IR 05000324/1997003
| ML20138J598 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 05/02/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20138J568 | List: |
| References | |
| 50-324-97-03, 50-324-97-3, 50-325-97-03, 50-325-97-3, NUDOCS 9705080271 | |
| Download: ML20138J598 (26) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION ll
Docket Nos: 50-325, 50-324
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i License Nos: DPR-71, DPR-62 Report No: 50-325/97-03, 50-324/97-03 '
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Licensee: Carolina Power & Light (CP&L)
Facility: Brunswick Steam Electric Plant, Units 1 & 2
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Location: 8470 River Road SE Southport, NC 28461 Dates: March 3 - 14 and April 2 - 4,1997 Inspectors: J. Lenahan, Reactor inspector M. Janus, Resident inspector D. Trimble, NRR Project Manager J. Mallanda, Beckman & Associates J. Williams, Beckman & Associates Approved by: H. Christensen, Chief, Engineering Branch Division of Reactor Safety l
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i i Enclosure 2 9705000271 970502 PDR O ADOCK 05000324 PDR
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EXECUTIVE SUMMARY Brunswick Steam Electric Plant Units 1 & 2 NRC Inspection Report 50-325/97-03, 50-324/97-03 This inspection included review of the licensee's engineering activities to support operation of the Brunswick plant, the environmental qualification (EQ) program and followup on previous inspection findings. The areas inspected included review of procedures, completed
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calculations for recently installed modifications, closeout of calculations and followup on corrective actions to resolve EQ problem Results:
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The licensee's design change control procedures complied with the requirements of 10 CFR 50.59, and 10 CFR 50, Appendix B, Criterion Il A violation was identified for failure to perform a reportable event evaluation for past operability of the Unit 2 reactor water cleanup system due to improper installation of seals on three Rosemount transmitter A violation was identified for failure to incorporate an engineering service request in a change to the UFSA ! -
In general the modifications packages were of good quality and would not degrade plant performance, safety, or reliabilit A minor weakness due to attention to detail was identified regarding deficiencies in documentation of design information in calculations and ESR packages.
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The licensee's Self Assessments in the Engineering Support Area were adequate in evaluating Engineering Suppor ,
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A weakness was identified for not properly documenting corrective actions to
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resolve findings from self-assessment The licensee's progress to correct the EQ program deficiencies was satisfactory.
- Equipment operability issues were appropriately evaluated through JCO' Enclosure 2
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REPORT DETAILS
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! 111. Enaineerina 1 E1 Conduct of Engineering E Desian Chance Control Processes
, Inspection Scope The inspectors reviewed the licensee's procedures which control the design change '
program, Observations and Findinas The inspectors reviewed the procedures listed below which control design and design changes to determine if the procedure implement the requirements of 10 CFR 50, Appendix B, Criterion ill and 10 CFR 50.59. The following procedures were reviewed:
EGR-NGGC-0001, Conduct of Engineering Operations, Rev. 2, dated February 3,1997 EGR-NGGC-0003, Design Review Requirements, Rev. O, dated June 3,1996 EGT!-NGGC-0005, Engineering Service Requests, Rev. 2, dated December 17, 1 1996, and Temporary Change No.97-006, dated February 17,1996 I EGR-NGGC-0006, Vendor Manual Program, Rev.1, dated August 6,1996 i
EGR-NGGC-0007, Maintenance of Design Documents, Rev. O, dated I December 17,1996 EGR-NGGC-0320, Civil / Structural Operability Reviews, Rev. O, dated 3 May 8,1996 l OENP-33.5, Quality Classification Analysis of Structures, Systems, and Components, Rev.10, dated February 4,1997 OENP-303, Preparation and Control of Design Analyses and Calculations, Rev.1, dated February 20,1996 OENP-1000, Brunswick Engineering Support Section Conduct of Operations, Rev. O, dated February 5,1997
, OIA-109, Performance of Nuclear Safety Reviews, Rev. 8, dated January 14,1997 The inspectors verified that the procedures adequately addressed: design inputs, design calculations, drawing changes, post-modification testing, control
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2 l of field changes,10 CFR 50.59 safety evaluations, training, and ALARA reviews. The inspectors noted that Temporary Change 97-006 was issued by the licensee to address the violation identified in NRC inspection Report number 50-325, 324/97-02 regarding design verification of safety related l configuration change engineering service requests (ESRs). Conclusions The inspectors concluded that the licensee's design change control procedures complied with the requirements of 10 CFR 50.59, and 10 CFR 50, Appendix B, Criterion 11 i E1.2 Review of Desian Chanaes and Modification Packaaes I Inspection Scope
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The inspectors reviewed randomly selected design change and modification packages to: (1) determine the adequacy of the safety evaluation screening and the 10 CFR 50.59 safety evaluations; (2) verify that the modifications were reviewed and approved in accordance with Technical Specifications and administrative controls; (3) verify that ,
applicable design bases were included; (4) verify that Updated Final Safety Analysis Report requirements were met; (5) verify that both installation testing and post modification testing requirements were specified so that adequate testing would be accomplishe Observation and Findinos The Engineering Service Requests (ESRs) discussed below were reviewed for adherence to procedures, codes and standards, commitments, and technical adequac ) ESR 9600311 Service Water Return Piping Replacement for the Diesel Generator Coolers This ESR replaced cement lined carbon steel service water return piping and components for the diesel generator jacket water coolers with 70-30 copper-nickel piping through the four day tank room, to the Unit 1 service water valve pit. A portion of piping downstream of a cavitating valve was replaced with stainless steel material. The inspectors reviewed stress calculations SA-SW-B053B, C, D and E-91072 and interviewed the licensee's engineers responsible for project engineering, piping analysis and inservice inspection and testing. The pipe stresses were low and within code allowable The design and construction showed careful consideration of differences in material properties such as modulus of elasticity, coefficient of thermal expansion, and galvanic corrosion for the materials used. Material changes were made at flanged joints which used dielectric flange kits consisting of flange face gaskets, and sleeves for bolts and washers under the regular bolt
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flat washers. The expansion joint and flanges were on an 18 month inspection cycle. After a review of the ESR, piping analysis, and design drawings, the inspectors walked down the completed installations in the diesel generator and service water buildings. The new piping was installed in accordance with the construction drawings and exhibited good design and workmanship. Piping penetrations through masonry wall construction were designed to only exert a vertical force on the wall The analysis referred to ANSI B31.1 and used material properties from a B3 Workbook. However the analysis states the strength and integrity of the p'iping i systems within Inservice inspection (ISI) boundaries shall be verified by pressure testing the joints and associated piping in accordance with ASME i Section XI. UFSAR paragraph 3.2.2 System Quality Group Classification, referenced Specification 248.117, Specification for installation of Piping Systems, Pressure Ratings, Material, and Code Requirements. Inspection j requirements were determined from Specification 248-117 and system P& ID '
diagrams which designate the ENP-16 boundaries. Pressure test requirements shall be per ASME Section XI for piping within the ENP-16 boundaries. Piping outside the ENP-16 boundaries (with the exception of fire protection piping)
shall be tested per ANSI B31.1. Subsequent code interpretation allow Section j XI hydrostatic testing rather than the provision to hydrostatic test to the original !
