IR 05000456/1998201

Insp Rept 50-456/98-201 on 980316-20,0330-0410 & 0420-24. No Violations Noted.Major Areas Inspected.Team Selected & Reviewed Relevant Portions of Ufsar,Ts,Drawings, Calculations,Mod Packages & Procedures

ML20248L388
Person / Time
Site:	Braidwood
Issue date:	06/01/1998
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20248L376	List: IR 05000456/1998201 ML20248L374
References
50-456-98-201, NUDOCS 9806110120
Download: ML20248L388 (34)

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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Docket No.:

50-456 Ucense No:

NPF-72 Report No.:

50-456/98-201 Licensee:

Commonwealth Edison Company Facility:

Braidwood Nuclear Generating Station, Unit 1 Location:

Braceville, IL Dates:

March 16 - 20,1998, March 30 - April 10,1998, and April 20 - 24,1998 Inspectors:

S. Matur, Team Leader, NRR R. Bradbury, Contractor *

P. Bienick, Contractor *

R. Gauthier, Contractor *

M. Hug, Contractor *

D. Schuler, Contractor *

(* Contractors from Stone & Webster Engineering Corp.)

Approved by:

Donald P. Norkin, Chief Operating Reactor Inspection Support Section inspection Program Branch Division of inspection and Support Programs Office of Nuclear Reactor Regulation

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9906110120 990601 PM ADOCK 05000456

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TABLE OF CONTENTS

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EAg2 EXEC UTIVE S U M MAR Y...................................................... I E 1 Cond uct of Enginee ring................................................... 1 E1.1 Inspection Objectives and Methodology.............................. 1

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E1.2 Auxiliary Feedwater System....................................... 1 i

E1.2.1 Mechanical Design Review.................................. 1 E1.2.2 Electiical Design Review................................... 5 E1.2.3 Instrumentation and Control Design Review..................... 8 E1.2.4 System Walkdown........................................ 9 E1.3 Safety lnjection System....

....................................10 E1.3.1 Mechanical Design Review................................. 10 E1.3.2 Electrical Design Review................................. 14 E1.3.3 Instrumentation and Control Design Review.................... 19 E1.3.4 System Walkd own....................................... 23 E1.4 UFSAR and Design Documentation Review.......................... 23 E1.4.1 Scope of Review........................................ 23 E1.4.2 Inspection Finding s...................................... 23 X1 Exit Meeting............

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APPENDIX B............................................................ B-1 APPENDIX C........

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EXECUTIVE SUMMARY A design inspection of Unit 1 of the Braidwood Nuclear Generating Station (BNGS) was performed by the Special inspection Section of the Office of Nuclear Reactor Regulation (NRR)

from March 2,1998, through April 24,1998, including on-site inspections during March 16-20, March 30-April 10, and April 20-24,1998. The inspection team consisted of a team leader from NRR and five engineers from Stone & Webster Engineering Corporation.

The purpose of the inspection was to evaluate the capability of the selected systems to perform the safety functions required by their design bases, to assess adherence to the design and licensing bases, and to evaluate consistency of the as-built configuration with the updated safety analysis report (UFSAR). The team selected the auxiliary feedwater (AFW) and safety injection (SI) systems including their support systems based on their importance in mitigating design basis accidents at BNGS. The engineering design and configuration control section of Inspection Procedure 93801 was followed for this inspection. The team selected and reviewed relevant portions of the UFSAR, Technical Specifications (TS), drawings, calculations, modification packages, procedures, and other associated plant documents. The team also conducted field walkdowns and observed a simulator training demonstration.

The team identified the following weaknesses in the design and operation of the AFW system:

the potential for exceeding the AFW system discharge piping design pressure when operating on minimum flow and during overspeed testing of the AFW pump diesel engine driver had not been evaluated; the pressure relief cap in the AFW pump diesel engine coolant system had not been tested to demonstrate that it will perform satisfactorily in service; and the operating procedures did not provide instructions to ensure AFW pump minimum flow if the common retum valve on the minimum flow line failed to open when operating with suction from the essential service water system.

The corrective actions for safety-related 4.16kV circuit breakers failures have not been timely.

These circuit breakers had a history of failure starting in 1988 at Byron station. The necessary modifications have not been implemented and the breaker refurbishment has just been started.

There has been no corrective action for an electrical separation design deficiency identified in 1992 in the control circuitry for the non safety-related turbine bearing oil pump. The non safety-related pump is automatically reloaded onto a safety-related bus after it is removed from the bus by a Si signal.

The team identified several components that were not being tested to demonstrate that they will perform their safety function. These components included: a time delay relay (K11) in the starting circuitry of the diesel-driven AFW pump that allows the pump to start and build up F.s lube oil pressure; an emergency diesel generator (EDG) circuit breaker contact function that would prevent load shedding of the safety-related buses on undervoltage when powered by the EDGs; and a check valve in the rsfueling water storage tank (RWST) level instrument tubing.

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pressure relief valve that was installed incorrectly such that the valve could be isolated from the sample cooler it was designed to protect, and a portion of the condensate storage tank level

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transmitter process tubing that was n6t heat traced.

The team identified several test procedure discrepancies. These included: a test procedure for reverse now testing of check valves in the RWST outlet piping that specified acceptance criteria i

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which could be satisfied with the check valves not fully closed; a reactor coolant system isolation valve leakage test procedure that had been inappropriately revised to remove the use of flow-limiting orifices during testing; surveillance procedures that allowed diesel-driven AFW pump i

I battery to be tested at temperatures below the design value and did not specify that the battery should be recharged prior to retum to service; and lack of a procedural requirement to trend and document station battery performance test results.

The rec. commendations in vendor manuals for periodic testing oNafety-related station battery charger circuit breakers and for periodic replacement of the tare um capacitors in the EDG

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control circuit relays were not being performed.

The team identified a number calculation errors and discrepancies. These included: room heat load calculations that did not include all the correct heat inputs; an analysis that was the basis for

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an engineering judgement in a calculation which determined the acceptability of an emergency procedure was not checked or verified; and an AFW pump battery sizing calculation with errors in the input data. However, the conclusions in these calculations were not adversely affected due to these errors. The team also identified several discrepancies in the UFSAR and drawings.

The licensee's evaluation of NRC generic communications had not been thorough in incorporating pump degradation in the analysis of pump-to pump interaction under minimum flow conditions or in identifying and investigating the NBFD relays installed in the plant. During the preparation for this inspection, the licensee identified that the NRC information notice on AFW piping overpresurization due to pump overspeed had not been evaluated.

The licensee issued problem identification forms and initiated corrective actions, as appropriate, for the concems identified by the team. Notwithstanding the open items from this inspection, the team concluded that the AFW and Si systems were capable of performing their safety functions, and the systems generally adhered to the design and licensing bases.

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9 E1 Conduct of Engineering E1.1 Inspection Objectives and Methodology The purpose of the inspection was to evaluate the capability of the selected systems to perform the safety functions required by their design bases, to assess adherence to the design and licensing bases, and to evaluate the consistency of the as-built configuration with the updated final safety analysis report (UFSAR). The team selected, for inspection, the safety injection (SI)

and auxiliary feedwater (AFW) systems and their support systems on the basis of their importance in mitigating design basis accidents at the Braidwood Nuclear Generating Station -

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The inspection was performed in accordance with the soplicable portions of inspection Procedure (IP) g3801, " Safety System Functional Inspection." The engineering design and configuration control section of the procedure was the primary focus of the inspection.

i The open items resulting from this inspection are included in Appendix A and acronyms used in

. this report are listed in Appendix C.

E1.2 Auxiliary Feedwater System E1.2.1 Mechanical Design Review E1.2.1.1 Scope of Review.

The mechanical design review of the AFW system included design and licensing documentation review, system walkdowns, and discussions with the licensee's system and design engineers.

The team reviewed: applicable portions of the UFSAR, technical specifications (TS), flow diagrams, equipment drawings, vendor documents, plant modifications, operating procedures, surveillance procedures, and applicable analyses and calculations. The evaluation included portions of the diesel engine fuel oil supply system and the diesel engine exhaust piping required to support operation of the diesel-driven AFW (DDAFW) pump. In addition, the team observed a i

demonstration of the loss of coolant accident (LOCA) scenario on the plant simulator, and observed the quarterly surveillance tests for both AFW pumps.

E1.2.1.2 Inspection Findings E1.2.1.2(a) AFW System Performance l

UFSAR section 10.4.g.3.1 identified the main feedline break (FLB) as the event which established the design capacity of a single AFW pump, and UFSAR Table 10.4-6 provided a

- summary of the assumptions used for the associated design conditions. During the inspection Table 10.44 was being revised to specify a flow requirement of 450 gpm to three steam generators from an AFW pump during a FLB. The ability of the AFW system to deliver at least 450 gpm was demonstrated in calculation PSA-B-g7-17, " Byron /3raidwood Maximum and Minimum AFW Pump Curve Development," Revision 0. Both Revision 0 and Revision 1 of calculation CN-TA-g7-008, " Byron /Braidwood Feedline Break Analysis," demonstrated acceptable results based on 150 gpm flow to each steam generator. The team concluded that the AFW

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The team reiewed the results of test and performance procedures 1BwVS 0.5-3.AF.1 1, " Unit 1 Motor-Driven Auxiliary FW Pump ASME Quarterly Surveillance," Revision 3, performed on December 19,1997, and 18wVS 0.5-3.AF.12 " Unit 1 Diesel Driven Auxiliary FW Pump ASME Quar 1erly Surveillance," Revision 5, performed on January 2,1998. These tests demonstrated that the pumps met the TS requirements. Although both AFW pumps were demonstrated operable by verifying acceptability at minimum flow conditions, the licensee had determitted that testing at a minimum flow of 85 gpm might not have been sufficient to detect performance differences at higher flow conditions. The licensee determined this condition to be acceptable since the potential reduced performance would be above the pump performance used in calculation PSA-B-97-17, which determined the AFW flow used in the various accident anulyses.

The licensee issued Problem Identification Form (PlF) 457-201-96-109800 to develop a program to observe pump performance trending at full flow conditions as tracked by Nuclear Tracking System (NTS) item number 457 201-96-109802. The team concluded that the AFW pumps were capable of performing their intended function.

The team also reviewed the diesel engine and associated auxiliaries, including the engine exhaust cooling system and fuel oil supply, and determined them to be acceptable except as discussed in section E1.2.1.2(d).

