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                                                EliCLOSURL2                                                i
                                                                                                          :
                                  U.S. NUCLEAR REGULATORY COMMISSION
                                                  REGION IV
      -
          - Docket Nos.:      50-313;50-368
          License Nos.:      DPR-51; NPF-6
          Report No.:        50-313/97-21; 50 368/97-21
            Licensee:        Entergy Operations, Inc.
            Facility:        Arkansas Nuclear One, Units 1 and 2
]
          - Location:        Junction of Hwy,64W and Hwy.333 South
                              Russellville, Arkansas
            Dates:            October 27-31 and November 10-14,1997
,          inspectors:      M. Runyan, Team Leader
                              P. Goldberg, Reactor inspector
                              R. Bywater, Reactor Inspector
                              G. Kalman, Project Manager
'
          Accompanying      D. Denver, Consu' tant
            Personnel
4
          Approved By:      T. Stetka, Acting Chief Engineering Branch, Division of Reactor Safety
'
          Attachment 1:      Supplemental Information
l          Attachment 2:      List of Documents Reviewed for Unresolved item 50-313/97201-06
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      9902180126 900206
;      PDR      ADOCK 05000313
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      G                  PDR
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                                                                              _ _ _ _ _ _ - _ __ - _-_-__
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                                                  -2-
                                                                                                          *
                                        EXECUTIVE SUMMARY
                                  Arkansas Nuclear One, Units 1 and 2
                            NRC Inspection Report 50-313/97-21; 50 368/97 21
  Engineerina
  -
        Within a review of selected documents, the licensee's 10 CFR 50.59 safety evaluation
        program appeared to be functioning satisfactorily, with the exception that some
        documented word searches of the Updated Safety Analysis Report for impact were
        inaccurate. However, some apparent failures to implement the program were identified
        (Sections E2,1 and E8.22).
  .
        The failure to procedurally incorporate a vendor recommended maximum emergency
        feedwater flow limit of 1500 gpm to a single steam generator into plant operating
        procedures was identified as a violation of 10 CFR Part 50, Appendix B, Criterion V
        (Section E8.7).
  *
        The failure to properly verify the piping configuration design inputs to a design calculation
        was identified as the first example of a violation of 10 CFR Patt 50, Appendix B,
        Criterion lli (Sec' ion EP.8).
  *
        Overpressure of the turbine-driven emergency feedwater pump downstream piping
        required further reviews by the licensee. The existing analysis used pump performance
        crit 6 ia that was less than the manufacturer's specified criteria (Section E8.9).
  -
        The failure to properly verify the design adequacy in four ciesign calculations was
        identified as additional examples of a violation of 10 CFR Part 50, Appendix B,
        Criterion 111 (Section E8.11).
  *
        The lack of a review by the licensee of a sample of existing design calcriations to
        determine if the types of past problems identified by the licensee existed within other
        calculations was identified as a weakness (Section E8.11).
    -
          The lack of over current trip testing of Unit 1 molded case circuit breakers required
          further review by the NRC (Section E8.16).
    -
        - Although the NRC identified five discrepancies in the Final Safety Analysis
          Report (FSAR), the licensee's ongoing review of the FSAR was considered sufficient to
          correct these specific examples and to resolve the resulting generic concern. Therefore,
          enforcement discretion was taken in accordance with the NRC Enforcement Policy
          (Section E8.18).
                          -        ___
 
                                                                                  . _ - _ _ -          - _ _ _ _ _ - _ _ _
  .
  9
                                                                  -3-
    .
            The failure to provide measures to ensure the correct translation of seismic design
            requirements to instrument tubing installations and the maintenance of this configuration
            during plant operations in Unit 1 was identified as an example of a violation of 10 CFR
            Part 50, Appendix B, Criterion lli (Section E8.20),                                                            ,
    .
            Three apparent violations were identified associated with activities involving the removal
            of the borated water storage tank vacuum relief valve for maintenance and testing on
            December 4,1996. The apparent violations involved the failure to perform proper safety
            evaluations as described in 10 CFR 50.59, the failure to perform the vacuum relief valve
            testing under the plant conditions as prescribed in the test procedure in accordance with
            10 CFR Part 50, Appendix B, Criterion V, and the failure to utilize the correct design
            control processes (temporary alterations) for the temporary cover configurations placed
            over the valve flange in accordance with 10 CFR Part 50, Appendix B, Criterion 111
            (Section E8.22).
    -
            The failure to properly check the adequacy of an engineering report that failed to account
            for instrument error in a calculation of vortexing in the borated v<ater storage tank was
            identified as an example of a violation of 10 CFR Part 50, Appendix B, Criterion 111
            (Section E8.23).
    Plant Succort
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    =
            During a plant walkdown, the NRC identified a fire barrier deficiency and the licensee
            took prompt compensatory actions as required (Section F2).
                                                                                                                            1
                                            _        _ - _ _ _ _ _ _ _ .
 
    _ - - _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ -                                -_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _              _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _
  .
  .
                                                                                                                                          -4
                                                                                                                                  Report Details
                                                    Summarv of Plant Status
                                                      Units 1 and 2 were operated at 100 percent power during the inspection.
                                                                                                                              ll1. Enaineerina
                                                      E2 Engineering Support of Facilities and Equipment (37001)
                                                      E2.1    10 CFR 503,J Imolementation
                                                          a.  Insoection Scoos (37001)
                                                              The inspectors reviewed the licensee's program to change plant design, revise
                                                              procedures, and conduct tests and experiments without prior NRC approval as described
                                                              under 10 CFR 50.59. The licensee's program was reviewed for compliance with
                                                              regulatory requirements and implementation of the program was evaluated by reviewing
                                                              three 10 CFR 50.59 evaluations. The inspectors reviewed the following documents:
                                                              *
                                                                      ANO-1 10 CFR 50.59 Summary Report for 1996, deted May 22,1997
                                                              -
                                                                      ANO 210 CFR 50.59 Summary Report for 1996, dated June 19,1997
                                                                50.59 Evaluations:
                                                                *      ANO-1, Procedure Change 95-7023, " Reactor Coolant Pump Starting
                                                                      Temperature"
,
                                                              *      ANO-1, Condition Report 1-96-0667, " Reactor Coolant Pump Lift Pump Disabled"
                                                              *-      ANO-2, Design Change Package 94-2007," Replace Charging Pump              .
                                                                      Downstream Check Valves CVC-22A, B, and C"
,                                                              Procedure 1000,131, "10 CFR 50.59 Review Program," Revision 3
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                                                          b.  Observations and Findings
                                                              -Within the review of the listed documents, the inspectors did not identify any
!-                                                            programmatic deficiencies in the licensee's 10 CFR 50.59 program. It was noted that
                                                              the licensee chose to evaluate the " margin of safety" of proposed changes by evaluating
                                                              the changes only against the basis section of the Techn! cal Specifications. This practice
L                                                              had been previously questioned by the NRC in Inspection Report 50-313:368/97-04 and
!
,
                                                                                                        . - ,                        , , -    -
 
.
.
                                                  ~5-
      found to be in compliance with 10 CFR 50.59 (however, a violation for a failure to
      perform a safety evaluation was issued). No deficiencies in the " margin of safety"
      evaluations were identified by the inspectors during review of the 10 CFR 50.59
      evaluations. Although currently in compliance, the licensee indicated that a more
      encompassing design basis for margin of safety determinations will be implemented in
                                                                                                    *
      the future.
      While review of the?e safety evaluations indicated appropriate programmatic controls,
      some apparent failures to implement the 10 CFR 50.59 program were identified as
      discussed in Section E8.22 of this report.
      The inspectors also identified problems with the licensee's search capability for the
      U) dated Final Safety Analysis Report. For example, Safety Evaluation 96-141 stated,
      "The lube oil pumps are not describad in the Technical Specifications, Safety Analysis
        Report, or operating license." The inspectors determined that this was in error. The lube
        oil pumps were discussed in Section 4.2.2.6 of the Updated Final Safety Analysis Report.
        The inspectors identified similar problems with the safety evaluatici determinations for
        the procedure changes. The inspectors discussed these findings with the licens6e and
        concluded that the automated search methodologies used to search for key words in
        licerwag basis documents were not always effective in finding relevant reference
        matenal and that the techt,ical reviewers were not ensuring that the scope of review was
        adequate. The licensee agreed with these conclusions and informed the inspectors that
        they had also identified this issue in the past year and had revised the safety evaluation
        procedure to address the deficiencies. The inspectors review of Procedure 1000,131
        indicated that improvements had been made in the Lensee's process to ensure that
        adequate automated searches were performed and that they were reviewed by a second
        individual.
  c.  Conclusions
        Within the review of selected documents, the licensee's 10 CFR 50.59 safety evaluation
        program, when implemented, appeared to be functioning satisfactorily, with the
        exception of Updated Final Safety Analysis Report key word searches. However, some
        apparent failures to implement the program were identified as discussed in
        Section E8.22 of this report.
  EB Miscellaneous Engineering issues (92903)
  E8.1  (Ocen) Unresolved item 50-313/9623-01: Consideration of Multiole Hot Short Actuations
        Backaround
        This issue addressed the unplanned energizaiion of motor-operated valves (MOVs)
        during a control room fire. This event, caused by a control cable shorting to an
          energized conductor, is known as a hot short. Under certain configurations, a hot-
          shorted MOV could stroke with the torque or limit switch removed from the control circuit,
          causing the valve to stallinto its open or closed seat. Depending on the capacity of the
                                                                                                  .
            _ _ - . - _ _ _ _ _ _ . - - _ _
 
                                        - - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
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  .
  .
                                                                        3
    motor actuator and the strength of the valve, this action could damage the valve
    sufficiently such that plant operators would not be able to reposition the va!ve using the
    valve's manual handwheel. The licensee's emergency operating procedures addressing
    a control room fire and subsequent evacuation of the control r00m assumed that muual
    control of accessible valves was available The licensee interpreted existing regvations
    tu require postulation of only one hot short during a control room fire, whereas the NRC
    maintained that multiple hot shorts, which is defined as more than one MOV having a hot
    short et the same time, were within the scope of the regulations. Independent from the
    resolution of this matter, at the request of the NRC, the licensee agreed to perform a
    calculation to determine which valves would be potentially susceptible to hot short
    damage. This was the anly aspect of this issue reviewed during the current inspection.
    Insoector Followoo
    The inspectors rev.ewed Engineering Report 97-R-0004-01, "Engineenng Information
    Notice 92-18 Evaluation," Revision 1 PC-1; Calcalation 96-E-0065-01, " Unit 1 Safety
    Related Valve Survivability Evaluation Under Stall Thrust / Torque, Motor Operated
    Valves," Revision 1; and Calculation 96-E-0066-01, " Unit 2 Safety Related Valve
    Survivability Evaluation Under Stall Thrust / Torque, Motor Operated Valves," Revision 0.
    The inspectors reviewed Calculation 96-E-0065-01. The purpose of this calculation was
    to determine operability with reduced margin for a select group of safe shutdown MOVs
    under stall thrust / torque conditions. The evaluations were intended to reflect one-time
    operation under the stated conditions and were not intended for use as design input in
    other analyses or calculations. An important assumption stated that the anelvsis was
    valid only for the time it was performed since actual test data was used in soine
    evaluations. The results showed that 37 of the 66 valves evaluated would be able to be
    manually positioned to their safe position following postulated hot short conditions.
    Twenty-nino valves were susceptible to damage. The inspectors determined that the
      licensee had used a valid methodology for determining the maximum torque and thrust
      that an MOV would experience during a hot short event, with one exception For MOVs
    that did not have previous measurements of stem friction, the licensee used a plant
      averaoe stem factor correlated to a stem friction coefficient vf 0.1342. The use of a
      bounding stem friction coefficient for the untested valve population, approximately 0.08,
      would have resulted in calculated torques and thrusts as much as 50 percent higher than
      those reported in the calculation. However, based on a qualitative judgement by the
      inspectors, application of the lower stem factors would not have significantiv changed the
      number of valves determined as potentially incapable of being manually repositioned.
      Engineering Report 97-R-0004-01 addressed multiple hot shorts, but in situations where
      two or more valves were determined to be unable to be manually repositioned, the report
      stated that it wat not credible that both or all would fail at the same time in the same
      direction. The inspectors noted that this reasoning reiterated the licensee's position that
      multiple hot shorts were not required within the regulations.
        Based on the results of the calculations and engineering report listed above and in light
        of the 29 valves in Unit 1 analyzed to be susceptible to hot shorts, the inspectors
 
        -  -          -    -          - - _ .    - -    - . - -        .    _  --- - - .            - - .
  ':
;    .
,
                                                          -7-
                concluded that come safety functions could be lost if mu:tiple hot shorts occurred in a
                particular manner. This was due to the fact that some of the motor operated valves may
                be subjected to loads during a hot short that would render them incapable of being
                manually stroked to their safe shutdown position. In some cases, as in the Unit 1 high
              - pressure injection system, a total of eight valves would have to hot short in the same
                direction to defeat the safety function of the system. However, the Unit 1 shutdown
                cooling function could be lost with the hot short failure of only two vu.ves, CW1428 and        -
                CV 1429, decay heat cooler outlet valves. Both valves were catulated to be potent! ally
                incapable of being manually repositioned following a hot short event.
;              The above discussion was illustrative of the potential impact of multiple hot short events
                and was not a complete listing of all potentially affected systems. This item will remain
                open pending further reviews by the NRC program office to define the regulatory
                requirements for this matter.
          E8.2  (Closed) Violation 50-313/96027-03: Failure to Notify the NRC Within 1 Hour of
                Declaration of Notification of Unusual Event
L              Backaround
                On October 17,1996, the licensee declared a Notification of Unusual Event in
                accordance with its Emergency Plan for n fire in the Unit i reactor building that lasted
!              more than 10 minutes. The licensee notified the NRC Operations Center of a fire in the
                reactor building within 1 hour of the declaration of the Notification of Unusual Event.
'
                Howover, the licensee did not inform the Operations Center that it had made a
                declaration of one of the Emergency Classes (Notice of Unusual Event) specified in its
                Emergency Plan in accordance with 10 CFR 50.72(a)(1)(l).
                Insoector Followuo
                The licensee determined that the cause of the violation was human error. The shift -
L              engineer who made the initial contact with the NRC Operations Center provided details
l              of the fire but neglected to inform the NRC that a Notification of Unusual Event had been
'
                declared.
                The licensee conducted training with notification communicators about this event and
                NRC notification requirements. Additionally, the licensee reviewed applicable
                procedures and revised Procedure 1903.011, " Emergency Response / Notifications,"
                Revision 22, to require verbatim transmission of iaformation included on
                Form 1903.011Y, " Initial Notification Message," Revision 22, to the NRC Operations
                Center officer followed by transmitting the form to the NRC Operations Center vi-
l.              facsimile. The form explicitly identified the emergency classification for the event. The
!                inspectors verified the licensee's corrective actions had been completed and considered
                these actions adequate to prevent recurrence of this violation.
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  _  , - -          ..    - . . -_- - - -            . .-            . - -    - - - . _- . -    -- -
    .
    .
*
                                                              -  8
            E8.3 (Closed) Licensee Event Report 50 313/96-009f Fire in the Reactor Buildina Durina
,
                    Heatuo Resulted from a Cracked Weld in an Oil Line on a Reactor Coolant Pumo Motor
i
'
                    Backaround
                    During a Unit i plant heatup on October 17,1996, a fire occurred in insulation around the
                    m ain feedwater nozzle ring on the Train B once through steam generator. A cracked
                    wekiin the discharge line of the Train B reactor coolant pump lube oil lift pump resulted
                    in oil spray being introduced into the insulation. The hot piping caused the oil to ignite. A
                    Notification of Unusual Event was declared, and fire brigade responders extinguished the
                    fire approximately 16 minutes after discovery The NRC conducted a specialinspection
                    to review the circumstances of the fire, continued safety of the plant, and corrective
                    actions taken by the licensee. The results of the inspection are documented in NRC
                    Inspection Report 50-313; -368/96-27. The !!censee initiated Condition Report CR-1-36-
                    0567, conducted an event evaluation, and documemed its evaluation results in a root
                    cause analysis report.
                    The licensee's event evaluation concluded that oil intrusion into the insulation caused a
                    wicking effect that lowered the auto-ignition point of the oil to a lower than expected
                    temperature.
.
                    The NRC conducted a special inspection documented by NRC Inspection
'
                    Report 50-313; -368/96-27, which identified problems with the licensee's implementation
                    of 10 CFR Part 50, Appendix R, fire protection requirements and that the licensee had
                    failed to identify and correct problems that were noted prior to the fire. These were the
                    subject of an NRC Escalated Enforcement Action (EA 96-512). The licensee developed
                    and implemented many corrective actions to correct the deficiencies in these areas.-
                    They are discussed in Sections E8.4 and E8.5 of this report.
                    Insoector Followuo
                    This item is closed based upon the followup and discussion in Sections E8.4 and E6.5 of
                    this report, which address the escalated enforcement items associated with this event.
          - E8.4 (Closed) Enforcement Action 50-313:-368/96512-01013- Inadeauate Lube Oil Collection
                    Systems for Reactor Coolant Pumos
,
                    Backaround
                    During the NRC special inspection following the October 17,1996, Unit 1 reactor building
                    fire (NRC Inspection Report 50-313; -368/96-27), t1e NRC reviewed the tube oil
                    collection systems for all of the reactor coolant purnps in both Units 1 and 2. The
4
                                              y                n  m m.e t    ~-
 
    _____--                                _ - - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ - - _ _
l  .
  .
                                                                                                        .g.
            inspectors determined that the lube oil collection systems for the UnP.1, Train B reactor
            coolant pump and all of the Unit 2 reactor molant pumps had not been designed to
            collect oilleakage from all pressurized and unpressurized locations. Also, the Unit 1
            Trains A C, and D reactor coolant pumps had inadequate leakage collection from
            some unpressurized locations. The deficiencies were determined to be in violation of
              10 CFR Part 50, Appendix R, Section Ill, requirements.
            Immediately following the fire, the licc nsee placed the unit in cold shutdewn to evaluate
            the effects of the fire and water suppression on plant equipment. No significant damage
            occurred other than to some insulation. The damaged insulation was repaired or
            replaced as necessary. Other actions taken prior to restarting the unit included installing
            a new shroud around the Train B reactor coolant pump motor and extending oil collection
              drip trays for all of the reactor coolant pump motor:, to provide coveragv. af potential
              leakuge points. Due to remaining concems regarding the integrity of othar portions of
            the Unit 1, Train B motor lube oil piping, the licensee implemented administrative controls
              to prohibit operation of the associated high pressure lift oil pumps unless a fire watch
              was present. The motor manufacturer provided confirmation that the absence of tube oil
              injection following an inadvertent pump trip would not adversely affect the coastdown
              characteristics of the pump. This was used as supporting information in a 10 CFR 50.59
              safety evaluation to conclude that operation without the lube oil!!ft pump in standby did
              not constitute an unreviewed safety question.
              In Unit 2, the licensee performed tube oil collection sysiem modifications following an
              unplanned reactor shutdown in November 1996. The NRC performed walkdowns of both
'
              units and reviewed the modifications. These activities were documented in NRC
              Inspection Reports 50-313; -368/96-27, 50-313; -368/96-07, and 50-313; -368/96-08.
              The NRC concluded that with the exception of the remote lube oil ti:: aystems for both
              units, the tube oil collection system modifications es'.ablished compliance with 10 CFR
              Part 50, Appendix R, requirements.
              The Unit i remote oil fill lines (Tygon tubing) to the top of each motor were not provided
              with a leakage collection system as required by Appendix R requirements. The licensee
              provided administrative controls to prevent operation of the system until a modification
              and any necessary regulatory exemption was applied for and obtained. The Unit 2
              remote oil system was designed differently, using stainless steel tubing and stainless
                steel flexible hose, but no oil collection system was provided. The licensee described the
                Unit 2 system in detail and proposed compensatory administrative controls for its use in
                an exemption request to the NRC dated December 23,1996. The NRC approved the
                exemption request in a letter dated June 14,1937. The exemption allowed the licensee
                to use the remote oil fill system without a collection system contingent upon the
                licensee's implementation of several compensatory measures prior to each use.
                The licensee responded to the Notice of Violation in a letter to the NRC dated May 9,
                1997. Corrective actions were identified in this document, as well as the licensee event
                report associated with this event, and sevaral condition reports. Some of these were
                completed and reviewed previously as discussed above; others were reviewed during
                this inspection.
 
