IR 05000454/2003004

IR 05000454-03-004, 05000455-03-004, on 05/05-23/2003; Exelon Generation Company, LLC; Byron Station, Units 1 and 2; Safety System Design and Performance Capability Inspection

ML031920333
Person / Time
Site:	Byron
Issue date:	07/03/2003
From:	Dave Hills NRC/RGN-III/DRS/MEB
To:	Skolds J Exelon Generation Co
References
IR-03-004
Download: ML031920333 (41)

Text

uly 3, 2003

SUBJECT:

BYRON STATION, UNITS 1 AND 2 NRC INSPECTION REPORT 50-454/03-04(DRS); 50-455/03-04(DRS)

Dear Mr. Skolds:

On May 23, 2003, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Byron Station, Units 1 and 2. The enclosed report documents the inspection findings, which were discussed on May 23, 2003, with Mr. S. Kuczynski and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. Specifically, this inspection focused on the design and performance capability of the auxiliary feedwater and DC power systems to ensure that they were capable of performing their required safety-related functions.

Based on the results of this inspection, there were three NRC-identified findings of very low safety significance, of which two involved violations of NRC requirements. However, because these violations were non-willful and non-repetitive and because they were entered into your corrective action program, the NRC is treating the issues as Non-Cited Violations in accordance with Section VI.A.1 of the NRCs Enforcement Policy.

If you contest the subject or severity of a Non-Cited Violation, you should provide a response within 30 days of the date of this inspection report, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region III, 801 Warrenville Road, Lisle, IL 60532-4351; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Byron Station. In accordance with 10 CFR 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 (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Dave E. Hills, Chief Mechanical Engineering Branch Division of Reactor Safety Docket Nos. 50-454; 50-455 License Nos. NPF-37; NPF-66 Enclosure: Inspection Report 50-454/03-04(DRS); 50-455/03-04(DRS)

w/Attachment: Supplemental Information cc w/encl: Site Vice President - Byron Byron Station Plant Manager Regulatory Assurance Manager - Byron Chief Operating Officer Senior Vice President - Nuclear Services Senior Vice President - Mid-West Regional Operating Group Vice President - Mid-West Operations Support Vice President - Licensing and Regulatory Affairs Director Licensing - Mid-West Regional Operating Group Manager Licensing - Braidwood and Byron Senior Counsel, Nuclear, Mid-West Regional Operating Group Document Control Desk - Licensing M. Aguilar, Assistant Attorney General Illinois Department of Nuclear Safety State Liaison Officer State Liaison Officer, State of Wisconsin Chairman, Illinois Commerce Commission

SUMMARY OF FINDINGS

IR 05000454/03-004(DRS), 05000455/03-004(DRS); Exelon Generation Company, LLC; 05/5-23/2003; Byron Station, Units 1 and 2; Safety System Design and Performance Capability Inspection.

This report covers a 3-week announced baseline inspection of the design and performance capability of the auxiliary feedwater and DC power systems. The inspection was conducted by regional engineering specialists with mechanical consultants assistance. Three Green findings associated with two non-cited violations were identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC)0609, Significance Determination Process (SDP). Findings for which the SDP 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 July 2000.

A. Inspector-Identified and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green.

A finding of very low safety significance was identified involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, that related to the design basis requirement to maintain auxiliary feedwater instrumentation piping water solid, not being correctly translated into specifications, drawings, procedures, or instructions. This resulted in a void developing in the piping to the suction pressure transmitters 1(2)PT-AF055, which perform a safety-related function to sense low suction pressure and initiate a swap over to the essential service water system on loss of the condensate storage tank.

The finding was more than minor because a lack of coordination between design requirements and procedural guidance affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance because it did not represent an actual loss of a safety function as the automatic switchover would still have occurred prior to the pumps losing suction pressure. (Section 1R21.2)

Green.

A finding of very low safety significance was identified associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, that related to the coordination, content, and control of design basis engineering calculations. Specifically, the inspectors identified a number of concerns related to the coordination, content, and control of existing calculations (including the failure to coordinate calculation inputs and assumptions as existing design basis calculations are revised or as additional calculations are originated), the use of incorrect or unsupported inputs or assumptions in design basis calculations, the absence of calculations to support some aspects of the current design basis, the failure to appropriately supercede certain calculations or to denote other calculations as historical documents, and, in certain instances, errors in existing calculations. As a result of these issues, the current design basis calculations, as well as the existing calculation control processes, may not be adequate to ensure that the design basis will continue to be maintained. Although none of the specific deficiencies identified during the inspection resulted in immediate operability concerns, it was concluded that the auxiliary feedwater system design basis was not being adequately controlled by the existing calculations nor by the licensees processes for coordination and control of the calculations.

This finding was more than minor based on the potential that the lack of adequate control and quality of design basis calculations could result in the ability of the auxiliary feedwater system to perform its safety functions to be degraded. Design basis calculations were routinely used in support of design changes, operating procedures, test acceptance criteria, and operability determinations. This finding is assessed as Green because it did not represent an actual loss of the auxiliary feedwater systems safety function. (Section 1R21.2)

Green.

A finding of very low safety significance was identified involving not maintaining a commitment to the NRC to have placards on the main control board. The placards provided guidance to operators to ensure the auxiliary feedwater pumps had sufficient recirculation flow prior to reducing flow to the steam generators below 100 gpm [gallons per minute], such that the pumps remained protected from being run at shutoff conditions that would have resulted in pump damage.

This finding was more than minor because this lack of guidance could have affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance because it did not represent an actual loss of a safety function. (Section 1R21.2).