Construction Code (B31.1).
Review of the calculations disclosed the following issues:
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The analysis contained the following statement: "The new expansion 4 joint has tie rods; therefore a bellows thrust load need not be analyzed." !
However this conclusion was not immediately obvious and should have I been quantified. The analysis shows low resulting loads and the l bellows pressure-temperature rating was sufficiently higher than the I system design conditions to substantiate qualification, but none was '
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Qualification of the expansion joint also used vendor data received via ;
telephone contact with product representatives which doubled the i catalog stated traverse allowaole movement from 1/4" to %". The ;
piping analysis showed a maximum requirement of 0.269" The vendor l followed up the telephone contact with a facsimile message. After the i inspector questioned this product data, the licensee obtained newer catalog data which substantiates the use of %".
l 2) ESR 9500665 Vital Header - Replacement with Copper-Nickel Pipe Material l This ESR replaced cement lined carbon steel piping in line numbers 2-SW-117-6-157 and 2-SW-133-6-157 with copper-nickel material in the reactor building service water system. The inspectors reviewed stress calculations ;
SA-SW-108 and SA-SW-294 which covered the piping replacement. While the i allowable stress and elastic modulus for the new replacement piping materials I
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were lower than the onginal piping (carbon steel) materials, the limited nature of the replacement along with low existing stress levels in the piping ensure that sufficient design margin would remain post-modification. The new piping was pressure tested to ASME Section XI requirements, and dielectric insulators were installed at the connecting pipe flanges to minimize corrosio ) ESR 9501423 Correct Erosion Problem with Line 1-SW-140-16-17A This ESR replaced an existing copper-nickel section of piping which had been in service for eight years, downstream of the 1-SW-V382 throttle valve, with stainless steel. The new piping consists of a flanged section between dissimilar metals separated by dielectric insulators. This portion of the RBCCW service water system stress analysis was not adversely affected by the type 316L stainless steel seamless pipe replacement since the stainless steel was compatible with the copper-nickel piping. The stainless steel had an allowable stress almost twice the copper-nickel piping,and a weight and coefficient of thermal expansion less than copper nickel. The inspectors reviewed stress calculations SA-SW-76111/762/763-0001 and PS-SW-763-0001 which cover this section of the piping. Pressure testing used ASME Section XI Code Case N-416- ) ESR 9500026, Replace Turning Gear Motor This ESR replaced the existing 2 speed turning gear motor with a main motor and a piggy back motor. This was a direct replacement recommended by General Electric, the turbine manufacturer. Since the original motor was being replaced by 2 motors, an analysis of the electrical distribution system was required. This modification also revised the control logic since there would be 2 separate motor starter The inspectors reviewed the 10 CFR 50.59 safety evaluation and determined that it was adequate and evaluated the impact of the design change. The responsible engineer obtained design (basis) impact statements from the Electrical, Mechanical and Civil disciplines. A human factors evaluation was performed since the ESR affected a control room pane Electrical Engineering issued Load Change Approval Memo BNP-918, dated 3/27/95, that contained a standard matrix that was completed to document the calculations affected by this ESR. The matrix incorrectly identified Attachment B of Calculation BNP-E-8.010, Revision 0,10/25/95, as the basis for the breaker settings for the 40 horsepower main turning gear motor. The setting for this breaker was actually specified in Attachment HH. The matrix indicated that the setting should be at least 700 amperes. However, the conclusion in Memo BNP-918 indicated that the nominal instantaneous breaker setting must be equal to 700 amperes. The Design Change Backup Form BNP-E-2.012-0001, dated 3/29/95, also specified setting of 700 amperes for the existing
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breaker. The licensee could not explain differences in breaker settings specified in memo BNP-918, equal to 700 amperes, versus set to at least 700 amperes as stated in the matrix.
Calculation BNP-E-8.010, Revision 0, Attachment HH, showed a setting of 2 which is equal to 830 amperes. The installation instructions also stated that the breaker setting shall be adjusted to setting number 2. The breaker setting was in accordance with the CP&L Calculation BNP-E-2.012, Revision 4, dated 6/30/95, BOP Protective Device Sizing Calculation. The overload heater sizes were selected in accordance with GE Selection Guide GET-2681L, which was not referenced in the modification package or Load Change Approval Memo BNP-918. However, the inspectors considered the specified breaker settings to be adequate, although the ESR documents were not clear and specific regarding the settings.
5) ESR 9400241. Transformer Fault Pressure Relay Annunciation Does Not Function The fault pressure alarm circuit for the Main, Unit Auxiliary, Start-Up and Caswell Beach transformers did not function when a fault pressure existed since the lockout relay in the circuit actuated faster that the annunciator could react to the alarm. The purpose of this ESR (modification) was to install an auxiliary relay with a seal-in contact to actuate the annunciator. The 10 CFR 50.59 evaluation which was performed for the design change was adequate.
Engineering Evaluation EE BNP-DC-032 was initiated by the licensee to evaluate the auxiliary electrical distribution system changes proposed by ESR 9400241 to ensure that implementation of the proposed changes would not have an adverse impact on the electrical distribution system. The changes were evaluated for their impact on the voltage /short circuit current / load flow analysis, coordination / protection, Appendix R analysis, LOCA/ Station Blackout DC Load analysis, and the Diesel analysis.