!
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                                                - 10-
      Insoector Followuo                                                                            0
      The inspectors reviewed documents, procedures, an'.8 interviewed personnel to
      determine the adequacy and completeness of the Sceasee's corrective actions. Also, the
      inspectors contacted the c:gnizant fire protection engineer in the NRC Office of Nuclear
      Reactor Regulation regarding the adequacy of the Unit 2 lube oil collection system.
      Documents reviewed included: portions of the Updated Final Safety Analysis Report;
      Condition Report 196-0507, its associated Root Cause Analysis Report and corrective
      action documents (including 10 CFR 50.59 safety evaluations and screenings);
      Procedure 1102.002, " Plant Startup," Revision 59; Procedure 1103.006, " Reactor
      Coolant Pump Operation," Revision 20; Procedure 1107.001, " Electrical System
      Operations," Revision 50; Procedure 1107.004, " Battery and 125V DC Distiibution,"
      Revision 8; and Procedure 1504.001, " Visual Inspection of the Unit 1 & 2 RCP's Oil            .
      Colloction System," Revision 4.
                                                                                                    {
      Modifications were made to the tube oil collection systems for all reactor coolant
      pumps in both units. These were determined by the NRC to be acceptable for meeting
      10 CFR Part 50, Appendix R, requirements. To ensure the continued operability of the
      systems, the licensee imp!*mented Procedure 1504.001, for inspection of the systems
      immediately prior to refueling outages and prior to startup.
      The licensee placed administrative controis on operation of the Unit 1 Train B reactor
      coolant pump lift oil pumps. This war, done because of uncertainty about oil system
      integrity during lift pump operation on the Jeumont Industrics motor Procedure
      1103.006 was revised to require that the Train B reactor coolant pump not be started or
      stopped during normal operations when the reactor coolant system temperature was
      greater than 350 degrees F. Operation of the lift oil system was al! owed, but only if two
      fire brigade-trained operators were present in the reactor building to identify oil leakage,
      inform the control room, and extinguish any resulting fire.
                                                                                                        '
      The norrr.al design of the oillift system provided an autostart of the lift pump when the
        reactor coolant pump tripped. The administrative controls discussed abova prevented an
        autostart from occurring. Therefore, if a reactor coolant pump trip occurred, the pump
      would coast down without injection of lubrication onto the motor thrust bearing. The
        inspectors reviewed the licensee's 10 CFR 50.59 Safety Evaluation 96-141, " Evaluation
        of Placing RCP #32B HP Lube Oil Pump (P-808 and P-638)in Pull-to-Lock," and safety
        evaluation determinations for associated procedure changes. Supporting documentation
        for the licensee's evaluation included a letter from the vendor documenting that reactor
        coolant pump coastdown time would not be adversely affected for at least 15 seconds
        without oil lift pump operation. The relevant accident analyses for consideration in the
        safety evaluation were the various sizes of loss of coolant flow accidents, and the
        evaluation determined that the lack of lift pump operation had no impact. The inspectors
        concluded that the safety evaluation provided an acceptable basis for concluding that
        reactor coolant pump operation without availability of tube oilinjection aid not constitute
        an unreviewed safety question.
)
4
                                                                                                          1
          _ _ _ _ _ _ _ _
 
!.
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                                                      -11
            The licensee initiated an evaluation of the tube oil system piping configuration to ensure
            that no additional questionable welds remained. The licensee and the motor supplier
            determined that the lube oil piping sysiem was constructed in accordance with applicable
            commercial standards with the use of qualified welders and adequate quality controls.
            No other similar weld failures were reported in more than 200 other reactor coolant pump
            motors manufactured by the motor suppiier. One difference in the subject configuration
            was the addition of a backup direct current powered lube oillift pump and associated
            piping. To evaluate this further, the licensee developed a plan to perform inservice visual
            inspections of the piping at operating pressure, collect vibration data, and perform data
            analysts. The licensee considered this option for addressing piping concerns preferable
            to others considered, which included piping replacement, weld replacement, and
            volumetric inspections with repair of rejectable welds. After discussing these options
            with the licensee, the inspectors considered the licensee's approach to evaluating the
            acceptability of the lube oil piping configuration to be adequate. Administrative controls
            on lift pump operation would remain in effect during future operating cycles pending
            completion of vibration analyses and continued visual inspections to identify any
            remaining weld failures.
            The inspectors determined that the licensee had completed corrective actions to
            ensure that its reactor coolant pump oil collection systems were in compliance with
            10 CFR Part 50, Appendix R requirements and that proposed corrective actions were
            either completed or in progress with responsive completion dates.
    E8.5 { Closed) Enforcement Actions 50-313/96512-01023 and 50-313/96512-01033: Failure to
            ldentify Sianificant Conditions Adverse to Quality and Take Promot Corrective Action
            Relatina to Oil Accumulation in Pioe Insulation.
            Backcround
'
            The NRC (and the licensee) determined that there were previous opportunities to identify
            that a problem existed with the reactor coolant pump oil lift system pr'ar to the fire event.
            These opportunities were missed, in part, because a condition report was not initiated to
            thoroughly evaluate the extent of the potential problem. Some of the opportunities to
            identify the scope of the problem included: discovery of a crack in the lift oil pump
<
            discharge line, the magnitude of the crack and inadequacy of the original weld, discovery
            of oil on the side of the steam generator during reactor building walkdowns during plant
            heatup, discovery of oil puddled under the reactor coolant pump, and identification of an
            excessive amount of smoke present in the reactor building during heatup.
            To correct these deficiencies in problem identification, communications, and
              understanding of oil-insulation interaction consequences, the licensee revised
              procedures and conducted training.
              10soector Followuo
              The inspectors reviewed documents, procedures, and interviewed personnel to
              determine the adequacy and completeness of the licensee's corrective actions.
                                - _-____
 
0
,
                                                  -12-
        Documents reviewed included: Procedure 1000.104, ' Condition Reporting and                        j
        Corrective Actions," Revision 13; Procedure 1015.036, ' Containment Building Closeout,'
                                                                                                          '
        Revision 5; Procedure 2102.002, ' Plant Heatup," Revision 10; Procedure 1102.002,
        ' Plant Startup,' Revision 61; Procedure 1102.010, 'P! ant Shutoown and Cooldown,"
        Revision 41; Procedure 2102.010, " Plant Cooldown,' Revision 30; Procedure 1203.034,              ,
        ' Smoke, Fire, or Explosion," Revision _10; Procedure 2203.034, " Fire or Explosion,"
        Revision 5; and training outlines and attendance records for Condition Report 196-567.
        The licensee's training appeared to adequately address the circumstances and causes
        of the fire and provided information on cr,uses of insulation fires; uses and limitations of
        material safety data sheets; importance of proper delivery, receipt, and accuracy of
        communications; and lessons learned in response to oil spills and observations of
        smoke.
        The condition reporting procedure _was revised to clearly address identified deficiencies
        in fire protection system components including oilleaks and spills. Items of this type that
        meet the definitions in the procedure now require the initiation of a condition report. The
        other procedures provided requirements to perform inspections for oilleakage and
        excessive sn.oke and provided actions to take in response to identified adverse
        conditions. These actions could include terminating plant heatup, posting fire watch
        personnel with fire fighting equipment, contacting the fire protection engineer, and othar
        appropriate act.ans. The procedures also contained guidance on the effects of oil
        soaked insulation on the auto-ignition temperature of the oil.
        The inspectors concluded that the licensee's corrective actions were adequate to
        improve staff understanding of the mnsequences of oilleaks, and the importance of
        communications when adverse conditions are identified, provide actions to prevent fires
        from occurring, provide additional requirements for inspections, and provide actions to
        take when adverse conditions are identified.
  E8.6 (Closed) Insoection Followuo item 50-313/97201-01: Licensee's Actions to Revise the
        Technical Soecification Bases fo,- the Minimum Water Volume in the Condensate
          Storcae Tank
          3ackaround
          The minimum water level in the Unit 1 safety-related condensate storage tank was based
          on the capability to provide a sufficient supply of water to permit the initiation of the decay
          heat removal system within 4 hours or less of an accident initiation. However, the actual
          period of time required to initiate decay heat amoval may be substantially greater than 4
          hours; The licensee acknowledged that this text was ir, accurate and proposed to revise
          the existing Technical Specification bases to eliminate (probably) reference to the decay
          heat removalinitiation and to state that the minimum condensate storage tank level was
          based on the volume that was required for 30 minutes of emergency feedwater operation
          before manual switchover of the pump suction to the service water system.
 
                                                                                                __
*
.
                                                        13-
      Insoector Followuo
      The licensee stated that they were in the process of neveloping the improved Standard
      Technical Specifications and that the appropriate condensate storage tank bases would
      be included in the initial submittal. The licensee stated that they planned to submit the
      revised Technical Specification design base along with the revised Technical
      Specifications by the end of 1998. The inspectors verified that there was no safety
      concern related to the minimum condensate storage tank water level and that there was
      no need on the part of tne licensee to revise this level. However, a clarification of the
      Technical Specification basis for the level was appropriate.
  E8.7 (Closed) Unresolved item 50-313/97201-02: Licensee's Failure to incoroorste Maximum
      Emergency Feedwater Flow limits into Plant Procedures to Monitor and Preclude
      Exceedina Recommended Limits
        Backaround
      A 1991 Babcock and Wilcox Company engineering report (92-R-1019-01) identified that
      the maximum emergency feedwater flow rate to preclude damage to the Unit 1 steam
        generator tubes was 1500 gpm. The NRC determined that the licensee had not revised
        plant procedures to incorporate this upper flow limit. Specifically, Procedure 1010.002,
        "hansient History / Transient Cycle Logging,"in effect at the time was not revised. The
        NRC reviewed operator logs and identified two occurrences where the emergency
        feedwater flow exceeded the 1500 gpm limit.
        Insoector Followuo
        The inspectors reviewed Condition Report CR-1-97-0081, dated March 27,1997. which
        the licensee generated to resolve the maximum emergency feedwater flow issue. A
        licensee contractor prepared an analysis which determined that the maximum feedwater
        flow rate to prevent steam generator tube damage was 2214 gpm. The inspectors
        reviewed the calculation and determined that the higher flow limit was adequately
        justified.
        The licensee stated that the upper flow limit was not included in operating procedures
        because high emergency feedwater flows would cause a cooldown, during which
        operators would take procedural actions to correct the high flow rate problem. However,
        the inspectors determined that the licensee had not previously evaluated the effects of
        the high flow on the steam generator tubes for the periods of time before operator actions
        would restore an acceptable flow rate.
          The inspectors reviewed Procedure 1010.010, " Unit One Transient Cycle Logging and
          Reporting," Revision 1, which was issued to provide guidance for determining excessive
          emergency feedwater flow and provide specifications for initiation of a condition report.
          This procedure superseded Procedure 1010.002. The inspectors determined that the
          Procedure 1010.010 required an engineering evaluation if the upper limit of the          '
          emergency feedwater flow was exceeded.
        .
                    ..  .
                                        _ _ _ _ _ - _ _
 
                                          -                - _ _ _ - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ ___
*
.
                                                  -14-
        During the inspection, the inspectors reviewed in detail Framatome Technologies final
        draft calculation 321229968-00 and Framatome Technologies Calculation 32-5000450-
        00,"ANO Auxiliary Feedwater Capacity - Phase 2." The purpose of these calculations
        was to determine the limit for emergency feedwater flow to preclude unacceptable steam
        generator tube damage or wear. The results showed that a flow rate of 2214 gpm
        resulted in less turbulence induced vibration than at full power conditions. As part of this
        review, a conference call was held with Framatome Technologies and licensee
        personnel during which they responded to questions from tM inspectors. Based on their
        review and the information presented during the conference call, the inspectors
        determined that the analysis provided in these calculations demonstrated that
        emergency feedwater .peration at flow rates up to 2214 gpm for limited periods of time
        would have no significant effect on steam generator performance. The inspectors
        verified that design of the emergency feedwater system would not resun la flow rates in
        excess of this amount.
        The inspectors determined that the licensee failed to procedurally incorporate a vendor
        recommended maximum emergency feedwater flow limit of 1500 gpm to a single steam
        generator. As a result, flow rates in excess of this amount were permitted to occur for
        short periods of time. Later, in response to the NRC's questions, these excess flow rates
        were shown to be acceptable.
        10 CFR Part 50, Appendix B, Criterion V, requires that activities affecting quality be
        prescribed by and performed in accorJance with instructions, procedures, or drawings.
        The licensee's failure to incorporate vendor recommended flow rate limits into
        procedures was identified as a violation of 10 CFR Part 50, Appendix B, Criterion V
        (50-313/9721-01),
        Conclusions
        The licensee failed to incorporate a vendor specified emergency feedwater flow limit into
        plant procedures, resulting in the failure to evalus'e the effects of exceeding the vendor
        recommended flow limit during two previous periw of plant operation. Concerns
        related to possible damage to the steam generator tunes were adequately resolved by a
        subsequent arialysis demonstrating the acceptability of higher flow limits.
  E8.8 (Closed) Unresolved item 50-313/97201-03: EFW Pioina Confiauration aifferences
        Backaround
          in review of Calculation 82-D-2086-02, which evaluated net positive suction head
        requirements for the motor-driven and turbine-driven emergency feedwater pumps, the
          NRC identified that the suction piping configuration used in the calculation to determine
          pressure drops was not consistent with the piping configuration identified in the piping
          isometric drawings. The pressure drops in the piping were negligible and the effect on
          the results of the calculation due to the differences was small. This item was opened
          pending action to revise the calculation and take any appropriate generic actions.
                                                                                                                                    _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _
 
                                                                                                          '<
  .
I
  ..
                                                        15 -
            Inspector Followup
            The licensee initiated Action item Requests L97-0198 and L97-0285 to revise
            Calculation 82-D-2086-02 to address the as-built configuration. The inspectors verified
            that these action items had been assigned to the appropriate engineering organization
            for completion. The licensee established a Uue date for completion of the calculation
            revision of March 1,1998. Based on the minhal impact on the results of the calculation -
            expected by incorporating the as-built piping configuration differences.in the calculation,
            the inspectors considered the licensee's schedule for completion adequate. The broader
                                                                                                            3
            issue of identified deficiencies in the licensee's maintenance of calculations is addressed
            in the inspectors' followup of Unresolved item 50 313/97201-06 (Section E8.11).-
              10 CFR Part 50, Appendix B, Criterion Ill, " Design Control," states, in part, that design
            control measurss, such as design reviews, shall provide for verifying or checking the
            adequacy of design. The licensee's failure to properly verify that the design inputs for
            Calculation 82 D-2086-02 were consistent with the installed piping configuration was
            considered as the first example of a violation of 10 CFR Part 50, Criterion lli (50-
            313/9721-02).
            Conclusion
            The licensee's failure to properly verify the design inputs to Calcu;ation 82-D-2086-02    ^i
          - was identified as a violation of 10 CFR Part 50, Appendix B, Criterion Ill.
    E8.g (Ocen) Unresolved item 50-313/97201-04f 1: *n=*a Pinino Pr===nre.ggi .
              Temperature Specifications
                                                                                                            ,
              Background
              During review of Calculation 88-E-0100-16, which evaluated pressure and temperature
          ; limitations of emergency feedwater system piping, the inspectors identified several
              discrepancies. For example, in the determination of maximum emergency feedwater
              discharge pressure limitations, the calculation used a pump suction alignment from the
              condensate storage tank instead of the service water system as input. The use of the
              service water system as a source would result in a higher emergency feedwater system
              discharge pressure than if suction were taken from the condensate storage tank.-
              The licensee initiated Engineering Request 973848 for the design engineering staff to
              review and revise Calculation 88-E-0100-16. The rrque'st was assioned a completion
              date of December 31,1997.'
            - _ _ _ _ - _ _ _.                                                                                l
 
                                                                                              ___
  .
  .
i
                                                13
      Insoector Followup
      The inspectors reviewed the licensee's followup action item response to the original
      inspection concern. The response identified that the design pressure of the emergency
      feedwater system discharge piping was 1850 psig. At rated speed of the turbine-driven
      emergency feedwater pump (3795 rpm) and with the suction aligned to the condensate
      storage tank, the calculated discharge pressure at shuto# head was 1732 psig. When
      the system was aligned to the service water system, the calculated discharge pressure at
      shutoff head was 1842 psig, resulting in a margin from the design pressure of only 8 psi.
      However, the ca!culation did not account for speed controller tolerances, nor did the
      calculation assume a failure of the speed controller. If a failure of the speed controller
      occurred, procedural guidance was provided for an operator to take manual control of the
      turbine-driven emergency feedwater pump. This procedure instructed the oper; tor to not
      exceed a pump speed of 3900 rpm to avoid overpressurization of the discharge piping.
      The licensee determined that at a pump speed of 3900 rpm, the resulting maximum
      discharge pressure was 1844 psig when suction was aligned to the condensate storage
      tank and was 1932 psig when suction was aligned to the service water system.
      Therefore, manual operation of the turbine-driven emergency feedwater pump at the
      maximum procedurally-allowed speed would result in a maximum discharge pressure            >
      that exceeded the design pressure of the piping.
      The licensee's response stated that the utilization of the 1932 psig maximum ressure as
      a basis for discharge piping design pressure was not considered because manual pumr>
      control while aligned to the service water source would require an equipment failure in an
      already infrequent emergency plant condition.
      The inspectors had a concern with this position. The condensate storage tank was
      designed to provide a tomado-protected 30-minute supply of water to the emergency
      feedwater system. The 30 minutes allowed the operations staff time to transfer the
      emergency feedwater supply to the service water system. However, if the turbine-driven
      speed control was out of tolerance or failed and the pump was operated manually per
      procedure, the maximum discharge pressure could exceed the design pressure of the
      turbine-driven pump discharge piping.
      The licensee's response also addressed historical mp performance during testing at a
      pump speed of 3920 rpm. Analysis of the test results in Calculation 80-D-10838-102A
      concluded that the discharge pressure would not have exceeded the 1850 psig design
        pressure, even if suction were aligned to the service water system at maximum pressure,
        because the pump was not operating at full potential. However, the pump had sufficient
        capacity to meet minimum specifications and, therefore, was operable.
        The inspectors discussed these issues further with the licensee and were informed that
        the original calculation would be reviewed to determine the extent that higher-than
        nominal pump speeds need to be considered, both from a speed controller tolerance
        perspective and a manual operation perspective. Additionally, since the pump appeared
        to be under performing with respect to its certified pump curve, using the pump curve for
    ,                                                        _ _ _ _ ___
                                                                      _ _ _
 
  .
f
  .
                                                  -17
          piping design pressure determination may be overly conservative. The inspectors were
          unable during this inspection to evaluate this issue further to determine whether the
          pump was performing in an acceptable manner with respect to the pump curve.
          However, the inspectors noted that the pump was meeting its Technical Specification
          limits. The licensee'e due date for completion of this review and revision to the
          calculation was December 31,1997. During the exit meeting conducted January 6,
          1998, the licensee indicated that the due date for this effort had been extended to
          January 15,1998.
          The inspectors considered this item open pending further NRC review of the licensee's
          ongoing evaluation and review of the performance of the turbine-driven emergency
          feedwater pump.
    E8.10 (C10MLUDIggived item 50-313/97201-05: Evaluation of EFW Pumo Room
          Environmerd
          Backaround
          The licensee identified an environmental qualification problem in the Unit 1 EFW pump
          room, which houses the turbine-driven and motor-driven EFW pumps, Steam traps
          located in the steam supply lines to the turoine-driven pump, which were not designed for
          seismic loads, were vented to the atmosphere of the room. A failure of one these traps
            (two in the high pressure piping and two in the low pressure piping) would cause a
            continuous b!owdown of hot steam in the room, raising the ambient temperature and
            humidity, and possibly challenging the operability of environmentally-sensttive equipment
            in the room. The licensee took interim measures to leave one steam trap isolated and to
            permanently open a door to the room. The licensee then calculated the impact of a
            steam trap failure under this configuration assuming that operators could isolat; the
            steam blowdown within 1 hour. The calculation (87-E-0026-09, Revision 0) co%ded
            that temperature in the room wouid peak at 149.5 degrees F for a brief time, but that the
c          steady state temperature would be 134 degrees F. Since the limiting long-term
            temperature for the room was 148 degrees F, the licensee concluded that the EFW
            system remained operable. The inspectors agreed with the licensee that humidity
            resulting from the event would not adversely affect equipment critical to the function of
            the EFW system.
            This item was opened to track licensee efforts to make modifications to resolve this issue
            permanently and to make necessary revisions to related design basis documents. The
            licensee also indicated that an assessment of past operability of the EFW room would be
            performed.
      --
 
    .
!  *
                                                -18-
      Insoector Followuo
      in review of this finding, the licensee determined that the original design of the EFW
      steam traps was inadequate for both Units 1 and 2. The safety implications for Unit 2
      were mitigated because the steam traps exhausted outside the EFW room. The licensee
      decided to modify EFW steam traps on both units: Unit 1 during Refueling Outage 1R14
      (Spring 1998) and Unit 2 during Refueling Outage 2R13 (first quarter 1999).
      The licensee performed a past operabmty assessment and concluded that the EFW
      system was operable in the past considering the room door closed, all traps in operation,
      and a 1 hour operator response to isolate the failed trap.
                                                                                                        ,
      The inspectors reviewed Root Cause Aralysis Report CR C-97-0048, " Unqualified
      Steam Trap Lines in the EFW Turbines' Steam Supply Lines." In this document, the
      licensee determined that the architect / engineer (A/E) for the design of the plant assumed
      that failure of the steam trap lines would not reduce the functional capability of the EFW
      systems. However, the A/E had evidently focused on sufficiency of the resulting supply
      of steam to the turbine and not on the environmental consequences of the failure.
      The inspectors questioned the assumption that operator action to isolate steam would
      occur within 1 hour of a steam trap failure, an assumption essential to the past operability
      status of the system. The licensee stated that an operator surveillance of the room
      (checking lube oil flows) occurs by procedure soon after each automatic start of the EFW
      system, which would presumably result in identification of the steam trap failure.
      However, since no time parameter was specified for this surveillance, it could be                  '
      indefinitely delayed if other complications were to occur. The inspectors asked about the
      effect of a response if the failure occurred after this surveillance effort. In this case, since
      no remote indication of temperature in the room is available, it did not seem reasonable
      to assume that steam isolation within one hour would occur. The licensee stated that the
      fire detection system would probably alarm given the amount of steam that would exist in
      the room. On one previous occasion, in 1992 (CR 1-92 0361), the deluge valve tripped
      open due to an excess amount of steam from the steam trap drain lines. However, on
      further discussion with the fire control engineer, the inspectors learned that the fire
      detection device in the EFW room is not sensitive to steam and would only possibly
      alarm in the event of a condensation buildup on the device. The inspectors determined
e
      that the fire detectors were not a reliable indication of a steam leak.
r
      The evaluation cf past operability indicated that, with the room door closed, the room
      temperature would reach 164 degrees F 1 hour after the steam trap failed, it seemed
        questionable whether an ooerator could enter an anwonment this harsh to pen ~orm the
        required valve isolation. In response to the inspectors' concern, the licensee generated
        a new action item (Item Number 9) for CR C-97-0048 to address the issue of accessibility
        as it pertained to the past operability determination. The licensee recalculated the room
        temperature profile using more realistic assumptions, including condensing heat transfer,
        and found that the room condition at 1 hour would be 133 degrees F at a relative
        humidity of 57 percent. This corresponded to a wet bulb globe temperature (WBGT) of
        121.5 degrees F. At this WBGT, the licensee, based on US Navy
  i
                                                                                                          1
                      ~                                          .    _ _ _ _ - _ _ _ _ - _ _ _ _ .
 