Licensee-Identified Violations

No findings of significance were identified.

REPORT DETAILS

1R21 Safety System Design and Performance Capability

Introduction Inspection of safety system design and performance verifies the initial design and subsequent modifications and provides monitoring of the capability of the selected systems to perform design bases functions. As plants age, the design bases may be lost and important design features may be altered or disabled. The plant risk assessment model is based on the capability of the as-built safety system to perform the intended safety functions successfully. This inspectable area verifies aspects of the mitigating systems cornerstone for which there are no indicators to measure performance.

The objective of the safety system design and performance capability inspection is to assess the adequacy of calculations, analyses, other engineering documents, and operational and testing practices that were used to support the performance of the selected systems during normal, abnormal, and accident conditions. The inspection was performed by a team of inspectors that consisted of a team leader, two Region III inspectors, and two mechanical consultants.

The auxiliary feedwater and DC [direct current] power systems were selected for review during this inspection based upon:

• having a high probabilistic risk analysis ranking;

• being high safety significant maintenance rule systems; and

• not having received recent NRC review.

The criteria used to determine the systems performance included:

• applicable Technical Specifications;

• applicable Updated Final Safety Analysis Report (UFSAR) sections; and

• the systems design documents.

The following system and component attributes were reviewed in detail:

System Requirements Process Medium - water, electricity Energy Source - electrical power, air Control Systems - initiation, control, and shutdown actions System Condition and Capability Installed Configuration - elevation and flow path operation Operation - system alignments and operator actions Design - calculations and procedures Testing - flow rate, pressure, temperature, voltage, and levels Components The auxiliary feedwater pumps, including the minimum recirculation flow line, and DC batteries were selected for detailed review during the inspection. These components were specifically reviewed for component degradation due to the impact that its failure would have on the plant.

.1 System Requirements

a. Inspection Scope

The inspectors reviewed the UFSAR, Technical Specifications, drawings and available design basis information to determine the performance requirements of the auxiliary feedwater and DC power systems. The reviewed systems attributes included process medium, energy sources, and control systems. The rationale for reviewing each of the attributes was:

Process Medium: This attribute required review to ensure that the auxiliary feedwater pumps would supply the required flow to the steam generators following design basis events. To achieve this function, the inspectors verified that the auxiliary feedwater system would be able to supply the required flow rates to the steam generators from the non-safety-related condensate storage tank and the safety-related essential service water system.

Energy Sources: This attribute required review to ensure that the auxiliary feedwater pumps would start when called upon, and that appropriate valves would have sufficient power to change state when so required. To achieve this function, the inspectors verified that the interactions between the auxiliary feedwater pumps and their support systems were appropriate such that all components would start when needed under normal or standby electrical power. The DC batteries were verified to ensure they could supply the required voltage to safety related components following design basis events.

Controls: This attribute required review to ensure that the automatic controls for starting the auxiliary feedwater pumps, and associated system components, were properly established. Additionally, review of alarms and indicators was necessary to ensure that operator actions would be accomplished in accordance with the design.

b. Findings

No findings of significance were identified.

.2 System Condition and Capability

a. Inspection Scope

The inspectors reviewed design basis documents and plant drawings, abnormal and emergency operating procedures, requirements, and commitments identified in the UFSAR and Technical Specifications. The inspectors compared the information in these documents to applicable electrical, instrumentation and control, and mechanical calculations, setpoint changes, and plant modifications. The inspectors also reviewed operational procedures to verify that instructions to operators were consistent with design assumptions.

The inspectors reviewed information to verify that the actual system condition and tested capability were consistent with the identified design bases. Specifically, the inspectors reviewed the installed configuration, the system operation, the detailed design, and the system testing, as described below.

Installed Configuration: The inspectors confirmed that the installed configuration of the auxiliary feedwater and DC power systems met the design basis by performing detailed system walkdowns. The walkdowns focused on the installation and configuration of piping, components, and instruments; the placement of protective barriers and systems; the susceptibility to flooding, fire, or other environmental concerns; physical separation; provisions for seismic and other pressure transient concerns; and the conformance of the currently installed configuration of the systems with the design and licensing bases.

Design: The inspectors reviewed the mechanical, electrical, and instrumentation design of the auxiliary feedwater and DC power systems to verify that the systems and subsystems would function as required under accident conditions. This included a review of the design bases, design changes, design assumptions, calculations, boundary conditions, and models as well as a review of selected modification packages.

Instrumentation was reviewed to verify appropriateness of applications and setpoints based on the required equipment function. Additionally, the inspectors performed limited analyses in several areas to verify the appropriateness of the design values.

Testing: The inspectors reviewed records of selected periodic testing and calibration procedures and results to verify that the design requirements of calculations, drawings, and procedures were incorporated in the system and were adequately demonstrated by test results. Test results were also reviewed to ensure automatic initiations occurred within required times and that testing was consistent with design basis information.

.1 Auxiliary Feedwater (AF) Suction Pressure Instrumentation

Introduction:

The inspectors identified a finding of very low safety significance involving a Green Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the inspectors identified that the design bases requirement to maintain AF instrumentation piping water solid was not correctly translated into specifications, drawings, procedures, and instructions.

Description:

The inspectors conducted a walkdown of the AF system on May 6, 2003, and observed that some instrument piping for the diesel driven AF (DDAF) pumps was installed with a large inverted U shaped loop. This piping was to the suction pressure transmitters 1(2)PT-AF055, which perform a safety-related function to sense low suction pressure and initiate a swap over to the essential service water system (SX) on loss of the condensate storage tank (CST) for the AF pumps.