The inspectors reviewed this evaluation and questioned the actual maximum operating voltage that would be available at the relays when the 125V DC System was on equalizing charge. The maximum operating voltages for the new relays was indicated as either 135.1V DC or 137.5V DC for relays 63FPX i and 63FPY respectively. The 63FPX relay was rated for a maximum of 52.8V l
DC and was in series with a 800 ohm resistor. Therefore, the licensee '
calculated a maximum DC voltage of 135.1V DC. The maximum operating voltages for the 63FPX and 63FPY relays were obtained verbally via a telephone conversation with individuals identified by name only with the relay manufacturers. The licensee indicated in the analysis that equalizing voltage on the 125V DC System could be as high as 140V DC for a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 4 or 5 times per year. This information was obtained by the responsible engineer via a verbal conversation with the system engineer. The inspectors questioned the use of verbal information instead of utilizing written test procedures or test informatio I i
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! 6 Preventive Maintenance Procedure OPM-BAT 004, Revision 5,6/1/95, indicates that the equalizing charge should be set at 139.8V to 140.3V DC, not taking into account any meter inaccuracies. This procedure also noted a W
caution that if battery charger voltage exceeded 142 volts, the battery charger
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Input breaker may trip. This indicated that the equalizing voltage could be as ;
high as 142V DC. The equalizing charge could be maintained for as long as !
j - 74 hours8.564815e-4 days <br />0.0206 hours <br />1.223545e-4 weeks <br />2.8157e-5 months <br />. In a query of the maintenance system, equalizing of the 125V !
batteries was performed quarterly (4 times per year). It may also be performed during an outage so equalizing could be performed a maximum of f 5 times per yea The inspectors determined that even though the voltage at the batteries may be higher than the maximum rating of the relays for approximately 5% of the i
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time, the impact on these relays was negligible since these relays are normally de-energized. The licensee was not able to produce additional information to support the maximum operating voltages of the relays. However, the life of the relays would not be affecte ) ESR 9400539, Installation of the Unit Auxiliary Transformer Backfeed Logic and Generator No-Load Disconnect Switch The purpose of this modification was to improve the availability of offsite power if the Startup Transformer is lost. Brunswick was originally licensed with one startup transformer per unit. In 1991 NRC determined that Brunswick did not conform with GDC-17 regarding the availability of the alternate power sourc In addition to a mechanical disconnect switch being installed in the isolated phase bus duct, an additional ground fault relay was added for alarming in the control room for a ground fault on the ungrounded isolated phase bus duct. A number of logic circuits were also revised to allow backfeed through the i generator breaker with the main generator not operating. In order to operate in the backfeed mode, a number of keylock switches were added to transfer controls to this mode. The keys are controlled by operation The electrical design impact indicated that the addition of the ground fault relay (GE HGA) was evaluated in Design Change Backup Form (DCBF) BNP-E-6.071-0002, dated 5/18/95, and DCBF BNP-E-6.075-0002, dated 5/18/95, ,
and was found to be acceptable. The civil design impact evaluated the additional weight of the no-load disconnect switch and supports in the Turbine Building and was acceptable. The 10 CFR 50.59 safety evaluation was adequat The inspectors reviewed DCBF BNP-E-6.071-0002 and DCBF BNP-E-6.075-0002 and requested backup information for relay resistances and voltage levels indicated in the evaluation. The references were not shown in the ES The analysis performed in these two DCBFs was identical. The catalog for the GE lAV52D ground fault relay indicated a resistance of 7.0 ohms for the ampere tap. The analysis used 8.3 ohms. This change did not change the results / conclusions of BNP-E-6.071-0002/BNP-E-6.075-000 . . . ~ . - - - - - - - . - - - - . .
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7) Modification 90-025,'AC MOV Overload Protection Division ll of the Nuclear Boiler, Recirculation, Containment Atmospheric Control and Standby Gas Treatment Systems The purpose of this modification was to review the present motor overload protection for motor operated valves as a follow up to the Motor-Operated Valve Task Force that recommended motor protection for the safety-related motor operated valves. When the overload heater size was determined, the licensee verified adequate cable size and breaker ratings and settings for short circuit protection and proper coordinatio The inspectors reviewed the 10 CFR 50.59 evaluation and' determined that it !
was adequate. The inspectors also noted that some of the breakers that were I substituted for existing breakers were not qualified as Category I breaker The licensee had performed a design impact evaluation as part of the modification that allowed replacing the installed breakers with DOR qualified breakers until Category I breakers were available. Regulatory Guide 1.89 allows the DOR qualified equipment to be replaced with other DOR qualified equipment to meet installation and operation schedules but can only be installed until upgraded Category I equipment becomes available. The new breakers were purchased as Q and Category 11. These breakers were on hold awaiting qualification documentation from General Electric per the Brunswick Supply inventory System. The licensee established a material evaluation ;
number to track the qualification documentation, which had not been received I by Brunswick at this time. This open item was being tracked within the EDB The inspectors reviewed the breaker testing criteria established in the modification for the testing of the magnetic trips for the newly installed breakers. The tolerances listed for testing the breakers reference NEMA AB4- .
1991 and Calculation BNP-E-8.082, Revision 0, dated 11/2/94, for the breakers I rated 10 amperes and below and NEMA AB4-1991, Table 5-4, for the breakers rated above 10 amperes. The inspectors were unable to confirm these test ,
requirements using these references. When questioned, the licensee was not l able to confirm the testing criteria referenced without talking directly to the j responsible engineer. The licensee explained that the engineer riecided to use more conservative testing ampacities than those indicated in the references since these were new breakers. Therefore, the testing criteria statement in the j modification were misleading and the testing criteria could not be determined '
with the references provided. The inspectors concluded that testing criteria utilized by the licensee was acceptable, although it was not clearly documented in the ESR document ) ESR 9600108, Replace Relay KT106 The purpose of this modification was to wire an unused normally closed contact of the KT106 relay in parallel with the presently used normally closed contact in response to an Operational Experience Feedback Report. This '
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modification was initiated due to the failure of KT106 relays at other nuclear plants that caused inadvertent reactor scrams. Normally closed contacts had failed open and caused the turbine valves to clos l The inspectors reviewed the design inputs and system impact evaluation which referenced ESR 9501480 written for Unit 2 and determined that these evaluations were adequate. There was no impact on the seismic response of the control room panel or on electrical loads. The inspectors reviewed the 10 '
CFR 50.59 evaluation and testing criteria and determine they were adequat ) ESR 9500593, Replace Unit 1 Refuel Bridge Main Hoist Power Cable Broken conductors in the main hoist cable on the refueling crane had caused temporary loss of the crane. The replacement cable (Perfect -A-Flex) had been designed for use on cable reels and was identified as a cable similar to the cross-linked polyolefin insulated, hypalon jacketed cables similar to the cable qualified for and used at Brunswick for safety-related applications. The replacement cable had been tested to meet the flame test requirements of IEEE Standard 383-1974. The inspectors questioned the similarity of the Perfect-A-Flex cable with the cross-linked polyolefin cables used for safety-related applications. The licensee provided environmental qualification package QDP No. 6 that was approved for use at Brunswick for safety-related applications and had a similar jacket and insulation material as the Perfect-A-
. Flex cable. The inspectors concluded that the replacement cable was acceptabl The inspectors reviewed the 10 CFR 50.59 evaluation and determined it was adequate. Since the work to replace the cable was to be performed on the refueling floor, the licensee performed an ALARA pre-approval walkdown. It was determined that a Health Physics representative must be present during the installation when the craft was working on the mast due to the high radiation levels and the possibility of contamination. Testing was performed in accordance with previously approved Procedure OPM-CRN00 ) ESR 9500545, Installation of Interposing Relays to improve Voltage Adequacy at MCCs The purpose of this modification was to initially installinterposing relays in selected MCC circuits since, during a system degraded grid voltage scenario, some MCC contactor coils would not pick up in safety-related circuits. New contactor coils were purchased and installe The inspectors reviewed the 10 CFR 50.59 evaluation. The licensee completed Engineering Evaluation Report 93-0176, Revision 2, dated 12/1/95, to confirm the operability of the MCCs prior to the change out of the contactor coils. Testing of the coils showed that the pick up voltages published by the manufacturers were conservative and that, during the degraded grid voltage scenario, the contactor coils would function as required. The inspectors
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reviewed the engineering evaluation report DCBF BNP-E-1.012-0002, dated 7/28/95, and DCBF BNP-E-1.013-0001, dated 7/28/95, that updated the .