E1.2.1.2(b) AFW Piping Design and Maximum Operating Pressva The AFW system was tested during operation at minimum flow conditions at discharge pressures significantly higher than those at pump rated flow conditions. The team identified the following inconsistencies in calculations related to the piping design pressure and the maximum pressure determined during system tests:

Calculation AF-TH03, " Auxiliary Feedwater Piping Overpressurization," Revision 0,

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determined the highest pressure in the AFW discharge piping as 2138 psig when operating with suction from essential service water (SX) system. This calculation assumed a pump TDH of 4689 ft, which is approximately 2030 psi. This pressure was below the 2167.5 psid pump differential pressure determined during the pump surveillance test on January 2,1998. The team estimated that the maximum piping pressure could be 2396 psig if the suction and discharge pressure gage errors and SX system supply pressure were considered.

Calculation AVFP-01," Verification of Piping Design for AF System," Revision 0, utilized

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inputs from calculation AF-TH03. It also used a 2080 psig piping design pressure w;thout any basis and determined that maximum allowable working pressure for the 6 inch schedule 120 piping as 2367 psig. This pressure was below the maximum pressure that could be reached during operation as discussed above.

Calculation AF-MP-02,"AF System Testing Condition Design Pressure," Revision 0, used

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a pump TDH of 4650 ft or 2013 psi, which was still below the unadjusted TDH determined during the surveillance test.

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Calculation AF-MP-01," Verification of AF System Overpressure Protection," Revision 1, i

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referenced a system design pressure of 1750 psig, a piping analysis pressure of 2080 l

psig, and a maximum pressure of 2038 psig during test conditions. The maximum l

pressure was below the pressure that could exist during operation.

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The team also noted that the acceptance criteria provided in 18wVS 0.5-3.AF.1-1, " Unit 1 Motor-Driven Auxiliary FW Pump ASME Quarter 1y Surveillance," Revision 3, and 18wVS 0.S 3.AF.1-2,

" Unit 1 Diesel Driven Auxiliary FW Pump ASME Quarteriy Surveillance," Revision 5, specified an alert value for pump TDH of 2357 psid that could result in pressures approaching 2550 psig considering SX system supply and maximum instrument error.

Paragraphs NC-3612.2 and ND 3612.3 of Section 111 of the ASME Code allowed the piping stress limits to be exceeded by up to 20 percent during 1 percent of the operating period. This allowable increase was still inadequate to accommodate the estimated pressures. Also, the team noted that the piping contained ANSI 900lb flanges at the flow transmitter locations. These

flanges require special consideration since they are considered comporients and not subject to

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the same stress li nits as the piping.

The licensee st:.ed that on the basis of an informal evaluation the existing flanges and piping

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were capable of withstanding pressures higher than those determined in the above calculations.

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The licensee issued PlFs A1998-01439 and 01482 to address the discrepancies identified above, and proposed placing a restriction dur;ng pump testing such that the flange loading limit would q

not be exceeded. (Inspection Follow-up item 50-456/98-201-01)

The licensee's original evaluation of the NRC Information Notice (IN) 90-45,"Overspeed of the Turbine-Driven Auxiliary Feedwater Pumps and Overpressurization of the Associated Piping

Systems," dated July 6,1990, determined that the conditions described in the IN did not apply i

because only motor or diesel engine pump drivers were provided for AFW pumps. In preparation for this inspection, the licensee reevaluated the IN and determined that it was applicable since the diesel engine driver utilizes a governor control system and is subjected to overspeed testing.

The DDAFW pump discharge pressure could reach 2600 psig at a diesel engine speed of 1925

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rpm which is higher than the rated speed of 1820 rpm. The licensee issued PlFs A1998-00619

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and 00981 to document this issue. NTS item 457-201-98-CAQS0022304 will track actions to formalize the maximum pressure calculation and determine whether future testing should be modified. (sFl 50-456/98-201-02)

E1.2.1.2(c) Diesel Engine for AFW Pump The dinsel engine cooling system is a closed system provided with automatic makeup water from the demineralized water (DW) system. The DW system is connected to the cooling system expansion tank via a non safety-related pressure reducing valve and a safety-related solenoid operated valve (SOV) that opens automatically on low tank level. In the event the SOV failed to reseat tightly, operation of the pmssure relief cap on the tank would be required to prevent overpressurizing and potential failure of the diesel engine cooling system. In December 1997, the SOV had experienced seat leakage which filled the expansion tank, caused the pressure relief cap to open, and spilled water on the AFW equipment. The vendor had recommended that the pressure relief cap be periodically tested for pressure and vacuum capability, as stated on the pressure relief cap drawing attached to calculation AF-SSL-02, " Safety System Limit for TSH-AF 147," Revision O. However, neither the SOV nor the pressure relief cap was included in any periodic test program. The licensee stated that the tank level was checked daily as part of operator rounds, and that these rounds did identify the leaking SOV. The team agreed that the operator rounds would detect a leaking SOV, but noted that this surveillance did not ensure that the pressurr *elief cap would operate as required. The licensee issued Engineering Request

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(ER) 9800389 to evaluate the requirement to test the pressure relief cap. The test program did not include testing of the pressure relief cap to demonstrate that it will perform satisfactory in service. (URI 50-456/98-201-03)

E1.2.1.2(d) Net Positive Suction Head The not positive suction head (NPSH) for the AFW pumps for all operating modes including at the time of switchover from the condensate storage tank (CST) to the SX system as determined in calculation PSA-B 97-14,'" Evaluation of New Technical Specification Levels for Byron and Braidwood Stations," Revision 0, was acceptable.

E1.2.1.2(e) AFW Pump Minimum Flow Operation The AFW pumps are provided with minimum flow retum lines for pump protection during low flow conditions. In the event the CST is unavailable, SX water is supplied to the pumps and a separate line common to both pumps retums SX water to the SX system. This line contains one normally-closed, fail-closed valve 1 AF024 which receives non safety-related instrument air (IA) to open. The team noted that a failure of valve 1AF024 to open could go unnoticed by the operators since there was no valve position indication in the control room. This issue was addressed at Byron in 1980, and permanent labels were installed at the auxiliary flow control stations and at the remote shutdown panel alerting the operators to maintain the AFW pump flow above the required minimum. These changes were not implemented at Braidwood. The licensee issued PlF A1998-01337 to address this concem, and proposed the use of procedural i

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or other steps to direct the operators to ensure a minimum AFW flow path was always available.

(IFl 50-456/98-201-04)

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E1.2.1.2(f) AFW Pump Room Design Temperature l

The team reviewed the capability of the cooler in the DDAFW pump room to limit the maximum room temperature to the design value of 120* F. Calculation L-VA-809, " Review Heat Capacity Verification Test for Diesel Driven Aux Feedwater Pump Rm 1B & 28," Revision 1, estimated the l

cooling load at design conditions and concluded that the maximum room temperature would be l

109.4 * F, based on an expected heat load of about 350,000 BTU /hr. The team reviewed

calculation VA 100, "ESF Pump Cubicle Energy Calculation," Revision 2, and noted that the calculated design heat load for the DDAFW pump room was 644,668 BTU /hr. The team determined that' calculation L-VA-809 was based on the diesel heat load during a test performed

while the DDAFW pump was in the minimum flow recirculation mode which was less than the j

heat load from the diesel engine when the pump was operating at design flow. The licensee's

preliminary evaluation determined that the room temperature would not exceed 120* F with the design heat load and issued an NTS item to PIF A1998-01240 to revise calculation L-VA-809 to incorporate the design heat load and to review other pump room cubicle cooler heat capacity verification calculations for a similar error and to revise these calculations as needed. (IFl 50-456/98-201-05)

The team also reviewed the design temperature for the MDAFW pump area. Calculation VA-101, " Aux Building Energy Load for EL. 330,346,364, and 383," Revision 4, stated the normal design temperature for this area as 122'F; calculation L-VA-430, " Auxiliary Bldg HVAC System

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Minimum Air Flow," Revision 2, identified the normal maximum design temperature of 133'F for the pump area; and calculation VA-102, " Aux. Bldg. Energy Load Celes. for EL. 330'; 346'; 364';

383'; 401' & 426 in Abaormal Condition," Revision 3 determined a temperature of 135'F for the

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post-LOCA condition with reduced auxiliary building ventilation. The auxiliary building area was not a harsh environment, and the licensee stated that the AFW pump motor was qualified to operate in the temperature environment in the area. The licensee issued PlF A1998-01155 to resolve the differences in the normal design temperature calculations.

E1.2.1.2(g) AFW System Modification The team reviewed modification M20-1-90-008, " LOOP Seal / Vent for Auxiliary Feedwater Pumps Suction Line." The inputs, assumptions, and method of analysis used for this modification were appropriate and the modification was consistent with the design and licensing bases. The safety evaluation was adequate.

E1.2.1.3 Conclusions The team concluded that the mechanical design of the AFW system was adequate to ensure supply of high pressure auxiliary feedwater to the steam generators. The AFW system performance information presented in the TS and UFSAR was consistent with the system design and operations documents.

The team identified a potential for exceeding the AFW discharge line design pressures when operating on pump minimum flow, and a potential overpressurization condition in the AFW system during diesel engine overspeed testing. The pressure relief cap on the diesel engine coolant expansion tank was not tested in accordance with the vendor recommendations. The AFW system operating procedure did not address maintaining AFW pump minimum flow when

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operating with suction from the SX system in the event of failure of the retum valve to open. The

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team identified errors in the heat capacity verification calculation for the DDAFW pump room cooler.

E1.2.2 Electrical Design Review E1.2.2.1 Scope of Review

I The team reviewed the ability of the electrical power supplies to the AFW system to enable the l

system to perform its design functions under normal and accident conditions. This evaluation addressed AC electrical bus loading, DDAFW pump starting and DC battery loading, protective device coordination, and modifications.

E1.2.2.2 Inspection Findings E1.2.2.2(a) AFW System Electrical Distribution l

To verify that the appropriate electricalloads had been accounted for in the design basis, the team reviewed the following calculations: 19AQ-68, " Division Specific Degraded Voltage Analysis," Revision 4; BRW-97-0340-E/BYR 97-193, " Battery Sizing for the Byron and Braidwood Diesel Driven Auxiliary Feedwater Pump and the Byron Diesel Driven Essential Service Water Pump," Revision 0; 19-T-6, " Diesel-Generator Loading During LOOP /LOCA Braidwood Units

I 1&2," Revision 2; and 4391/19AN-3, " Protective Relay Settings for 4.16kV ESF Switchgear,"

Revision 14. All major loads were accounted for within the calculations for normal and accident conditions. The AFW pump motor was sized to provide the required brake horsepower for j

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motor. The team also verif;ed that the emergency diesel generators (EDGs) were capable of powering the AFW motors during station blackout and accident conditions without exceeding their ratings.