The inspectors observed that there were no high point vents and questioned how these lines were assured to be water solid. The licensee determined that there was no periodic procedure to vent these lines to ensure they remained water solid. The licensee conducted ultrasonic testing and found voids at the high point towards the pump side of the lines on both units. The licensee calculated that the largest void would lower the swap over setpoint by approximately 2.3 psi [pounds per square inch]. The licensee vented the lines to ensure operability of the system.

The licensee subsequently evaluated the effect of the setpoint bias on operability and concluded that the system would have performed its safety function.

Analysis:

Evaluation of this issue concluded that it was a design control deficiency resulting in a finding of very low safety significance (Green). The deficiency was due to the licensees failure to maintain these suction instrument lines in a water solid condition.

This finding was determined to be greater than minor because this lack of coordination between design requirements and procedural guidance affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

The finding was assessed through Phase I of the significance determination process.

The inspectors agreed with the licensee's position that, despite the loss of margin in the swap over setpoint, the system would perform its safety function. Therefore, the inspectors concluded that the finding was a design deficiency that did not represent an actual loss of a safety function and the issue screened out as having very low safety significance or Green.

Enforcement:

10 CFR Part 50, Appendix B, Criterion III, "Design Control," requires, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.

Contrary to the above, as of May 6, 2003, the design basis requirement to maintain safety-related instrument piping water solid was not correctly translated into specifications, drawings, procedures, or instructions for the AF system. Specifically, a void was discovered in safety-related instrument piping that lowered the setpoint for AF suction swap over to the safety-related SX water supply. Because the licensee entered the condition into their corrective action system as condition report (CR) 157954, this violation is being treated as a Non-Cited Violation consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 50-454, 455/2003004-01).

.2 Calculation Issues

Calculation Coordination, Content, and Control

Introduction:

The inspectors identified a finding of very low safety significance involving a Green Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, that related to the coordination, content, and control of design basis engineering calculations. Specifically, the inspectors identified a number of concerns related to the coordination, content, and control of existing calculations including the failure to coordinate calculation inputs and assumptions as existing design basis calculations are revised or as additional calculations are originated, the use of incorrect or unsupported inputs or assumptions in design basis calculations, the absence of calculations to support some aspects of the current design basis, the failure to appropriately supercede certain calculations or to denote other calculations as historical documents, and other minor errors in existing calculations. As a result of these issues, the inspectors determined that the current design basis calculations, as well as the existing calculation coordination and control processes, may not be adequate to ensure that the design basis will continue to be maintained. Although the specific deficiencies identified during the inspection did not result in immediate operability concerns, the inspectors concluded that the AF system design basis was not being adequately controlled by the existing calculations nor by the licensee's processes for coordination and control of the calculations.

Discussion: The inspectors review of the licensees calculation indices identified 60 calculations addressing the hydraulic design of the AF system. A number of these calculations appeared to be redundant. The inspectors requested the licensee to identify which of these calculations were the design basis hydraulic calculations for the AF system. In the response to this request, the licensee provided a list in excess of 30 calculations. The inspectors noted that the large number of calculations complicated the licensees ability to maintain the design basis, perform design and modification activities, and perform operability determinations given the fragmented design information.

During the inspection, the inspectors noted a number of calculation deficiencies as well as deficiencies in the licensee's processes for coordination and control of the calculations. The licensee initiated individual CRs, as appropriate, to ensure that each of these conditions will be addressed by the corrective action program. The following discussion includes examples of deficiencies identified during the inspection in individual calculations and in the calculation coordination and control processes.

Calculation Coordination - The inspectors identified a concern related to the control and coordination of existing calculations. As shown in the following examples, conditions were identified where design basis calculations were not based on current input data, were based on assumed inputs in lieu of actual plant conditions, were not consistent with other design basis calculations, or were not revised when appropriate to reflect a change in input data. In response to these concerns, the licensee initiated a number of CRs to address the individual issues. The number of the fragmented design basis documents, and inconsistent inputs and assumptions, which were not properly linked together, underscored the inspectors concern about the licensees acceptance of the results of the hydraulic calculation that were not conservative in respect to the licensing analysis and their ability to perform correct and timely operability determinations. The fragmented design information could result in the implementation of modifications or operational decisions that would have a negative impact on the plant safety and conformance to the licensing basis.

• Calculations PSA-B-97-31, Auxiliary Feedwater System Operation Analyses for Byron and Braidwood Stations, and PSA-B-98-01, Auxiliary Feedwater System Design Basis Analyses for Byron and Braidwood, established maximum and minimum design flow rates for the AF system based on the predicted variations in the system resistance and pump performance. These calculations predicted a flow of 146 gpm [gallons per minute] to each of the three intact steam generators for the feed water line break accident. This value was less than the 151 gpm used in the feed water line break analysis and reported in the UFSAR Sections 10.4.9.3.2 and 15.2.8.2. The 146 gpm flow rate was based on the maximum allowable degraded AF pump performance. The licensee initiated CR 19541 to address this issue. At the time of the inspection, no operability issues were associated with this condition since the AF pumps were exhibiting very little degradation.