calculations due to replacement of the contactors in the applicable I compartments. These documents were adequate to support both continued operation and the permanent plant modification j 11) Review of 10 CFR 50.59 Safety Evaluations The inspectors reviewed the licensee's evaluations for the design changes ,
listed below to determine if the changes introduced an unreviewed safety .i question. The inspectors determined that the evaluations were adequate in ;
that they satisfied the requirements of the applicable regulation (10 CFR l 50.59). None of the evaluations identified an unreviewed safety question. The I design changes reviewed were as follows: {
ESR 94-00539 - No Load Disconnect to Accomplish UAT Backfee ;
ESR 96-00353 - Power Uprate Turbine Controls Modificatio ,
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ESR 96-00546 - Disable B32-PS-N018 A & ' Conclusions in general, the modification packages were of good quality and would not i degrade plant performance, safety, or reliability. With the exception of some i documentation deficiencies and lack of attention to details, the modification packages contained sufficient specifications, drawings and procedures to be properly installed and tested. The licensee's 10 CFR 50.59 evaluations were completed in accordance with NRC requirement E1.3 Environmental Qualification Inspection Scope The inspectors reviewed the licensee's Environmental Qualification (EQ)
program, specifically their corrective actions to respond to findings identified during Self-Assessment numbers 95-0041 and 96-0271 and the violations '
identified in NRC Inspection Report number 50-325,324/96-1 Observations and Findinas 1) Background The inspectors reviewed the status of the licensee's corrective actions to resolve problems identified in the EQ program. The following issues were discussed with the licensee's EQ Task Force Manage . _ _ . _ __ _ _. __ _ . _ . _ - _ _ . . _ . _ _ _ . - . . _ _
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Corrections to the Equipment Data Base System (EDBS) and corrections to the EQ equipment list.
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Updating of Qualification Data Packages (ODPs).
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Revision of the Reactor Building Environmental Report (RBER).
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Status of the walkdown inspections being performed to determine if equipment required to be EQ is installed in accordance with the QDPs.
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Status of the seven previously identified JCOs and resolution of the
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technicalissues required for closeout. These JCOs address operability of PASS, thread sealants, associated circuits, the MCCs, fuses,
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Tripoint pressure switche The discussions disclosed that the licensee's actions were on schedule to correct the program deficiencies. The EDBS system and EQ list has been updated. The licensee is in the process of updating the QDPs. An Architect-Engineer firm was recently retained to assist in the QDP updates to meet the scheduled completion date of December,199 ) EQ Equipment Walkdowns I Corrective actions for the escalated enforcement actions identified in NRC Inspection Report numbers 50-325, 324/96-14 included performance of walkdown inspections to determine if EQ equipment was installed in accordance with the QDPs. During the EQ equipment walkdowns, three Rosemount transmitters (numbers 2G31-FT-N012, NO36, and N041) with improperly installed seals were identified on the Unit 2 reactor water cleanup system. The inspectors reviewed the EQ component field verification data sheets which document the licensee's inspections performed on January 28, 1997. The seals had been installed at the terminal box end of the flexible conduit instead of adjacent to the instrument itself as the QDP required. The flexible conduit is not considered qualified to provide a moisture tight barrier and prevent moisture from intruding into the Rosemount transmitters. This problem was documented on CR 97-00436 which was issued on January 29, 1997, after the field data was evaluated by licensee engineers. Licensee engineers ;etermined that the remaining Rosemount transmitters which were required to be EQ were properly sealed. The licensee removed the Unit 2 RWCU from senrice and issued three work requests to correct the problem and return the system to service. The inspectors reviewed the work requests, numbers WR/JO 96-AJMG3, -AJMJ4, and AJMJ5, which were initiated to install the seals at the proper location. The inspectors examined the three instruments, Rosemount transmitter numbers 2-G31-FT-N012, -N036, and-N041 and verified new seals had been installed adjacent to the instruments. The inspectors walked down the Units 1 and 2 reactor buildings on elevations 20 and 50 and verified that Rosemount transmitters designated as EQ were properly seale l The inspectors identified numerous Rosemount transmitters that were installed with no seals adjacent to the instruments. The inspectors checked ten of the instruments and determined through review of EDBS that the instruments were not required to be environmentally qualified. However the licensee determined that some of these I instruments may involve an associated circuits issue. The inspectors reviewed ESR I numbers 9600704 and 9600715 which the licensee prepared to evaluate excluding these components from the EQ program.
Review of Condition Reports disclosed that an issue was identified regarding an unqualified EQ seal configuration in replacement of an ASCO solenoid valve in ESR 9501767. This problem, which was identified on October 16, 1996, was documented in CR 96-03272, dated October 17,1996. The i procedures in the ESR specifically directed implementing personnel to install a seal at the junction box end of a Oexible conduit, instead of at the instrument.
The flexible conduit was not qualified. The licensee's corrective actions include revising the installation instructions in the ESR to obtain an EQ qualified installation and to perform walkdowns to determine if similar unqualified installations existed in the plant. These walkdowns were included .
with those discussed abov l The inspectors questioned the licensee regarding past operability of the RWCU system and whether this problem had been evaluated for reportability under 10 CFR 50.73. These discussions disclosed that the problem was evaluated and determined not to be significant enough to be reported as a four hour repor ;
However, review of the 50.73 evaluation for this issue and discussions with !
licensee engineers disclosed that the licensee failed to perform the required reportability evaluations. CP&l. procedure ORCl-06.1, Reportable Event Evaluation Criteria and Processing, Revision 13, dated June 20,1995, requires potentially reportable events to be evaluated to determine if they are required to be reported to NRC under 10 CFR 50.73. The failure to perform the evaluation was identified as Violation item 324/97-03-01, Failure to Perform a Reportable Event Evaluation for Past Operability of the Unit 2 RWCU due to improperly Installed Seals for Rosemount Transmitters.