E1.2.2.2(b). Calculations

The DDAFW pump stading motor is powered by a 24Vdc battery system. The battery system consists of two nickel-cadmium battery banks and two 120/24Vdc chargers. Either'of the two battey banks is capable of providing the necessary power to start the diesel engine via the starting motors. The following discrepancies were identified by the team during the review of calculation BRW-97-0340 E/BYR 97-193, Revision 0:

In Section 3 of the calculation it is stated that the 13 Vdc sizing criteria for the starting

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l motor was based on testing performed on the AFW diesel engine at Byron station. The referenced testing was performed in December 1984. Documentation from the starter motor supplier dated May 22,1990, attached to the calculation stated that a minimum of 20Vdc was required for the starting motor. During the inspection, the licensee contacted the motor supplier and confirmed that the motor would perform its function at a voltage of 13Vdc. The team noted that the effects of the 13Vdc starting voltage on the qualified life of the starting motors and the increased current on the installed cables from the battery to each starter motor had not been evaluated. The licensee stated that these concems

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would be addressed in the calculation revision.

UFSAR Section 8.3.2.1.2 stated that each battery had sufficient capacity to run the diesel

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through four cranking cycles of 5 seconds each before the cranking timer timed out. The team identified that the calculation evaluated 3 cranking > ycles of 5 seconds each and one cranking cycle of 3 seconds. The licensee stated that the calculation would be revised to reflect the four cranking cycles of 5 seconds each as stated in the UFSAR.

The team noted the temperature derating factor of 1.0, which corresponds to a battery

temperature of 77' F used in the calculation. However, the calculation assumed a battery temperature of 65" F which corresponded to a temperature derating factor of 1.05. The licensee stated that the calculation would be updated to reflect the correct temperature derating factor.

The team noted that incorrect capacity rating factors were used in several places in the

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The licensee initiated PlF A1998-01066 to correct the discrepancies in the calculation. The team determined that the battery could perform its safety function after accounting for the above discrepancies in the battery sizing calculation. (IFl 50-456/98-201-06)

E1.2.2.2(c) Testing and Surveillance Procedures During the review of calculation BRW-97-0340-E/BYR 97193, the team noted that the design of the AFW battery was based upon an ambient temperature of 65' F. However, the team identified that the operator rounds sheet indicated that a room temperature of 60" F was specified as acceptable. Procedures 1BwOS 7.1.2.1.a.1-2, Diesel Driven Auxiliary FW PMP Monthly Surveillance " and 1BwoS DC W4, "24 VDC Aux Feed PP 1B Batt Bank A & B Weekly Surveillance," permitted surveillance to be performed with the room temperature below the

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design requirement of 65* F. The licensee initiated PlF A1998-01043 to resolve this issue and issued instructions that the operator rounds sheets be changed to reflect the design basis battery ambient temperature of 65" F.

The team reviewed procedures 1BwVS 7.1.2.3.C-1, " Auxiliary Feedwater Diesel Prime Mover Performance Surveillance," Revision OE1, and 18wOS 7.1.2.1.a.1-2, " Diesel Driven Auxiliary Feedwater Pump Monthly Surveillance," Revision 2, and noted that the batteries were utilized to demonstrate that the DDAFW pump starting motor would start under emergency conditions. The procedures did not require that the batteries be recharged prior to retum to service. The licensee initiated PlF A1998-01263 to address the battery depletion during testing and. maintenance operations and to revise the surveillance procedures. (IFl 50-456/96-201-07)

E1.2.2.2(d) Modification Review The team reviewed Exempt Change E20-1-96 282-010, " Replacement of Westinghouse Type AR-3 relays with Westinghouse DC Contactors in Breaker Closing Circuits for the Auxiliary Feedwater Pump 1 AF01PA," and the associated safety evaluation BRW-SE-1998-104 dated January 26,1998. This modification was initiated because safety-related 4.16kV DHP circuit

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l breakers had failed to close on several occasions at the Byron station and the AR-3 relay had also failed. The containment spray (CS) pump motor circuit breaker 2A at Byron failed to close on three separate occasions in 1988. Based upon discussions with the vendor, the licensee replaced the AR-3 relay with DC contactors in the closing circuits of all safety-related 4.16kV DHP circuit breakers at Byron. However, in December 1995, the modified CS pump 2A circuit

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breaker at Byron Power Station failed to close.

The licensee stated that the AR-3 relays in the safety-related 4.16kV DHP bre akers at Braidwood had not been replaced, that modification requests to do so were initiated in 1996, and that modifications for some safety-related 4.16kV DHP breakers have been issued. The actual cause of the failure of the breakers to close had not been established. The licensee has initiated a

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refurbishment program for the breakers and had refurbished 4 of the 79 safety-related 4.16kV DHP breakers as of the time of the inspection. About 27 breakers were scheduled for refurbishment in 1998. A completion date for the ESF breaker refurbishment has not been identified. Conditions adverse to quality of these breakers had not been promptly corrected and the causes of the breaker failures have not been determined (Unresolved item 50-456/201-98-08)

The team reviewed Exempt Change E20-193-293 001, " Addition of Cooling Fans to Unit Substation 132X,134X, and 134Y," Revision 1, and noted that drawing 20E-0-4001, " Station One Line Diagram," Revision V, had not been updated to reflect the increase in transformer rating resulting from the modification. The licensee initiated PlF A1998-01207 to evaluate this concem and initiated field change request (FCR) 980042 to revise the drawing.

E1.2.2.3 Conclusions The electrical design for components that performed the normal and accident functions of the AFW system supported the design basis functions of the system. The electrical system provided independent, redundant, safety-related power to the electrical AFW loads. The team identified a lack of timeliness in corrective actions for 4.16kV circuit breaker failures and discrepancies in several test procedures and calculations.

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L E1.2.3 instrumentation and Control Design Review E1.2.3.1 Scope of Review The team evaluated the ability of the instrumentation and controls for the AFW system to support the system in the performance of its safety functions. The team reviewed sections of the UFSAR, applicable TS,9ow diagrams, uncertainty calculations, and calibration data. System walkdowns were also caducted. The review included the instrumentation for the automatic switchover of the AFW pump suction from the CST to the SX system.

E1.2.3.2 Inspection Findings E1.2.3.2(a) Diesel-Driven Auxiliary Feedwater Pump Controls-The team reviewed the DDAFW pump control logic to verify that the time delay relays were -

adequately tested. The logic contained appropriate startup and shutdown interlocks including set points. The DDAFW pump startup panel circuitry shown on drawing 20E-1-4030AF12,

'" Schematic Diagram Auxiliary Feedwater Pump 1B (Diesel-Driven) Engine Startup Panel'

1 AF01J," Revision W, was reviewed for adequacy. The time delay relay K11, which provides a 10-second time delay of the low lube oil pressure trip function to allow the engine to start and establish oil pressure, was not adequately tested or periodically calibrated. One contact from the K11 relay is used to defeat the 10 psig low oil pressure trip and the 10 psig low oil pressure alarm to allow the engine to start following an engine' start signal. The licensee stated that the relay was originally screened as an " alarm only" relay when all safety-related time delay relays were screened for testing applicability during the review required by NRC Generic Letter (GL) 96-01, Testing of Safety-Related Logic Circuits," dated January 10,1996.

The licensee determined that lack of calibration and performance tracking of the K11 relay was not an operability concem since the relay was functionally tested during the monthly diesel run and if the time delay relay were set significantly shorter than the required 10 seconds the diesel would not start. The licensee also stated that the only K11 relay function that was not tested either implicitly or explicitly was the opening of contact 3-5 when the relay de-energizes. If this contact failed to open, the low oil pressure trip would always be defeated when the DDAFW

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pump is running; but the pump would still start and perform its safety function. The licensee also stated that the low oil pressure alarm would still alert the operator to a low oil pressure condition

- and that the monthly start also confirmed that the actual time delay adequately defeated the low oil pressure trip to allow a successful start. The team was concemed that if the time delay relay drifted outside of the required 10-second time delay setting due to the lack of calibration and performance trending, the DDAFW pump might not start when required. The test procedures did not include the testing of K11 relay and its contact 3-5 to demonstrate that they will perform satisfactorily. The licensee issued PIF A1998-01073 to include proper testing of the time delay relay in the appropriate surveillance procedure. (URI 50-456/201-98-09)

E1.2.3.2(b) Air Accumulators for AFW Pump Discharge Valves The AFW pump discharge valves AFWOO4A and B are normally-open, air-operated valves that are closed only during surveillance testing of the AFW pumps. Although the valves are normally open, they also receive a Si signal to open. On a loss of instrument air, a source of backup air is available to the valve operators from accumulators installed locally. The ability to open valves AFWOO4A and B using air from the accumulators in the event of a loss of instrument air was not

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periodically tested. The team also observed that there was no maintenance procedural requirement to periodically blowdown the tanks to remove the collected moisture in the accumulators. The licensee stated that the instrument air system has a dowpoint of -40 * F which precluded water from condensing in the accumulators. The dowpoint was verified quarterly. The licensee initiated ER 9800838 to evaluate additional maintenance and testing activities for the backup air supply to the valves.

E1.2.3.2(c) Modifications The team reviewed setpoint/ scaling change requests SSCR 89-206, SSCR-89-237, and SSCR-95-027, and modifications E20-1-93-238 and M20-1-94-003, and determined that these were consistent with the design basis and that the associated safety evaluations were acceptable.

E1.2.3.3 Conclusions The team concluded that the design of instrumentation and controls for the AFW system was adequate to support the safety functions of the system and was consistent with the licensing bases.

Time delay relay K11 in the starting circuitry of the DDAFW pump was not included in the relay testing procedures. Failure of this relay to operate as designed could prevent startup of the pump. The function of an air accumulator which provided back-up air to open valves AFWOO4A/B on a loss of instrument air was not tested.