• The licensees AF system review performed prior to this inspection identified an over pressurization of the AF discharge piping. This finding was documented in CR 157062. The inspectorss review of this CR identified that the licensee did not consider the allowable pump performance above nominal

+7.5% curve developed in calculation PSA-B-97-17, Byron/Braidwood Maximum and Minimum AFW Pump Curve Development. The results of the unit 2 DDAF pump surveillance, 2BVSR AF-3, Simultaneous Start of Both AF Pumps with Flow to the Steam Generators, identified that this pump performance at the nominal flow rate was on the +7.5% curve. The licensee initial response attributed this to measurement inaccuracy, since pump developed head cannot be higher than the one shown on the certified vendor curve at the same flow rate. The inspectors, however, determined the certified vendor curve was developed at a pump speed of 3570 rpm, whereas step 4.3.12 of procedure 2BVSR 5.5.8.AF.1-2, ASME Surveillance Requirements for the Diesel Driven Auxiliary Feedwater Pump, required setting the AF pump speed to 3646-3794 rpm. The licensee determined that although the overpressure value would be higher, the conclusions of the CR 157062 would be still valid, since the final pressure would not exceed the American Society of Mechanical Engineers (ASME) Code over pressurization allowable, hence no detailed evaluation of the pressure boundary components was required. The conclusions of the licensees evaluation was based on application of the +7.5%

pump curve. The inspectors, however, noted the application of the pump affinity laws for the nominal pump speed of 3720 rpm [revolutions per minute] would yield a higher developed head than the one predicted for the +7.5% pump curve (at the same flow rate). The licensee concurred with the inspectors assessment and initiated CR 160059 to address the overpressure condition. The licensee concluded that there was no operability concern based on a similar evaluation performed by Braidwood Station which evaluated the effect of the discharge pressure on all components at elevated pump speed. The inspectors independent verification confirmed that the Byron Station pump discharge pressures would be bounded by the Braidwood Station conditions and was considered acceptable.

• The inspectors review of the following calculations identified that these calculations were based on the maximum AF flow rate of 990 gpm.

• SX1-87, "Available NPSH for AF Pump when Supplied from SX System;

• SX1-88, Available NPSH for AF Pump when Supplied from SX System;

• PAS-B-91-14, Evaluation of New CST Technical Specification Levels for Byron and Braidwood Stations; and

• PSAG-138, Available NPSH for AF Pump when Supplied from SX System.

This flow rate was less than the maximum flow predicted in calculations PSA-B-97-31 and PSA-B-98-01. The higher flow rates result in the increase of the piping suction losses and the pump internal entrance losses. These increases lead to the decrease in the available net positive suction head (NPSH)and the increase of the required NPSH. Hence, for the same suction static pressure, the flow increase could lead to the condition where the required NPSH was less than the available NPSH, which would result in a cavitation condition.

The licensee concurred with the inspectors finding that the NPSH calculation did not address the higher flow rates and initiated CR 159779 to address this issue.

The evaluation determined that there was no operability concern, since the maximum flow rate conditions were predicted to exist (by calculations PSA-B-97-31 and PSA-B-98-01) during elevated static suction pressure conditions, thus the available NPSH was greater than the required NPSH. The inspectors noted that the design hydraulic calculations (PSA-B-97-31 and PSA-B-98-01) did not address that the operational controls for AF were based exclusively on the steam generator level without any restrictions on the flow.

Additionally, a loss of offsite power event during the final stages of normal plant cooldown could result in the most challenging NPSH condition. The AF pump would receive an automatic initiation signal while the level in the CST would be at a low level, the steam generators would be at a low pressure, and all flow controlled valves would go to the full open position due loss of the non-safety related instrument air. The licensee initiated CR 160098, which agreed that the current operating procedures did not provide the maximum flow limitations and stated that it might be prudent to include maximum flow limitations in our procedures to provide the operations department with an upper limit on flow.

The licensee determined that there was no operability concern based on the operator training, operating procedures, high flow alarms, and indications that exist to provide the operator with various indications of high flow conditions. The inspectors concurred with the licensees operability assessment. Additionally, the inspectors independent evaluation of the analytical value for the automatic suction switchover from the CST to the SX determined that there was sufficient margin to prevent any cavitation potential based on the current value of this setpoint, including accounting for the air void in the instrument sensing line.

Incorrect or Unsupported Inputs and Assumptions - The inspectors identified concerns related to incorrect or unsupported inputs and assumptions in the existing calculations.

As shown in the following examples, conditions were identified where design basis calculations were not based on appropriately documented inputs or assumptions.

• Calculation PSAG-138 evaluated the available NPSH to the AF pumps assuming various scenarios of silt blockage of the pump suction piping from the SX system. The inspectors identified that the calculation used non-conservative input values for L/D [length over diameter] of 12 for a 45-degree pipe elbow and L/D of 14 for a 90-degree elbow. The licensee subsequently determined the correct L/D values to be 16 and 30, respectively, and that the discrepancy did not affect the conclusions in the calculation, as the postulated silt blockage scenarios would not be realistic based on the results of programs in place to prevent silt buildup in the piping. The licensee documented this discrepancy in CR 158958.

• Calculation PSA-B-97-17 incorrectly indicated the AF005 valve's were non-safety grade in section 2.1. The Passport D031 panel indicated the AF005 valve's were safety-related. UFSAR 10.4.9.1.1 indicates the AF005 valve's were safety category I. The licensee documented these inconsistencies in CR 157904.

• During a plant cooldown following a reactor trip, the AF pumps could be placed in runout flow condition if no flow control actions were taken. Current procedures focused on maintaining steam generator level with minimal reference to pump protection. Design basis hydraulic calculations (e.g. PSA-B-97-31 and PSA-B-98-01) did not consider actual plant operation. The licensee documented this issue in CR 160098.

• The NPSH required values for the DDAF pumps used in design basis hydraulic calculations (e.g., PSA-B-97-31 and PSA-B-98-01) did not include an adjustment for pump speed. The licensee documented this issue in CR 160105.