3) Review of ESR 9700087 After the problem discussed above was identifbd, the licensee adjusted their EQ equipment walkdown schedule to inspect instrumentation and other components which required seals to protect the equipment from moisture intrusion. The inspections identified approximately 80 ASCO Tripoint pressure switches with improperly installed conduit seals (i.e. seals were installed at terminal box end of the flexible conduit). This problem was documented in CR 97-00508. A JCO was issued in ESR 9700087 on February 3,1997, to address the as-found conduit seal configuration for the ASCO Tripoint pressure switches and other similar components, such as excess flow check valves, which also required conduit seals. The inspectors reviewed the JCO and questioned licensee engineers regarding a temperature discrepancy in the JCO regarding the qualification of the NAMCO limit switches, and whether a short
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d circuit in the excess flow check valves would be an associated circuits issu After performing a walkdown inspection in the Units 1 and 2 reactor buildings, the inspectors also questioned licensee engineers regarding the type and identification of the flexible conduit installed. The inspectors noted during the i inspection that at least two different types of flexible conduit had been installed
. and some flexible conduit had been painted so that identification of the type / materials was not possible. Discussions with licensee engineers disclosed that the conduit had been purchased as non-safety related material In response to the inspectors' questions, and questions from other NRC staff, the licensee revised the JCO and issued ESR 9700087, Revision 1 on February 12,1997. The revised JCO only addresses the ASCO Tripoint pressure switches and provides additional specific test data that shows they can be qualified with the existing seal configuration. Other types of components were not included in the revised JCOs since additional inspections by licensee EQ personne! have not identified any new seal installation problems. The licensee will install new seals adjacent to the instrument, as required by the QDP, as a long term corrective action. The preliminary schedule is for this work to be completed by July 199 Review of Revision 1 of ESR 9700087 disclosed that the licensee has performed a detailed evaluation of flexible conduit to determine its effect on the operability of the ASCO instruments. This evaluation included identification of all types of flexible conduit possibly used at Brunswick and the effect of the accident temperatures on the integrity of the various types of flex conduit The licensee determined that the flexib!e conduit was qualified to prevent any moisture intrusion into the ASCO Tri-Point Pressure Switches. However, as !
stated above, the long term corrective action will be to install seats in the !
correct locatio ) Review cf the Reactor Building Environmental Report The inspectors reviewed Revision 5 of the Reactor Building Environmental Report (RBER), dated October 2,1996. The purpose of this report was to confirm environmental pressure and temperature profiles including the power uprate madification and other modifications implemented since Revision 4 of the RBER was issued. The RBER was prepared using the guidelines of NUREG - 0588. The licensee is in the process of updating the QDPs to include consideration of the data from RBER, Rev. 5. The licensee has also performed a review of the impact of the revised environmental data on emergency operating procedures (EOPs). The affected EOPs are the .
secondary containment control procedures, EOP-03-SCCP, Revision 5, and l EOP-01-UG, Revision 24, and EOP-01-SEP-04, Revision 6, Reactor Building HVAC Restart Procedure. The RBER shows a revised upper temperature limit of 310' F. The current secondary containment control procedures specify an operator action to initiate a reactor scram based on a reactor building temperature of 295* F to protect EQ qualified equipment. The current procedure is conservative. Discussions with licensee engineers disclosed that j i
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the procedures may be revised to reflect the higher temperatures when the QDPs are updated to include consideration of the higher temperature c. Conclusions The inspectors concluded that the licensee's progress to correct the EQ program deficiencies was progressing satisfactory. Equipment operability issues were appropriately evaluated through JCOs. Additional followup inspections will be performed to review and inspect EQ issues. A violation was identified for failure to perform a reportable event evaluation for past operability of the Unit 2 RWCU due to improverly installed seals for Rosemount Transmitter E.2 Engineering Support of Facilities and Equipment a. Scope The inspectors reviewed the licensee's system for processing and evaluating vendor information and the licensee's system for processing information reported by vendors in accordance with 10 CFR Part 2 '
b. Observations and Findinas The inspectors reviewed the handling of information received from various vendors in the form of Services information Letters (SILs), Rapid Information Communication Services Ir. formation Letters (RICSILs), Technical information Letters (TILs), and other forms of vendor communication. This process was controlled by CP&L procedure EGR-NGGC-6, Vendor Manual Program. Vendor information was received by or routed to the vendor manual coordinator at each site. On receipt, the vendor manual '
coordinator initiates an ESR for initiation of engineering review and evaluation for applicability to the Brunswick site. The ESR was screened and assigned to the appropriate system engineer for disposition. The inspector reviewed the system and the evaluation of a recent SIL, and found it to have been properly reviewed and dispositioned. The inspector noted that the vendor manual coordinator was knowledgeable of the system and proces ;1 The inspectors also reviewed the process for evaluation of notifications in accordance with 10 CFR Part 21, Reporting of Defects and Noncompliance. This process was defined within Regulatory Compliance Instruction ORCI-6.1, Reportable Event Evaluation Criteria and Processing, and Plant Program Procedure OPLP-4, Corrective Action Management. These two procedures provided the definition and guidance for the identification of potential Part 21 issues. Attachments 4,5, and 6 to ORCl-6.1, contain the specific evaluation and screening guidance necessary to determine if an event meets the reportability requirements of 10 CFR part 21. OPLP-4, defines the subsequent process for review and evaluation of the identified problem, including the issuance and disposition of the related Condition Report (CR) identifying the proble The inspectors reviewed the process and determined that it provided adequate guidance for the identification of these issue .
in review of the Part 21 process, the inspectors noted that while the procedural guidance existed for the identification and evaluation of potential Part 21 issues at the plant, similar detailed guidance for the handling and disposition of Part 21 notifications received by the plant did not exist. This process was handled and tracked by the Part ,
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21 coordinator in Regulatory Affairs, who serves as the clearing house for the receipt, evaluation and tracking of Part 21 notifications. On receipt of a Part 21 notification, the report was screened for applicability to Brunswick. If applicable, a CR was initiated to track the evaluation of the item. The issue was then assigned to the appropriate group for further evaluation and disposition. Final determination and closure of the Part 21 was then retumed to Regulatory Affairs for documer,tation and retention in the Part 21 database. The inspector reviewed this process, reviewed the evaluation of a recent Part 21, and found the personnel involved to be knowledgeable and competent in their processing of this information. The inspector also noted that the licensee was currently developing a procedure to formalize this proces I Conclusion The inspectors reviewed the processes the licensee utilized for the handling, 1 identification and reporting of vendor information and Part 21 notifications. The
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inspectors determined that the licensee had established acceptable procedures for the receipt, evaluation and disposition of vendor supplied information. Additionally, the licensee had a defined process for the identification evaluation and disposition of potential Part 21 issues raised on site. An adequate process existed for the evaluation and disposition of Part 21 notifications received by the plant from offsite sources such as vendors or other facilities. However, this process was not contained in a procedure. The licensee was in the process of developing a procedur E5 Engineering Staff Knowledge and Qualification E5.1 Trainina and Qualification of System Enaineers Inspection Scope The inspectors reviewed the licensee's program for training and qualification of plant (system) engineering personne Observations and Findinos The inspectors discussed the licensee's program for training and qualification of engineers with the BESS training coordinator. The inspectors reviewed the status of the system engineers' qualification program. This review disclosed that, for assignment of primary system engineers, more than half of the systems are assigned to fully certified system engineers, while the remaining are assigned to system engineers who are in the process of becoming either qualified or certified on their assigned systems. The inspectors reviewed the schedule for completion of the certification process for plant engineers on the remaining systems. The schedule shows that the majority of the primary system engineers will be fully certified on their systems by the end of 199 .