                                                                                                _ _ _ _ _ _ _ _ - _ _ _ _ _ _
    ,
    .
                                                19-
      Document OPNAVINST 5100.19C, Changa 1, dated May 15,1990, determined that
      strenuous work at a maximum duration of 30 minutes could occur in this environment.
      The amount of time needed to isolate a failed steam trap would be only several minutes.
      The inspectors noted that the shapa of the temperature profile suggested that the
      temperature would peak out at something less than 140 degrees F at a relative humidity
      of less than 60 percent. This suggested that the room would be fit for the necessary
      operntor's actions to take olace and that equipment operability would not be threatened.
      Dunng the inspection, the inspectors reviewed Calculation 87 E-0026-09, ''EFW Pump
      Ruom Tempaniture Profiles." This calculation provided an analysis of the heat up of the
      EFW puu : u sum!ng various failed op9n steam traps in addition to the normal thermal
      load proviued by the EFW pumps and other equipment. The analysis was carried out
      with an ANO developad and maintained software package,"RHUR," which calculates
      roora heat up from art >ltrary sources. RHUR was p ogrammed in a ' macro' language
      executed in a spreadsheet program, SYMPHONY. It was documented as a software
      package and was controlled and maintained under the software control program. The
      RHUR model devebped for thlS analysis was constructed by substantially modifying the
      " macro" statements that make up the RHUR software package.
'
      The inspector reviewed the analysis provided in the calculation and determ!ned that the
      results of the four cases presented seemed reasonable and conservative for the given
      configurations, in response to the NRC concern, the licensee provided additional
      supporting analysis in the form of an alternate calculation using the GOTHIC software
      program that substantially verified the results presented in Calculation 87-E 0026-09.                                  (
      The inspectors considered that the licensee had adequately established thJ past
      operability of the EFW room equipment. Based on the less severe conditions of having
      the room door removed, the insoectors also considered this in9rmation to support the
      licensee'e assessment of current operability However, the inspectors identified a
      weakness in the licensee's original response to this issue in that it frJted to document
      consideration of the ability of an operator to perform an assumed operation in a harsh
      environment. A tecond weakness was the lack of verification of the software model
      used to calculate the postulated EFW room thermal transient. Although considered good
      enginee% practice, licensee procedt,res did not specifically address this validation.
      The licensee was in the process of developing a modification for both units to mitigate
      the effects of pressure boundary failures and the blowdown of live steam from steam
      traps in the EFW rooms, while preserving adequate condensato removal. Through
      discussions, the inspectors learned that the modification w jd reroute steam trap piping
      to exhaust outside the EFW room and upgrade the re          a be safety related and
        seismically qualified. The inspectors determinM 'N          modification would resolvo the
        hardware concerns associated with this issue.
  4
        .__ _ _ _ _ _ _ - _ _ _ _ _
 
  0
  .
                                                      20-
    E8.11 (Cloand) Unresolved item 50-313/9720106: Desion Control Weakness
          Bsckaround
          This issue involved several calculation deficiencies. The examples included calculations
          containing plant configuration discrepancies as well as calcult.tions that required
          enhancement, updating, or to be superseded. The specific examples were as follows:
          *        Calculation 88 E-0086 01,"lE Bulletin 88 04 Review for P7A and P7B Minimum
                    Flow Evaluation," did not adequately document the effect of two parallel EFW
                    pumps operating on minimum recirculation.
          *        Calculation 92 E 0077-04, " Unit 1 EFW System Pump Performance
                    Requirements," used an incorrect level in tank T-418 for EFW actuation due to a
                    loss of offsite power following a tornado.
          *
                    Calculation 92 E 0021-01," Emergency Nty Cycle and Battery Sizing
                    Calculation," was not revised to reflect plant modifications that resulted in
                    changes in the continuous and locked rotor current for EFW Valves CV 2620 and
                    CV 2627, and a modification that resulted in an additional load due to the reactor
                    building spray puinps close and trip circuits.
          *        Calculation 80-D-1083A42, "EFIC DC Valve Torque Calculation Under Reduced
                    Voltage Condition," was not superseded as required.
          The actions implemented or planned by the licensee included those to correct the
          specific examples and to improve design control programs to prevent recurrence.
          Insoector Followuo
          The inspectors reviewed Licensee information Requests (LIRs) L97 0289 and 0290 as
          well as Condition Reports CR C-97-0058 and CR-C 97-0059.
'
            LIR L97 289 addressed the failure to updato Battery Loading Calculation 92 E-0021-01,
            following iniplementation of Design Change Packages (DCPs) 92-1003 and 93-008. The
            licensees's response to the inspection report (letter 1CAN099703, dated September 22,
            1997), indicated that this calculation would be superseded before January 15,1998.
            F sever, at the time of this inspection, this calculation had already been updated and
              (as reviewed by the inspectors and found acceptable, in discussions regarding this
            inconsistency, the licensee explained that the oversight resulted from the existence of a
            large backlog of dats that had not been entered into the Design Configuration
            Information Management System (DCIMS). This condition resulted in a failure to identify
            affected calculations when engineers accessed the DCIMS system to perform searches
 
                                _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ . _ _            _ _ - _ - _ - _ _ _ _ - _ _ _ _ _ __
  .
  .
                                                                                              21-
    based on key words related to a particular modification or calculation revision. The
    licensee indicated that this backlog had been eliminated and that current practice would
    preclude new backlog formation. The inspectors confirmed that no backlog existed and
    that the DCIMS was up to-date at the time of this inspection. In additia, the inspectors
    reviewed Memorandum ANO 97 00225 issued to the engineering staff that reinforced the
    existing procedures and guidance related to calculation updates for load changes.
    The second issue within LIR L97 289 dealt with Calculation 80-D 1083A 02, *EFIC DC
    Valve Torque Calculation Under Reduced Voltage Condition" which should have been
    identified as superseded. The licensees's response to the inspection report indicated
    that this calculation will be superseded before January 15,1998. Discussions with
    licensee personnel and review of the applicable procedure, OP 5010-015, " Engineering
    Calculations," indicated that clear direction existed to ensure that superseded
    calculations are identified, The cause of the specific instance was attributed to a backlog
    of data to be entered in the primary tracking tool, DCIMS, that had previously existed. As
    discussed above, the inspectors confirmed that no backlog existed at this time of this
    inspection.
    LIR L9/ 290 addressed Calculations 88 E-0086-01 and 92 E-0077-04, both of which
    needed enhancement for various reasons. The licensees's response to the inspection
    ecport indicated that these calculations would be revised before March 1,1998.
    10 CFR Part 50, Criterion lil, * Design Control," states that measures shall be established
    to ensure that applicable requirements and the design basis are correctly translated into
    specifications, drawings, procedures, and instructions.
    The licensee's failure to properly account for design information in the
    Calculations 88 E-0086 01,92 E 0077-04,92 E-002101, and 80-D-1083A-02 was
    considered as additional examples of a violation of 10 CFR Part 50, Appendix B,
    Criterion 111 (50-313/9721.~1).
    The inspectors review of Condition Report CR C 97-0058 indicated that the condition
    report dealt with prioritization and completion of the Upper Level Documents (ULD) and
    ULD-related discrepancies. The actions taken as a result of the condition report included
;  determining completion priority and establishing and communicating to the engineering
'  organization the philosophy for use of the ULDs. The inspector reviewed the actions
    taken on these issues and confirmed that they were appropriate.
    The inspectors review of Condition Report CR-C 97 0059 !ndicated that this condition
    report dealt with completion of the Design Configuration Documentation (DCD) Project
I  and resolution of discrepancies identified during this project. The project was completed
l  in December 1994; however, approximately 100 discrepancies remained unresolved.
    The discrepancies were originally screened for significance and operability, but this CR
    recognized that recently there was greater interest in and higher standards being applied
    to design configuration issues, in addition, several other recent CRs have been initiated
    to reexamine previously closed discrepancies that were found to have inadequate
    resolution by current expectations. The CR-C 97-0059 evaluation established an action
!
                _  _                                                                            _____,
 
  .
  *
                                                                                                            :
                                                    22
        to priontize review of the remaining open discrepancies. This action was completed. In
        addition, previously closed discrepancies have been reviewed to ensure that they are
        adequate. This completed action led to further actions to evaluate 44 previously closed            l
        items. These subsequent actions were not available for review at the time of this                  )
        inspection.                                                                                        '
        In addition to the specific observations and related corrective actions reviewed above,
        the inspectors reviewed other condition reports, procedures, guidelines, and
        programmatic evaluations that impact the control of calculations. These are listed as
        Attachment 2 to this report. Several recent calculations and design change packages
        were reviewed to assess the results of ongoing improvement efforts on current
        calculation practices. These actions addressed broader issues associated with
        calculation processes and control.
        The licensee initially identified a concem regarding the adequacy of their calculation
        processes in Condition Report CR C 96-0060 initiated in March 1996. The most
        significant outgrowth of efforts to resolve this concern led to the formation of a Quality
        Action Team (QAT) to study calculation process issues. The team's final report,
        published in September 199d, identified a large number of concerns (56) organized into            ,
        the following broad categories:                                                                    l
        *        Calculation in processing time was too long creating backlogs and delays for
                  engineering personnel.
        *        The DCIMS data base contained inaccurate, incomplete, not up-to-date or
                  erroneous information.
        *        The calculation preparation, revision, and approval process requirements were
                  not clearly and consistently defined, documented, or understood by engineering
                  personnel.
t
        *        Retrieval of existing calculations was sometimes difficult.
!        +        Inconsistent interpretation of standards for legibility of calculation documents
l                resulted in delays for processing calculations through quality assurance records
l                storage.
l
        *        Data may be buried within a calculation with its relationship to other processes
                  not defined,
        *        Program engineers were not alwa ys aware when other engineers were revising
,
l                calculations that impact their program.
;                                                                                                          -
        In reviewing the QAT action plan, the intpectors determined that the proposed corrective
        actions were appropriate and that they r hould provide significant calculation process
        improvements when fully implemented in addition, the interim compensatory actions
    . .-              -          - - _ - . - .          - - . - - - . -                ,          - - . .
 
                          . _ _ -
                                    _ _ . _ . . _ _                        _ _ _      _      _
.
.
                                            23-
  that were implemented in the short term appeared to be of sufficient quality to reduce the
  occurrence of calculation deficiencies.
  In addition to reviewing the licensee's action plan, the inspectors reviewed a sample of
  approximately 30 calculations generally chosen at random. These are listed in
  Attachment 2. The inspectors found that the calculations were thorough, generally clear,
  and well documented. Assumptions and other data sources were clearly identified e7d
  referenced. The few exceptions were found in some older calculations performed during
  original plant design and construction, approximately 20 years ago.
  The inspectors focused on the licensee's process for assigning status to calculations.
  This included the use of new, pending, as installed, amending, and superseded status
  levels. Although the inspectors found the use of the amending status somewhat
  confusing and cumbersome, no erroneous cross references were found in the sample
  reviewed. For example, there are m limits to the number of calculations that can amend
  a given calculation. Nor is there a time limit beyond which amending calculations would
  require incorporation. In practice, however, the only calculations found with a large
  number of amending calculations were the electricalloading calculations. In this case,
  the licensee's practice was to assign a coordinator or owner for these calculations. The
  coordinator was responsible for incorporating all amending calculations once each cycle.
  The incorporated calculations were then assigned a status of superseded. The
  inspectors reviewed and found this to be correctly performed for a recently issued
  revision to Calculation 92 E 0021-01. " Emergency Duty Cycle and Battery Sizing
  Calculations." In addition, as mentioned above, the licensee instituted a Calculation
  Revision Notice (CRN) process that allows minor changes to calculations to be issued
  and more easily tracked. This process was designed to reduce the number of amending
  calculations to no more than nine in practice, based on a sample review, the inspectors
  found no examples with more than two CRNs.
  In spite of an apparently comprehensive effort to improve the calculation process, the
  inspectors recognized a potential inconsistency in the way the licensee handled the
  issue. The licensee's quality assurance team identified a concern related to the
  preexistent backt0g in the DCIMS data base that had apparently been the cause of
  several NRC-identified discrepancies. In addition, other persistent problems within the
  calculation process were identified by the team. However, the licensee did not perform a
  sample review of existing calculations to determine whether the DCIMS problem, in
  addition to the other identified problems listed above, had created a situation where a
  large number of calculations currently in effect may contain technical errors as a result of
  these problems. The inspectors considered the lack of review of a sample of
  calculations to represent a weakness in the licensee's overall response to the issue.
  The inspectors reviewed the capability and use of the DCIMS database. It was a very
  useful and powerful tool for searching for needed information concerning calculations,
  evaluations, and reports.
            _    _
 
  ,.  .                _            _                                    ._      .__
    ,
;
L
    .
                                                          24
                Conclusips
                A violation of 10 CFR Part 50, Appendix B, Criteilon lil, was identified conceming
                discrepancies identified with four licensee calculations. The lack of a review of existing
                calculations to determine if past problems identified by the licensee affected other
                calculations of record was identified as a weakness.
        E8.12 (CJosed) Insoection Followeo item 50-382/9720107: Licensee's Actions for Revising
                the Modification Procedure to Address Field Routed Installations
                Background
                The NRC identified that the layout configuration of certain installed conduits was different
                from that shown on the conduit and cable tray layout drawings. The NRC also identified
                that there were no drawings or records of the seismic support details for the installed
                conduit and cable tray supports. The licensee stated that this was due to conduits being
                field routed.
                The NRC noted that conduits attached to the wall of the emergency feedwater pump
                room contained redundant cables supported on the same unistrut support. The licensee
                stated that the supports were designed to withstand seismic design loads and were
                passive. In addition, the licensee stated that redundant trains could be supported from a
                common support from a seismic standpoint provided Appendix R and other criteria were
                addressed.
                The licensee's modification procedure discussed a constructability walkdown involving
                the modification engineer, modification supervisor, craft supervisor, and quality engineer,
                During discussions, the licensee stated that they would document the decisions arrived
                at during the walkdown by including these in the form of work steps in the controlled work
                package. The licensee stated that this would include conduit routing and seismic support
                details that would be documented as sign-off steps. The need to revise the modification
                procedure to address field routed installations was identified as Inspection Followup
                Item 50 313/9720107.
                Insoector Followuo
                The inspectors reviewed the licensee's response to the inspection followup item
                documented in Letter 1CAN099703, dated September 22,1997 (response letter to NRC
                Inspection Report 50-313/97-201). The inspectors also interviewed licensee personnel.
                The inspectors requested that the licensee supply details concerning how the Appendix
                R requirements were met for the redundant cables supported on the same unistrut in the
                emergency feedwater pump room.
                The inspectors rewewed draft Calculation 85-E-0086-1, " Safe Shutdown Capability
                Assessment," Revision 3, and found that the turbine driven emergency feedwater pump
                cable and the motor driven emergency feedwater pump cable had a 1 hour fire barrier
                installed. In addition, the inspectors found that an NRC exemption was granted for
 
                                                                      .-            - . - - .
  .
  .
                                                    25-
          configurations having less than 20 feet of separation. The inspectors reviewed licensee
          Letters 1CAN068706, dated June 24,1987, and 1CAN048708, dated April 22,1987,
          which requested the exemption for the cables. The exemption, granted in an NRC letter
          dated October 26,1988, contained a stipulation that a 1 hour fire wrap was necessary in
          addition to a fire suppression system. During a walkdown, the inspectors determine)
          that there was a fire suppression system in the pump room and that the cable was
          wrapped with fire wrap.
          The inspectors reviewed the revised Procedure 6030.005, " Control of Modification
          Work," Revision 3, Change PC-1. The inspectors noted that the procedure was revised
          to include instructions for the development of a single line drawing depleting the types of
          supports used, the support spacing, and the conduit layout. The procedure required that
          the draw!ag be included with the controlled work package.
    E8.13 Ogied) Insoection Followuo item 50 313/97201-08: Licentee's Actions to Perform a
          Revised Cable Pullina Calculation Usino Additional Desian Information Obtained from
          the Cable Vendor
          flitCMIQMDd
          The NRC identified that Procedure / Work Plan 6030.112, " Installation of Raceway
          Systems," used for the installation of raceways, included a table of conduit bend radil for
          manufacturers' standard bends but not for long radius bends. The NRC asked the
          licensee which cables required long radius bends as a result of the minimum bending
          radius and maximum sidewall pressure restrictions during pulling. The licensee provided
          a calculation that compared the cable minimum bending radil to the bend radil of the
          conduit that was appropriate for a single cable. The NRC questioned the values for the
          redil on the inside surface of the conduit bends as well as the accuracy of the cable
          diameters and maximum cable bend radii. The licensee stated that they would obtain
          updated cable vendor data to verify the conduit bend dimensions and would revise the
          calculation accordingly. The NRC reviewed the preliminary calculation and noted that a
          few largo diameter cables would require long radius bends. This item was opened
          pending licensee actions to obtain additional design information from the cable vendor in
          order to perform a revised cable pulling calculation,
          insoector Followuo
          The inspectors reviewed the licensee's response to the inspection followup item in
          Letter 1CAN099703, dated September 22,1997, and interviewed licensee personnel.
          The inspectors also requested a listing of condition reports involving damage to cables
          that resulted from cable pulling; however, no such conditions were found.
-                                                                          -
 
      . - . - - - . -        . - .- .- - - - - . -                                            _ -.- --. - . - -                  - - - -
    .
                                                                                                                                            !
                                                                                                                                            '
;  .
                                                                                  26
                            The licensee stated that updated cable information would be obtained from cable                                '
                            vendors for approximately 10 percent of the cable codes listed in the cable minimum
                            bend radius review table. The licensee expected to receive this information by mid.
                            January,1998. The licensee also stated that the updated cable information obtained
                            would be used to revise the calculation for each of the applicable cable codes. The                            e
                            licens*a stated that the cable calculations would be revised by March 1998.
                            The inspectors determined that the licensee's corrective actions were adequate for the
                            current closure of this item based on the fact that the condition report search had not            _
                            found any cables damaged by cable pulling and that the actions necessary to revise the
                            cable code calculations were being tracked and had acceptable completion dates.
                      E8.14 (Closed) Insoection Followun item 50-313/97201 11: Verification of Calculation Changes
                            Associated with DC Batteries
                            Backaround
                            The NRC identified that Calculation 92 E 0021-04," Battery DOS and 007 Recharge
                            Time," Revision 0, was not revised to show the reduced charging time resulting from a
                            change out of the Unit 1200 amp battery chargers with new 400 amp chargers. An
                            amending calculation,93 D 1010-04, " Recharge Time for D06 and D07 with Changes
                            per DCP 931010," Revision 0, calculated the revised charging time. This item was
                            opened pending incorporation of the amending calculation information into the parent
                            calculation,
                            insoector Followuo
                            The inspectors determined that this condition did not represent an actual prob'sm
                            because the battery recharge times were not required by license conditions and did not
                            affect battery capacity. In addition, as discussed in Section E8.11 of this report, thers
                            were no licensee procedures requiring incorporation of amending calculations into the
                            parent calculation within a specific time period. Based on discussions with the licensee,
                            the inspectors determined that information provided in amending calculations, but
                            missing from the parent calculation, should not result in a quality concern because
                            annotations provided in the parent calculation positively identified the existence of the
                            amending calculations.
                            The inspectors verified that amending Calculation 93 D 1010-04 had been incorporated
                            into Calculation 92 E 002104. The inspectors observed that the revised recharge times
                            were based on the updated 400 amp recharge capacity.
I '
d
                                          m_ _ - - _ _ . _ . . , ..,,_,_--m =--. - . , _ , _ ,            -      ---,--r-- -- - m.        -
 
                                                      -__.                  .__      . ..      .  ._      _
.
.
                                                          27
      E8.15 (Closed) Insoection Followuo item 50 313/97201 12: Review of Current Desion BM!s for
              Bvoassina Emergency Diesel Generator (EDG) Protective Trio
              Background
              For emergency starts of the Unit 1 EDGs, the licensee did not bypass those protective
              functions delineated by Regulatory Guide 1.9, " Selection, Design, and Qualification of
              Diesel-Generator Units Used as Standby Onsite Electrical Power Systems at Nuclear
              Power Plants," to be bypassed under accident conditions. This was because the
              licensee was committed to Safety Guide 9, " Selection of Diesel Generator Set Capacity
              for Standby Power Supplies," which does not specify protective functions to be
              bypassed. The concern was that the licensee's setup could result in a higher probability
              of EDG loss during accident conditions. This item was left open pending review by the
              NRC program office,
              insoector Fougsp
              Based on discussions with the NRC program office, the inspectors determined that the
              licensee's adherence to Regulatory Guide 1.9 was satisfactory.
      E8.16 (Ooen) U.Q[esolved item 50-313/97201 14: Lack of Testina Unit 1 Molded Case Circuit
              Breakers
              Background
              The Updated Final Safety Analysis Report identified in Section 8.3 that the Class 1E
              electrical distribution system was designed to meet the requirements of IEEE Standard
              3081971, *lEEE Standard Criteria for Class 1E Electric Systems for Nuclear Power
              Generating Stations ' Section 6.3 of this standard required periodic testing at scheduled
              intervals to demonstrate that components that were not exercised during normal
              operation of the station were operable. The standard further stated,'The specific tests
              and the frequency at which they are performed depend upon the specific components
              installed, their function, their environment, and the fact that they are in the maintenance
              program of the unit (s). Illustrative examples of tests are given in Table 2." Table 2 of the
              standard identified switchgear and one of the tests identified for illustrative purposes for
              that type of component was an overcurrent trip test.
                In contrast to these requirements, the licensee did not have a program to perform
                periodic overcurrent trip testing of its Unit 1, Class 1E, molded-case circuit breakers.
                The licensee developed a position paper that concluded that IEEE 308-1971 was not
                applicable to molded-case circuit breakers. This conclusion was based on an
                interpretation that Table 2 of the standard identified four tests for 600 V and below
                switchgear: operation test, mechanical inspection, overhaul, and overcurrent trip test.
  _ __                                                                                                  _
                                                                _ - . - - .
 
- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _                            _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ______________ _ ___
                                                                                                                                                                                                                              i
                    .                                                                                                                                                                                                        :
                    .
                                                                                                                                                                                                                          28-
                                                                                    Due to the physical construction of molded caso circuit breakers, mechanicalinspection
                                                                                    and circuit breaker overhaul were not possible. Therefore, the licensee concluded that
                                                                                    the perindic tests in Table 2 were not applicoble to molded-case circuit breakers but were
                                                                                    rather intended for load center switchgear.
                                                                                    The inspectors considered the licensee's interpretation of Table 2 of IEEE 3081971 to
                                                                                    be in error, based on the following points:
                                                                                    *      Table 2 identified illustrative periodic tests for Class 1E components.
                                                                                    .
                                                                                            The identified tests were not an exhaustive or restrictive lists nor was Table 2                                                  i
                                                                                            intended to exclude components for which it was not physically possible to                                                        l
                                                                                            perform all of the identified tests.
                                                                                                                                                                                                                              )
                                                                                    *      The user of the standard was expected to evaluate which of the identified tests
                                                                                            (or others) were appropriate for each component.
                                                                                    As a result of its operating experience assessments for NRC's Information Notices 92 51,
                                                                                    * Misapplication and inadequate Testing of Molded-Case Circuit Breakers" and 93-64,
                                                                                    * Periodic Testing and Preventive Maintenance of Molded-Case Circuit Breakers," the
                                                                                    licent,ee initiated a practice of performing electrical testing of molded case circuit
                                                                                    breakers in the receipt inspection process and performing mechanical exercising of the
                                                                                    breakers on a periodic basis. However, they did not implement the practice of periodic                                                    ;
                                                                                    electrical testing. Information Notice 93-64 referred to Electrical Power Research
                                                                                    Institute Report NP 7410," Breaker Maintenance." Volume 3 of this document," Molded
                                                                                    Case Circuit Breakers," recommended electrical testing of molded case circuit breakers
                                                                                    on a 4 to 6 year periodicity.
                                                                                    Prior to the inspection documented in NRC Inspection Report 50-313/97 201, the
                                                                                    licensee performed a self assessment that identified that periodic electrical testing of
                                                                                    molded-case circuit breakers was not performed. The licensee established a task group
                                                                                    to determine if a testing program, replacement program, or a combination of both should
                                                                                    be adopted.
                                                                                    Insoector Followuo
                                                                                    The inspectors conducted interviews with personnel and reviewed the licensee's position
                                                                                    paper on molded-case circuit breaker testing and plans for developing a periodic testing
                                                                                    program for molded-case circuit breakers.
                                                                                    The licensee identified that they performed several activities related to safety related
                                                                                    molded-case circuit breakers to venfy that they will function under design basis
                                                                                    conditions and that the breakers are maintained accordhg to vendor requirements.
 
  ,
  *                                                                                                      F
>
                                                      2g.
          These activities included: testing Unit 2 electrical penetration breakers in accordance
          with the Technical Specifications, tet, ting all new breakers prior to installation, preventive
          maintenance inspection and cleaning of breakers and panels, manual cycling of some
          breakers, and performing periodic thermography inspections of 480 V molded case
          circuit breakers. The licensee stated that these testing activities have historically
          resulted in very low failure rates.
          The licensee informed the inspectors that they had undertaken a project to develop a
          periodic electrical testing program for molded case circuit breakers. Specifically, the
          licensee stated that a program would be developed to test the Unit 1 containment
          electrical penetration breakers and the program would be implemented in the refueling
          outage scheduled for the spring of 1998. At the time of the inspection, the licensee had
          identified a sample population of 64 molded-case circuit breakers for inclusion in the
          program, of which, approximately 12 wcre scheduled for testing in the next refueling
          outage. Additionally, the licensee stated that it would categorize the remaining
          safety related Unit i breakers by risk significance and develop and begin implementation
          of a testing program by the end of 1998. With respect to Unit 2, the licensee identified
          that safety related molded-case circuit breakers that were not included in the Technical-
          Specification-required testing program would be evaluated for risk significance and a
          testing program would be developed and implemented by the Unit 2 refueling outage
          scheduled in 1999.
          Based upon the previous record of satisfactory Technical Specification-required electrical
          testing experience of Unit 2 molded-case circuit breakers and other maintenance and
          testing activities the licensee has implemented, the inspectors did not consider the
          licensee's lack of a periodic testing program for Unit 1 molded-case circuit breakers an
          immediate safety concern. However, the licensee's practices will be reviewed further
          with the NRC's program office to determine if the licensee's reference to IEEE 308-1971
          in its Updated Final Safety Analysis Report imposed a regulatory requirement to
          implement an electrical testing program for molded-case circuit breakers. This
          unresolved item was left open pending review of this issue by the NRC program office.
    E8.17 (Closed) Unresolved item 50-313/97202-15: Lack of includino all Safetv-Related Fuses
          in Proce&dC
            BacharWad
            The NRC identified that the licensee had not adequately resolved previous concerns
            re;ated to incorrect electrical fuse installations. The licensee had established a like for-
            like replacement policy for installed fuses but had physically verified only 4 percent of the
            fuses in Unit 1 to be correct in all aspects. The fuse control procedure required an
            engineering evaluation only for blown fuses or when replacement fuses were different
            from those installed. All other 3pplications involved a like for like replacement without an
            engineering review. Under these conditions, an existing incorrect fuse would be
            replaced with a like fuse. The NRC considered the like-for like fuse replacement policy
            to be inappropriate considering the small percentage of fuses validated.
                                                          _                          _
 
    -  .          _ _ _. _ _ _ _ _._ _ _ _ _ _ _ _ _ . ._ _ _ _ _                                                                    _ _ ._ _
  .
  *
;                                                                                                                                                ,
                                                                                -  30-
                    Insoector Followup
                    In response to the NRC finding, the licensee proposed to add the following DC
                    distribution equipment to the fuse control program by June 30,1998: D01, D02, D15,
                    D16, and D17 for Unit i and 2001,2002,2D26,2D27,2D41,2D42,2D51, and 2D52 for
                    Unit 2. Also, as a long term corrective action, the licensee stated that whenever a design
                    change was r ade to a safety related circuit protected by a fuse, a review of tha fuse list
                    would be performed by engineering to verify, correct, or add fuses to the procedure as
;                  necessary.
>
                    The licensee reviewed their condition report database for fuse-related discrepancies
                    occurring since January 1,1993. Seven condition reports during this period addressed
                    wrong size or type of fuse. Only one of these condition reports affected Unit 1, and this
,                    fuse was installed in a nonsafety related circuit. One was in a nonsafety circuit common
                    to both units and five affected Unit 2. Of these five, only three affected safety related
                    circuits. None affected circuit operability. Based on these facts, the licensee concluded
                    that the fuse control program was being effectively managed. The inspectors determined
                    that it was unlikely that a pervasive fuse control problem existed and, therefore, did not
                    consider the matter to represent a safety concern.
        E8.18 (Closed) Unresolved item 50-313/97201-16: FSAR Disneoancies
                    Backaround
                    The NRC identified five discrepancies in the license's FSAR, mostly involving
,                    Table 7-11 A, "R.G.1.97 Post Accident Monitoring Variables." The discrepancies                                              ~
                    involved inconsistencies between the table and design documents concerning instrument
                    ranges and display configurations. One additional discrepancy concerned a note in
                    FSAR Section 14.2.2.1.3.1 stating that both steam generators would be isolated in the
,                    event of a steamline break; whereas, in actuality, only one steam generator would be
                    isolated.                                                                                                                  i
                    Insoector Followun
!                    Prior to the identification of these discrepancies the licensee was in the midst of a
                    comprehensive FSAR review for both units. This program was described to the NRC in
                    Letter CNRO 97/00010 "EOl Licensing Basis Assessment and UFSAR Review
                    Initiatives," dated May 7,1997. The licensee stated that this effort would resolve the
,                  generic concerns iaised by this item since this review effort would identify the same
                    types of errors discovered by the NRC. The specific errors were to be reviewed and
                                                                        _
                    corrected using the ticensing document change process. These corrections were to be                                        ,
                    made effective in the 1998 submittal of the ANO Unit 1 FSAR.
                    The inspectors reviewed Letter CNRO-97/00010 and noted that it referred to a generic
                    review of the Unit 1 and Unit 2 FSAR for accuracy. The inspectors reviewed a document
                    entitled, " Safety Analysis Report Upgrade Project," that explained the scope of this effort,
                    currently scheduled for completion in July 1998. Based on review of this material and
;
                                                                                                                                                ,
    ,,      ,-,s.v          .,                                    ,_r.-  --
                                                                              ..,y,  ,-- -y , . w, - - - ,, -, -- - - , - __. , , ,,--- ---,-
 
  .
  .
                                                      -31-
          discussions with licensee personnel, the inspectors concluded that the project as
          pirnned would resolve the generic concerns related to this issue. At the time of this
          inswa ine licensee was approximately 50 percent complete with this project and had
          identified 895 potential discrepancies. Of these,391 vea strictly editorial, leaving 504
          items having potential significance. However, the licensee stated that many items may
          subsequently be deterrnined not to be FSAR errors since the reviewers have been
          instructed to docuinent everything that might be a discrepancy.
          The inspectors reviewed licensing Information Requests L97-0213, L97 0326, and
          L97 0310 and confirmed that the specific errors identified by the NRC were scheduled to
          be corrected in the next FSAR update.
          The inspectors determined that the licensee's efforts were sufficient to correct the
          specific and generic concerns associated with this inspection finding.
            10 CFR 50.71(e) states that the updated FSAR shall be revised to include the effects of
          all changes made to the facility or procedures as described in the FSAR. The licensee's
          failure to properly update the FSAR regarding the five discrepancies described above
          was considered a violation of 10 CFR 50.71(e). However, because of the favorable
            progress being made by the licensee's FSAR review team and the fact that a description
          of this project has been placed on the docket, and the NRC's view that the discrepancies
            identified by the NRC likely woule. have been identified by the licensee's program, the
            NRC is exercising discretion in accordance with Section Vll.B.3 of the Enforcement
            Policy and is not taking formal enforcement action for this finding.
    E8.19 (Closed) Unresolvnd item 50-313/97201-17: Ineffective Resolution of Previous
            Corrective Actions
            Backaround
            During a plant walkdown, the NRC noted that two in-parallel check valves located in the
            steam supply piping to the turbine-driven emergency feedwater pump (P7A) were
            continually chattering because of steam leakage through downstream solenoid valves.
            The NRC's concern was that wear from this mechanical action could eventually result in
            failure of the check valves to perform their reverse seating function,
            insoector FollQEuD
            The Wspectors were informed that leakage through the solenoid valves had been a
            problem at ANO since at least 1988 when CR 1-88 0015 identified this problem. The
            licensee had performed maint3 nance on the valves on several occasions, but the
            leakage problem persisted. At some point, the licensee determined that the small
            amount of leakage was not detrimental to the system, since the check valves were not
            experiencing abnormal wear and the heating of the downstream piping was considered
              beneficial in reducing the potential for condensate-induced turbine overspeed events.
1            Then, the licensee deferred any plan for modification. However, in October 1996,
              the licensee identified accelerated wearing of elastomeric surfaces on one of the
 
.
.
                                                    32-
      two solenoid valves in the EFW steam lines. As documented in Condition
        Report CR 196 0569, electricians found the O-ring on the operator of Solenoid
      Valve SV 2663 to be hard and brittle following its installation on July 14,1992. The
        licensee's equipment qualification program stated that this 0 ring was qualified for
        60 years of service but did not consider seat leakage to be a normal condition for the
        establishment of qualified life. The licensee determined that the solenoid valve remained
        operable with the damaged 0-ring. Based on the O ring failure, the qualified life of the
        O rings was being revised to be from 2 to 8 years depending on the service condition.
        While the licensee considered a modification to the solenoid valves, they were also
        considering installation of motor-operated globe valves and relecating the valveb closer
        to the turbine.
        The check valves performed a safety related function by isolating the piping from a
        faulted steam generator. The licensee stated that one of these valves would be
        disassembled and inspected during the next refueling outage in the Spring of 1998. In
        February 1995, the other check valve was inspected and found structurally sound with            ,
        slight indications of scoring. During past tests of the check valve's reverse seating
        capability, the results have been acceptable.
        The inspectors were informed that the check valves had chattered since their installation
        in 1986, even before the solenoid valves began to leak, because of the presence of a
        drain / strainer orifice that creates a constant steam demand of 1600 pounds-mass per
        hour. Since the check valves require iOOO pounds-mass per hour to remain constantly
        open, the solenoid valve leakage couid actually reduce the chattering if it were of
        sufficient magnitude. However, the current chatter rate of the check valves was 10 taps
        per second, indicating that the solenoid leakage was well less than 3400 pounds mass
        per hour.
        The inspectors determined that the licensee had adequately ensured that EFW system
        operabililty had not been compromised by the solenoid leakage problem. The inspectors
          recognized that the check valve chattering was partNily independent of the leakage
          problem and that the leakage had been observed tc to beneficial for preventing
          condensation in the EFW turbine steam supply line Gese facts mitigated any safety
          concerns re' < ed to this condition.
  E8.20 (Closed) Unresolved item 50-313/97201-18: Lack of Seismic Sucoort of OTSG Pressure
          Transmitters
          (Closed) Unresolved item 50-313/97201-19: Lack of Desian Basis for Sucoort of Steam
          Generator Instrument Sensina Lines
          Backaround
          The NRC observed that a clip designed to restrain the isolation valve and sensing line for
          Steam Generator Pressure Transmitter PT 2667B was missing. The concern was that
          the line could fail during a seismic event, potentially affecting the operability of the steam
                  _ _ _ _ _ _ _ _ _ .
 
.
.
                                            33-
  generator pressure controlloop. The NRC also observed that the corresponding clips on
  the three other redundant pressure instrument channels in the same vicinity were loose.
  The licensee initiated Condition Report CR 197-0058 to determine the operability status
  and potential generic concerns. Before the end of the inspection, the licensee reported
  to the NRC that the as found configuration was operable and that a walkdown of other
  instrument tubing restraints indicated that the mounting clip problem was not widespread.
  In the Unit 1 EFW pump room, the NRC found a 10-foot unsupported span of instrument
  tubing for a pressure tranomitter (PT/PI 2811). This is a safety related sensing line that
  monitors the discharge pressure of the turbine driven emergency feedwater pump. A
    m. mum unsupported span of 2.5 feet was generally specified in the seismic drawing
  details. The licensee issued Condition Report CR 107-0074 and, subsequently,
  determined that the as-found configuration was operable. The licensee determined that
  a generic issue existed and elevated CR 197-0074 to a significant condition, requiring a
  root cause determination. The licensee performed a detailed walkdown of the EFW,
  Decay Heat / Low Pressure injection, and High Pressure injection systems and identified
  the following conditions;
  *
          Several installations were undocumented or unanalyzed, such as tubing supports
          exceeding the 30-inch unsupported span and use of nonstandard instrument
          supports.
  *
          Lack of design bases or criteria for determining instrument line slopes.
  *
          Drawing discrepancies between drawings and the as built condition.
  *
          Missing or incorrect tags on instrument valves.
  Based on these findings, the licensee expanded the scope to include other nystems. For
  example, Condition Report CR-197 0087 was later opened to document insidequately
  supported emergency diesel generator instrument lines. In all,37 instrument
  installations were found in Unit 1 that did not conform to either seismic or instrument
  installation requirements as defined in the standard drawing details. Design engineering
  deiermined that none of the 37 discrepancies created an inoperable condition.
  Insoector Followuo
  The inspectors reviewed Root Cause Analysis Report CR 197-0074," Instrument Tubing
  not Mounted in Accordance with Seismic Design Requirements," dated March 24,1997.
  The licensee identified the following root causes for this event:
  (1)    Inadeouate_trainina: Personnel involved in the installation and maintenance of
          instrument tubing were not adequately trained or sensitized to seismic and
          configurations requirements.
                                                                                _ _ _ _ - _ _ _ _
 
                    . _ .      . .    ._.          -      -  -- ..              .      ..      -
  .
  .
                                            34-
    (2)      Work oractics Personnelinstalling and maintaining instrument tubing either did
            not use required configuration documentation during work activities or did not
            follow these documents correctly.
    (3)      Desian confiaura@n and analygls: Design documentation and prints were
            inadequate or not easily accessible.
    The licensee walked down 40 instrument installations in Unit 2 and found that the quality
    of the installations was much better than in Unit 1. Only four of the installations
    inspected had notable deficiencies involving damaged mounting clips or brackets
    resulting from removal and reinstallation to support work on nearby equipment.
    The licensee performed the following corrective actions: (1) the individual discrepancies
    were repaired as necessary, (2) a letter was issued on April 4,1997, to site groups
    reiterating instrument tubing requirements, and (3) design engineering reconfirmed the
    initial operability assessments fiom the tubing walkdowns. Additionallong term actions
    included the following: (1) develop a scope for additional walkdowns, (2) determine
    whether existing seismic tubing mounting requirements can be relaxed, (3) evaluate
    enhancements to design drawings associated with tubing, (4) evaluate the work planning
    processes that involve instrument tubing or the removal of tubing for interference
    purposes, and (5) provide training on seismic and configuration requirements for
    instrument tubing.
    The inspectors discussed the status of the proposed long term actions discussed above
    with the licensee. A calculation was in the final stages of review that redefined the
    acceptable support span lengths for various configurations. Once this calculation was
    complete, the licensee was planning to perform additional walkdowns of those safety-
    related components in Unit 1 that were potentially susceptible to instrument tubing
    discrepancies. This effort was to be completed by May 1998. The licensee was also
    planning to provide personnel training and to clarify drawings that contained confusing
    tubing support information.
l  The inspectors determined that the licensee's completed and proposed corrective
    actions (each of which were assigned completion due dates) were sufficient to address
    the issues raised by this item and considered the overall response to be strong,
    particularly considering that none of the identified discrepancies had been evaluated to
    affect operability.
    10 CFR Part 50, Appendix B, Criterion 111, " Design Control," states that the design control
    measures shall provide for verifying or checking the adequacy of the design.
    The licensee's failure to provide measures to ensure the correct translation of seismic
    design requirements to instrument tubing installations and the maintenance of this
    configuration during plant operations in Unit 1 was identified as a third example of a
    violation of 10 CFR Part 50, Appendix B, Criterion ill (50-313/9721-02).
!
1                                              -.    _-.                            _              _
 