• Calculation PSA-B-97-14, Evaluation of New CST Technical Specifications Levels for Byron and Braidwood Stations, was developed using piping friction losses calculated based on two AF pumps in operation at 990 gpm each, as well as pressure loss values measured with one AF pump in operation at 720 gpm and at 990 gpm. The measured values for one pump operation at 720 and 990 gpm were selected as the more conservative of referenced individual test values for Byron and Braidwood Stations. Referenced measurements for the Braidwood Station were indicated to be the more conservative values and were used in PSA-B-97-14. However, the inspectors identified that the referenced Braidwood information was not available in the calculation's reference materials.

Also the measurements for Byron Station were referenced in PSA-B-97-14, but the inspectors identified that the referenced Byron information was not available in the reference materials. The licensee had initiated CR 154658 previously in relation to this issue and other calculations during the licensees focused area self assessment.

Lack of Available Calculations to Support Aspects of the Current Design Basis - The inspectors identified the following examples of design basis requirements or information that were not supported by available calculations. These conditions also appear to be related to the deficiencies in calculation coordination and control.

• The inspectors identified there was not an approved calculation supporting the determination of the required minimum volume of 420 gallons of fuel oil in the DDAF pump day tank. The licensee provided an initial simple calculation of the required volume that appeared acceptable, and indicated a formal approved calculation would be developed. The licensee documented this issue in CR 159411.

• The inspectors identified there was no supporting documentation to verify the installer's notes on Stewart & Stevenson Drawing 62242 (Commonwealth Edison Drawing 62240-1, Sheet 2, Installation Drawing - 16V-149T1, Auxiliary Feedwater Pump Drives, Revision D) were met in relation to the relative installed locations of the day tank and the DDAF pump diesel engine. Note 2 stated, No point in the fuel system including off skid piping and day tank should exceed fuel pump suction elevation which is 55 inches above the bottom of the skid. The note was not met based on the as-built configuration. The licensee's initial evaluation concluded that existing administrative controls and surveillance requirements to monitor the diesel systems provided adequate compensatory measures to meet the intent of the note. The licensee documented this issue in CR 160119 and indicated an engineering evaluation will be performed to formally document the acceptability of the installed configuration.

• The UFSAR stated that the SX booster pump (1/2SX04P) was capable of providing the required SX flow to the DDAF pump diesel engine in the event the SX pumps were lost. The inspectors questioned how the SX system would be able to provide adequate suction pressure to the booster pump under this condition. The licensee documented in CR 159208 that no formal design calculation existed to document the acceptability of the various operating modes for the booster pumps nor to document the available NPSH to justify the ability of the pump to adequately perform under normal, abnormal, and accident conditions. The licensee's initial evaluation concluded that a margin of approximately 36 feet did exist between the required and available NPSH.

• The inspectors identified there was no specific calculation to address hydrogen generation by the DDAF pump Ni-Cad battery banks. This would verify that the ventilation in the DDAF pump rooms was adequate to maintain hydrogen concentration less than 4%. The licensee documented this issue in CR 157816.

Calculations not Superceded or not Designated as Historical - The inspectors concluded that it was difficult to identify the status of calculations, and to determine if a calculation was a current design basis calculation. Several calculations contained outdated information, but were not superceded or designated as historical. This provided the potential for incorporating erroneous information into a new calculation. In response to this concern, the licensee initiated the following CRs to capture the individual instances:

• CR 157922 Braidwood calculation ATD-0054 on ESW system cooled lube oil heat exchangers - Not applicable for Byron

• CR 159305 Loop seal criteria calculation AF-91 - Should be historical

• CR 159770 Medium energy line break calculation AFW-KG1 (AF005 valve area) - Not voided

• CR 160022 Calculations AF-TH06 and MAD 89-0175 contain outdated information - Should be historical

Analysis:

Evaluation of this issue concluded that it is a design control deficiency resulting in a finding of very low safety significance (Green). The design control deficiency was due to a licensee performance deficiency in that certain design calculations either were not adequately coordinated, contained incorrect or unsupported inputs and assumptions, did not exist, contained errors, or that the current, historical, or superceded status of certain calculations was not maintained accurate and up-to-date.

The Mitigating Systems Cornerstone was affected due to the potential for the AF system's capability to provide heat removal function being degraded by this condition.

No other cornerstones were degraded as a result of this issue.

The inspectors determined that this finding was associated with design control attributes and affected the objective of the Mitigating Systems Cornerstone to ensure the capability of the AF system to respond to initiating events to prevent undesirable consequences, and is therefore greater than minor. The lack of adequate coordination, control, content, status, and quality of design basis calculations had the potential to result in the ability of the AF system to perform its safety functions or to be degraded.

Design basis calculations were routinely used in support of design changes, operating procedures, test acceptance criteria, and operability determinations.

The finding was assessed through Phase I of the significance determination process. A review of the system calculations identified a number of deficiencies, however, they did not result in immediate operability concerns. This provided reasonable assurance that there was not an actual loss of system function due to this condition. Therefore, this issue was screened out of the significance determination process as Green.

Enforcement:

10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.

Contrary to the above, as of May 23, 2003, the design basis of the AF system were not correctly translated into plant documents, in that certain design calculations lacked adequate coordination, control, content, or status, and in certain instances the design basis calculations contained errors or were not available to verify that the AF system design basis capability was maintained.

Because of the low safety significance of this issue and because it is in the licensees corrective action program, the issue is being treated as a Non-Cited Violation, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 50-454, 455/2003004-02).