E1.2.4 System Walkdown The team performed several walkdowns of the AFW system to verify that the system configuration was consistent with the design basis and the UFSAR. The system configuration was generally consistent with the AFW and Main Feedwater flow diagrams. The team identified the following discrepancies :

The sensing elements for the Jacket water temperature switches and indicator were

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installed to measure the outlet and not the inlet temperature as shown on drawing 62241,

"6V149T1 Auxiliary Feedpump Drive Engine Cooling System," Revision B. The installation is the preferred arrangement, as it would readily detect potential overheating of the Jacket water. The licensee issued PIF A1998-01329 to correct the drawing error.

A seismically supported light fixture located in front of cubicle 7 of Division 12 engineerir:g

safety features (ESF) switchgear was tied with a wire to hangar CC506. No maintenance or temporary tag was attached to the installation. The licensee initiated PlF A1998-01312 to correct this discrepancy.

A non safety related cable was in contact with safety-related conduit C1 A5116 in the

Division 12 ESF switchgear room. The licensee initiated PlF A1998-01314 to address this condition and to provide directions for separating safety-related conduit from non safety-related cable.

The CST level transmitter 1LT-CD051 is located in an und* jround room next to the

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CST. A portion of the 1/4 inch process tubing between the sevel transmitter and the CST outlet pipe was not adequately insulated or heat traced. Drawing M-39, Sheet 1,

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• Diagram of Condensate (Make-up and Overflow) Unit 1," Revision AR, showed heat tracing along the entire length of the tubing. The team was concemed that water in this section of tubing could freeze and adversely affect the CST level indication. The licensee could not determine the reason for the lack of heat tracing. The licensee stated that other instruments, such as the pump suction pressure, could be used to compute the CST water level. Although the CST and its level instrumentation are non safety-related, the CST level indicated in the control room is monitored in accordance with TS 3.7.1.3. (IFl 50-456/201-98-10)

The team noted an unauthorized metal bar and plate that were attached to the conduit

containing electric cable for 1LT-CD051 and rusted mounting bolts on the instrument support column. The licensee issued AR 980023447 to correct these deficiencies.

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E1.3 Safety injection System E1.3.1 Mechanical Design Review E1.3.1.1 Scope of Review The mechanical design review of the Si system included design and licensing documentation review, system walkdowns, and discussions with the licensee's system and design engineers. The team reviewed: applicable portions of the UFSAR, technical specifications (TS), flow diagrams, equipment drawings, manufacturers' information, plant modifications, operating procedures, surveillance procedures, and applicable analyses and calculations. In addition, the team observed a LOCA scenario demonstration on the simulator.

E1.3.1.2 Inspection Findings E1.3.1.2(a) SI System Performance The team reviewed the available licensing, design, and operations documents related to the capability of the Si system to provide adequate emergency core cooling flow under accident conditions. The team also reviewed the arrangement of the two SI pumps and associated valves and their power supplies and verified that the Si system safety functions would be performed in case of a single active failure of a component or power supply.

The team reviewed calculation RSA-B-93-04, " Byron /Braidwood ECCS Flow Verification," Revision 0, that determined ECCS flows for use in the LOCA analysis, and found it acceptable. These flow data were incorporated for use in the LOCA accident analysis in RSA-B-94-09, " Byron /Braidwood ECCS Flow Calculation for LOCA Analysis," Revision 1. The team also reviewed the surveillance test procedures for the SI, RHR, and charging pumps and procedures for check valve stroke tests and verified that the acceptance criteria were consistent with calculation RSA-B-93-04. The team also reviewed the results of the latest performance of these tests and verified that the acceptance criteria were met.

E1.3.1.2(b) Not Positive Suction Head The team reviewed calculation MAD 88-169, " Fuel Pool Storage & NPSH," Revision 0, which

. determined the available NPSH for the ECCS pumps in the injection mode using the RWST as a water source. The calculation assumed elevation (EL) 404.83 ft as the 0% RWST elevation,

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instead of EL 401.667 ft (which is the top elevation of the suction pipe) as determined in calculation SETH-1, " Refueling Water Storage Tank (RWST) Level Setpoints," Revision 4. However, the conclusions of the calculation were still valid and adequate NPSH was available to the ECCS pumps because of the available margins in NPSH. During the preparation for this inspection, the licensee had ident%d otiser discrepancies in this calculation and had issued PlF A1998-00892 to correct them.

The team reviewed calculation CS-5, "NPSH for RHR & CS Pumps," Revision 3, which verified that the available NPSH tor the RHR pumps was adequate during the recirculation mode and concluded that suffic%nt margin was available for the RHR pumps during recirculation.

E1.3.1.2(c) SI System Valve Operation and Testing The team reviewed the applicable TS requirements, surveillance test procedures, and results of the last test of the SI system valves. The team determined that the Si system valves functioned in accordance with the design and operation requirements. With the exception of the following items, the twsting of the Si system valves was sufficient to verify that the valves were capable of performing the required functions under accident conditions.

Information Notice (IN) 91-56, " Potential Rsidioactive Leakage to Tanks Vented to Atmosphere,"

identified potential problems resulting from leakage of post-LOCA recirculating containment sump water through isolation valves to the RWST and out the tank vent, thus contributing to off-site and control room dose. The licensee's evaluation of this IN, documented as NTS item 456-103-91-05600, concluded that this problem did not exist at Braidwood since the RWST was vaated to the auxiliary bui: ding filtered ventilation system. The valves preventing leakage to the RWST are the CV, RHR, and SI pump suction check valves CV8546, RH8958A and B, and Sl8926 respectively, if the associated motor-operatert valves were open because of the valve operating sequence in procedures or a single failure to close, these check valves would have to prevent back flow to the RWST. The Si m'.

mum flow isolation valves, Sl8814 and SlS920, also p* vent leakage of containment sump water to the RWST. These valves were tested for closure but were not leakage tested.

The need to test closure of check valves was identified in September 1996, as documented in PlF 456-201 96-1953. The team reviewed the closure tests performed for CV8546 and Sl8926 in October 1997 in accordance with procedure BwVS 0.5-2.SI.2-3, " Safety inj6 tion System Check Valve Stroke Test," Revision 9. Each check valve closure was tested by pmssurizing the

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appropriate pump (CV or SI) suction piping with RHR pump flow and verifying that the pressure in l

the suction piping was above a certain acceptance criterio" The team determined that the closure

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test acceptance criterion could be met even if the check valve was not fully closed. The licensee agreed with the team, and initiated PlF A1998-01475 to determine the actions necessary to properly test closure of these valves. The licensee determined the check valves were operable on the basis of previous valve testing, inspection, and valve exercising. Testing of closure of check valves CV8546 and Sl8922 did not demonstrate that they will perform satisfactorily in service. (URI 50-456/201-98-11)

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The licensee refer *ed to page E.77-3 of Revision 6 of the UFSAR where it was stated that the RWST and associated piping were excluded from the leak reduction program for systems thet could

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pots.itiaily contain radioactive fluids after a LOCA and stated that the SER supplement 1 issued by j

the NRC confirmed such exclusion on page 9-5. Therefore, leakage testing of the check valves and

- minimum flow isolation valves between the RWST and the ECCS pumps was not required. The j

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t team believed that the NRC SER allowed exclusion of RWST and its associated piping from the leak reduction program because the UFSAR stated that highly contaminated water is prevented from entering the RWST. This issue of leakage testing has been referred to the NRC staff for further review. (IFl 50-456/98-201-12)

Engineering Request (ER) 9501896,.,oncluded that leakage measurement of reactor coolant system isolation valves by opening valves 1/2RH028A/B,1/2Sl044,1/2S1053,1S1052, or 2S1050 in the vent / drain lines and collecting the leakage in a graduated cylinder, must use a trmporary 0/8 inch inside diameter orifice in the line to restrici the loss of reactor coolant in order for the evolution to be bounded by the existing UFSAR design basis analysis. The current procedure BwVS 4.6.2.2-1,

" Reactor Coolant System Pressure Isolation Valve Leakage Surveillance," Revision 14, did not require orifices to be installed when using valves 1/2RH028A/B for leakage measurements. The licensee stated that Revision 9 of the procedure added steps to install orifices on all the vent / drain valves used for leakage measurement and the steps to install the orifices for the RH028 valves were removed in Revision 12. No justification for removal of the steps was available. The licensee determined that past tests without the orifice did not present a condition outside the design basis and initiated PlF A1998-01231, which proposed that the revision history of BwVS 4.6.2.2-1 be further researched to determine why the requirement for the orifice was removed, to evaluate the inclusion of the requirement, and to revise procedure BwVS 4.6.2.2-1. (IFl 50-456/98-201-13)

E1.3.1.2(d) St Accumulator Filling, Draining, Venting, and Pressurizing

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During a self-assessment in preparation for this inspection, the licensee had initiated PlF A1998-00670, "Si Accumulator piping safety / seismic classification," dated February 19,1998, to identify that the remotely operated valves Sl8878A-D, Sl8877A D, Sl8875A-D, and S10943 used to fill, drain, vert, and pressurize the Si accumulators were not in the IST program and could not be relied upon to close in the event of a LOCA during these activities. These valves also fomi a boundary between Category IB and Category llD piping. The resolution of the PlF was that the equipment was l

currently operable and that no chuges to the procedures were needed since the procedures allowed adjustment of the water or nitrogen inventory in only one accumulator at a time and a best estimate evaluation had determined that the peak clad temperature (PCT) would be less than 2200 F if only two accumulators were credited. The corrective actions for the PlF were to determine if a LCO should be eniered during the above evolutions, determine the appropriate testing requirements for the valves involved, and sample other systems to determine if other similar conditions existed.

l During the inspection, the licensee performed an enluation of the scenarios where failure of the subject valves to close during a LOCA could adverself affect the results of the accident analyses.

This evaluation concluded that the loss of accumulator water inventory through a failed valve would i

result in an increase in PCT of less than 25 F and the loss of nitrogen pressure through a failed I

valve would result in a PCT increase of about 90* F. The licensee initiated NTS item 456-201-98-CAQS-0046405 to revise the accumulator venting procedure to ensure that one of the two vent valves in series would be closed to preclude an open vent path during venting.

