ML20138J598
| 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) | |
See also: IR 05000324/1997003
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U. S. NUCLEAR REGULATORY COMMISSION
REGION ll
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Docket Nos: 50-325, 50-324
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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
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i Enclosure 2
9705000271 970502
O ADOCK 05000324
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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 problems.
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 Ill.
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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 transmitters.
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A violation was identified for failure to incorporate an engineering service
request in a change to the UFSAR.
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In general the modifications packages were of good quality and would not degrade
plant performance, safety, or reliability.
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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 Support.
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A weakness was identified for not properly documenting corrective actions to
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resolve findings from self-assessments.
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The licensee's progress to correct the EQ program deficiencies was satisfactory.
- Equipment operability issues were appropriately evaluated through JCO's.
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Enclosure 2
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REPORT DETAILS
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! 111. Enaineerina
1 E1 Conduct of Engineering
E1.1 Desian Chance Control Processes
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s. Inspection Scope
The inspectors reviewed the licensee's procedures which control the design change '
program,
b. 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
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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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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).
c. 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 111. 1
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E1.2 Review of Desian Chanaes and Modification Packaaes I
a. 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
accomplished.
b. Observation and Findinos
The Engineering Service Requests (ESRs) discussed below were reviewed for
adherence to procedures, codes and standards, commitments, and technical
adequacy.
1) 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 allowables.
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 walls.
The analysis referred to ANSI B31.1 and used material properties from a B31.1
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 %".
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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
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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 corrosion.
3) 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
4) 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 starters.
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 panel.
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
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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
hours 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 information.
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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
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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 year.
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 affected.
6) 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 source.
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 operations.
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
adequate.
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 ESR.
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 0.2
ampere tap. The analysis used 8.3 ohms. This change did not change the
results / conclusions of BNP-E-6.071-0002/BNP-E-6.075-0002.
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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 coordination.
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 breakers.
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 EDBS.
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 documents.
8) 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 close.
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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 adequate.
9) 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
acceptable.
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-CRN001.
10) 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 installed.
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 modifications.
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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 Backfeed. ;
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ESR 96-00353 - Power Uprate Turbine Controls Modification. ,
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ESR 96-00546 - Disable B32-PS-N018 A & B. '
c. 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 requirements.
E1.3 Environmental Qualification
a. 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 50-325,324/96-14.
b. 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 Manager.
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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,
- marathon terminal blocks and improperly installed seals for ASCO
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Tripoint pressure switches.
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,1997.
2) 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 sealed.
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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 above. 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 report. ;
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
-
\
.
12
d
circuit in the excess flow check valves would be an associated circuits issue.
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 materials.
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 1997.
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 conduits.
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 location.
4) 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
.
I
13
the procedures may be revised to reflect the higher temperatures when the
QDPs are updated to include consideration of the higher temperatures.
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
Transmitters.
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 21. '
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 process. ;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 problem.
The inspectors reviewed the process and determined that it provided adequate
guidance for the identification of these issues.
.
14
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 ,
'
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 process. I
c. Conclusion
The inspectors reviewed the processes the licensee utilized for the handling, 1
identification and reporting of vendor information and Part 21 notifications. The
'
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 procedure.
E5 Engineering Staff Knowledge and Qualification
E5.1 Trainina and Qualification of System Enaineers
a. Inspection Scope
The inspectors reviewed the licensee's program for training and qualification of
plant (system) engineering personnel.
b. 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 1997.
.
4
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
{
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 area. I
c. Conclusions l
The inspectors concluded that the licensee's program for training and
qualification of system engineers meets NRC requirements.
E.6 Engineering Organization and Administration
E6.1 BESS Calculation Reconciliation Plan
a. 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 compromised. ;
The inspectors reviewed approximately 5 percent of 1056 total open or
unresolved calculations.
b. 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 calculations.
Category A1 calculations were identified as void, or void and superseded. An
administrative change was required to void / supersede these calculations.
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 function.
Category B2 calculations were listed in NRCS as being impacted by a change
document but no specific plant modification calculation was developed. The
I
.
.
16
1
licensee identified this effort as both administrative and technical. At the time l
of the inspection there were 58 open calculations in Category B2.
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,
12/31/94.
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:
i
-
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.
,
-
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 calculation.
-
Some attachments did not contain the calculation number and, in some
cases, contained other calculation numbers.
l
l
The inspectors determined that the calculations were rolled up appropriately
except for the following items:
-
Calculation SA-E11-545, Revision 0 - Appendix E contained 4 pages.
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 E.
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 E.
-
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 form.
- 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 sheet.
.
.
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17
c. Conclusions i
!
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 addressed.
E6.2 UFSAR Review
a. 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 Amendment.
b. 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: '
-
In a number of cases the latest revision of the drawing at the time of
the amendment was not consistently shown on each UFSAR diagram.
-
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
-
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 change.
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18
-
Plant Drawing Correction Traveler PDC 90-0216, dated 11/10/90, was
incorporated into Drawing F-03028, Revision 18, dated 11/21/90.
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
issued.
-
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 9600017.
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 question.
1
c. Conclusions '
The licensee has identified problems in maintaining the UFSAR up-to-date.
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.
E.7 Quality Assurance in Engineering Activities
E.7.1 Licensee Self Assessments
a. Inspection Scope
The inspectors reviewed two self assessments which were performed in the
engineering support area during 1996.
,
- . _-, -- . .. _ _ _ . ._ . - -- - .- - .
.
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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
- -
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 .tage. l
-
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 weaknesses.
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 actions.
-
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 area.
Significant findings included that system engineering effectiveness has
'
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20
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
established.
c. 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
programs.
E.8 Miscellaneous Engineering issues
E.8.1 (Closed) Unresolved item 50-325,324/96-14-02, UFSAR Environmental Data l
Discrepancies.
1
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 UFSAR.
E.8.2 (Closed) Unresolved item 50-325,324/96-14-03, Effect of RBCCW System
Operability on PASS.
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 6.8.3.3.
_
_
4
.
1
21
The PASS is defined as a Regulatory Guide 1.97 Category 3 system which
{
specifies that the system should be high quality commercial grade selected to ;
withstand the specified service environment. PASS is not safety related. ;
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 Failure.
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 24,1995.
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
_. . ._ _ . . _ _ - . _ _ . _ . . _ _ _ _ _ . .__ _ _ _ _ . __ _
'
. . .
'.
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
'
could obtain and analyze a sample within the three hour time period specified
in their commitment to NRC.
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
- Specifications.
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 licensee.
The licensee did not identify any materials used during the inspection as proprietary
information.
PARTIAL LIST OF PERSONS CONTACTED
'
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
personnel.
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. .
e
23
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
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
. . .
.
24
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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