.3 Auxiliary Feedwater Minimum Recirculation Line

Introduction:

The inspectors identified a Green finding of very low safety significance involving not maintaining a commitment to the NRC to have placards on the main control board that provided guidance to operators to ensure the AF pumps had sufficient recirculation flow prior to reducing flow to the steam generators below 100 gpm, such that the pumps remained protected from being operated at shutoff conditions.

Discussion: Based on a review of the AF system and the minimum flow recirculation lines, the inspectors were concerned with the ability of the system and licensee personnel to protect the AF pumps from damage due to deadheading. Although the normal system lineup had normally open valves to both the steam generators and through the recirculation lines to the CST, this would not be the case for all scenarios where the AF system was required to provide a source of water to the steam generators. If the AF suction source automatically switched to the safety-related SX supply, the valves in the recirculation line back to the CST would close and the valves in the recirculation line to the SX would have to open. As discussed in section 1R21.2.b.4 of this report, these valves and their functions were not periodically tested per the inservice testing (IST) program to ensure they would function when called upon. There was also no position indication for air-operated valve AF024 or recirculation flow indication on the main control panel. The emergency operating procedures also did not discuss ensuring there was adequate flow through the AF pumps at all times to prevent damaging the pumps through shutoff conditions. The licensee identified that the alarm response procedure, BAR 1(2)-3-B6, AF Pump Auto Start, did have directions for the operator to verify recirculation flow after a pump start via a computer point and re-verify this condition every hour the pump continued to run. The procedure also directed the operators to monitor AF pump operation locally. The inspectors were concerned that there still existed ample time between re-verifying recirculation flow that the AF pumps could be damaged if the recirculation flow path closed due to a failure of AF024. Since AF system operation was based on controlling steam generator level, some operators stated to prevent overfilling the steam generators they would probably shutoff flow by closing the control valves rather than securing the AF pumps. This response could lead to subsequent pump failure if the recirculation line and its components did not function as required.

A similar concern arose out of a 1986 NRC inspection where valve AF024 could fail under certain scenarios and the recirculation flow path would no longer be available to protect the AF pumps. In a letter to the NRC, dated December 15, 1986, which stated To assure this flow [85 gpm minimum required for safe pump operation] is provided, Byron Station will place permanently affixed labeling at the AFW flow central stations on the main control board panels 1PM04J and 2PM04J and each units remote shutdown panel. These labels will alert the operators of the minimum flow requirements for the AFW pumps and direct them to verify minimum recirculation flow before reducing pump discharge flow below this limit. The proposed labels read as follows: VERIFY $ 85 GPM RECIRC FLOW (A = F2333)/(B = F2334) PRIOR TO THROTTLING FLOW TO STEAM GENERATORS BELOW 100 GPM. However, several years ago during a process of removing unnecessary operator aids from the main control room, these placards were removed without knowing they had been installed as part of a commitment to the NRC to resolve a previous concern. The placards remained installed at the remote shutdown panels. As a result of this issue, the licensee installed new placards on the main control board to provide guidance to the operators based on the previous commitment to the NRC. In addition, several Byron operating procedures will incorporate additional guidance to the operator in identifying that AF flow to the steam generators is being reduced to less than 85 gpm and ensure adequate recirculation flow discharge and/or recirculation valve realignment similar to the changes made at Braidwood resolving a similar 1998 issue identified by the NRC.

Analysis:

Evaluation of this issue concluded that it was a commitment control deficiency resulting in a finding of very low safety significance (Green). The deficiency was due to the licensees failure to maintain a commitment to the NRC to have placards on the main control board that provided guidance to operators to ensure the AF pumps had sufficient forward or recirculation flow to prevent pump damage at shutoff conditions.

This finding was determined to be greater than minor because this lack of guidance could have affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

The finding was assessed through Phase I of the significance determination process.

The inspectors determined there were a number actions or controls that would have to fail to result in pump damage. These included; based on operator training, the operators should know that they need to ensure recirculation flow prior to reducing flow to the steam generators below the 100 gpm requirement to protect the pumps; guidance in alarm response procedure to verify recirculation flow and monitor pumps locally; and the control valves to the steam generators and the recirculation valve to the CST, AF022A/B, were fail open air-operated valves, such that on a total loss of instrument air, there should be an adequate flow path for the AF pumps. As such, there was reasonable assurance that the system would perform its safety function. Therefore, the inspectors concluded that the finding did not represent an actual loss of a safety function and the issue screened out as Green.

Enforcement:

This issue involved the licensee failure to meet a commitment to the NRC in providing the operators with adequate guidance to ensure the AF pumps would not be damaged due operating them at shutoff conditions. Since this was not a regulatory requirement, no violation of regulatory requirements occurred. The licensee entered the event into its corrective action system as CR 159833.

Because of the low safety significance of this issue and because it is in the licensee's corrective action program, the issue is being treated as a Green Finding (FIN 50-454, 455/2003004-03).

.4 Inservice Testing of Auxiliary Feedwater Minimum Recirculation Line Valves

Introduction:

The inspectors identified that the valves in the minimum recirculation lines for the AF pumps were not included in the licensees IST program although the valves appear to perform a safety function in protecting the pumps from being operated at shutoff conditions. This is an unresolved item (URI) pending further review of the facilitys licensing basis with respect to this issue.

Discussion: The valves in the minimum recirculation line perform a safety function to protect the AF pumps from overheating and flow instabilities by providing a minimum flow path either to the CST or the SX discharge pipe when AF flow to the steam generators was throttled to a low or no flow condition. Without an adequate flow path above the minimum required by the pump vendor, damage to the pumps will occur in a short period of time.

