIC function hmpt MSIV ciosupe does nat occur Opemton initiate suppression

pot cooling . Opmtops eventually

perform ve5sel level contd in accordance

with preceduren

"CI operation

done wit! be considered

as RCIC unavailability

is part of the ease definition.

"CI will auto start on low reactor warer level (LL2, 105"). HPCI will trip when level reaches the high mor Ievel trip setpoint at

2W'. bvel may continue to cycle between 105 an8 206" until such time that opereton have had time to assess plant

conditions

and complete any ether more important

actions. Manual actions to controi level in specified

170" to 2QO" kvel control band would probably take place

early in the event. However, it is not needed to sssurne them actions until after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into the event.

This event dws not involve a small break LOCA, but it may involve a loss of CrOyweH cdlng. Drywell heating and pressurization

to above 2 psig may or may not occur. Operators

would place a! lewt one loop of RHR in suppmsion

pool cooling at

f 0 minutes. With RHR in suppreasion

pool cooling snd the reactor

not depressurized via

SRV6, suppression

pool tempemtiares would not

inmase to a value where overriding

the WCI high suppression

poot level transfer Insmmmtation

is allowed. Note that if RHR suppression

pl coaiing is not started, WCI would eventually

be opting with the suctien lined up to the suppssim pod and the suppmsim pool water temperature

above the value allowed for "CI operation, Continued

operation

of HPCI tends to depressurize

the vessel BS it removes steam from the reactor and 88 it inject8 low temperature

water into the vesscl. Although it ia possible that continued

HPGI operation could

reduce vessel pressufe to below the "Cf isolation

setpoint prior to any automatic

suction transfer for small breaks, this is not expected for the case being considered

here. Much of the coolant leaving the

vessel will be discharged

to the main condenser

in this cwe. One potential

initiator

for this event would be a loss of condensate

system pnssurc boundary inte@ty ar loss of condensate

sysrern flow path. For this case it is appropriate

to assume that the high suppmsim pool KPCI level instrument

setpoint is not reached prior to the CST depletion

that would be expected after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into

the event. AH pmeters aasoeisted

with the suctim transfer are the sme as for Case 3. Either the IPCI flow rare would be reduced acceptably

or HPCI wouid be operated at full flow for Only 60 seconds. For the full flow cwe, no air would reach

the pump during the last injection

with 6ST suction and

the HPCK suction swap would then be completed

prior to the next Hp(31 injetion.

This provides the protection

that is ncedd to prevent continued

HpeI opratim with the sUCtiim lined up to a depleted em.

-~ ' KICT-09-2B83

03:48 PPI BRUNSWICK

REG BFF 9104573014

AK 106230-10

Operability

Review Page I7 of 28 mere are no specific limitations.

As long as operators

comply with pedure requirements

as they m gained to do, ?he setpoint is adequate to supp~fl the PfPCI licensing

basis functions

and can be consided operable with no compensatory

actions. Technical

Specification

3.5.1, Table 3.3.5.1 Technicd Specification

Bw B 3.3.5.1 WSAR 5.4.6,6.2.2.3,6.3.1.2.1.6.3.2.8,6.3.3.5.5,9.2.6.2

EGR-NGGe-0019, Engineering

Operability

Assessment

N]RC Inspection

Manual, Part 9900: Technical

Guidance §TSlOO.II%

and sm 100P.STS h%C Infomath Notice 97-76 dated 10/23/97:

Crediting

of Operator Actions in Place of Automatic

Actions and Modifications

of Operator Actions, Including

Response Times GL91-18 rev. 1 * SAE.WGE§TR-LOQcA

Analysis Submittal, dated March 29 1989 h?ZW31624P.