15 Individual training schedules have been developed for all NED engineers which document required training and the scheduled completion dates for the ,
training. The inspectors also reviewed the qualifications of engineers in the
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EQ task force. These included four CP&L direct employees and six temporry i contractor employees. The records indicated that the personnel involved in the !
EO program were well qualified in the EQ are I Conclusions l The inspectors concluded that the licensee's program for training and qualification of system engineers meets NRC requirement Engineering Organization and Administration E6.1 BESS Calculation Reconciliation Plan inspection Scope in 1995 the licensee issued Procedure EGR-NGGC-0304, Maintenance of Design Documents, Revision 0, effective 11/30/95, which required updating l Category A calculations prior to modification turnover to operations or within 10 I days of closure of non-modification ESRs. Prior to the new procedure,1056 calculations existed with operable plant change documents outstanding against them that were then considered overdue. The licensee initiated a plan to eliminate this backlog in a controlled manner to ensure that the administrative details were complete and accurately capturW in the Nuclear Revision Control System (NRCS) arid that the technical accuracy had not been compromise ;
The inspectors reviewed approximately 5 percent of 1056 total open or unresolved calculation Observations and Findinas l The licensee established 5 categories for these calculations to resolve this issue. Three categories, A1, A2, and A3 were identified under the general category of reserved or voided calculations and two categories, B1 and B2, were identified under the general category of existing approved calculation Category A1 calculations were identified as void, or void and superseded. An administrative change was required to void / supersede these calculation Category A2 calculations were identified as reserved for plant change documents that were closed or operable. Therefore, an administrative roll up was required. Category A3 were reserved calculation numbers with no references to change documents. The calculations (reserved numbers) were l canceled. Category B1 calculations were identified as outstanding changes with plant modification calculations as part of the modification. The base calculations were updated by including the plant modification calculation as an attachment to the base calculation. This was an administrative functio Category B2 calculations were listed in NRCS as being impacted by a change document but no specific plant modification calculation was developed. The I
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1 licensee identified this effort as both administrative and technical. At the time l of the inspection there were 58 open calculations in Category B i The inspectors reviewed a sample of 39 calculations to ensure that l administrative personnel were not making technical changes and that the l resulting calculation packages wer 2erstandable. The licensee indicated that the calculations were updateo, if necessary, using Procedure OENP-303, J Preparation and Control of Design Analyses and Calculations, Revision 0, 12/31/9 I The inspectors noted that it was very difficult to follow the review process for i some of these calculations since some details were not addressed in the procedure. The following items were observed:
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The list of affected pages were not always updated to reflect Revision 0 for each sheet even though the calculation was considered a Revision 0 level calculation.
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Some calculation sheets indicated that the sheet was preliminary (in !
some cases, there was no indication of preliminary, final, or void) even l though the calculation was considered a final calculatio Some attachments did not contain the calculation number and, in some cases, contained other calculation number l l
The inspectors determined that the calculations were rolled up appropriately except for the following items:
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Calculation SA-E11-545, Revision 0 - Appendix E contained 4 page However, the Table of Contents indicated 3 pages and the cover sheet to Appendix E showed 2 pages. The licensee advised the inspectors that the last page of Appendix E was actually a walkdown redline sheet for a different calculation that should not have been part of Appendix Therefore, the Table of Contents was correct. The licensee removed the redline sheet from the calculation and revised the number of pages on the cover sheet of Appendix Calculation PS-E11-002, Revision 1 - A design verification sheet for an attached calculation to the base calculation did not have the date that the responsible engineer signed this sheet. The design verifier did sign and date this sheet correctly. The licensee had the responsible engineer initial and date the form and the design verifier also initialed I and dated the for Calculation MSR-0001, Revision 0 - The cover sheet to Attachment D indicated the wrong number of pages in this attachment. The licensee cc octed the cover shee .
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17 Conclusions i
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In general, the reconciliation plan adequately resolved the outstanding l calculations in the hacklog. However, the inspectors noted some minor !
discrepancies in the calculations due to a lack of attention to detail, which the licensee addresse E6.2 UFSAR Review Inspection Scope The inspectors reviewed a portion of Chapter 8.3.1 of the UFSAR and the UFSAR Chapter 8.3.1 and 8.3.2 drawings to determine if the licensee was I maintaining the UFSAR in accordance with the latest plant configuration. The text review was limited to a review of the Diesel Generator protective relaying )
and a comparison of the text to the one line diagrams. The inspectors mmpared the drawings in the UFSAR that were issued with Amendment 13, dated 11/21/95, with the one line diagrams that were the latest revisions at the time of the Amendmen Observations and Findinas l The text of the diesel generator protective relaying matched the latest one line diagrams. In general, the UFSAR drawings matched the plant configuration at the time of the last UFSAR amendment. However, the following items were i noted by the inspectors: '
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In a number of cases the latest revision of the drawing at the time of the amendment was not consistently shown on each UFSAR diagra Plant Drawing Correction Traveler PDC 91-1021, dated 9/27/94, was not incorporated into Figure 8.3.2-6. The licensee stated that it was not necessary to incorporate this change into the figure since the traveler was marked "No" for "FSAR Change Required?". The inspectors reviewed Adverse Condition Report (ACR) 94-02082, dated 12/2/94, that determined that a number of UFSAR figures were technically inaccurate. One of the causes identified was the lack of a safety review of PDCs that had been issued. Therefore, the PDC process was indicated as weak. The corrective actions for the ACR did not include a review of previously issued PDCs to determine their impact on the UFSA Plant Drawing Correction Traveler PDC 91-1017, dated 1/6/95, was not incorporated into Figure 8.3.2-4. The engineer who checked the PDC stated that it was not necessary to change the UFSAR figure since it was not a functional chang .