The minimum recirculation lines for the pumps were as follows:

• Suction source from CST: The recirculation line tees off the discharge line of each AF pump going through a flow orifice and then normally open (fail open)air-operated valve 1(2)AF-022A/B. The recirculation lines for each train then come together and go through the common non-safety-related check valve 1(2)CD220, before returning to the condensate storage system header, which can either go back to the CST or to the suction of the AF pumps.

• Suction source from SX: The recirculation line tees off the discharge line of each AF pump going through a flow orifice, check valve 1(2)026A/B, and then the recirculation lines for each train come together and go through the common normally closed (fail closed) air-operated valve 1(2)AF-024 (valve opens on the automatic switchover to service water when valves 1(2)AF-022A/B start going closed). The recirculation line then go through the common check valve 1(2)SX194 and goes into the discharge pipe for the SX system.

Based on this design, the following valves perform a function in protecting the safety-related AF pumps, which appear to meet the scoping criteria of the ASME Code,Section XI, which implements OM-10, Section 1.1, Scope.

• Check valves 1(2)026A/B perform an active open function to provide the minimum flow path when the AF pumps take a suction from the SX water system and valves 1(2)AF-022A/B are closed. Testing requirements for exercising the valves to open position would be OM-10, section 4.3.2, Exercising Tests for Check Valves.

• Air-operated valves 1(2)AF-024 perform an active open function to provide the minimum flow path when the AF pumps take a suction from the SX water system and valves 1(2)AF-022A/B are closed. Testing requirements for stroke timing the valves to open position would be OM-10, section 4.2.1, Valve Exercising Testing.

• Check valves 1(2)SX194 perform an active open function to provide the minimum flow path when the AF pumps take a suction from the SX water system and valves 1(2)AF-022A/B are closed. Testing requirements for exercising the valves to open position would be OM-10, Section 4.3.2.

• Air-operated valves 1(2)AF-022A/B perform a passive open function to provide the minimum flow path when the AF pumps take a suction from the CST and an active open function to provide the minimum flow path after the valve is closed upon switchover to SX system and non-safety-related air fails. The loss of air would close valves 1(2)AF-024 and valves 1(2)AF-022A/B would be required to open (fail open on loss of air) to provide a minimum flow path for the AF pumps.

Testing requirements for stroke timing and fail safe testing the valves to open position would be OM-10, Section 4.2.1.

• Check valves 1(2)CD220 perform an active open function to provide the minimum flow path when the AF pumps take a suction from the CST. Since these valves are not in ASME Code class piping, they need not be included in the IST program. However, these valve do perform a function and should be included in a testing program to verify the valves will function as required .

The licensees position on the AF minimum recirculation lines was that they were not part of Byrons design or licensing basis and as such, do not need to be tested. This was based on operators would always maintain forward flow to the steam generators.

They depended on operator training to prevent them from fully shutting the AF flow control valves. However, discussions with some operations personnel indicated that they would shut the control valves rather than secure the pumps, which was the only other way to stop flow if the level in the steam generators exceed the prescribed band.

There was no guidance in the emergency operating procedures as to how the operators should maintain steam generator level. There was a step in the alarm response procedure to verify there was recirculation flow through the computer point in the control room when the alarm comes in and every hour the pumps were operating.

However, if recirculation flow was lost during the one hour time frame, and the flow controls valves were closed, the pumps would fail. One of the actions as a result of the 1986 inspection, as discussed in Section 1R21.2.b.3 of this report, was to make a commitment to the NRC to install placards on the main control board and remote shutdown panel to provide guidance to the operators to maintain adequate flow through the AF pumps. These placards were no longer installed on the main control board as they were removed in an effort to cleanup the clutter on the main control board without recognizing the previous NRC commitment. The placards were reinstalled during the inspection.

The inspectors identified the following references in licensee documents where the recirculation lines are discussed:

• UFSAR 10.4.9.1.1 - These pumps normally take suction from and have a recirculation line back to the condensate storage tank, which are Safety Category II, Quality Group D (not seismic)

• UFSAR 10.4.9.1.2 - The auxiliary feedwater system must be capable of functioning for extended periods...

• UFSAR 10.4.9.2.1 - Delivered capacity (each) 890 gpm exclusive of minimum flow

• UFSAR 10.4.9.3.1 - ... the auxiliary feedwater pumps have a design capability of 990 gpm (including 100 gpm for minimum flow) at 3350 feet net developed head.

• TS Bases B 3.7.5 - The AF System consists of a motor driven AF pump and a diesel driven pump configured into two trains. Each pump provides 100 percent of the required AF capacity to the steam generators, as assumed in the accident analysis. The pumps are equipped with independent recirculation lines to prevent pump operation against a closed system.

• NRC Bulletin No. 88-04, Potential Safety-related Pump Loss, required licensees to review mini-flow lines to ensure flow capacity is adequate to prevent pump damage during operation and testing. The licensee responded to the bulletin, dated July 11, 1988 and February 21, 2989, indicated that the capacity of the mini-flow lines was acceptable to the pump vendor to prevent pump damage and that testing under min-flow conditions has not degraded the pumps or reduced their ability to operate as designed.

Based on some of the statements identified above, it appears that the licensee is taking credit for the recirculation lines and as such, the valves in these lines may need to be tested in accordance with the Code requirements discussed earlier.

Related issues have been raised previously at both Byron and Braidwood, but their resolutions did not address why the valves were not included in the IST program.