I E1.3.1.2(e) St Pump operation at Minimum Flow

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in letters to the NRC dated February 21,1989, and June 30,1989, the licensee responded to NRC Bulletin 88-04 and stated that the safety injection, charging (high head safety injection), and auxiliary feedwater pumps were not susceptible to the potential for dead heading due to pump-to pump interaction. However, calculation CWBS-C-149, dated June 16,1988, which calculated weaker SI

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I interaction. However, calculation CWBS-C-149, dated June 16,1988, which calculated weaker Si and charging pump flows did not take into consideration the pump degradation allowed by Section XI of the ASME Code as required by the NRC bulletin.. The licensee issued PlF A1998-1259 to re-perform the calculation. With regard to the AFW pumps, the licensee stated that thwe were no scenarios where both AFW pumps are run aligned to minimum flow. However, the team pointed out that emergency proceciure 18wEP-O allowed operation of both AFW pumps at minimum flow to control steam generator level. The licensee issued PIF A1998-01395 to include the AFW and charging pumps in the minimum pump flow evaluations. Because the latest test results for the put.ips showed that the performances of both pumps in each system were close to each other, the j

team did not have imrnediate operability concems. (IFl 50-456/98-201-15)

i E1.3.1.2(f) Interfacing Systems Review The SI system interfaces with the heating, ventilating, and air conditioning (HVAC), IA, component cooling water (CCW), and equipment and floor drainage systems. With the exception of the i

following items, the team found that the interfaces supported the design basis of the SI system.

- The team reviewed the capability of the coolers in the SI pump rooms to limit the room temperature to the design value of 122 F. Calculation L-VA-811," Heat Capacity Verification Safety injection Room 1 A/B & 2A/B," Revision 1, determined the est; mated cooling load at design conditions with the pump motor heat load based upon a pump recirculation flow test and concluded that the maximum temperature of the SI pump 1 A room would be 106.6 * F. The team noted that the calculation did not consider the hot piping in the room during post-LOCA recirculation and the additional motor heat load when operating at design flow compared to the minimum recirculation flow condition during the test. The licensee performed a preliminary evaluation which determined that the design temperature would be acceptable with the additional heat load and issued PlF A1998-01240 to revise the calculation as necessary. This is another example of problems with room heat lund calculations. (IFl 50-456/98-20105)

Calculation L-VA-811 also stated that the test data for the 1B Si pump room was in error and did not estimate the maximum room temperature and recommended a retest. The licensee stated that capacity verification had been demonstrated through tl'e te: ting program developed after issuance of GL 89-13, * Service Water System Problems Affectir'g Gefety-Related Equipment," and that calculation L-VA-811 would be revised to disc.ist coolet capacity verification for the 1B SI pump room.

UFSAR Sections 9.2.2.2.2.2 and 9.2.2.4.1 stated that relief valves were installed in the CCW system to relieve the volumetric expansion which would occur in the cooled components if the CCW were isolated and heat added to the isolated portion of the CCW system. The relief valve provided

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for the non safety-related sample cooler panel 1PS29) was incorrectly installed so that closure of I

the CCW supply and return isolation valves CC9477A and CC9448B for the cooler panel would

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have isolated the cooler from the relief valve. This arrangement was shown on UFSAR Figure 9.2-3. The licensee issued PIF A1998-01215 and the proposed solution included placing caution

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cards on the cooler inlet r'id outlet isolation valves to ensure that the outlet isolation valve is verified I

to be open when the inlet valve is closed unless the process fluid side is isolated, and determining corrective actions for complying with ANSI B31.1 code for the non safety-related piping. (IFl l

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E1.3.1.2(g) SI System Modifications

- The team reviewed five modifications for the Si system. These modifications were: DCP 9700096-(

. E-20-2-97-225, that enhanced the containment sump screen filtering capability; DCP-9700322-E20-

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0-97-298, that modified the containment sump screen support structure; DCP 8800434-M20-1-88-

' 085, that added a support clamp to the accumulator fill line to reduce vibration during filling; CP 9000183 P-20-90-021, that replaced Kerotest diaphragm type valves with Anchor Darling valves; and DCP-95-00174-E-20-1-95-237 which changed the gear ratio in the Limitorque operators for valves Sl8801 A and Sl88018.

The team identified no concems with the engineering analyses and safety evaluations for these

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modifications.

E1.3.1.3 Conclusions The mechanical design of the Si system was adequate to provide emergency core cooling fiow under accident conditions. The SI system performance information presented in the UFSAR was consistent with the system design and operations documents.

The team identified an inadequate procedure for lesting closure of check valves. The reactor j

coolant system isolation valve leakage surveillance valve leakage procedure had been '

<j inappropriately revised to delete use of flow limiting orifices during testing. The team also identified a CCW relief valve for a cooler panel that was installed cutside the cooler isolation valves and discrepancies in Si pump room cooler capacity verification calculation. The licensee's response to NRC Bulletin 88-04 did not consider the effect of ASME Code allowed pump degradation on pump-to-pump ir.teraction during minimum flow conditions.

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E1.3.2 Electrical Design Review E1.3.2.1 Scope of Review The team reviewed the ability of the electrical power supplies to the Si and interfacing systems to enable them to perform their design functions under accident conditions. This evaluation addressed electrical AC bus loading, DC battery loading and distribution, protective coordination, EDG loading, and plant modifications.

E1.3.2.2 Inspection Findings E1.3.2.2(a) SI System Electrical Distribution The team reviewed calculations for EDG loading, degraded voltage, protective relay settings for

- 4.16kV ESF switchgear, short circuit analyses, and sizing the 125Vdc safety related batteries and battery chargers. The team verified that all major Si system electrical loads were accounted for within these calculations. The team verified that each EDG had sufficient capacity to supply power

. to the minimum engineered safeguards equipment during an accident. The team also reviewed breaker protective coordination for the Si pump motors and determined that the niotors were property protected.

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The team identified the following discrepancies in the reviewed calculations:

Calculation 19-D-21 " Calc. for Sizing the 125Vdc Safety-related Batteries," determined the

battery loading from the safety-related DC loads for a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> battery duty cycle. The calculation did not include the breaker clo. sing loads for those breakers which were required to be sequenced onto the ESF buses during a LOOP /LOCA condition. The licensse demonstrated to the team that the batteries had sufficient capacity to support breaker closure, although the calculation did not include the brosker closing loads. The licensee initiated FIF A1998-01493 to address these discrepancies.

The Heated Junction Thermocouple Microprocessor Chassis (HJTC) was tested to show

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that it would operate at voltages less than 115 + 10% V AC. The testing established that the system was capable of operating at a voltage down to 90V AC at 72' F ambient for a maximum of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Calculation 19-AQ-69," Evaluation of the Adequacy of the 120VAC Dist. Circuits Volt. at Degraded Voltege Setpoint," Revision 9, determined that the operating voltage for the HJTC could be 96Vac, but did not demonstrate that the environmerital conditions were within the test conditions at 90Vac. The licensee reviewed calculation 19AQ-69 and determined that the operating voltage for the HJTC would be at least 103.5V AC, which was included in the manufacturer's tests. The licensee issued PlF A1998-01495 to document this discrepancy and revise the calculation.

The team questioned the assumption of negligible resistance contribution from switches,

breakers, and relays used in calculation 19-AQ-72, " Calc. for 125 Vdc Voltage Drop Calculation," Revision 6. The licensee provided testing data from Dresden station for AC components which showed that these resistances were negligible. The team also noted that

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TS Section 4.8.2.1.2c provided a limit of 150 x 10 ohms for each battery cell-to-cell resistance and that these resistances were not included in the calculation. The battery surveillance and manufacturer's testing have shown the cell-to-cell resistances to be much lower than the TS limit; however, the team was concemed that additional voltage drop may occur during a LOOP /LOCA and sufficient battery voltage may not be cvailable to operate the required loads. Because the existing batteries will be replaced during the next outage, this concem should be addressed in the modificat;on.

Calculation BRW-97-0473-E/BYR97-225, " Circuit Breaker Trip Settings - 125 VDC and 250

VDC Distribution Centers," Revision 6, evaluated breaker frame sizes 150A and 250A; however, the drawings for these breakers,20E-1-4010G, "125V DC Distribution Center Bus i

111 & 112 (1DC05E & 1DC06E) 250V DC Non-Safety-Related MCC 123 (1DC09E),"

i Revision B, and 20E-1-4010H, "125V DC ESF Distribution Panel 111 (1DC05EA)," Revision A, showed the breakers to be frame sizes 100A and 225A. The licensee evaluated the breaker curves for the installed devices, determined that they would operate within the engineering design requirements, and issued PlF A1998-01472 to resolve the discrepancy,

E1.3.2.2(b) EveNaNs of Generic Problem Notifications Illinois Power Company had issped a 10 CFR Part 21 notification regarding the use or resistances

- for untinned copper conductors instead of resistances for tinned copper conductors in calculations.

Untinned copper has lower resistance values, and therefore, would be less conservative when used in calculations for power arni ontrol circuits containing tinned copper conductors. The licensee identi*ied that this issue appiied to its electrical calculation s, and issued PlF A1998-00304 on January 20,1998. The licensee evaluated this issue and determined that the electrical calculations

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were sufficiently conservative to assure that the additional increase in resistance would have i

minimal effect on the results. The team agreed with this evaluation. The corrective actions included the identif; cation of the population of calculations affected and establishing an action plan to make the necessary revisions.

Information Notice 91-45,"Possible Malfunction of Westinghouse ARD, BFD, and NSFD Relays, and A200 DC and DPC 250 Magnetic Contactors," dated July 5,1991, identified that these relays might malfunction due to an epoxy compound used in coils becoming semi-fluid when the coil was energized for extended periods. The licensee evaluated this notice and concluded that no NBFD relays were installed at Braidwood. In response to the team's questions, the licensee determined that NBFD relays were installed as "Y" relays on all medium voltage switchgear. The licensee issued PIF A1998-01483 to document and investigate this discrepancy. (IFl 50-456/98-201-17)

E1.3.2.2(c) luodifications The team reviewed 12 electrical modifications. Except for the items discussed below, the modifications were consistent with tha design basis and the associated safety evaluations were generally well documented. However, the team noted that the following safety evaluations were not thorough:

The safety eveluation dated February 18,1994, for modification M20-1-92-007 did not provide the basis for the reduction of the battery duty cycle for the safety-related station batteries from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The licensee stated that the reduction in duty cycle had been understood and approved by the NRC as demonstrated in the SER for TS amendment 5S/47. The taani noted that the SER only discussed the relocation of the information on the design duty cycle from the TS to the UFSAR, but did not explicitly approve the design basis duty cycle change from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The licensee stated that this safety evaluation would be reviewed and the supporting bases provided for the change, in the design duty cycle.