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Plant Drawing Correction Traveler PDC 90-0216, dated 11/10/90, was incorporated into Drawing F-03028, Revision 18, dated 11/21/9 However, UFSAR Figure 8.3.2-7 was not updated to the latest drawing revision (Revision 22, dated 11/10/94) when Amendment 13 was issue NRCS did not list ESR 9600017 as an open document against Figure 8.3.1-9. The licensee initiated CR 97-01039 to identify this item and corrected ESR 96-00017 and NRCS on 3/10/97 to note that UFSAR figure 8.3.1-9 was affected by the ESR. The licensee determined that the preparer and safety reviewers of the ESR incorrectly used a temporary modification as the basis for the safety evaluation for ESR 9600017. The temporary modification was evaluated in ESR 9400712, but did not identify a change in breaker labeling revised by ESR 9600017. CP&L Procedure OIA-109 specifies the requirements for performance of safety reviews. Section 5.4 of OIA-109 requires use of Attachment C of OlA-109 to classify items affected by any proposed activity / change to the facility. Paragraph 5.4.1 of OIA-109 and Question 1 of Attachment C to OIA-109 asks whether the item (ESR)
requires a revision to the UFSAR. This question was incorrectly answered "No" when Attachment C was completed for ESR 960001 The failure to correctly complete the safety review was identified to the licensee as Violation item 325,324/97-03-02, Failure to Incorporate an Engineering Service Request in a Change to the UFSAR. The licensee initiated CR 97-01039 to document this problem. The inspectors determined that the ESRs related to this issue did not involve an unreviewed safety questio Conclusions '
The licensee has identified problems in maintaining the UFSAR up-to-dat However, the licensee has initiated a program to update the UFSAR. The next amendment to the UFSAR was scheduled to be completed in April,1997. A violation was identified for failure to incorporate an Engineering Service Request in a change to the UFSAR. Quality Assurance in Engineering Activities E.7.1 Licensee Self Assessments Inspection Scope The inspectors reviewed two self assessments which were performed in the engineering support area during 199 ,
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19 b. Observation and Findinas The inspectors reviewed the self assessments listed below to determine the adequacy of the assessments and the adequacy of the corrective actions. The .
self assessments reviewed were as follows: l
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Self-Assessment 96-03760, Electrical Distribution Change Control, was performed on August 21 and 22,1996 by personnel from the Corporate 'j Nuclear Engineering Department. The scope included a review of the l voltage drop and load flow study of the AC electrical distribution system 1 for Brunswick. The self assessment identified 3 findings that needed to be addressed by site engineering. Another item for management consideration was also identified. BESS issued CAPS 96-03789, 11/14/96, to respond to the self assessment findings. The action plan addressed all of the findings in a reasonable manner including Finding 3 which addressed a non-conservative number listed in a calculation for the minimum criteria voltage for MCC 1XA-2. The plan included a review of the equipment connected to this MCC to ensure e mthis equipment was not affected by correcting this minimum .tag l
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Self-Assessment NED 96-02, Design Control Unit Assessment, was l performed by personnel from the Corporate Nuclear Engineering Department. The scope of the assessment was to review the functions of the Design Control Units (DCU) at each CP&L site. This self ;
assessment was initiated at the request of the Engineering Support l Section managers at Brunswick; Harris and Robinson to assess the effectiveness of the DCU and to identify areas where improvement was needed. The self assessment identified 1 issue and 3 weaknesse The issue indicated that the Design Control Unit at Brunswick was being used as staff augmentation personnel for BESS when there was a excess of work for the BESS. The weaknesses were as follows: The DCU evaluation criteria differ between plants, resulting in a widely varying depth and scope of ESR reviews; NGG common procedures are not on the information search system, and no personnel rotation plan exists for DCU personnel. Although Condition Reports were not written at the time of the assessment to track the closure of items identified during the self assessment, corrective actions have been initiated to correct the findings. Some of these issues are being addressed in the licensee's reorganization as described in BESS Organizational Proposal, Revision 2, dated February 24,1997,and revisions to applicable procedures. The licensee initiated three CRs, numbers CR 97-00968, -00969, and -00970, on March 11,1997 to document corrective action Integrated Performance Self-Assessment was performed in Summer, 1996, using the NRC IPAP process. The inspectors reviewed the Engineering Section of the report and findings identified in this are Significant findings included that system engineering effectiveness has
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been reduced by over emphasis on system engineers becoming qualified to perform design work, that self assessments have not been effective in consistently producing improvements, and that a clear sustained improvement in quality of engineering work has not been establishe Conclusions The self assessments were adequate in evaluating the identified scope of the assessments. However the licensee did not follow the process of issuing CRs in a timely manner to document findings in the DCU assessment. This issue was not identified as a violation since the purpose of the self assessment was to identify weakness in the DCU organization and the licensee initiated I corrective actions to resolve the findings. However the failure to properly j document the issues was identified as a weakness. The findings of the I licensee's IPAP indicate continuing problems with their corrective action program Miscellaneous Engineering issues E.8.1 (Closed) Unresolved item 50-325,324/96-14-02, UFSAR Environmental Data l Discrepancie Review of updated environmental data for the reactor building disclosed that i the temperature data contained in the UFSAR did not reflect current plant conditions. The resident inspectors documented a similar issue pertaining to the drywell temperature data in NRC Inspection Report numbers 50-325, 324/96-05 as Unresolved item number 325,324/96-05-02. The licensee has updated the Reactor Building Environmental Report as part of the corrective actions to address the civil penalties / violations for the EQ program deficiencies identified in NRC Inspection Report numbers 50-325, 324/96-14. The updated environmental data will be an input into the QDPs. The UFSAR will be amended by the licensee to reflect the revised environmental data. The licensee has initiated a project to revise the UFSAR. Since the discrepancies in the environmental data were identified by NRC as examples of the EQ violations and correction of the discrepancies are part of the licensee"s corrective actions, URI number 325,324/96-14-02 will be closed. The licensee has initiated a program to correct and update the UFSA E.8.2 (Closed) Unresolved item 50-325,324/96-14-03, Effect of RBCCW System Operability on PAS The installation of the Post Accident Sampling System (PASS) was required by NUREG 0737 and by an Order dated July 10,1981 for implementation and maintenance of Three Mile Island action items. PASS is addressed as item ll.B.3. Requirements for training of personnel, maintenance of PASS equipment, and procedures for sampling and analysis from the PASS are specified in Technical Specification 6.8. _
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21 The PASS is defined as a Regulatory Guide 1.97 Category 3 system which
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specifies that the system should be high quality commercial grade selected to ;
withstand the specified service environment. PASS is not safety relate ;
However, the licensee did commit to provide some environmentally qualified {
valves in the system and to design the system to meet seismic design ;
requirements. These issues were reviewed during the inspection documented l in NRC Inspection Report numbers 50-325, 324/96-14. Review of operating procedures and the system design documents for PASS showed that the Reactor Building Closed Cooling Water System (RBCCW) is required to be operable to obtain samples from PASS. NRC Criterion 3 for design of the post l accident sampling system states " Reactor coolant and containment 1 atmosphere sampling during post-accident conditions shall not require an i isolated auxiliary system (e.g., the letdown system, reactor water cleanup system (RWCU)) to be placed in operation in order to use the sampling system. In response to questions from NRC, the licensee stated in a letter to NRC dated January 28,1983, Subject: NUREG 0737 Item ll.B.3 Post-Accident Sampling implementation Submittal, that they would comply with the above criterion. The response states sample availability does not depend upon operation of any isolated system. In Enclosure 1 to a letter from NRC to the PASS Owners' Group, dated May 3,1990, under item 19, an isolated auxiliary system was defined as a system which could transport highly radioactive fluids outside of primary containment. Therefore, the RBCCW system was not considered an isolated system per the above referenced Criterion 3. The inspectors concluded that design of PASS met the requirements of RG 1.97 and the licensee's commitments to NRC.