The first time this issue was raised was in NRC Inspection Report 454/84040. Starting on page 11 of that report, two issues were identified with valve 1AF024. Both scenarios started with a seismic event that required the AF suction supply to switch over to the SX system from the CST. The first scenario would fail the 1AF024 based on a fire in the 1B AF pump room. The second scenario would fail the 1AF024 due to not performing periodic testing. Unresolved items were opened for each scenario, 454/86040-05 and 454/86040-06, respectively. The first item was closed in NRC Inspection Report 454/87038 based on NRC regulations for fire protection not requiring the licensee to design against a fire and concurrent tornado or seismic event. For this scenario, the resolution was appropriate. The second item was closed in NRC Inspection Report 454/88020 based on the valves mitigating a beyond design basis accident. This appeared to be based on the closure of another unresolved item, 454/86040-02, concerning the DDAF pump SX booster pump (1SX04P), where the NRR determined that the loss of all AC [alternating current] power was a beyond design basis accident. The scenario for the second item, however, was the loss of the non-seismic CST and offsite power due to a tornado/seismic event, which will then automatically start AF. This scenario would not be considered a beyond design basis accident. As such, the initial concern that periodic testing was not performed on the 1AF024 valves was still valid.

The close-out of the second item contained a reference to a letter from L. Olshan to H. Bliss, dated September 15, 1988, which was the NRC Safety Evaluation for approval of the licensees first Ten-Year IST Program. The safety evaluation did not provide any specific mention that it was acceptable to not include the DDAF pump SX booster pump or 1AF024 in the IST program. Safety evaluations for IST programs do not specifically review IST programs to ensure the scope of the program was acceptable. That function has been left to the inspectors reviewing the program at the sites. The function of the safety evaluation was to review the licensees relief request where the licensee has determined for one reason or another that they can not meet the Code requirements and need to conduct alternative testing. As an aside, the DDAF pump SX booster pumps were now included in the licensees IST program.

The Braidwood issue was raised in NRC Inspection Report 456/98201, which noted in Inspection Follow-up Item 456/98-201-04 that the failure of 1AF024 could go unnoticed as there was no indication in the control room for this valve and it failed closed on a loss of air. This item was closed in NRC Inspection Report 456/99013 based on the licensee adding guidance to their procedures to not reduce flow to the steam generators to less than 85 gpm without ensuring adequate recirculation flow. There was no mention or review as to whether the recirculation valves needed to be in the IST program.

Based on the previous issues identified in inspection reports, there was no adequate evaluation of whether these valves in the recirculation lines need to be included in the IST program and the conclusion reached in closing the issue appeared to be in error (not a beyond design basis accident). Based on these observations, it appeared that the valves in these lines may need to be tested in accordance with the Code requirements discussed earlier.

Analysis:

Although the inspectors concluded that the valves perform a safety function that would appear to require their inclusion in the IST program, the licensee had presented information that based on their evaluation, the recirculation lines were not in their licensing basis, such that testing was not required. In addition, related issues concerning the recirculation lines had been previously identified and resolved in NRC inspection reports. Based on this information, this will be considered an unresolved item (URI 50-454, 455/2003004-04) pending further review to determine if the recirculation lines are within the licensing basis with respect to this issue.

Enforcement:

The enforcement aspects of this issue will be determined after the evaluation of the unresolved item.

.3 Components

a. Inspection Scope

The inspectors examined the auxiliary feedwater pumps, valves associated with the minimum recirculation lines, diesel driven auxiliary feedwater pump batteries, and the 125 VDC batteries to ensure that component level attributes were satisfied. The attribute selected for review was component degradation.

Component Degradation: This attribute was verified through review of component repair histories and review of corrective action documents. The inspectors reviewed the attribute to verify the licensee was appropriately maintaining components in the auxiliary feedwater and DC power systems.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

a. Inspection Scope

The inspectors reviewed a sample of auxiliary feedwater and DC power systems problems that were identified by the licensee and entered into the corrective action program. The inspectors reviewed these issues to verify an appropriate threshold for identifying issues and to evaluate the effectiveness of corrective actions related to design issues. In addition, condition reports initiated on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problem into the corrective action system. The specific corrective action documents that were sampled and reviewed by the inspectors are listed in the attachment to this report.

b. Findings

No findings of significance were identified.

4OA6 Meetings

Exit Meeting The inspectors presented the inspection results to Mr. S. Kuczynski, and other members of licensee management at the conclusion of the inspection on May 23, 2003. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Exelon Nuclear, LLC

B. Adams, Engineering

E. Blondin, Design Engineering

J. Drowley, Design Engineering (Corporate)

D. Drawbaugh, NRC Coordinator

D. Flowers, Business Operations Manager

B. Grundmann, Regulatory Assurance Manager

K. Hansing, Nuclear Oversight Manager

B. Jacobs, Engineering

S. Kuczynski, Plant Manager

R. Lopriore, Site Vice President

K. Passmore, System Engineering

B. Perchiazzi, Engineering

R. Randels, Design Engineering Manager

D. Sargent, Engineering

M. Shah, Design Engineering

D. Spitzer, PED Manager

S. Stimac, Operations Manager

Nuclear Regulatory Commission

R. Skokowski, Senior Resident Inspector

P. Snyder, Resident Inspector

A. Stone, Chief, Branch 3, Division of Reactor Projects, RIII

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

50-454, 455/03-04-01 NCV Failure to maintain auxiliary feedwater instrumentation piping water solid (Section 1R21.2)

50-454, 455/03-04-02 NCV Design basis calculations contained errors or did not exist (Section 1R21.2)

50-454, 455/03-04-03 FIN Commitment to have placards on the main control board concerning minimum flow for the auxiliary feedwater pumps not maintained (Section 1R21.2)

50-454, 455/03-04-04 URI Auxiliary feedwater recirculation line valves not in inservice testing program (Section 1R21.2)

Discussed

None.

Attachment