The team reviewed safety evaluation BRW-SE-1997-1655 for modification M20-1-96-001, "ESF 125 VDC Battery and Rack Replacement Modification," scheduled for installation during the 1998 refueling outage. The safety evaluation stated that the cross-tie capability between Units 1 and 2 was to be increased from 100 amps to 200 amps, and that the batteries would be capable of supplying the additionalload. Section 7.5.1 of IEEE 946-1992,"iEEE Recommended Practice for the Design of DC Auxiliary Power Systems for Generating Stations," required that for designs in which the battery is equalized while connected to the load and includes an :nverter with normal ac power supply (transformer / rectifier unit), the ac supply should be designed so that it supplies the inverter while the charger is at equalize voltage. The licensee stated that the current mode of operation did not equalize the station batteries. The licensee further stated that the battery charger analysis for the planned station battery replacement would be re-evaluated tc determine the impact of the 200 amp cross-tie load and the performance of the inverter when equalizing the batteries.

The licensec issued PlF A1998-01469 to address these concems.

E1.3.2.2(d) Electrical Separation Note (t) on UFSAR Table 8.3-5 states that the turbine bearing oil pump is powered from the class 1E 480-V switchgear, however, it automatically trips on a safety injection sinal concurrent with a loss of offsite power. The team noted that, upon receipt of a Si signal, the motor is stripped from l

the bus, but was designed to be automatically loaded back onto the bus after the load shedding had

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l Electrical Distribution Safety Functional Inspection (EDSFI), and initiated NTS item 925-20-92-00500 on September 14,1992, to track modification to the control circ Jit. This design did not meet the intent of the UFSAR commitments relative to electrical separation requireme.ts in Regulatory Guide 1.72, Revision 2, since non safety-related loads are stripped from safety-related buses to achieve electrical separation. Automatically reloading the motor on to the bus would present a condition where a single failure of a circuit breaker to operate could allow a fault in the non safety-related motor to propagate to safety-related equipment. In 1996, the licensee issued PlF 456 201-96-2708 to document this condition and performed an operability assessment. Exempt change E20-1-97-

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277 was issaed on April 7,1998, to revise the turbine bearing oil pump trip control circuitry to prevent automatic reloading on the bus after the pump was stripped. The modificat:on was

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scheduled to be installed in the next refueling outage. A design deficiency in complying with electrical separation requirements had not been promptly corrected. (URI 50 456/98-201-18)

E1.3.2.2(e) Uninterruptible Power Supply (UPS)

The 120V DC inverter provides power for the 120V AC instrumentation in each ESF division.

Redundancy within the inverter aquipment is established by a normal 480V AC input feed from a

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Claes 1E motor control center (MCC) and a reserve 125V DC feed from the 125V DC distribution

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bus in the same ESF division. The normal AC power supply for the inverter is transformed from 480V AC to approximately 135V DC, which is a higher voltage than the DC reserve feed. An auctioneering circuit allows the DC supply to power the inverter when its voltage exceeds the input from the AC bus. The 'nput transformer of the inverter was set on the 483V AC tap. The battery charger sizing calculawns did not include the ir.verter loads during normal or accident conditions.

The team determined that the inverter could be powered by DC under certain conditions anr lot by the normal AC source as designed. The licensee performed a review of daily operator rounds sheets and determined that the inverters were be ing powered from the normal power AC source.

The licenses could not confirm, however, by using the analytical model, that the inverters would be powered from the AC power supply under all conditions. The team was also concemed that the output voltage of the charger could be high enough to become the primary supply to the 120V DC inverters. No engineering limits were placed on the charger output voltage settings to prevent this from occurring. The licensee issued PlFs A1998-01469 and PlF A1998-01468 to address these concems. (IFl 50-456/98-201-19)

E1.3.2.2(f) Testing and Surveillance Procedures UFSAR Section 8.3.1.2 states that the electrical system is designed to prevent automatic load shedding of the emergency power buses once the onsite sources are supplying power to all sequenced losds on the buses after a loss of power. The load shed interlock feature uses the "b" contact of the respective diesel generator breaker. This interlock defeats the load shedding feature while the loads are being fed from the onsite source. The load shed feature is reinstated when the diesel generator l'reaker is open and the loads are fed from the offsite source. The function of the

"b" contact was not tested. The licensee stated that this potential weakness in the testing program had been identified in 1993 and again as part of the GL 96-01 review, and would be addressed during that review. (IFl 50-456/98-201-20)

The vendor manual for the safety-related 125V DC battery chargers. L-0520, " Instruction Manual, Three Phase Thyristor Controlled," dated June 29,1979, stated in Section 6.6 that "As a result of qualification to IEEE-323, the following maintenance program must be adhered to:... circuit breakers must be operated at least every 6 months. The breakers should have voltage applied and

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be operated under load." The team determined that tha circuit breakers supplied with the battery l

chargers were not tested in accordance with the vendor manual recommendations. The licensee stated that the breakers were cycled without load dering the performance of BwHS 4009-003,

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" Clean and inspect Station Battery Chargers," Revision OE1, and that this cycling met the intent of the vendor manual. Procedure BwHS 4009-003 did not, however, contain any specific steps requiring that these circuit breakers be cycled. The team noted that the lack of 125V DC breaker testing was also considered a testing weakness during the EDSFl conducted in 1993. (URI 50-456/98-201-21)

The team reviewed surveillance procedures associated with the 125 VDC battery system and identified the following discrepancies (IFl 50-456/98-201-07):

1BwOS 8.2.1.2.b 1, "125V DC ESF Battery Bank and Charger 111 Operability. Quarterly

Surveillance," Revision 7, stated in Precaution D.8. " Water added to a cell should not be less than 50* F unless it is added for the purpose of maintaining electrolyte level over plates and separators." The licensee could not provide the basis for adding water to the cells at a temperature less than the battery minimum temperature of 60* F.

1BwOS 8.2.1.2.b-! and TS Table 4.8-2 (ATT) allowed the specific gravity of 1.280 in each

individual cell which exceaded the manufacturer's allowed range of 1.300 +.005. The licensee contacted the manufacturer during the inspection who stated that a specific gravity of 1.280 was acceptable.

18wOS 8.2.1.2.b-2, "125V DC ESF Battery Bank and Charger 112 Operability Quarteriy

Surveillance," Revision 7, was performed on February 23,1998, and battery cell temperatures were recorded. Step F.1 on the surveillance data r,heet had the

"ThermometeriDigital Dens ID:" marked N/A. The licensee stated that the digital voltmeter that was utilized during the surveillance was capable of recording temperatures and that the block may have been inadvertently checked N/A. The licensee stated that this would be reviewed.

The team noted that the table on page D-5 of procedure 18wOS 8.2.1.2.b-1 did not comply

with the requirements of IEEE 450-1995, which stated that 1 point (.001) should be added to the specific gravity reading for each 3' F above 77* F and 1 point subtracted for each 3* F below 77* F. The tabic within the surveillance procedure did not require that the specific gravity be adjusted for the first 2 F of deviation from 77* F. The licensee stated that electronic hydrometers were used to read the specific gravities which automatically corrected for temperature. The licensee initiated PlF A1998-01332 to revise the table in the surveillance procedure.

Calcuistion 19-D-21 Identified that the battery sizing criteria was based on a final minimum

battery voltage of 107.9Vdc. The safety evriustion for modification M20-1-92-007, dated February 18,1994, also confirmed this voltage. However, step 3.12 of surveillance procedure 1BwVS 8.2.1.2.d-111, " Unit One 125 Volt Btry Bank Operability - Stry Srve Test,"

Revision 1, specified an acceptance criterion of 107.88Vdc.

The team noted that a formal evaluation of battery degradation was not being performed. This evaluation would compare the overall battery capacity determined during the preceding test to the capacity determined during the present test and determine the rate of degradation. TS 4.8.2.1.2 stated that this evaluation should be performed so as to determine if an accelerated testing

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Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations," which required battery capacity trending. The TS required the testing frequency to be ivessed I? signs of degradation occurred. The team noted that battery testing procedure 19wVS 8.2.. 2.e-111, "125 V Battery Bnk and Charger Operability 5 Yr. Batt Capacity," Revision 1, did not include verification of reduction in battery capacity required by the TS.

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The licensee stated that an informal review was performed and that this review met the TS and IEEE-450 requirements. However, no records were maintained of battery test result trending.

E1.3.2.3 Conclusions The electrical design for components that performed the normal and accident functions of the Si system supported the design basis functions of the system. The electrical system provided independent, redundant, safety-related power to the electrical Si loads.

An electrical separation design deficiency that allowed !.' e turbine bearing oil pump motor to be automatically loaded onto the ESF bus after a loss of offsite power was not corrected in a timely manner. The potential for operating the battery chargers to power inverters existed and this additional load was not considered in the battery sizing calculation. Evaluation of IN 91-45 was not thorough in that it was incorrectly determined that NSFD relays were not installed in the plant and the reported relay problems were not evaluated. The team identified several calculation and surveillance procedure discrepancies and two safety evaluations that were not thorough in justifying the acceptability of modifications.

E1.3.3 Instrumentation and Control Design Review E1.3.3.1 Scope of Review Tne team evaluated the ability of the instrumentation and controls for the SI system to support the system in the performance of its safety functions. The team reviewed sections of the UFSAR, applicable TS, flow diagrams, uncertainty calculations,' and calibration data. The installation of selected l&C components was inspected during walkdowns.

E1.3.3.2 Inspection Findings E1.3.3.2(a) SI System Switchover from injection 'o Recirculation Mode The team reviewed procedure 18wEP ES-1.3, " Transfer to Cold Leg Recirculation Unit 1," Revision 1 A, and observed a demonstration of the procedure on the plant simulator. Step 2 of this procedure requires verification of adequate containment sump water level by checking that the bottom four lights of level switches LS-0940A and LS-0941 A were lit. The lights indicate that adequate NPSH was available to the RHR pumps before isolating the flowpath from the RWST. The switches are safety-related and seismically mounted, but are not qualified to operate in the harsh environment in the containment during a LOCA, and therefore, could not be relied upon to provide a reliable indication of water level in the containment sump during a LOCA.

The licensee evaluated the failure modes of the level switches and concluded that level switch failure would not mislead the operators because calculation SI-90-01, " Minimum Water Volume Available for Containment Recirculation Sump Flooding," Revision 6, had conservatively determined

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proper operation of the RHR pumps at the time of switchover, and that backup indications from qualified containment floor water level indicators 1Ll-PC006 and 1LI-PC007 would be available.