i In certain accident scenarios, the RBCCW system will not be available since the RBCCW pumps will trip and service water which is used to cool RBCCW will be isolated. The RBCCW system is not a safety related system and was not required to be single failure proof. The inspectors discussed restoration of RBCCW following an accident with operations personnel and reviewed operating procedures: 1-OP-21 and 2-OP-21, Unit 1 and Unit 2 RBCCW System Operating Procedures, and AOP-16.0, RBCCW System Failur Although the RBCCW is not required to safely shut down the reactor following a LOOF or LOCA, the licensee has procedures and operators are trained to restore RBCCW, when conditions permit. In addition to PASS, RBCCW also provided cooling water for the spent fuel pools, drywell chillers, the reactor water cleanup system and CRD pumps. None of these systems are required to shut down the reactor following an accident. There are attemate sources of cooling available for the spent fuel pool. These were evaluated by NRC during an assessment conducted March 6 - 10,1995, which was documented in a report attached to a letter from NRC to CP&L, dated May 24,199 The inspectors also reviewed the evaluation performed by licensee engineers to determine the effect of restart of the RBCCW pumps on the diesel generators. The evaluation was documented in the following calculations:
DBCF numbers BNP-E-7.0004-0005, BNP-E-7.007-0005, and BNP-E-7.010-0005. The calculations showed that the RBCCW pumps will not overload the . ._ _ . . _ _ - . _ _ . _ . . _ _ _ _ _ . .__ _ _ _ _ . __ _
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l 22 diesels if they were restarted one hour after an event. The inspectors reviewed training performed by E&RC personnel to practice obtaining samples from PASS on December 12,1996. Licensee personnel demonstrated they
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could obtain and analyze a sample within the three hour time period specified in their commitment to NR The inspectors concluded that use of RBCCW to cool the samples obtained )
from PASS complies with the PASS design criteria, and meets the licensee's l
3 commitments to NRC for design and operation of PASS. The inspectors also concluded that the licensee complies with the requirements of the Technical
- Specification V. MANAGEMENT MEETINGS l
The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on March 14 and April 4,1997. The licensee acknowledged the findings presented. Dissenting comments were not re ,eived from ;
the license The licensee did not identify any materials used during the inspection as proprietary informatio PARTIAL LIST OF PERSONS CONTACTED
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Licensee J. Cannon, Supervisor, Electrical Systems, BESS W. Campbell, Vice-President, Brunswick J. Franke, Superintendent, BOP Systems, Brunswick Engineering Support Section (BESS)
J. Gawron, Manager, Nuclear Assessment Section J. Gee, Supervisor, Configuration Control, Design Control, BESS L. Grzeck, Project Engineer, BESS K. Jury, Manager, Regulatory Affairs W.' Levis, Director, Site Operations J. Lyash, Manager, BESS R. Lopriore, Plant Manager R. Miller, Superintendent, Design Control, BESS C. Pardee, Manager, Operations R. Schlichter, Manager, Environmental and Radiation Control S. Tabor, Senior Speciali" Regulatory Compliance J. Titrington, Supervisc : 2.54f Review, Licensing M. Turkil, Manager, Lice... sing and Regulatory Programs R. Williams, Manager, EQ Task Force, BESS Other licensee employees included office, maintenance, engineering, and chemistry personne . _ . . . - - - - _
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NRC E. Brown, Resident inspector M. Janus, Resident inspector J. Lenahan, Reactor inspector C. Patterson, Senior Resident inspector D. Trimble, NRR Project Manager J. Mallanda, Beckman and Associates J. Williams, 'Beckman and Associates INSPECTION PROCEDURES USED IP 37550: Engineering IP 92903: Followup - Engineering ITEMS OPENED, CLOSED, AND DISCUSSED l
Opened 50-324/97-03-01 VIO Failure to Perform Reportable Event Evaluation for Past Operability of the Unit 2 RWCU due to improperly Installed Seals in Rosemount Transmitters (Paragraph E1.4)
50-325, 324/97-03-02 VIO Failure to incorporate an Engineering Service Request in a Change to the UFSAR (Paragraph E6.2)
Closed 50-325, 324/96-14-02 URI UFSAR Environmental Data Discrepancies (paragraph E8.1)
l 50-325, 324/96-14-03 URI Effect of RBCCW Operability on PASS l (Paragraph E8.2) I ACRONYMS AC -
Alternating Current ACR -
Adverse Condition Report ALARA -
As Low as Reasonably Achievable ANSI -
American National Standard i ASME -
American Society of Mechanical Engineers i BESS -
Brunswick Engineering Support Section '
BNP -
Brunswick Nuclear Plant CR -
Condition Report DC -
Direct Current DOR -
Division of Operating Reactors EDBS -
Equipment Data Base System
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1 EER -
Engineering Evaluation Report EQ -
Environmental Qualific ation ESR -
Engineering Service Request GE -
General Electric Company ISI -
Inservice Inspection IEEE -
Institute of Electrical and Electronic Engineers JCO -
Justification for Continued Operation LOCA -
Loss of Cooling Accident LOOP -
Loss of Offsite Power MCC -
Motor Control Center NED -
Nuclear Engineering Department NEMA -
National Electrical Manufacturers' Association PASS -
Post Accident Sampling System PDC -
Plant Drawing Change QDP -
Qualification Data Package RBCCW -
Reactor Building Closed Cooling Water RBER -
Reactor Building Environmental RG -
Regulatory Guide RWCU -
Reactor Water Clean-up System UFSAR -
Updated Final Safety Analysis Report URI -
Unresolved item VIO -
Violation V -
Volts WR/JO -
Work Request ,
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