      .
      *
i
                                                              35-
                    Conclusions
                    The licensee's failure to provide measures to ensure the correct translation of seismic
                    design requirements to instrument tubing installations and the maintenance of this
                    configuration during plant operations in Unit 1 was identified as a violation.
              E8.21 (Gosed) Unresolygd item 50-313/9720120: Inadeauate Work Plan for the Controlof
                    Post-maintenance Testina
                    Backomund
                    The NRC identified a misaligned limit switch on Valve CV-4804, rector building vent.
                    The misalignment caused the actuator cam to drag on the limit switch arm toward the
                    end of the valve stroke instead of riding on the roller bearing. Repeated operation in this
                    configuration could have resulted in a faulty valve position indication. Although this
                    particular application was not safety related, there existed a concern that other safety-
                    related applications may have a similar misalignment problem.
                    The licensee recognized that the post modification plan and job order for this valve
                    (00954697) did not include a post maintenance test to verify proper operation of the limit
                    switch cam assembly. The licensee reopened this job order and accompanying work
                    plan with the intent to add steps to reconfigure and test the assembly. One other valve in
                    Unit 1 (CV-1845, quench tank sample valve) had a similar mounting design, and a similar
                    revision was scheduled for its work plan. The licensee performed a review for other
                    similar limit switch configurations but did not identify any other of these configurations in
                    either Units 1 or 2.
                    Insoector Followup
                    The inspectors reviewed the work plan revisions discussed above and observed that
                    steps were added to verify that the limit switch rollers were in proper contact with the
                    cams. The licensee considered that this verification in the work plan coupled with the
                    existing post maintenance test that ver;fies smooth motion and proper valve position light
                      sequencing would be sufficient to ensure that the condition would not recur (i.e., no
                      revision to the post modification plan was considered necessary). The inspectors
                      observed that the post maintenance testing problem,in this instance, did not affect
                      safety related equipment and determined that the licensee had adequately addressed
                      the specific and generic implications of this finding.
              E8.22 (Closed) Unresolved item 50-313/972121: Inadeauate 50.59 Review Associated v@
                      the Removal of the Unit 1 BWST Vacuum Breaker and Followuo Corrective Action
  ,
                      Backaround
                      This item hvolved an inadequate 10 CFR 50.59 safety evaluation performed by the
                      licensee for a temporary covering placed over the borated water storage tank vacuum
                      relief valve flange. The evaluation was inadequate, because it did not consider the air
                                            _
                                                                                                                  '
    ..  .. .    ..        ..
                                                            .
                                                                      . .  .
                                                                                ______ ____________
 
.
.
                                            36-
  flow restricting effects of a mesh screen that was an integral part of the temporary
  covering. The following is a summary of key events:
  *      On December 4,1996, while Unit 1 was at full power, the iicensee removed the
          8-inch diameter vacuum relief valve from the borated water storage tank for a
          surveillance test. When the valve was removed, the licensee installed a white
          plastic bag over the 8 inch diameter open tank flange to act as a foreign material
          exclusion (FME) cover. Two slits, each approximately 8 inches long, were cut in
          the plastic cover to allow air flow. This work was done using Job Order
          00952943. The job order stated that if suction for the tank was necessary, a
          supplemental vent path was provided by a 4 inch tank overflow line that was
          always open during plant operations.
  *      On December 5,1996, the system engineer observed that the tank flange was
          covered with what he believed to be a solid FME cover and initiated Condition
          Report 106 0663 to document the potentialinoperability of the borated water
          storage tank. Subsequently, the system engineer determined that the tank was
          operable since the FME cover contained two slits that would permit tank pressure
          and vacuum relief. The plastic bag was installed for approximately 30 hours.
          The licensee subsequently ter*ved the plastic bag and installed a blank flange
          on the tank flange with a 1 inch gap between the flange and tank flange to permit
          pressure and vacuum relief. Around the 1 inch opening, a mesh screen was
          placed to prevent foreign material introduction into the tank.
  *      Condition Report 196-0663 also documented the discovery of a through wall
          crack on the inlet flange of the removed vacuum relief valve and assessed the
          operability of the tank. The operability assessment did not address tank
          operability while the plastic bag was installed as an FME cover. The assessment
          only addressed the flange and 1 inch gap and stated that the blank flange with
          the 1 inch air gap installed over the tank flange would provide sufficient air
          movement to vent the tank during rapid drain downs. At this time, any air flow
          restriction from the installed mesh screen was not considered,
                                                                                              .
    *      On December 12,1996, the gap dimension was changed from 1 to 3-inches
          under Job Order 00952943. This was done to increase the venting capacity
          based on engineering judgement that the 1 inch gap may have been marginal.
    *      On January 22,1997, Significant Condition Report 197-0019 was initiated and
            identified that a safety evaluation was not performed for se plastic bag installed
            over the tank flange for approximately 30 hours. In addition, the condition report
            identified that the procedure for testing the valve (Procedure 1306.034, " Testing
            of Unit 1 Pressure Vacuum Relief Valve PSV 1617, PSV 2423, and PSV 1412,"
            Revision 2) did not allow removal of the vacuum relief valve during power
 
  ._ _ __ __ _ __- _ _ _ _.. _ __ _ _ _ _ _ _ _ . _ . . _ . _                                                                                        _ _ _
      *                                                                                                                                                            -
      .
.
                                                                                                      37                                                            -
.                                              operations and that the safety evaluallon for the procedure did not include
l                                              consideration for valve removal at powcr. The condition report included an                              _
                                                                                                                                                                    >
                                              operability assessment for the tank when the plastic t sg was installed on the tank
i                                            flange. The licensee sta;ed that the tank was operable based on engineering
                                              judgerrent that the plastic bag would break before an internal vacuum would
                                                                                                                                                                    '
,
                                              cause damage to the tank.
                                                                                                                                                                    '
                                    +        On February 3,1997, the licensee issued Condition Report CR 197-0031 to
                                              address the previously made assumption that the 4-inch overflow line would, by
                                              itself, provide sufficient vacuum relief. The conclusion of this effort was that,
<                                              contrary to prior judgement, the 4 inch line was not large enough to prevent the
                                              formation of an internal vacuum in the tank during rapid draindown.                                                  .
                                    *        On February 11,1997, the licensee prepared Temporary Alteration 971-001 that
                                              documented the removal of the vacuum relief valve and the installation of the
i                                              blank flange with a 3 inch gap over the tank flange. The temporary alteration was
                                              prepared approximately 2 months after the actualinstallation. Temporary
                                              alterations were not issued for the previous two temporary coverings (plastic bag
                                              and 1 inch gap flange connection). Temporary Alteration 971001, which                                                >
                                              contained the first 10 CFR 50.59 evaluation performed in association with these                                      .
                                              events, concluded that the borated water storage tank was operable with the                                          ;
                                              3 inch gap blank flange installation.                                                                                j
'
                                    *        On March 4,1997, in response to concerns from the NRC, Condition Report 1
,
                                              96-0663 was amended to incluc's a discussion of the air flow blockage caused by
                                              the mesh screen over the 1 inch gap. After considering the screen restriction, the -
                                              licensee concluded that the borated water storage tank was still operable under
                                              this configuration.
,
                                    insoector FolloWMD
                                    The inspectors reviewed the documents referenced above and had discussions with
                                    licensee personnel, in review of the facts, the inspectors identified several concerns in
                                    addition to those addressed in the previous inspection, as discussed below.                                                    <
                                    Procedure issu_es
'
                                    The inspectors found, in agreement with Significant Condition Report CR-197 0019, that
                                                                                                                                                                    '
                                    the licensee's use of Procedure 1306.034 during power operations was improper. This
                                    procedure was originally written for use during refueling outage conditions as indicated                                      ;
                                      by paragraph 4.3.1 of the procedure, which stated that inspection of the valve would be
                                      performed each refueling outage. The 10 CFR 50.59 evaluation for the procedure did
4
                                      not address performing maintenance or testing of the vacuum relief valve while the plant
                                    was at power. Also, the procedure did not address temporary provWons for vacuum
                                      relief, which would be an essential element for online maintenance of the vacuum relief
                                      valve. Although one of the corrective actions from the condition report was for the
                                                                                                                                                                    ,
            ,--v--e    -    - ,,,,,                y      ,-e , , , . ,-, ~ . ~ , - , ., , - ,,-r ,    - , , - , , - - , ,yw --e , ,e~ n.- -,--a, n-      -rm.,.
 
  .
  .
                                              38-
    licensee to revise Procedure 1300.034 to allow valve removal and maintenance of the
    VMye while at power, this revision had not been corr,pleted at the time of this inspection.
    10 CFR Pr.rt 50, Appendix B, C6terion V, require', thst activities affecting quality.shall be
    oresenbod by docum6ntsd procedures of a type appropriate to the circumstances and
    shall be accomplished !,1 accordance with these procedures. The licensee's failure to
    comply with the plant condition prerequisites of Procedure 1306.034 was identified as an
    apparchtTiolent.,n (50 341iW21-03).
    Desian Controlinna
    The inspectors determined that the licensee had performed temporary changes to the
    borated water storage tank without the use of proper design control measures. The
    licensee removed the tank vacuum relief valve and sequentially installed three temporary
    covering configurations, each of which constituted a change to a safety related
    con'oonent, without use of the temporary alteration procedure (a temporary alteration
    wat eventually prepared for t,1e third configuration, but this was issut j approximately 2
    months after installation).
    The inspectors reviewed Procedure 1000.028," Control of Temporary Alterations,"
    Revision 20. Attachment i to this procedure listed the criteria for requiring a temporary
    alteration, which included a situation where a design change was made to safety-related
    equipment without the removal of the equipment from service. Using this criteria, the -
    inspectors concluded that, because the borated water storage tank was not taken out of
    service for these evolutions, a temporary alteration was required. The use of a
    temporary alteration was intended to ensure that plant changes were subjected to design
    control measures commensurate with the original design.
      10 CFR Part 50, Appendix B, Criterion til, requires those design changes, including field
    changes, shall be subject to design control measures commensurate with those applied
    to the original design. The licensee's failure to provide design control measures
<
    commensurate with initial design for three temporary changes made to the borated water
    storage tank vacuum relief valve was identified as an apparent violation
    (50-313/9721 04).
      Operability issues
      The plastic bag was attached to the tank flange in a manner such that the 8-inch
'
      ventilation slits may not have been pulled into the opening during an event (e.g., a safety
      injection event due to a large nreak loss of coolant accident that would cause a rapid
      tank draindow). Therefore, the venting function of the line could have been defeated.
      This would have left the 4-inch overflow line as the only venting source for the tank,
      which, according to the licensee's calculations, was not adequately sized to prevent a
      vacuum from forming in the tank during a rapid tank draindown. Consequently, the
      inspectors questioned the operability of the borated water storage tank during the 30
                            -__-
 
  .
  .
t                                              39
    hours that the plastic bag was attached to the vent flange. Because of a lack of
    redundancy, an inoperable borated water storage tank would render the entire
    emergency core cooling system inoperable. The licensee performed several tests and
    calculations in an attempt to support the original engineering judgement that the tank
    was operable in this configuration. Following the onsite inspection period; the licensee
    provided the following evaluations and calculations for NRC review;
    *        *BWST (borated water storage tank) Past Operability," internal memorandum
              from Don Phillips to Charles Tyrone, dated November 25,1997.
    *        Calculation 97 E-02100 01, "BWST Venting Capability Through 4 inch Overflow,"
              Revision 0, December 3,1997.
    *        Calculation 97 E 021101, "BWST Level Analysis," Revision 1, December 4,
              1997.
    .
              Calculation 97 E-0209-01," Evaluation of BWST Tank T 3 for Potential Vacuum
              Load in Support of Operability Determination associated with CR-196-0663,"
              Revision 0, December 8,1997.
    The licensee concluded that borated water storage tank was operable during the period
      of time the plastic bag was attached to the vacuum relief valve flanga. However, the
      analyses clearly demonstrated that the borated water storage tank was in a degraded    I
      condition and could have plastically deformed above the water level in the tank in the
      event of a large or medium break loss of coolant accident. This deformation, according
      to the licensee's evaluation, would not have prevented the operation of the emergency
      core cooling system as designed. Also, the effect of the internal vacuum on the tank
      level instrumentation could have caused an early transfer of emergency core cooling
      water system suction from the borated water storage tank to the reactor building sump.
      This early transfer would have reduced the net positive suction head margins for the
      emergency core cooling system pumps. The licensee determined that this would not
      have rendered any of these pumps inoperable.
'
      10_CFR 50 59 Issues
      The inspectors determined that the licer,see had failed to evaluate whether temporary
      changes made to the borated water storage tank had created an unreviewed safety
    - question.
      10 CFR 60.59, " Changes, Tests and Experiments," permits the licensee to make
      changes to the facility and to procedures as described in the Safety Analysis Report
      without prior Commission approval provided the change does not involve a change to the
      Technical Specification and does not involve an unreviewed safety question. A
                                                                                                !
 
  .
  .
                                                    -4 0-
    proposed change, test, or experiment shall be deemed to involve an unreviewed safety
    question if a possibility for an accident or malfunction of a different type than any
    evaluated previously in the safety analysis report may be created. The licensee shall
    maintain records of changes in the facility, and these records must include a written
    safety evaluation that provid6s the bases for the determination that the change does not
I  involve an unreviewed safety question.
    The inspectors identified the following two examples where the licensee made changes
    to the borated water atorage tarik vacuum relief line without adequately eva!uating
    whether an unreview6d safety question existed:
    *        The licensee reinoved the existing vacuum relief valve and covered the tank
              flange with a plaatic bag taped to the tank. In the absence of a safety evaluation,
              the licensee used engineering judgement to consider the tank operable under this
              configuration. However, when questioned, the licensee had to perform extensive
              testing and analysis beicte concluding that the original engineering assumption
              was correct and that the tank was operable. The installation of the plastic bag
              appeared to create the possibility for an accident or a malfunction of a different
              type than any evaluated previously in the Safety Analysis Report and may have
              created an unreviewed safety question.
    *        The licensee installed a blind flange separated by a 1 inch gap from the tank
              flange and installed a foreign material exclusion screen over the gap. The
              unreviewed safety question evaluation of this configuration failed to consider the
              air flow restriction created by the foreign material exclusion screen.
    The failure to properly determine whether an unreviewed safety question existed as
t
    required by 10 CFR 50.59 was identified as an apparent violation (50-313/9721-05).
    Conclusion
    Three apparent violations were identified associated with temporary coverings placed
      over the borated water storage tank vacuum relief valve flange, when this valve was
      removed on December 4,1996, for testing and maintenance. The apparent violations
      involved the failure to perform proper safety evaluations as described in 10 CFR 50.b9,
      the failure to perform the vacuum relief valve testing under the plant conditions as
      prescribed in the test procedure in accordance with the requirements of 10 CFR Part 50,
      Appendix B, Criterion V, and the failure to utilize the correct design control processes
      (temporary alterations) for the temporary cover configurations placed over the valve
      flange in accordance with 10 CFR Part 50, Appendix B, Criterion Ill.
                      ,.        ..      . . .  .. .
                                                                  .
                                                                                      ..
                                                                                                  .
                                                                                                    -o
 
    -_.    . _.    -          .                . _ - . ---- - -      -
  0
  0
                                                          41-
    E8.23 (Cloigd) Unresglyed item 50-313/9720122: Revised Calculations Associated with
          BWST Vonexing and Pumo NPSH
          Background
          The licensee prepared an engineering report,93 R 1002-01, "ANO-1 BWST Outlet
          Vortex Suppressor," dated February 5,1993, to examine the potential for vortexing in the
          BWST resulting in a loss of net positive suction head (NPSH) for the emergency core
          cooling pumps. The conclusion of the report was that the ECCS pumps would have
          sufficient NPSH. Within this report, the licensee identified that it had failed to consider
          the effect of instrument error in the vortexing calculation. This meant that the actual level
          in the tank at the time of switchover from the BWST injection mode to the recirculation
          mode may have been less than the nominal value, bringing the hydrodynamic system
          closer to a state of cavitation and air entrainment than had been originally assumed. The
          licensee issued Condition Report CR 197-0039 to investigate this condition. Within this
          report, the licensee concluded that the ECCS system would have remained operable.
          The licensee concluded that only the high pressure injection pumps would be susceptible
          to an NPSH problem but only at low tank levels near the switchover point to the
          recirculation mode. At this time in the accident, the high pressure injection (HPI) pumps
          could be shut down because the low pressure injection pumps alone could provide
          adequate core cooling. Within CR 197-0039, the licensee also was to complete
          Revision 1 to Engineering Report 93 R 1002 01 and to determine whether other
          calculations may have been affected by a similar failure to consider instrument error.
          Insoector Followup
          The inspectors reviewed the documents listed above and discussed the issue with
          licensee engineers. The inspectors determin. d that this instrument error was only
          reflected in the vortexing calculation and did not affect actualinstrument readings.
          10 CFR Part 50, Criterion lil, " Design Control," states that the design control measures
,        shall provide for verifying or checking the adequacy of design, such as by the
l        performance of design reviews.
          The licensee's failure to check properly the adequacy of Engineering
          Report 93 R 1002-01 was identified as a fourth example of a violation of 10 CFR Part 50,
          Appendix B, Criterion lli (50 313/972102).
          Conclusioni
          The licensee's failure to check the adequacy of an engineering report properly that failed
          to account for instrument error in a calculation of vortexing in the borated water storage
          tank was identified as a violation.
1
                                                                            _ _ -.
 
  .
  .
l                                                    -4 2-
    E8.24 [Gosed) Unresolved item 50 313/97202 23: InadtSuate 10 CFR 50.59 Evaluation
          6110 dated With BWST Releastts
          Background
          This issue involved a problem witi        10 CFR 50.59 safety evaluation performed for
          Temporary Alteration 97-1001, whi n is discussed in Section E8.23 of this report. In
          assessing the potential post loss of coolant accident (LOCA) radioactive release through
          the 3 inch gap foreign material exclusion (FME) covedag on the flange of the borated
          water storage tank (BWST) vent, the licensee addressed only offsite exposure and did
          not address control room exposure, which could ba considerably more significant. Also,
          the safety evaluation did not refer to the licensee's response to information Notice 9156,
          * Potential Radioactivity Leakage to Tank Vented to Atmosphere "
          Insoector Fe20Wup
          The licensee revised the 10 CFR 50.59 evaluation for Temporary Alteration 971-001 to
          include a discussion of control room dose resulting from potential backleakage to the
          BWST. In the evaluation, the licensee stated that no accident consequence analysis
          takes credit for holdup or filtering of radioactive releases to the BWST, and, therefore,
          back leakage to the BWST is beyond the licensing basis. Based on recent test results of
          the BWST supply and return valves, the licensee stated (in the revised safety evaluation)
          that little or no backleakage would be expected to occur post LOCA. Consequently, the
          licensee determined that there would be no change to the estimated control room
          dosage as a result of the temporary alteration.
          The licensee reviewed its response to information Notice 9156 and determined that the
          conclusions reached as a result of Temporary Alteration 97-1001 were not affected by
          this iirmation notice.
          The inspectors considered the lack of reference to control room dose in the original
            10 CFR 50.59 evaluation to be an oversight, but observed that two mitigating factors
            existed. First, the original evaluation did not quantify a release rate based on the
            assumption that the release would be negligible. Had the release been quantifiable, an
            assessment of control room dose would have been necessary. Also, the fact ' hat no
            credit was taken in the licensee's accident analysis for containment of BWST radioactive
            leakage relegated the issue to a beyond-licensing basis status, The inspectors
            considered the licensee's response to this issue to be adequato.
                                                                                                      .
 