Dunng the simulator demonstration, the team noted that the containment ficor water level indicators, l

l 1Ll-PC006 and PC007, indicated a level of 15 inches at the time the operators were executing step 2 of procedure BwEP ES-1.3. If the four sump levellights were not lit and a levelless than 22

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inches was indicated, the operators were directed to go to step 1 of 18wCA-1.1, " Loss of l

Emergency Coolant Recirculation Unit 1," Revision 1 A, to restore emergency coolant recirculation capability. The licensee reviewed the bounding calculations for minimum and maximum post-LOCA

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containment water level, and determined that the expected water level would be. greater than 22 inches and the simulator model for containment water level was incorrect. The licensee issued PlF

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A1998-01262 to provide a basis for the containment water level set points and update the simulator model for post-LOCA containment water level.

In preparation for this inspection, the licensee performet a self-assessment and had issued PlF A1998-00711 to document that an appropriate calculation for the expected post-LOCA containment water level at the time of the suction switchover did not exist. The licensee had also identified an inconsistency between '.he emergency operating procedure (EOP) setpoint basis document and procedure BwCA-1.1 in that the checking of the bottom four containment sump level lights in step 10 b. would not indicate that adequate water level for tae suction of the CS pumps was available and issued PIF A1998-00712 to document this issue. The licensee determined that the procedure

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was still adequate. The licensee alts bsued PIF A1998-00813 as a result of the problems identified concoming the EOP set points and identified a need for development of a enntrolled EOP setpoint program. (IFl 50-456/98-201-22)

Calculation SITH-1, *Refuelirq Water Storage Tank (RWST) Level Set points," Revision 4, contained a data sheet and acceptance criteria that were used during simulator drills to verify that the switchover of Si from injection to recirculation could be performed before the RWST reached a level which could cause adverse suction conditions for the ECCS pumps. The data sheet was based on flow rate calculations that considered pump flow and backflow from the RWST directly to -

the containment sumps when the RHR pump suction was simultaneously open to the RWST and the sumps. The backflow was a major factor in the development of the data sheet since it was about 50 percent or more of the total flow from the RWST depending on valve alignment. The calculation stated that this backflow rate was a conservative engineering judgement that did not require an approved calculation due to the level of conservatism used but provided no basis for this judgement in the calculation. In response to the team's questions, the licensee retrieved the basis for the engineering judgement which was an informal bounding engineering analysis that included developing a piping model and checking the model against Westinghouse results for backflow during a large break LOCA. The team's review of this analysis determined that the flow rates calculated were not conservative because it did not maximize flow from the RWST. The licensee performed a preliminary re-evaluation of the calculation considering other conservatism in the assumptions used for pump flows and valve opening times, and concluded that the acceptance criteria on the data sheet used to verify adequate switchover time were still acceptable. The licensee initiated PIF A1998-01440 to revise the calculation. Section 4.0 of procedure NEP-12-02,

" Preparation, Review, and Approval of Calculations," Revision 5, requires that the basis for engineering judgement shall be documented to permit the reviewer to verify the logic and adequacy of the preparer's engineering judgement. This procedure requirement was not followed. Also, the analysis that formed the basis for an engineering judgement was not verified or checked. (URI 50-456/98-201-23)

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f E1.3.3.2(b) Vendor Recommendation * on Component Replacement The protective relays moun'ed within.

switchgear for EDGs 20E-1-4020A and 20E-1-4021 A were type SA-1 generator differential relays used for the differential current relay circuit that tripped the EDG circuit breaker on generator differential overcurrent. The team questioned whether the electrolytic capacitors on the relays had a replacement schedule, because they were known to degrade over time and adversely affect circuit performance.

The team noted that product manual LO297 *4160-6900 V Switchgear and Bus Duct Volume 2,"

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dated November 21,1996 (Braidwood number BwAP 1340 5T3, Revision 0) required that all relays should be checked once a year to catch the electronic component failures which occur on a random basis and that the tantalum capacitors may have a common mode failure characteristic and should be checked visually for symptoms of electrolyte leakage every year and replaced if necessary. The manual also stated that the tantalum capacitors should be replaced every ten years.

The SA-1 relays were checked for function, operation, and calibration every 36 months in accordance with BwHS 4002-066, * Periodic Protective Relay Calibration," Revision 3. No routine replacement program had been established for the tantalum capacitors. The licensee stated that no adverse trends in the performance of tantalum capacitors, transistors or potentiometers had been noted during periodic maintenance of these relays, and issued PlF A1998-01346 to evaluate the SA-1 relay maintenance requirements. Activities affecting the performance of the SA-1 relays had not been prescribed in the procedure. (URI 50-456/98 201-24)

E1.3.3.2(c) Volume Control Tank Level The team reviewed the operation of the level control instrumentation (1LT 112 and 1LT 185) for the volume control tank (VCT) and the suction conditions for charging pumps 1CV01PA and 1CV01PB during switchover from the VCT to the RWST upon a Si signal. The VCT normally provides water for the charging pumps. Upon a Si signal, the charging pump suction switches over to the RWST.

There was no design documentation that demonstrated that a vortex would not form and entrain air in the charging pump suction during high pump flow before closure of VCT outlet valves or at the low-low level setpoint in the VCT.

During the inspection, calculation SAE/FSE-C-CAE/CBE/CCE/CDE-0196, "VCT Vortex Potential,"

Revision 0, was prepared by Westinghouse to evaluate the potential for vortexing and air entrainment during switchover of the charging pump suction from the VCT to RWST upon receipt of a SI signal. This calculation also evaluated the acceptability of operation at the low-low level setpoint in the VCT. The calculation concluded that there was adequate volume available in the VCT to prevent air entrainment during switchover as well as at the low-low level. The team agreed with the conclusions of the calculation. The team also reviewed the available calibration data on the VCT level transmitters and determined that they were acceptable.

E1.3.3.2(d) RWST Level Transmitters During a LOCA, the draw down of water from the RWST would result in a partial vacuum in the tank air space due to the arrangement of the vent piping. Check valve 1Sl8969F in the common line connecting the " low" side of the level transmitters is intended to prevent air from the RWST tunnel leaking through to the level transmitter reference legs. The eneck valve is normaP closed and

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opens intermittently to drain condensate that may have co!!ected in the common level transmitter reference legs.

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The potential partial vacuum in the RWST air space during a LOCA sensed by the "high" side of the four transmitters might not also be sensed by the reference leg (" low" side) should check valve 1Sl8969F fail to close. A check valve failure could relieve the vacuum applied to the low side of the transmitters and result in a RWST level signal that would indicate a lower level than actually in the tank. This erroneous level signal could cause switchover from injection to recirralation before the RWST had discharged the volume of water to the containment used in design basis analyses. This check valve was not in the IST program or the preventive maintenance program. Since the pressure difference between the RWST air space and the RWST tunnel was near zero during normal operation, failure of the check valve to close would not be detected by abnormal level indication. The licensee issued PlF A1998-01278 to evaluate the appropriate testing for the check valve. The inservice testing program did not include reverse flow testing of check valve 1Sl8969F which performs a safety function. (URI 50-456/98-201-25)

E1.3.3.2(e) Test Equipment Calibration Flow indicator 1F1-0121 A was used for inservice testing of charging pump 1CV01PA. Paragraph 4.6.1.1. and Table 1 of ASME/ ANSI OM-1987, Part 6, (Inservice Testing of Pumps in Light-Water Reactor Power Plants) specified the required instrument accuracy as +/- 2% of full scale for individual analog instruments. However, the allowable instrument tolerance for 1FI-0121 A was +/-

3% as documented in test report packages 1L-0459A and 2L-0459A of procedure BwlP 2500-006,

" Pressurizer Water Level Control and Charging Flow Loop," Revision 3.

The licensee reviewed the most recent calibration of 1F1-0121 A and determined that the indicators were within +/-2 %. The inservice test was conducted after the instrument was calibrated, and therefore, the licensee concluded that the last test was acceptable. The licensee issued PlF A1996-01428 to resolve this discrepancy.

E1.3.3.2(f) Modifications The team reviewed modifications E20-1-96-255, BRW-SE-1997-1556, and E20-1-96-245 and setpoint/ scaling change requests SSCR 96-002 and SSCR 95-009.

The modifications were consistent with the design basis and the associated safety evaluations were acceptable. SSCR 95-009 for Unit 1 accumulator level transmitter was identified as non safety-related, while SSCR 96-002 for the corresponding Unit 2 transmitter was identified as safety-related.

The correct designation on both SSCR's should have been non safety-related. The two SSCRs were otherwise adequately prepared. The licensee issued PIF A1998-01344 to correct this discrepancy.

E1.3.3.3 Conclusions The team concludea mat the instrumentation and controls for the SI system were capable of performing their design functions and were consistent with the licensing bases.

The team identified a discrepancy in the containment water level used in the simulator model for a LOCA/ LOOP scenario. The licensee had identified, during a self-assessment in preparation for this inspection, that the bases for containment water level set points in various EOPs were inadequately supported by existirig calculations. The results of an unverified analysis was used as engineering l

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judgement in an RWST level setpoint calculation. A capacitor in the protection relay in the EDG switchgear was not scheduled for periodic replacement as required by the vendor manual. Reverse flow testing of a check valve in the RWST level reference leg was not included in the IST program.

E1.3.4 System Walkdown

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The team performed several walkdowns of the accessible portions of the SI system and the RWST to verify that the system configuration was consistent with the design basis. The inspected portions of the Si system configuration were consistent with the system flow diagrams. The installed

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instrumentation configurations were consistent with the design and were appropriate for their

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application.

E1.4 UFSAR and Design Documentation Review E1.4.1 Scope of Review The team reviewed the UFSAR and drawings for the AFW and Si systems for consistency with the design and licensing basis.

E1.4.2 Inspection Findings The team identified the following discrepancies in the UFSAR:

Table 3.112 listed the maximum temperature for zone A11 as 133' F whereas the diesel-

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driven AFW pump room maximum was 122 F. All the components listed in zone A11 did not have the same maximum design temperature. The licensee initiated UFSAR Draft Revision Package (DRP) 7-198 to revise the table.

Section 9.4.5.1.1.a.8 incorrectly stated that all HVAC electrical components are powered

from ESF buses. The licensee initiated Design Basis initiative Tracking Number 945 to correct the UFSAR.

Table 9.2-4 appeared to list required flow rates but was described in the text as giving

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expected flow rates. The licensee issued PIF A1998-01424 to identify the discrepancy and DRP 7-208 to revise the table.