Brunswick

Steam Electric Plant, Units II & 2. SAFBWOESTR-LOCA

hsa-sf~Qulanr

Accident Analysis SWGESTR-LWA

Analysis Response to Request For Additional

Infomation, datal May 17,1989 NRC approval ledter and SER for SAFEWGESTR-LOGA

ANALYSIS, BRUMSWICK

STEAIW ELECTRIC PLANT, UNITS 1 AND 2, dated lune 1.1989 m Bmnswick Unite 1 and 2 Extended Power Wprate submittal

dated O8/09101 * NEDC-33039P, 'Ke Safety Analysis Report for Brunswick

Units 1 and 2 Extended Power Wprate * Ex& Pwcr Uprate Kcspensc to Request For Additionel

Infomation, dated 03/12@2 c m2001/BwR-19010

rn-02626 FP-02762 AB102456 * ESR 95-61733 Rev. 0 AI 15 BSR99-00062

P.32

OCl--B9--2003

0S1:48 PM BRUNSWICK

REG FlFF 9104573814

P.33 AR 106230-10

Operability

Review Page 18 of 20 Attachment

General inputs of CST volume determinations

are as foollows:

input Tank OD from Tank shell thickness, 1st ring Tank shell heigth, 1st ring Tank shell thickness, ring 2, 3 & 4 t-tPCVRC!C

nozzle (N-1) centerline

HkCt/RCi6

nozzle (N-1 j thickness

HPCVRCIC nOZle (N-1) SIZ& HPGllRClC

noule (N-1) ID Volumes to specific levels

Normal Low bevel per OP 31 2 Level needed for routine OPT-09.2 01-03.6 & UFSAR 9.2.6.2 req'd level Nominal drain down via CRD MZ (CR[a/cond)

i% N9 [CS) Nozzel bottom Top of first ring HPCI lnstr Max Setpoint adjusted for AR 102466 HPCI lnstr Nom Setpoint adjusted for AB 102456 HBCl lnetr Min Setpoint adjuijlasted

for AW 102456 HPCl lnlstr T/S adjusted for AR 102456 RCIC lnstr Max Setpoint adjusted for AR 102456 RCIC lnatr Nom Setpoint adjusted for Af? 102456 RCIC lnstr Min Setpoint adjusted for AR 102456 8616 lnstr TIS adjusted far AR 102458 HPCilRClC

Sucd Top HPCllRC1C

Suct Centerline

APP UA-04 5-7 Source FP 2626 FP 2626 FB 2626 FP 2626 FP 2626 FP 2628 FP 2626 FP 2626 Height (in) 40.0 39.5 39.0 38.5 36.0 35.5 35.0 34'5 31.5 24.0 Height (ft) 23.50 20.00 12.00 10.00 9.50 9.38 7.75 3.333 3.292 3.250 3.208 3.000 2.958 2.81 7 2.875 2.625 2.000 Value 52 ft 0.279 in 7.75 ft 0.25 in 2ft 16 in 0.5 in 15 In Volume Volume (e%) (gallons)

49,824 372712 42,403 317198 25,441 190310 21,208 158588 20,140 1 50667 16,428 12295O 7,066 52860 6,978 52205 6,890 52539 6,801 50878 6,360 47574 6,271 46914 6,183 46253 6,095 45592 6,566 41628 4,240 31716 19,875 i 481375 Note distances

above are referenced

to the tank bottom at plant eievarlon

20' 1.5' bl from fop of nozzle ID to HPCl Tech Spec 7.0 Volume, 10' to HPCl max setpoint 14,134 155727 Volume, 10 to HPCI Tech Spec 14,389 107710 Volume, 23.s' to HPCl Tech Spec 43,023 321834 Volume, 20 to HPCl Tech Spec 35,602 266320 Volume, 16' to HPCI Tech Spec 27,221 202876

.-*.I - 9 1 84 57 38 1 4 , OCT--89--2BE3

03:49 PPl BRUNSWICK

REG RFF AR 106230-10 Operability

Review Page 19 of 20 Attachment

EBB 6541733 Rev. 0 AI 15 was used to document the HPCI Suppression

Pool HI Level Instment bwis. The values and methods of this document were used to determine

the Containment

Inventory

increase assuming small break, HPCI plus RClC operation

at 4700 gpm until the HPCl Suppression

Pool Hi auto transfer Tech Spec level of -24" Is react4 assuming no operator actions. With Torus level starting at The Torus inventory

wouM be With Torus level ending at The Torus inventory

would be Torus inventory

increase ~iyweil spill over volume (rnax. no misc structures)