  .
  .
                                                  -43
                                          IV. PlanLSunpart
1
    F2  Status of Fire Protection Facilities and Equipment
(
    a. Insoection Scooe (92903)
        During a walkdown of the Unit i emergency feedwater system pump room, which was
        performed in support of the review of various items discussed in Section E, the
        inspectors inspected the condition of the electrical cable fire barrier material,
    b. Qhstryations and Findings
        The inspectors identified a deficiency in the fire wrap for Conduit EJ2029, which provided
        protection for the control circuitry for the turbine-driven emergency feedwater pump. On
        a vertical run of the conduit, a collar joining two sections of fire wrap had loosened and
        slid down. This resulted in about a 2 inch portion of the conduit being exposed.
        The inspectors reported the condition to the licensee and showed the deficiency to a fire
        protection engineer and an operator. The licensee agreed that the as-found condition
        was unacceptable and immediately declared the fire barrier inoperable. In accordance
        with the requirements of its fire protection program, the licensee verified the operability of
        smoke detection equipment and control room alarm capability and established an hourly
        fire watch. Condition Report 197-0313 was initiated to document the item and Job
        Request 928623 was initiated to repair the fire wrap.
        The inspectors asked the licensee whether there was a program for inspecting fire
        barriers and whether the fire wrap in the emergency feedwater pump room was included
        within the scope of the program. The licendee provided the inspectors with
        Procedure 1307.062, * Unit 1,1 Hour Cable Fire Wrap Inspection," Revision 0. The
        inspectors reviewed the procedure, discussed fire wrap inspection techniques and the
        process of dispositioning inspection observations with licensee personnel, and verified
        that inspection of the pump room was included within the scope of the procedure. The
        inspection of the sub; set ' ire wrap was last performed with satisfactory results in July
        1997. The procedure was required to be performed every 6 months. At the time of the
        inspection, no explanation was available for the cause of the deficiency. The licensee
        determined that a noncompliance did not exist, because the surveillance interval had not
        been exceeded and the damage could reasonably be assumed to have occurred
        following the most recent inspection.
        The inspectors considered the licensee's actons on this matter to have been
        appropriate,
 
                                                                                                                      I
  O
  e                                                                                                                  i
                                                            +44 *                                                    r
'
          c.  Conclusions
                The inspectors identified a fire barrier deficiency and the licensee took prompt
                compensatory actions as required.
                                              V. ManSment Meetings
          XI    Exit Meeting Summary
                The inspectors presented the inspection resu'ts to members of licensee management by
                telephone on January 6,1998. -The licensee acknowledged the findings presented.
                During this meeting, the licensee stated that it will expand its definition of margin of
                safety as it pertains to the performance of safety evaluations performed under
                10 CFR 50.59. This issue is further discussed in Section E2.1 of this report.                        i
                The inspectors asked the licensee whether any materials examined during the inspection                !
                should be considered proprietary. The licensee sta..J that one document reviewed by
                the inspectors contained proprietary information. The proprietary information from this
;              document was not discussed within this inspection report.
                                                                                                                      ,
4
                                                                                                                      1
    . -..    ,                    -              , . . , .              ._ ._                        . . - ,m.._,
 
  .
i
l
  .
                                            ATTACHMEliT 1
                                    SUPPLEMENTAL INFORMATION
                              PARTIAL LIST OF PERSONS CONTACTED
    Licemen
    G. Ashley, Licensing SupeMsor
    E. Blackard, Design Engineer
    M. Cooper, Licensing Specialist
    G. Dobbs, Electrical and instrumentation and Control Supervisor
    J. Howell, Design Engineer
    R. Hutchinson, Vice President, Arkanshs Nuclear One
    R. Lane, Director, Design Engineering
    D. Mims, Licensing Director
    D. Phillips, Nuclear Steam Supply Syatem Supervisor
    J. Richardson, Design Engineer
    C. Tyrone, Director of Design Engineering
    T. Waldo, Supervisor, Modifications
    C. Zimmerman, Plant Manager, Unit i
    Nf1C
    J. Melfi, Resident inspector
                                  INSPECTIQfiERQCIQUEES USED
      37001            10 CFR 50.59 Safety Evaluations
      92903            Followup of Engineering issues
                              ITEMS OPENED. CLOSED. AND DISCUSSED
    Ooened
    50-313/9721-01                VIO    Failure to incorporate EFW Flow Limits into Procedures
                                          (Section E8.7)
    50-313/9721-02                VIO    Design Calculation Deficiencies (Sections E8.8, E8.11,
                                          E8.20, and E8.23)
    50-313/9721-03                APV    Failure to Perform BWST Vacuum Relief Valve Testing
                                          Procedure Under Proper Plant Conditions (Section EB.22)
    50-313/9721 04                APV    Failure to implement Temporary Alterations for BWST
                                          Vacuum Relief Valve Removal (Section E8.22)
                                                                                                  l
                                                                                                  '
                                                                                  _ _ _ _
 
.
.4
                                              2-
    50-313/9721 05          APV  Failure to Perform 10 CFR 50.59 Evaluations for BWST
                                  Vacuum Relief Valve Removal (Section E8.22)
    Closed
    50-313/9627-03          VIO  Failure to Notify NRC Within 1 hour of Declaration of NUE
    50 313/96-009          LER  Fire in Reactor Building During Heatup Resulted from
                                  Cracked Weld in Oil Line
    50-313; 368/96512-01013 EA    inadequate Lube Oil Collection Systems for Reactor
                                  Coolant Pumps
    50 313/96512-01023, and EA    Failure to identify and Take Prompt Action to Lube Oil
    01033                        Leakage
    50-313/97201 01        IFl  Licensee's Actions to Revise the Technical Specification
                                  Bases for the Minimum Water Vo!ume in the Condensate
                                  Storage Tank
    50 313/97201 02        URI  EFW Flow Rates Exceeding B&W Flow Rates
    50 313/97201-03        URI  EFW Piping Configuration Differences
    50 313/97201-05        URI  Evaluation of EFWP Pump Room Environment
    50 313/97201 06        URI  Drawing and Calculation Revisions
    50-313/97201-07        IFl  Modification Work Procedure Revisions
    50 313/97201 08        IFl  Add;tional Vendor infonnation Necessary for Revised
                                  Caole Pulling Calculation
    50-313/97201 11        IFl  Verification of Calculation Changes Associated With DC -
                                  Batteries
    50 313/97201 12        IFl  Review of Current Design Basis for Bypassing EDG
                                  Protective Trip
    50-313/97201 15        URI  Lack of including all Safety Related Fuses in Procedure-
    50 313/97201 16        URI  FSAR Discrepancies
    50-313/97201 17        URI  Ineffective Resolution of Previous Corrective Actions
    50 313/97201-18        URI . Lack of Seismic Support of OTSG Pressure Transmitters
    50-313/97201 19        URI  Lack of Design Basis for Support of OTSG Instrument
                                  Sensing Lines
    50 313/97201 20        URI  Inadequate Work Plan for the Control of Post-
                                  Maintenance Testing
  .50 313/97201 21        URI  Inadequate 50.59 Review Associated with BWST Breaker
                                            .
                                                                                            /
 
  .
                  .            .        .
                                                              .
                                                                                                          _ _ _ _ _ _ _ _ _ _
                                                                                                                              ,
o
D
                                                                3-
  50-313/97201 22                        URI      Revised Calculations Associated with Vortexing and
                                                  Pump NPSH
  50 313/97201-23                        URI      Inadoquate 50.59 Evaluation Associated with BWST
                                                  Releases
  Discussed
  50-313/97201-04                          URI    Inadequate Piping Pressure and Temperature
                                                  Specifications
  50-313;                                  URI    Consideration of Multiple Hot Short Actuations
  -368/9623-01
                                                                                                                                1
  50-313/97201 14                          URI    Lack of Testing of Unit 1 Molded Case Circuit Breakers
                                                LIST OF ACRONYMS USED
    A/E                  Architect / Engineer
    BWST                  Borated Water Storage Tank
    CR                    Condition Report
    CRN                  Calculation Revision Notice
    DC                    Direct Current
    DCD                  Design Configuration Documentation
    DCIMS                Design Configuration Information Management System
    EDG                  Emergency Diesel Generator
    EFW                  Emergency Feedwater
    FME                  Fore;gn Material Exclusion
    FSAR                  Final Sa'ety Analysis Report
    MOV                  Motor-operated valve
    NPSH                  Net Positive Suction Head
    QAT                  Quality Assurance Team
    SIMS                  Safety Information Management System
      ULD                Upper Level Document
    WBGT                  Wet Bulb Globe Temperature
                                                                                                                                  :
    ._    _ _ _ _ _ _ . -
 
                                                  .-__ - __ -__-___ -_-_                            _ _ _ _ _ - _
e
o
                                              AIIACHMENT 2
                      LIST OF DOCUMENTS REVIEWED FOR ITEM 50 313/97-201-06
      CALCULATIONS
      Lalculation A-86," Items Discussed with Daleas R.S. of CE MSLB, ECCS and LOCA,"
      Revision O.
      Calculation Quality Action Team Final Reput, September,1996.
      Calculation 32-1229968-00, " Final Draft, ANO AFW Capacity - Phase 1".
      Calculation 32 5000450-G0, "ANO AFW Capacity, Phase 2," Revision 0.
      Calculation 80-D-1083A-02, "EFIC DC Valve Torque Calculation Under Reduced Voltage
      Condition," Revision 1.
      Calculation 80-D-2085-01, " Correlation of Teletector to Activity Release," Revision 0.
      Calculation 80-D-2085-02, " Provide Correlation Between MR/hr Reading Detector Activity Steam
      Line Result Table Method," Revision 0
      Calculation 85-D-1042-03, "AMSAC Instr Loop Errors & Setpoints," Revish15.
      Calculation 85-E-0053-20, " Combustible Loading Calc for Fire Area G," Revision 29.
                                                                                                                    '
      Calculation 86-E-0002-01, "EDG 1 & 2 Load Study," Revision 8.
      - Calculation 86-E-0020-01, "lE Station Battery 2D11 Duty Cycle'& Sizing Calc," Revision 7.
      Calculation 87 E-0026-09, "ErM Pump Room Temperature Profile," Revision 0.
      Calculation 88 E-0086-01, "lE Bulletin 88-04 Review for P7A and P78 Minimum Flow
      Evaluation," Revision 0.
      Calculation 89-E-0029-02, " Cal Factor Calc ANO-2 MS Line Rad Monitor," Revision 0.
      Calculation 89-E-OiO2-01, " Terminal Voltage Calc for DC MOV," Revision 8.
      Calculation 89-E-0144-01, "EDG Loading for 2A3 & 2A4 Bus," Revision 3.
      Calculation 90-E-0116-01, "ANO-2 EOP Setpoint Document," Re.wision 4.
      Calculation 91 E-0090-06, " Heat Load Determination for Post Accident for Rooms 95 98 99100
        104109110 & 149 for Post Accident Cooling," Revision 2.
        Calculation 92-E-0021-01, " Emergency Duty Cycle and Battery Sizing Calculations," Revision 4
                                                                                                                  :
  .mu                          _ __ ___-___
 
                                                                                                              _ _ _ _ _ _ _ _
                                                                                                                                !
,,
  9
                                                                    2-
                  Calculation 92-E-0077-04, " Unit 1 EFW System Pump Performance Requirements,' Revision 0,
                  - Calculation 93-D-1008-05, 'MOV Sizing Calc for CV-2620,* Revision O.
                  Calculation 93-R-1029-30,'ANO-1 SRR for Auxiliary Building HVAC System ABH V,'
                  Revision O.
                  Calculation 94-D-5033-01, * Removal of Local Fire Trouble Horns from D21 Breaker 34,"
                  Revision 0.
                  Calculation 94-E-006101," Tech Spec Allowable Outage Time (AOT) Utilizing Attemate AC                          ,
                  EDG," Revision O.
                  Calculation 95-D-1004-03, 'DCP 95-1004 Load Addition to RC1," Revision O.
                  Calculation 95-D-1011-01,"EDG Load Add for DCP 95-1011," Revision 1.
                    Calculation 95-D 7064-01, " Vital AC Panels & Instrument AC Panels," Revision 0.
                    Calculation 95-D-7072-01,"D01 Battery Loading Amendment," Revision O.
                    Calculation 96-E-0065-01, " Unit i Safety Survivability Eval Under Stall Thrust / Torque,"
                    Revision 1.
                                                                                                                                    ,
                    Calculation 96-R 0002-02, " Generic Letter 89-10 Program Valves Limit Switch Setpoint,"
                    Revision 1
                                                                                                                                    j
                    CONDITION REPORTS
                    CR-C-96-0060, re Calculation Room Control and Discrepancies, March 22,1997.
-                  CR-C-97-0053, re Calculation 92-D-2001-49 Approval, February 11,1997.
                    CR-C-97-0058, re Upper Level Document Discrepancies, February 13,1997.
                    CR-C-97-0059, re Design Configuration Documentation Discrepancies, February 13,1997.
                    CR-1-96-0184, re Seismic Qualification Calculations, May 23,1996.
                    CR-1-97-0017, re Accuracy Analysis Calculations for EFW Flow Transmitters, January 21,1997.
                    CR 2-97-0304, re Discrepancy in Instrument Setpoint Calculations 2LT-5659-1 and 2LT-5668-2,
                    May 28,1997
    ..
      .. . . . .                                                                                                              .
 
                                                  -      _--_ ___-_-__                    _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
  %l
'
  o
                                                    3-
    DESIGN CHANGES
    Design Change Package 94-1003, "Feedwater (MFW Venturi Replacement)," Revision 2
    Design Change Package 95-2001, " Control Room Emergency Chiller Installation - Air
    Conditioning Modification," Revision 1
    ENGINEERING STANDARDS
                                                                                                                                      '
    Engineering Standard GES-39, " Calculation Numbering Schemes and Codes," Revision 2
    LICENSING INFORMATION REQUESTS
    LIR L97-0035, ' Action items from ANO 50.54(f) Design Bases Assessment: Maintenance of
    Calculations [1.1,1.2),",
    LIR L97-0037, " Action items from ANO 50.54(f) Design Bases Assessment: DCIMS
    Improvements [1.6],",
    LIR L97-0201,"NRC AE Design Basis inspection," July 31,1997
    LIR L97 0202,'NRC AE Design Basis inspection," July 31,1997.
    LIR L97-0289, " Response to IR 97-202 AE Design Basis inspection," September 28,1997
    LIR L97-0290, " Response to IR 97-201 AE Design Basis Inspection," September 26,1997
    PROCEDURES-
    Procedure OP-5010-015," Engineering Calculations," Revision 1 PC1,
    Procedure OP-5010-019, " Computer Software Control," Revision 1 PC1
    MISCELLANEOUS
    Memorandum ANO-97-00225, M. Stroud to ANO Engineering Personnel," Reminder of Design
      Considerations for Possible Impact to Battery and EDG Lording Calculations," March 3,1997
      Design Configuration Documentation Discrepancies 2M7-07-01,02,03,07,08,09,12,14.
      Design Configuration Documentation Discrepancies RBS 02-05,06
                                                                                                                                        d
                                                            _ --__-_-_ _                                                            a
}}
}}

Latest revision as of 23:05, 1 January 2021

Insp Repts 50-313/97-21 & 50-368/97-21 on 971027-31 & 1110- 14.Violations Noted.Major Areas Inspected:Engineering & Plant Support.Nrc Identified Fire Barrier Deficiency & Licensee Took Promp Compensatory Actions as Required
ML20202E340
Person / Time
Site: Arkansas Nuclear  Entergy icon.png
Issue date: 02/06/1998
From:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20202E294 List:
References
50-313-97-21, 50-368-97-21, NUDOCS 9802180126
Download: ML20202E340 (50)
v • d • e


See also: IR 05000313/1997021

Text

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EliCLOSURL2 i

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

-

- Docket Nos.: 50-313;50-368

License Nos.: DPR-51; NPF-6

Report No.: 50-313/97-21; 50 368/97-21

Licensee: Entergy Operations, Inc.

Facility: Arkansas Nuclear One, Units 1 and 2

]

- Location: Junction of Hwy,64W and Hwy.333 South

Russellville, Arkansas

Dates: October 27-31 and November 10-14,1997

, inspectors: M. Runyan, Team Leader

P. Goldberg, Reactor inspector

R. Bywater, Reactor Inspector

G. Kalman, Project Manager

'

Accompanying D. Denver, Consu' tant

Personnel

4

Approved By: T. Stetka, Acting Chief Engineering Branch, Division of Reactor Safety

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Attachment 1: Supplemental Information

l Attachment 2: List of Documents Reviewed for Unresolved item 50-313/97201-06

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9902180126 900206

PDR ADOCK 05000313

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EXECUTIVE SUMMARY

Arkansas Nuclear One, Units 1 and 2

NRC Inspection Report 50-313/97-21; 50 368/97 21

Engineerina

-

Within a review of selected documents, the licensee's 10 CFR 50.59 safety evaluation

program appeared to be functioning satisfactorily, with the exception that some

documented word searches of the Updated Safety Analysis Report for impact were

inaccurate. However, some apparent failures to implement the program were identified

(Sections E2,1 and E8.22).

.

The failure to procedurally incorporate a vendor recommended maximum emergency

feedwater flow limit of 1500 gpm to a single steam generator into plant operating

procedures was identified as a violation of 10 CFR Part 50, Appendix B, Criterion V

(Section E8.7).

The failure to properly verify the piping configuration design inputs to a design calculation

was identified as the first example of a violation of 10 CFR Patt 50, Appendix B,

Criterion lli (Sec' ion EP.8).

Overpressure of the turbine-driven emergency feedwater pump downstream piping

required further reviews by the licensee. The existing analysis used pump performance

crit 6 ia that was less than the manufacturer's specified criteria (Section E8.9).

-

The failure to properly verify the design adequacy in four ciesign calculations was

identified as additional examples of a violation of 10 CFR Part 50, Appendix B,

Criterion 111 (Section E8.11).

The lack of a review by the licensee of a sample of existing design calcriations to

determine if the types of past problems identified by the licensee existed within other

calculations was identified as a weakness (Section E8.11).

-

The lack of over current trip testing of Unit 1 molded case circuit breakers required

further review by the NRC (Section E8.16).

-

- Although the NRC identified five discrepancies in the Final Safety Analysis

Report (FSAR), the licensee's ongoing review of the FSAR was considered sufficient to

correct these specific examples and to resolve the resulting generic concern. Therefore,

enforcement discretion was taken in accordance with the NRC Enforcement Policy

(Section E8.18).

- ___

. _ - _ _ - - _ _ _ _ _ - _ _ _

.

9

-3-

.

The failure to provide measures to ensure the correct translation of seismic design

requirements to instrument tubing installations and the maintenance of this configuration

during plant operations in Unit 1 was identified as an example of a violation of 10 CFR

Part 50, Appendix B, Criterion lli (Section E8.20), ,

.

Three apparent violations were identified associated with activities involving the removal

of the borated water storage tank vacuum relief valve for maintenance and testing on

December 4,1996. The apparent violations involved the failure to perform proper safety

evaluations as described in 10 CFR 50.59, the failure to perform the vacuum relief valve

testing under the plant conditions as prescribed in the test procedure in accordance with

10 CFR Part 50, Appendix B, Criterion V, and the failure to utilize the correct design

control processes (temporary alterations) for the temporary cover configurations placed

over the valve flange in accordance with 10 CFR Part 50, Appendix B, Criterion 111

(Section E8.22).

-

The failure to properly check the adequacy of an engineering report that failed to account

for instrument error in a calculation of vortexing in the borated v<ater storage tank was

identified as an example of a violation of 10 CFR Part 50, Appendix B, Criterion 111

(Section E8.23).

Plant Succort

'

=

During a plant walkdown, the NRC identified a fire barrier deficiency and the licensee

took prompt compensatory actions as required (Section F2).

1

_ _ - _ _ _ _ _ _ _ .

_ - - _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - -_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _

.

.

-4

Report Details

Summarv of Plant Status

Units 1 and 2 were operated at 100 percent power during the inspection.

ll1. Enaineerina

E2 Engineering Support of Facilities and Equipment (37001)

E2.1 10 CFR 503,J Imolementation

a. Insoection Scoos (37001)

The inspectors reviewed the licensee's program to change plant design, revise

procedures, and conduct tests and experiments without prior NRC approval as described

under 10 CFR 50.59. The licensee's program was reviewed for compliance with

regulatory requirements and implementation of the program was evaluated by reviewing

three 10 CFR 50.59 evaluations. The inspectors reviewed the following documents:

ANO-1 10 CFR 50.59 Summary Report for 1996, deted May 22,1997

-

ANO 210 CFR 50.59 Summary Report for 1996, dated June 19,1997

50.59 Evaluations:

Temperature"

,

  • ANO-1, Condition Report 1-96-0667, " Reactor Coolant Pump Lift Pump Disabled"
  • - ANO-2, Design Change Package 94-2007," Replace Charging Pump .

Downstream Check Valves CVC-22A, B, and C"

, Procedure 1000,131, "10 CFR 50.59 Review Program," Revision 3

l

b. Observations and Findings

-Within the review of the listed documents, the inspectors did not identify any

!- programmatic deficiencies in the licensee's 10 CFR 50.59 program. It was noted that

the licensee chose to evaluate the " margin of safety" of proposed changes by evaluating

the changes only against the basis section of the Techn! cal Specifications. This practice

L had been previously questioned by the NRC in Inspection Report 50-313:368/97-04 and

!

,

. - , , , - -

.

.

~5-

found to be in compliance with 10 CFR 50.59 (however, a violation for a failure to

perform a safety evaluation was issued). No deficiencies in the " margin of safety"

evaluations were identified by the inspectors during review of the 10 CFR 50.59

evaluations. Although currently in compliance, the licensee indicated that a more

encompassing design basis for margin of safety determinations will be implemented in

the future.

While review of the?e safety evaluations indicated appropriate programmatic controls,

some apparent failures to implement the 10 CFR 50.59 program were identified as

discussed in Section E8.22 of this report.

The inspectors also identified problems with the licensee's search capability for the

U) dated Final Safety Analysis Report. For example, Safety Evaluation 96-141 stated,

"The lube oil pumps are not describad in the Technical Specifications, Safety Analysis

Report, or operating license." The inspectors determined that this was in error. The lube

oil pumps were discussed in Section 4.2.2.6 of the Updated Final Safety Analysis Report.

The inspectors identified similar problems with the safety evaluatici determinations for

the procedure changes. The inspectors discussed these findings with the licens6e and

concluded that the automated search methodologies used to search for key words in

licerwag basis documents were not always effective in finding relevant reference

matenal and that the techt,ical reviewers were not ensuring that the scope of review was

adequate. The licensee agreed with these conclusions and informed the inspectors that

they had also identified this issue in the past year and had revised the safety evaluation

procedure to address the deficiencies. The inspectors review of Procedure 1000,131

indicated that improvements had been made in the Lensee's process to ensure that

adequate automated searches were performed and that they were reviewed by a second

individual.

c. Conclusions

Within the review of selected documents, the licensee's 10 CFR 50.59 safety evaluation

program, when implemented, appeared to be functioning satisfactorily, with the

exception of Updated Final Safety Analysis Report key word searches. However, some

apparent failures to implement the program were identified as discussed in

Section E8.22 of this report.

EB Miscellaneous Engineering issues (92903)

E8.1 (Ocen) Unresolved item 50-313/9623-01: Consideration of Multiole Hot Short Actuations

Backaround

This issue addressed the unplanned energizaiion of motor-operated valves (MOVs)

during a control room fire. This event, caused by a control cable shorting to an

energized conductor, is known as a hot short. Under certain configurations, a hot-

shorted MOV could stroke with the torque or limit switch removed from the control circuit,

causing the valve to stallinto its open or closed seat. Depending on the capacity of the

.