Section 10.4 stated with "each direct diesel engine-driven auxiliary feedwater pump". The

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wording incorrectly implied that there was more than one diesel engine-driven AFW pump.

The licensee stated this discrepancy will be corrected upon completion of the Design Basis initiative effort.

Table 8.3 5 for EDG loading contained historical data but explanatory note (x) did not clearly

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state this condition. Additionally, item #23 of the table was never installed. The licensee stated that the UFSAR table would be revised appropriately.

The above discrepancies had not been corrected and the UFSAR updated to ensure that the information included in the UFSAR contained the latest material.

The team identified the following drawing discrepancies:

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t The flow diagrams for the Reactor Coolant System, M-60-1B, Revision BC; M-60-2, Revision j

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AW; M-60-3, Revision AY; and M-60-4, Revision BA, incorrectly identified the CV pump

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connections to the cold legs as connections from the SI pumps. The licensee initiated ER9800799 to correct the drawings and the corresponding Unit 2 drawings.

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Drawing 20E-1-4010E, "125V DC ESF Distnbution Center Bus 112 (1DC06E) Part-2,"

Revision F, identified branch breakers supplying the normal and reserve control power to the

"125V DC Control Bus at 480V SWGR"; however, these branch circuit breakers were spares. The licensee stated that the drawing was not revised when the common ByroniBraidwood drawing was made unique to Braidwood. The licensee siso determined that this discrepancy was also applicable to drawings 20E-1-4010B and Braidwood Unit 2 I

drawings 20E-2-4010B and 20E-2-4010E, and initiated PIFs A1998-01265 and ER 9800672 to correct the drawings.

l The identifying number for the low oil pressure time delay relay was missing from drawing

20E-1-4030AF12, "Senematic Diagram Auxiliary Feedwater Pump 1B (Diesel-Driven) Engine Startup Panel 1 AF01J," Revision W. Tht Iicensee issued DCR No. 980066 to correct this discrepancy.

A note identifying the flow restrictors used to limit flow to the faulted steam generator was

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incorrectly shown on drawing M37, " Diagram of A'.txiliary Feedwater Unit 1," Revision 8B.

The licensee issued PlF A1998-01645 to correct this discrepancy.

The team also identified that the TS Bases for Section 3/4.8 was incorrect in that it referred to UFSAR Table 8.3-1 instead of Table 8.3-5 and identified the SX pump load as 1247 brake horsepower (bhp) instead of 1290 bhp in accordance with calculation 19-T-6, " Diesel-Generator Loading During LOOP /LOCA -Braidwood Units 1&2," Revision 2. The licensee issued PlF A1998-01503 for this discrepancy.

X1 Exit Meeting After completing the on-site inspection, the NRC conducted an exit meeting with the licensee on April 24,1998, that was open to public observation. During the meeting, the team leader presented the results of the inspection. A list of persons who attended the exit meeting is contained in Appendix B. The team reviewed licensee provided proprietary information during the inspection, but such proprietary information is not included in this report.

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APFENDIX A OPEN ITEMS This report categorizes the inspection findings as unresolved items and inspection follow-up items in accordance with the NRC inspection Manual, Manual Chapter 0610. An unresolved item (URI) is a

. matter about which more information is required to determine whether the issue in question is an acceptable item, a deviation, a nonconformance, or a violation. The NRC Region ill offee will issue any enforcement action resulting from the review of the identified unresolved items. An inspection follow-up item (IFI) is a matter that requires further inspection because of a potential problem, because specific licensee or NRC action is pending, or because additional information is needed that was not available at tha time of the inspection.

I item Number Finding

.T.ine IY.E.1 50-456/98-201-01 IFl AFW Piping Design Pressure (Section E1.2.1.2(b))

50-456/98-201-02 IFl AFW Pump Overspeed Testing (Section E1.2.1.2(b))

50-456/98-201-03 URI Diesel Engine Pressure Relief Cap Testing (Section E1.2.1.2(c))

50-456/98-201-04 IFl AFW Minimum Flow (Section E1.2.1.2(e))

50-456/98-201-05 IFl Room Heat Capacity Verification Calculations (Sections E1.2.1.2(f) and E1.3.1.2(f))

50-456/98-201 06

!FI DDAFW Battery Calculation Discrepancies (Section E1.2.2.2(c))

50-456/98-201-07 IFl Electrical Test Procedure Discrepancies (Section E1.2.2.2(c))

50-456/98-201-08 URI ESF Circuit Breaker Failures (Section E1.2.2.2(d))

50-456/98-201-09 URI Testing of K11 Relay (Section E1.2.3.2(a))

50-456/98-201-10 IFl Instrument Tubing Heat Tracing (Section E1.2.3.2(d))

50-456/98-201-11 URI Check Valve Reverse Flow Testing (Section E1.3.1.2(c))

50-456/98-201-12 IFl ECCS Leakage Through Valves (Section E1.3.1.2(c))

50-456/98-201-13 IFl Orifice Used in Test Procedure (Section E1.3.1.2(c))

50-456!98-201-14 IFl SI Accumulator Operations (Section E1.3.1.2(d))

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50-456/98-201-15 IFl ECCS and AFW Pump-to-Pump interaction (Section E1.3.1.2(e))

50-456/98-201-16 IFl CCW Relief Valve Location (Section E1.3.1.2(f))

50-456/98-201-17 IFl Evaluation of NBFD Relays (Section E1.3.2.2 (b))

50-456/98-201-18 URI Automatic Loading of Turbine Bearing Oil Pump (Section E1.3.2.2(d))

50-456/98-201-19 IFl Inverter Power Supply (Section 1.3.2.2(e))

50-456/98-201-20 IFl EDG Circuit Breaker Testing (Section E1.3.2.2(f))

50-456/98-201-21 URI Battery Charger Circuit Breaker Testing (Section E1.3.2.2(f))

50-456/98-201-22 IFl Containment Water Level Calculations and EOP Setpoints (Section E1.3.3.2(a))

50-456/98-201-23 URI Verification of Design input Analysis (Section E1.3.3.2(a))

50-456/98-201-24 URI EDG Relay Maintenance (Section E1.3.3.2(b))

50-456/98-201-25 URI RWST Level Instrument Check Valve (Section E1.3.3.2(d))

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o APPENDIX B EXIT MEETING ATTENDEES blafdE ORGANIZATION Commonwealth Edison T. Tulon Site Vice President J. Hosmer Vice President - Engineering J. Meister Engineering Manager G. Schwartz Station Manager R. Wegner Operations Manager L. Weber Shift Operations Supervisor T. Simpkin Regule. tory Assurance Supervisor F. Lentine Programs Engineering Supervisor B. Viehl Engineering Assurance Supervisor J. Phelan Staff Engineer J. Matthews System Engineer J. Tolar Mechanical Engineer

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G. Van Duyne Office Support C.'Dunn System Engineering Supervisor A. Ferko System Engineer G. Smith Consultant R. Brodsky Consultant i

P. McHall Westinghouse - Engineering Manager S. Mullins Project Engineer D. Galanis Station Support G. Norvil Quality & Safety Assessment J. Nalewajka Quality & Safety Assessment M. Kluge Quality & Safety Assessment R. Wunder Design Engineering T. VanWyck Design Engineering J. Zoelter Mechanical Engineenng q

K. Respotich Office Support j

K. Namors Office support Supervisor

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T. Bozan.

Office Support

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M. DiPonzio RPA J

D. Chrzanowski NLA J

i D. Radice Design Engineering Supervisor P. Donavin Design Supervisor-Byron W. Kouba Engineering Manager-Byron J. Panfill Lead Engineer P. Beinicke Consultant

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W. Marini Consultant

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N. Valos Operations -Zion

- A. Mahadevia Site Engineering l

D. Pierce Engineering Admin. Supervisor

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A. Nies Sargent & Lundy l

J. Krvavac Sargent & Lundy

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M. Riegel Quality & Safety Assessment Manager

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W. Dupuis Maintenance Staff Supervisor E. Hendrix Operations Staff Supervisor J. Walker Training Manager R. Graham Work Control Manager B. Humphries Westinghouse - Project Manager R. Byers Maintenance Manager J. Chovan Self Assensment Admin.

A. Haeger HP/ Chemistry Supervisor J. Bergner Group Leader R. Adams SED S. Butler Quality & Safety Assessment D. MacMahan Equipment Attendant J. Kmetz Equipment Attendant R. Stopka Equipment Attendant

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D. Jenco Equipment Attendant p

M. Gorski Engineer D. Lee Mechanical Design B. Momsen Senior Staff Engineer P. Vandevisse I & C Engineer D. Hieggalke Corrective Actions Program Manager K. Lawshe Site Engineering - Mechanical D. Hodge Designer / Drafter J. Gastouniotis Mechanical Engineer S. Shah Mechanical Engineer B. Acas Station Support - Mechanical Lead J. Muraida Station Support Engineering NBC D. Norkin Section Chief, NRR/PECB S. Malur Team Leader, NRR/PECB Z. Falevits Region lli Inspector C. Phillips Senior Resident inspector S. Bailey Project Manager, NRR R. Bradbury Contractor, SWEC P. Bienick Contractor, SWEC R. Gauthier Contractor, SWEC M. Huq Contractor, SWEC D. Schuler Contractor, SWEC OTHER T. Esper Resident inspector - lDNS

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APPENDIX C

LIST OF ACRONYMS USED AC,ac Attemating Current AFW Auxiliary Feedwater ANSI American National Standards institute ASME American Society of Mechanical Engineers bhp brake horsepower BTU British Thermal Unit CAR Corrective Action Record CCW Component Cooling Water CFR Code of Federal Regulations CS Containment Spray CST Condensate Storage Tank CV Chemical & Volume Control DC,dc Direct Current DDA.W Diesel-Driven Auxiliary Feedwater DRP Draft Revision Package DW Demineralized Water EDG Emergency Diesel Generator EDSFl Electrical Distribution Safety Functional Inspection EOP Emergency Operating Procedure ER Engineering Request ESF Engineered Safety Features ECCS Emergency Core Cooling System ELMS Electrical Load Monitoring System i

EWCS Electronic Work Control System FLB Main Feedline Break R.

feet GL NRC Generic Letter gpm Gallons Per Minute HJTC Heated Junction Thermocouple Microprocessor Chassis IEEE institute for Electrical and Electronic Engineers j

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