Endwd volume Plui sump volume Minus pedestal volume Total Injection

volume Or HPCi injection

flow rate Minimum standby total inventwy in CST (1 0') Tank volume at Hi Torus Transfer start Tank afeR near bottom Tank Level at HI Torus Transfer Or Top of HPCi nozzle ID (FP-02826)

Nozzle subinergence (U ) Ushg llmithg wive stroke rimes and no credit for flow r$duction

prim to end cf valve travel the level duction for the transfer will be 85 fOllOWS: E41-F041/!%42

stroke tlme TOM transfer time HPCl flow durlng transter C~T wlurne at end d valve motion Tank Level Nozzle submergence (U) E41 -F004 EilrOk8 flille *31 in 87140 eu ft -24 in 9a90 cuft 43160 gallons 5770 cun 7306 GU R loo CUB 585 cuft m1 cuft la11 CUR 94330 gal 4700 QPm 158588 gallcns 84257 gallons 8599 ft* 2120 w2 48.83 in 31.50 in 17.13 In 4.05 ft 70 8Bc 76 8s 154 see 12063 galllons 52194 gallon8 6978 w 39.50 in 8.00 In P. 34

AR iOg230-10

Operability

Review 0 1 2 3 Q $0 11 12 19 16 16 16 $7 18 .~ 18 20 21 22 l.1 7.00 8.95 8.88 8.84 8.78 8.73 6.87 6.67 6.81 6.48 8.34 6,'ZLl 8.24 8.18 8.13 &OB 6.M 5.87 5.91 8.86 6.81 6.75 5.64 6.88 6.53 5.48 8.43 5.37 5.52 5.26 6.21 5.16 6.W 5.05 4.w 4.84 4.88 4.e3 4.77 4.72 4.67 6.81 4.63 4.69 1.43 b.38 424 4.29 4.23 4.18 4.12 4.07 6.01 3.98 3.91 3.86 3.W IM) 3.a 8.82 e.@ 8.m 3.80 Pa¶ 1 1 1 1 1 I 1 1 1 1 1 1 1 1 9 1 1 1 1 1 1 1 1 1 I 11 1 ? 1 1 1 1 1 1 1 1 1 1 t 1 FWA vel 7.g 722 7.22 7.22 7.22 7.22 7.22 7.22 7.22 7.22 7.22 7.22 ?.?.E 7.22 7.22 7.22 7.22 7.22 7.11 7.22 7.22 7.22 7.P 7.22 7.21 7.22 7.22 7.22 7.92 7.22 7.22 7.22 7.22 7.22 7.21 9.22 7.72 7.22 722 ?.a 7.22 7.22 722 7.22 7.22 7.7.2 7.22 7.22 7.22 7.22 7.22 7.8 7.a 722 7.22 7.22 7,zz 7.22 7.22 7.E 7.22 72.2 7.22 FO42 POS 0.m 0.013 Q.028 0.038 0.061 0.064 0.0V 0.080 0.103 0.115 0.128 0.141 0.154 0.187 0.179 0.1% 0305 0.218 0.Pl 0.244 0.288 0.m 0.2W 0.308 0.321 0.333 0.346 0.358 0.372 0.386 0.M 0.410 0.423 0.436 Q.448 0.462 0.474 0.447 0.903 0.513 0.528 0.538 0.681 0.664 0.477 0.580 0.m 0.816 0,m 0.641 0.654 0.W 0.879 0.692 a.ms 8.718 0.73t 0744 0.75e 0.789 0.782 8.7115 0.258 Air DlSt ffl) ? 14 22 28 38 4a 51 50 55 72 78 87 Bl 1Qd 1 08 1t6 123 130 13? $44 152 18% 186 173 160 186 9% 202 208 217 224 291 91045930114

Page 20 of 20 P. 35