_ _ - . - _ _ _ _ _ _ . - - _ _

- - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -

'

.

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3

motor actuator and the strength of the valve, this action could damage the valve

sufficiently such that plant operators would not be able to reposition the va!ve using the

valve's manual handwheel. The licensee's emergency operating procedures addressing

a control room fire and subsequent evacuation of the control r00m assumed that muual

control of accessible valves was available The licensee interpreted existing regvations

tu require postulation of only one hot short during a control room fire, whereas the NRC

maintained that multiple hot shorts, which is defined as more than one MOV having a hot

short et the same time, were within the scope of the regulations. Independent from the

resolution of this matter, at the request of the NRC, the licensee agreed to perform a

calculation to determine which valves would be potentially susceptible to hot short

damage. This was the anly aspect of this issue reviewed during the current inspection.

Insoector Followoo

The inspectors rev.ewed Engineering Report 97-R-0004-01, "Engineenng Information

Notice 92-18 Evaluation," Revision 1 PC-1; Calcalation 96-E-0065-01, " Unit 1 Safety

Related Valve Survivability Evaluation Under Stall Thrust / Torque, Motor Operated

Valves," Revision 1; and Calculation 96-E-0066-01, " Unit 2 Safety Related Valve

Survivability Evaluation Under Stall Thrust / Torque, Motor Operated Valves," Revision 0.

The inspectors reviewed Calculation 96-E-0065-01. The purpose of this calculation was

to determine operability with reduced margin for a select group of safe shutdown MOVs

under stall thrust / torque conditions. The evaluations were intended to reflect one-time

operation under the stated conditions and were not intended for use as design input in

other analyses or calculations. An important assumption stated that the anelvsis was

valid only for the time it was performed since actual test data was used in soine

evaluations. The results showed that 37 of the 66 valves evaluated would be able to be

manually positioned to their safe position following postulated hot short conditions.

Twenty-nino valves were susceptible to damage. The inspectors determined that the

licensee had used a valid methodology for determining the maximum torque and thrust

that an MOV would experience during a hot short event, with one exception For MOVs

that did not have previous measurements of stem friction, the licensee used a plant

averaoe stem factor correlated to a stem friction coefficient vf 0.1342. The use of a

bounding stem friction coefficient for the untested valve population, approximately 0.08,

would have resulted in calculated torques and thrusts as much as 50 percent higher than

those reported in the calculation. However, based on a qualitative judgement by the

inspectors, application of the lower stem factors would not have significantiv changed the

number of valves determined as potentially incapable of being manually repositioned.

Engineering Report 97-R-0004-01 addressed multiple hot shorts, but in situations where

two or more valves were determined to be unable to be manually repositioned, the report

stated that it wat not credible that both or all would fail at the same time in the same

direction. The inspectors noted that this reasoning reiterated the licensee's position that

multiple hot shorts were not required within the regulations.

Based on the results of the calculations and engineering report listed above and in light

of the 29 valves in Unit 1 analyzed to be susceptible to hot shorts, the inspectors

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':

.

,

-7-

concluded that come safety functions could be lost if mu:tiple hot shorts occurred in a

particular manner. This was due to the fact that some of the motor operated valves may

be subjected to loads during a hot short that would render them incapable of being

manually stroked to their safe shutdown position. In some cases, as in the Unit 1 high

- pressure injection system, a total of eight valves would have to hot short in the same

direction to defeat the safety function of the system. However, the Unit 1 shutdown

cooling function could be lost with the hot short failure of only two vu.ves, CW1428 and -

CV 1429, decay heat cooler outlet valves. Both valves were catulated to be potent! ally

incapable of being manually repositioned following a hot short event.

The above discussion was illustrative of the potential impact of multiple hot short events

and was not a complete listing of all potentially affected systems. This item will remain

open pending further reviews by the NRC program office to define the regulatory

requirements for this matter.

E8.2 (Closed) Violation 50-313/96027-03: Failure to Notify the NRC Within 1 Hour of

Declaration of Notification of Unusual Event

L Backaround

On October 17,1996, the licensee declared a Notification of Unusual Event in

accordance with its Emergency Plan for n fire in the Unit i reactor building that lasted

! more than 10 minutes. The licensee notified the NRC Operations Center of a fire in the

reactor building within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the declaration of the Notification of Unusual Event.

'

Howover, the licensee did not inform the Operations Center that it had made a

declaration of one of the Emergency Classes (Notice of Unusual Event) specified in its

Emergency Plan in accordance with 10 CFR 50.72(a)(1)(l).

Insoector Followuo

The licensee determined that the cause of the violation was human error. The shift -

L engineer who made the initial contact with the NRC Operations Center provided details

l of the fire but neglected to inform the NRC that a Notification of Unusual Event had been

'

declared.

The licensee conducted training with notification communicators about this event and

NRC notification requirements. Additionally, the licensee reviewed applicable

procedures and revised Procedure 1903.011, " Emergency Response / Notifications,"

Revision 22, to require verbatim transmission of iaformation included on

Form 1903.011Y, " Initial Notification Message," Revision 22, to the NRC Operations

Center officer followed by transmitting the form to the NRC Operations Center vi-

l. facsimile. The form explicitly identified the emergency classification for the event. The

! inspectors verified the licensee's corrective actions had been completed and considered

these actions adequate to prevent recurrence of this violation.

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E8.3 (Closed) Licensee Event Report 50 313/96-009f Fire in the Reactor Buildina Durina

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Heatuo Resulted from a Cracked Weld in an Oil Line on a Reactor Coolant Pumo Motor

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Backaround

During a Unit i plant heatup on October 17,1996, a fire occurred in insulation around the

m ain feedwater nozzle ring on the Train B once through steam generator. A cracked

wekiin the discharge line of the Train B reactor coolant pump lube oil lift pump resulted

in oil spray being introduced into the insulation. The hot piping caused the oil to ignite. A

Notification of Unusual Event was declared, and fire brigade responders extinguished the

fire approximately 16 minutes after discovery The NRC conducted a specialinspection

to review the circumstances of the fire, continued safety of the plant, and corrective

actions taken by the licensee. The results of the inspection are documented in NRC

Inspection Report 50-313; -368/96-27. The !!censee initiated Condition Report CR-1-36-

0567, conducted an event evaluation, and documemed its evaluation results in a root

cause analysis report.

The licensee's event evaluation concluded that oil intrusion into the insulation caused a

wicking effect that lowered the auto-ignition point of the oil to a lower than expected

temperature.

.

The NRC conducted a special inspection documented by NRC Inspection

'

Report 50-313; -368/96-27, which identified problems with the licensee's implementation

of 10 CFR Part 50, Appendix R, fire protection requirements and that the licensee had

failed to identify and correct problems that were noted prior to the fire. These were the

subject of an NRC Escalated Enforcement Action (EA 96-512). The licensee developed

and implemented many corrective actions to correct the deficiencies in these areas.-

They are discussed in Sections E8.4 and E8.5 of this report.

Insoector Followuo

This item is closed based upon the followup and discussion in Sections E8.4 and E6.5 of

this report, which address the escalated enforcement items associated with this event.

- E8.4 (Closed) Enforcement Action 50-313:-368/96512-01013- Inadeauate Lube Oil Collection

Systems for Reactor Coolant Pumos

,

Backaround

During the NRC special inspection following the October 17,1996, Unit 1 reactor building

fire (NRC Inspection Report 50-313; -368/96-27), t1e NRC reviewed the tube oil

collection systems for all of the reactor coolant purnps in both Units 1 and 2. The

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inspectors determined that the lube oil collection systems for the UnP.1, Train B reactor

coolant pump and all of the Unit 2 reactor molant pumps had not been designed to

collect oilleakage from all pressurized and unpressurized locations. Also, the Unit 1

Trains A C, and D reactor coolant pumps had inadequate leakage collection from

some unpressurized locations. The deficiencies were determined to be in violation of

10 CFR Part 50, Appendix R, Section Ill, requirements.

Immediately following the fire, the licc nsee placed the unit in cold shutdewn to evaluate

the effects of the fire and water suppression on plant equipment. No significant damage

occurred other than to some insulation. The damaged insulation was repaired or

replaced as necessary. Other actions taken prior to restarting the unit included installing

a new shroud around the Train B reactor coolant pump motor and extending oil collection

drip trays for all of the reactor coolant pump motor:, to provide coveragv. af potential

leakuge points. Due to remaining concems regarding the integrity of othar portions of

the Unit 1, Train B motor lube oil piping, the licensee implemented administrative controls

to prohibit operation of the associated high pressure lift oil pumps unless a fire watch

was present. The motor manufacturer provided confirmation that the absence of tube oil

injection following an inadvertent pump trip would not adversely affect the coastdown

characteristics of the pump. This was used as supporting information in a 10 CFR 50.59

safety evaluation to conclude that operation without the lube oil!!ft pump in standby did

not constitute an unreviewed safety question.

In Unit 2, the licensee performed tube oil collection sysiem modifications following an

unplanned reactor shutdown in November 1996. The NRC performed walkdowns of both

'

units and reviewed the modifications. These activities were documented in NRC

Inspection Reports 50-313; -368/96-27, 50-313; -368/96-07, and 50-313; -368/96-08.

The NRC concluded that with the exception of the remote lube oil ti:: aystems for both

units, the tube oil collection system modifications es'.ablished compliance with 10 CFR

Part 50, Appendix R, requirements.

The Unit i remote oil fill lines (Tygon tubing) to the top of each motor were not provided

with a leakage collection system as required by Appendix R requirements. The licensee

provided administrative controls to prevent operation of the system until a modification

and any necessary regulatory exemption was applied for and obtained. The Unit 2

remote oil system was designed differently, using stainless steel tubing and stainless

steel flexible hose, but no oil collection system was provided. The licensee described the

Unit 2 system in detail and proposed compensatory administrative controls for its use in

an exemption request to the NRC dated December 23,1996. The NRC approved the

exemption request in a letter dated June 14,1937. The exemption allowed the licensee

to use the remote oil fill system without a collection system contingent upon the

licensee's implementation of several compensatory measures prior to each use.

The licensee responded to the Notice of Violation in a letter to the NRC dated May 9,

1997. Corrective actions were identified in this document, as well as the licensee event

report associated with this event, and sevaral condition reports. Some of these were

completed and reviewed previously as discussed above; others were reviewed during

this inspection.

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Insoector Followuo 0

The inspectors reviewed documents, procedures, an'.8 interviewed personnel to

determine the adequacy and completeness of the Sceasee's corrective actions. Also, the

inspectors contacted the c:gnizant fire protection engineer in the NRC Office of Nuclear

Reactor Regulation regarding the adequacy of the Unit 2 lube oil collection system.

Documents reviewed included: portions of the Updated Final Safety Analysis Report;

Condition Report 196-0507, its associated Root Cause Analysis Report and corrective

action documents (including 10 CFR 50.59 safety evaluations and screenings);

Procedure 1102.002, " Plant Startup," Revision 59; Procedure 1103.006, " Reactor

Coolant Pump Operation," Revision 20; Procedure 1107.001, " Electrical System

Operations," Revision 50; Procedure 1107.004, " Battery and 125V DC Distiibution,"

Revision 8; and Procedure 1504.001, " Visual Inspection of the Unit 1 & 2 RCP's Oil .

Colloction System," Revision 4.

{

Modifications were made to the tube oil collection systems for all reactor coolant

pumps in both units. These were determined by the NRC to be acceptable for meeting

10 CFR Part 50, Appendix R, requirements. To ensure the continued operability of the

systems, the licensee imp!*mented Procedure 1504.001, for inspection of the systems

immediately prior to refueling outages and prior to startup.

The licensee placed administrative controis on operation of the Unit 1 Train B reactor

coolant pump lift oil pumps. This war, done because of uncertainty about oil system

integrity during lift pump operation on the Jeumont Industrics motor Procedure

1103.006 was revised to require that the Train B reactor coolant pump not be started or

stopped during normal operations when the reactor coolant system temperature was

greater than 350 degrees F. Operation of the lift oil system was al! owed, but only if two

fire brigade-trained operators were present in the reactor building to identify oil leakage,

inform the control room, and extinguish any resulting fire.

'

The norrr.al design of the oillift system provided an autostart of the lift pump when the

reactor coolant pump tripped. The administrative controls discussed abova prevented an

autostart from occurring. Therefore, if a reactor coolant pump trip occurred, the pump

would coast down without injection of lubrication onto the motor thrust bearing. The

inspectors reviewed the licensee's 10 CFR 50.59 Safety Evaluation 96-141, " Evaluation

of Placing RCP #32B HP Lube Oil Pump (P-808 and P-638)in Pull-to-Lock," and safety

evaluation determinations for associated procedure changes. Supporting documentation

for the licensee's evaluation included a letter from the vendor documenting that reactor

coolant pump coastdown time would not be adversely affected for at least 15 seconds

without oil lift pump operation. The relevant accident analyses for consideration in the

safety evaluation were the various sizes of loss of coolant flow accidents, and the

evaluation determined that the lack of lift pump operation had no impact. The inspectors

concluded that the safety evaluation provided an acceptable basis for concluding that

reactor coolant pump operation without availability of tube oilinjection aid not constitute

an unreviewed safety question.

)

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-11

The licensee initiated an evaluation of the tube oil system piping configuration to ensure

that no additional questionable welds remained. The licensee and the motor supplier

determined that the lube oil piping sysiem was constructed in accordance with applicable

commercial standards with the use of qualified welders and adequate quality controls.

No other similar weld failures were reported in more than 200 other reactor coolant pump

motors manufactured by the motor suppiier. One difference in the subject configuration

was the addition of a backup direct current powered lube oillift pump and associated

piping. To evaluate this further, the licensee developed a plan to perform inservice visual

inspections of the piping at operating pressure, collect vibration data, and perform data

analysts. The licensee considered this option for addressing piping concerns preferable

to others considered, which included piping replacement, weld replacement, and

volumetric inspections with repair of rejectable welds. After discussing these options

with the licensee, the inspectors considered the licensee's approach to evaluating the

acceptability of the lube oil piping configuration to be adequate. Administrative controls

on lift pump operation would remain in effect during future operating cycles pending

completion of vibration analyses and continued visual inspections to identify any

remaining weld failures.

The inspectors determined that the licensee had completed corrective actions to

ensure that its reactor coolant pump oil collection systems were in compliance with

10 CFR Part 50, Appendix R requirements and that proposed corrective actions were

either completed or in progress with responsive completion dates.

E8.5 { Closed) Enforcement Actions 50-313/96512-01023 and 50-313/96512-01033: Failure to

ldentify Sianificant Conditions Adverse to Quality and Take Promot Corrective Action

Relatina to Oil Accumulation in Pioe Insulation.

Backcround

'

The NRC (and the licensee) determined that there were previous opportunities to identify

that a problem existed with the reactor coolant pump oil lift system pr'ar to the fire event.

These opportunities were missed, in part, because a condition report was not initiated to

thoroughly evaluate the extent of the potential problem. Some of the opportunities to

identify the scope of the problem included: discovery of a crack in the lift oil pump

<

discharge line, the magnitude of the crack and inadequacy of the original weld, discovery

of oil on the side of the steam generator during reactor building walkdowns during plant

heatup, discovery of oil puddled under the reactor coolant pump, and identification of an

excessive amount of smoke present in the reactor building during heatup.

To correct these deficiencies in problem identification, communications, and

understanding of oil-insulation interaction consequences, the licensee revised

procedures and conducted training.

10soector Followuo

The inspectors reviewed documents, procedures, and interviewed personnel to

determine the adequacy and completeness of the licensee's corrective actions.

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Documents reviewed included: Procedure 1000.104, ' Condition Reporting and j

Corrective Actions," Revision 13; Procedure 1015.036, ' Containment Building Closeout,'

'

Revision 5; Procedure 2102.002, ' Plant Heatup," Revision 10; Procedure 1102.002,

' Plant Startup,' Revision 61; Procedure 1102.010, 'P! ant Shutoown and Cooldown,"

Revision 41; Procedure 2102.010, " Plant Cooldown,' Revision 30; Procedure 1203.034, ,

' Smoke, Fire, or Explosion," Revision _10; Procedure 2203.034, " Fire or Explosion,"

Revision 5; and training outlines and attendance records for Condition Report 196-567.

The licensee's training appeared to adequately address the circumstances and causes

of the fire and provided information on cr,uses of insulation fires; uses and limitations of

material safety data sheets; importance of proper delivery, receipt, and accuracy of

communications; and lessons learned in response to oil spills and observations of

smoke.

The condition reporting procedure _was revised to clearly address identified deficiencies

in fire protection system components including oilleaks and spills. Items of this type that

meet the definitions in the procedure now require the initiation of a condition report. The

other procedures provided requirements to perform inspections for oilleakage and

excessive sn.oke and provided actions to take in response to identified adverse

conditions. These actions could include terminating plant heatup, posting fire watch

personnel with fire fighting equipment, contacting the fire protection engineer, and othar

appropriate act.ans. The procedures also contained guidance on the effects of oil

soaked insulation on the auto-ignition temperature of the oil.

The inspectors concluded that the licensee's corrective actions were adequate to

improve staff understanding of the mnsequences of oilleaks, and the importance of

communications when adverse conditions are identified, provide actions to prevent fires

from occurring, provide additional requirements for inspections, and provide actions to

take when adverse conditions are identified.

E8.6 (Closed) Insoection Followuo item 50-313/97201-01: Licensee's Actions to Revise the

Technical Soecification Bases fo,- the Minimum Water Volume in the Condensate

Storcae Tank

3ackaround

The minimum water level in the Unit 1 safety-related condensate storage tank was based

on the capability to provide a sufficient supply of water to permit the initiation of the decay

heat removal system within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or less of an accident initiation. However, the actual

period of time required to initiate decay heat amoval may be substantially greater than 4

hours; The licensee acknowledged that this text was ir, accurate and proposed to revise

the existing Technical Specification bases to eliminate (probably) reference to the decay

heat removalinitiation and to state that the minimum condensate storage tank level was

based on the volume that was required for 30 minutes of emergency feedwater operation

before manual switchover of the pump suction to the service water system.

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Insoector Followuo

The licensee stated that they were in the process of neveloping the improved Standard

Technical Specifications and that the appropriate condensate storage tank bases would

be included in the initial submittal. The licensee stated that they planned to submit the

revised Technical Specification design base along with the revised Technical

Specifications by the end of 1998. The inspectors verified that there was no safety

concern related to the minimum condensate storage tank water level and that there was

no need on the part of tne licensee to revise this level. However, a clarification of the

Technical Specification basis for the level was appropriate.

E8.7 (Closed) Unresolved item 50-313/97201-02: Licensee's Failure to incoroorste Maximum

Emergency Feedwater Flow limits into Plant Procedures to Monitor and Preclude

Exceedina Recommended Limits

Backaround

A 1991 Babcock and Wilcox Company engineering report (92-R-1019-01) identified that

the maximum emergency feedwater flow rate to preclude damage to the Unit 1 steam

generator tubes was 1500 gpm. The NRC determined that the licensee had not revised

plant procedures to incorporate this upper flow limit. Specifically, Procedure 1010.002,

"hansient History / Transient Cycle Logging,"in effect at the time was not revised. The

NRC reviewed operator logs and identified two occurrences where the emergency

feedwater flow exceeded the 1500 gpm limit.

Insoector Followuo

The inspectors reviewed Condition Report CR-1-97-0081, dated March 27,1997. which

the licensee generated to resolve the maximum emergency feedwater flow issue. A

licensee contractor prepared an analysis which determined that the maximum feedwater

flow rate to prevent steam generator tube damage was 2214 gpm. The inspectors

reviewed the calculation and determined that the higher flow limit was adequately

justified.

The licensee stated that the upper flow limit was not included in operating procedures

because high emergency feedwater flows would cause a cooldown, during which

operators would take procedural actions to correct the high flow rate problem. However,

the inspectors determined that the licensee had not previously evaluated the effects of

the high flow on the steam generator tubes for the periods of time before operator actions

would restore an acceptable flow rate.

The inspectors reviewed Procedure 1010.010, " Unit One Transient Cycle Logging and

Reporting," Revision 1, which was issued to provide guidance for determining excessive

emergency feedwater flow and provide specifications for initiation of a condition report.

This procedure superseded Procedure 1010.002. The inspectors determined that the

Procedure 1010.010 required an engineering evaluation if the upper limit of the '

emergency feedwater flow was exceeded.

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-14-

During the inspection, the inspectors reviewed in detail Framatome Technologies final

draft calculation 321229968-00 and Framatome Technologies Calculation 32-5000450-

00,"ANO Auxiliary Feedwater Capacity - Phase 2." The purpose of these calculations

was to determine the limit for emergency feedwater flow to preclude unacceptable steam

generator tube damage or wear. The results showed that a flow rate of 2214 gpm

resulted in less turbulence induced vibration than at full power conditions. As part of this

review, a conference call was held with Framatome Technologies and licensee

personnel during which they responded to questions from tM inspectors. Based on their

review and the information presented during the conference call, the inspectors

determined that the analysis provided in these calculations demonstrated that

emergency feedwater .peration at flow rates up to 2214 gpm for limited periods of time

would have no significant effect on steam generator performance. The inspectors

verified that design of the emergency feedwater system would not resun la flow rates in

excess of this amount.

The inspectors determined that the licensee failed to procedurally incorporate a vendor

recommended maximum emergency feedwater flow limit of 1500 gpm to a single steam

generator. As a result, flow rates in excess of this amount were permitted to occur for

short periods of time. Later, in response to the NRC's questions, these excess flow rates

were shown to be acceptable.

10 CFR Part 50, Appendix B, Criterion V, requires that activities affecting quality be

prescribed by and performed in accorJance with instructions, procedures, or drawings.

The licensee's failure to incorporate vendor recommended flow rate limits into

procedures was identified as a violation of 10 CFR Part 50, Appendix B, Criterion V

(50-313/9721-01),

Conclusions

The licensee failed to incorporate a vendor specified emergency feedwater flow limit into

plant procedures, resulting in the failure to evalus'e the effects of exceeding the vendor

recommended flow limit during two previous periw of plant operation. Concerns

related to possible damage to the steam generator tunes were adequately resolved by a

subsequent arialysis demonstrating the acceptability of higher flow limits.

E8.8 (Closed) Unresolved item 50-313/97201-03: EFW Pioina Confiauration aifferences

Backaround

in review of Calculation 82-D-2086-02, which evaluated net positive suction head

requirements for the motor-driven and turbine-driven emergency feedwater pumps, the

NRC identified that the suction piping configuration used in the calculation to determine

pressure drops was not consistent with the piping configuration identified in the piping

isometric drawings. The pressure drops in the piping were negligible and the effect on

the results of the calculation due to the differences was small. This item was opened

pending action to revise the calculation and take any appropriate generic actions.

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Inspector Followup

The licensee initiated Action item Requests L97-0198 and L97-0285 to revise

Calculation 82-D-2086-02 to address the as-built configuration. The inspectors verified

that these action items had been assigned to the appropriate engineering organization

for completion. The licensee established a Uue date for completion of the calculation

revision of March 1,1998. Based on the minhal impact on the results of the calculation -

expected by incorporating the as-built piping configuration differences.in the calculation,

the inspectors considered the licensee's schedule for completion adequate. The broader

3

issue of identified deficiencies in the licensee's maintenance of calculations is addressed

in the inspectors' followup of Unresolved item 50 313/97201-06 (Section E8.11).-

10 CFR Part 50, Appendix B, Criterion Ill, " Design Control," states, in part, that design

control measurss, such as design reviews, shall provide for verifying or checking the

adequacy of design. The licensee's failure to properly verify that the design inputs for

Calculation 82 D-2086-02 were consistent with the installed piping configuration was

considered as the first example of a violation of 10 CFR Part 50, Criterion lli (50-

313/9721-02).

Conclusion

The licensee's failure to properly verify the design inputs to Calcu;ation 82-D-2086-02 ^i

- was identified as a violation of 10 CFR Part 50, Appendix B, Criterion Ill.

E8.g (Ocen) Unresolved item 50-313/97201-04f 1: *n=*a Pinino Pr===nre.ggi .

Temperature Specifications

,

Background

During review of Calculation 88-E-0100-16, which evaluated pressure and temperature

limitations of emergency feedwater system piping, the inspectors identified several

discrepancies. For example, in the determination of maximum emergency feedwater

discharge pressure limitations, the calculation used a pump suction alignment from the

condensate storage tank instead of the service water system as input. The use of the

service water system as a source would result in a higher emergency feedwater system

discharge pressure than if suction were taken from the condensate storage tank.-

The licensee initiated Engineering Request 973848 for the design engineering staff to

review and revise Calculation 88-E-0100-16. The rrque'st was assioned a completion

date of December 31,1997.'

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Insoector Followup

The inspectors reviewed the licensee's followup action item response to the original

inspection concern. The response identified that the design pressure of the emergency

feedwater system discharge piping was 1850 psig. At rated speed of the turbine-driven

emergency feedwater pump (3795 rpm) and with the suction aligned to the condensate

storage tank, the calculated discharge pressure at shuto# head was 1732 psig. When

the system was aligned to the service water system, the calculated discharge pressure at

shutoff head was 1842 psig, resulting in a margin from the design pressure of only 8 psi.

However, the ca!culation did not account for speed controller tolerances, nor did the

calculation assume a failure of the speed controller. If a failure of the speed controller

occurred, procedural guidance was provided for an operator to take manual control of the

turbine-driven emergency feedwater pump. This procedure instructed the oper; tor to not

exceed a pump speed of 3900 rpm to avoid overpressurization of the discharge piping.

The licensee determined that at a pump speed of 3900 rpm, the resulting maximum

discharge pressure was 1844 psig when suction was aligned to the condensate storage

tank and was 1932 psig when suction was aligned to the service water system.

Therefore, manual operation of the turbine-driven emergency feedwater pump at the

maximum procedurally-allowed speed would result in a maximum discharge pressure >

that exceeded the design pressure of the piping.

The licensee's response stated that the utilization of the 1932 psig maximum ressure as

a basis for discharge piping design pressure was not considered because manual pumr>

control while aligned to the service water source would require an equipment failure in an

already infrequent emergency plant condition.

The inspectors had a concern with this position. The condensate storage tank was

designed to provide a tomado-protected 30-minute supply of water to the emergency

feedwater system. The 30 minutes allowed the operations staff time to transfer the

emergency feedwater supply to the service water system. However, if the turbine-driven

speed control was out of tolerance or failed and the pump was operated manually per

procedure, the maximum discharge pressure could exceed the design pressure of the

turbine-driven pump discharge piping.

The licensee's response also addressed historical mp performance during testing at a

pump speed of 3920 rpm. Analysis of the test results in Calculation 80-D-10838-102A

concluded that the discharge pressure would not have exceeded the 1850 psig design

pressure, even if suction were aligned to the service water system at maximum pressure,

because the pump was not operating at full potential. However, the pump had sufficient

capacity to meet minimum specifications and, therefore, was operable.

The inspectors discussed these issues further with the licensee and were informed that

the original calculation would be reviewed to determine the extent that higher-than

nominal pump speeds need to be considered, both from a speed controller tolerance

perspective and a manual operation perspective. Additionally, since the pump appeared

to be under performing with respect to its certified pump curve, using the pump curve for

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piping design pressure determination may be overly conservative. The inspectors were

unable during this inspection to evaluate this issue further to determine whether the

pump was performing in an acceptable manner with respect to the pump curve.

However, the inspectors noted that the pump was meeting its Technical Specification

limits. The licensee'e due date for completion of this review and revision to the

calculation was December 31,1997. During the exit meeting conducted January 6,

1998, the licensee indicated that the due date for this effort had been extended to

January 15,1998.

The inspectors considered this item open pending further NRC review of the licensee's

ongoing evaluation and review of the performance of the turbine-driven emergency

feedwater pump.

E8.10 (C10MLUDIggived item 50-313/97201-05: Evaluation of EFW Pumo Room

Environmerd

Backaround

The licensee identified an environmental qualification problem in the Unit 1 EFW pump

room, which houses the turbine-driven and motor-driven EFW pumps, Steam traps

located in the steam supply lines to the turoine-driven pump, which were not designed for

seismic loads, were vented to the atmosphere of the room. A failure of one these traps

(two in the high pressure piping and two in the low pressure piping) would cause a

continuous b!owdown of hot steam in the room, raising the ambient temperature and

humidity, and possibly challenging the operability of environmentally-sensttive equipment

in the room. The licensee took interim measures to leave one steam trap isolated and to

permanently open a door to the room. The licensee then calculated the impact of a

steam trap failure under this configuration assuming that operators could isolat; the

steam blowdown within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The calculation (87-E-0026-09, Revision 0) co%ded

that temperature in the room wouid peak at 149.5 degrees F for a brief time, but that the

c steady state temperature would be 134 degrees F. Since the limiting long-term

temperature for the room was 148 degrees F, the licensee concluded that the EFW

system remained operable. The inspectors agreed with the licensee that humidity

resulting from the event would not adversely affect equipment critical to the function of

the EFW system.

This item was opened to track licensee efforts to make modifications to resolve this issue

permanently and to make necessary revisions to related design basis documents. The

licensee also indicated that an assessment of past operability of the EFW room would be

performed.

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Insoector Followuo

in review of this finding, the licensee determined that the original design of the EFW

steam traps was inadequate for both Units 1 and 2. The safety implications for Unit 2

were mitigated because the steam traps exhausted outside the EFW room. The licensee

decided to modify EFW steam traps on both units: Unit 1 during Refueling Outage 1R14

(Spring 1998) and Unit 2 during Refueling Outage 2R13 (first quarter 1999).

The licensee performed a past operabmty assessment and concluded that the EFW

system was operable in the past considering the room door closed, all traps in operation,

and a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> operator response to isolate the failed trap.

,

The inspectors reviewed Root Cause Aralysis Report CR C-97-0048, " Unqualified

Steam Trap Lines in the EFW Turbines' Steam Supply Lines." In this document, the

licensee determined that the architect / engineer (A/E) for the design of the plant assumed

that failure of the steam trap lines would not reduce the functional capability of the EFW

systems. However, the A/E had evidently focused on sufficiency of the resulting supply

of steam to the turbine and not on the environmental consequences of the failure.

The inspectors questioned the assumption that operator action to isolate steam would

occur within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of a steam trap failure, an assumption essential to the past operability

status of the system. The licensee stated that an operator surveillance of the room

(checking lube oil flows) occurs by procedure soon after each automatic start of the EFW

system, which would presumably result in identification of the steam trap failure.

However, since no time parameter was specified for this surveillance, it could be '

indefinitely delayed if other complications were to occur. The inspectors asked about the

effect of a response if the failure occurred after this surveillance effort. In this case, since

no remote indication of temperature in the room is available, it did not seem reasonable

to assume that steam isolation within one hour would occur. The licensee stated that the

fire detection system would probably alarm given the amount of steam that would exist in

the room. On one previous occasion, in 1992 (CR 1-92 0361), the deluge valve tripped

open due to an excess amount of steam from the steam trap drain lines. However, on

further discussion with the fire control engineer, the inspectors learned that the fire

detection device in the EFW room is not sensitive to steam and would only possibly

alarm in the event of a condensation buildup on the device. The inspectors determined

e

that the fire detectors were not a reliable indication of a steam leak.

r

The evaluation cf past operability indicated that, with the room door closed, the room

temperature would reach 164 degrees F 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the steam trap failed, it seemed

questionable whether an ooerator could enter an anwonment this harsh to pen ~orm the

required valve isolation. In response to the inspectors' concern, the licensee generated

a new action item (Item Number 9) for CR C-97-0048 to address the issue of accessibility

as it pertained to the past operability determination. The licensee recalculated the room

temperature profile using more realistic assumptions, including condensing heat transfer,

and found that the room condition at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> would be 133 degrees F at a relative

humidity of 57 percent. This corresponded to a wet bulb globe temperature (WBGT) of

121.5 degrees F. At this WBGT, the licensee, based on US Navy

i

1

~ . _ _ _ _ - _ _ _ _ - _ _ _ _ .

_ _ _ _ _ _ _ _ - _ _ _ _ _ _

,

.

19-

Document OPNAVINST 5100.19C, Changa 1, dated May 15,1990, determined that

strenuous work at a maximum duration of 30 minutes could occur in this environment.

The amount of time needed to isolate a failed steam trap would be only several minutes.

The inspectors noted that the shapa of the temperature profile suggested that the

temperature would peak out at something less than 140 degrees F at a relative humidity

of less than 60 percent. This suggested that the room would be fit for the necessary

operntor's actions to take olace and that equipment operability would not be threatened.

Dunng the inspection, the inspectors reviewed Calculation 87 E-0026-09, EFW Pump

Ruom Tempaniture Profiles." This calculation provided an analysis of the heat up of the

EFW puu : u sum!ng various failed op9n steam traps in addition to the normal thermal

load proviued by the EFW pumps and other equipment. The analysis was carried out

with an ANO developad and maintained software package,"RHUR," which calculates

roora heat up from art >ltrary sources. RHUR was p ogrammed in a ' macro' language

executed in a spreadsheet program, SYMPHONY. It was documented as a software

package and was controlled and maintained under the software control program. The

RHUR model devebped for thlS analysis was constructed by substantially modifying the

" macro" statements that make up the RHUR software package.

'

The inspector reviewed the analysis provided in the calculation and determ!ned that the

results of the four cases presented seemed reasonable and conservative for the given

configurations, in response to the NRC concern, the licensee provided additional

supporting analysis in the form of an alternate calculation using the GOTHIC software

program that substantially verified the results presented in Calculation 87-E 0026-09. (

The inspectors considered that the licensee had adequately established thJ past

operability of the EFW room equipment. Based on the less severe conditions of having

the room door removed, the insoectors also considered this in9rmation to support the

licensee'e assessment of current operability However, the inspectors identified a

weakness in the licensee's original response to this issue in that it frJted to document

consideration of the ability of an operator to perform an assumed operation in a harsh

environment. A tecond weakness was the lack of verification of the software model

used to calculate the postulated EFW room thermal transient. Although considered good

enginee% practice, licensee procedt,res did not specifically address this validation.

The licensee was in the process of developing a modification for both units to mitigate

the effects of pressure boundary failures and the blowdown of live steam from steam

traps in the EFW rooms, while preserving adequate condensato removal. Through

discussions, the inspectors learned that the modification w jd reroute steam trap piping

to exhaust outside the EFW room and upgrade the re a be safety related and

seismically qualified. The inspectors determinM 'N modification would resolvo the

hardware concerns associated with this issue.

4

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0

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20-

E8.11 (Cloand) Unresolved item 50-313/9720106: Desion Control Weakness

Bsckaround

This issue involved several calculation deficiencies. The examples included calculations

containing plant configuration discrepancies as well as calcult.tions that required

enhancement, updating, or to be superseded. The specific examples were as follows:

  • Calculation 88 E-0086 01,"lE Bulletin 88 04 Review for P7A and P7B Minimum

Flow Evaluation," did not adequately document the effect of two parallel EFW

pumps operating on minimum recirculation.

  • Calculation 92 E 0077-04, " Unit 1 EFW System Pump Performance

Requirements," used an incorrect level in tank T-418 for EFW actuation due to a

loss of offsite power following a tornado.

Calculation 92 E 0021-01," Emergency Nty Cycle and Battery Sizing

Calculation," was not revised to reflect plant modifications that resulted in

changes in the continuous and locked rotor current for EFW Valves CV 2620 and

CV 2627, and a modification that resulted in an additional load due to the reactor

building spray puinps close and trip circuits.

  • Calculation 80-D-1083A42, "EFIC DC Valve Torque Calculation Under Reduced

Voltage Condition," was not superseded as required.

The actions implemented or planned by the licensee included those to correct the

specific examples and to improve design control programs to prevent recurrence.

Insoector Followuo

The inspectors reviewed Licensee information Requests (LIRs) L97 0289 and 0290 as

well as Condition Reports CR C-97-0058 and CR-C 97-0059.

'

LIR L97 289 addressed the failure to updato Battery Loading Calculation 92 E-0021-01,

following iniplementation of Design Change Packages (DCPs) 92-1003 and 93-008. The

licensees's response to the inspection report (letter 1CAN099703, dated September 22,

1997), indicated that this calculation would be superseded before January 15,1998.

F sever, at the time of this inspection, this calculation had already been updated and

(as reviewed by the inspectors and found acceptable, in discussions regarding this

inconsistency, the licensee explained that the oversight resulted from the existence of a

large backlog of dats that had not been entered into the Design Configuration

Information Management System (DCIMS). This condition resulted in a failure to identify

affected calculations when engineers accessed the DCIMS system to perform searches

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ - _ - _ - _ _ _ _ - _ _ _ _ _ __

.

.

21-

based on key words related to a particular modification or calculation revision. The

licensee indicated that this backlog had been eliminated and that current practice would

preclude new backlog formation. The inspectors confirmed that no backlog existed and

that the DCIMS was up to-date at the time of this inspection. In additia, the inspectors

reviewed Memorandum ANO 97 00225 issued to the engineering staff that reinforced the

existing procedures and guidance related to calculation updates for load changes.

The second issue within LIR L97 289 dealt with Calculation 80-D 1083A 02, *EFIC DC

Valve Torque Calculation Under Reduced Voltage Condition" which should have been

identified as superseded. The licensees's response to the inspection report indicated

that this calculation will be superseded before January 15,1998. Discussions with

licensee personnel and review of the applicable procedure, OP 5010-015, " Engineering

Calculations," indicated that clear direction existed to ensure that superseded

calculations are identified, The cause of the specific instance was attributed to a backlog

of data to be entered in the primary tracking tool, DCIMS, that had previously existed. As

discussed above, the inspectors confirmed that no backlog existed at this time of this

inspection.

LIR L9/ 290 addressed Calculations 88 E-0086-01 and 92 E-0077-04, both of which

needed enhancement for various reasons. The licensees's response to the inspection

ecport indicated that these calculations would be revised before March 1,1998.

10 CFR Part 50, Criterion lil, * Design Control," states that measures shall be established

to ensure that applicable requirements and the design basis are correctly translated into

specifications, drawings, procedures, and instructions.

The licensee's failure to properly account for design information in the

Calculations 88 E-0086 01,92 E 0077-04,92 E-002101, and 80-D-1083A-02 was

considered as additional examples of a violation of 10 CFR Part 50, Appendix B,

Criterion 111 (50-313/9721.~1).

The inspectors review of Condition Report CR C 97-0058 indicated that the condition

report dealt with prioritization and completion of the Upper Level Documents (ULD) and

ULD-related discrepancies. The actions taken as a result of the condition report included

determining completion priority and establishing and communicating to the engineering

' organization the philosophy for use of the ULDs. The inspector reviewed the actions

taken on these issues and confirmed that they were appropriate.

The inspectors review of Condition Report CR-C 97 0059 !ndicated that this condition

report dealt with completion of the Design Configuration Documentation (DCD) Project

I and resolution of discrepancies identified during this project. The project was completed

l in December 1994; however, approximately 100 discrepancies remained unresolved.

The discrepancies were originally screened for significance and operability, but this CR

recognized that recently there was greater interest in and higher standards being applied

to design configuration issues, in addition, several other recent CRs have been initiated

to reexamine previously closed discrepancies that were found to have inadequate

resolution by current expectations. The CR-C 97-0059 evaluation established an action

!

_ _ _____,

.

22

to priontize review of the remaining open discrepancies. This action was completed. In

addition, previously closed discrepancies have been reviewed to ensure that they are

adequate. This completed action led to further actions to evaluate 44 previously closed l

items. These subsequent actions were not available for review at the time of this )

inspection. '

In addition to the specific observations and related corrective actions reviewed above,

the inspectors reviewed other condition reports, procedures, guidelines, and

programmatic evaluations that impact the control of calculations. These are listed as

Attachment 2 to this report. Several recent calculations and design change packages

were reviewed to assess the results of ongoing improvement efforts on current

calculation practices. These actions addressed broader issues associated with

calculation processes and control.

The licensee initially identified a concem regarding the adequacy of their calculation

processes in Condition Report CR C 96-0060 initiated in March 1996. The most

significant outgrowth of efforts to resolve this concern led to the formation of a Quality

Action Team (QAT) to study calculation process issues. The team's final report,

published in September 199d, identified a large number of concerns (56) organized into ,

the following broad categories: l

  • Calculation in processing time was too long creating backlogs and delays for

engineering personnel.

  • The DCIMS data base contained inaccurate, incomplete, not up-to-date or

erroneous information.

  • The calculation preparation, revision, and approval process requirements were

not clearly and consistently defined, documented, or understood by engineering

personnel.

t

  • Retrieval of existing calculations was sometimes difficult.

! + Inconsistent interpretation of standards for legibility of calculation documents

l resulted in delays for processing calculations through quality assurance records

l storage.

l

  • Data may be buried within a calculation with its relationship to other processes

not defined,

  • Program engineers were not alwa ys aware when other engineers were revising

,

l calculations that impact their program.

-

In reviewing the QAT action plan, the intpectors determined that the proposed corrective

actions were appropriate and that they r hould provide significant calculation process

improvements when fully implemented in addition, the interim compensatory actions

. .- - - - _ - . - . - - . - - - . - , - - . .

. _ _ -

_ _ . _ . . _ _ _ _ _ _ _

.

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23-

that were implemented in the short term appeared to be of sufficient quality to reduce the

occurrence of calculation deficiencies.

In addition to reviewing the licensee's action plan, the inspectors reviewed a sample of

approximately 30 calculations generally chosen at random. These are listed in

Attachment 2. The inspectors found that the calculations were thorough, generally clear,

and well documented. Assumptions and other data sources were clearly identified e7d

referenced. The few exceptions were found in some older calculations performed during

original plant design and construction, approximately 20 years ago.

The inspectors focused on the licensee's process for assigning status to calculations.

This included the use of new, pending, as installed, amending, and superseded status

levels. Although the inspectors found the use of the amending status somewhat

confusing and cumbersome, no erroneous cross references were found in the sample

reviewed. For example, there are m limits to the number of calculations that can amend

a given calculation. Nor is there a time limit beyond which amending calculations would

require incorporation. In practice, however, the only calculations found with a large

number of amending calculations were the electricalloading calculations. In this case,

the licensee's practice was to assign a coordinator or owner for these calculations. The

coordinator was responsible for incorporating all amending calculations once each cycle.

The incorporated calculations were then assigned a status of superseded. The

inspectors reviewed and found this to be correctly performed for a recently issued

revision to Calculation 92 E 0021-01. " Emergency Duty Cycle and Battery Sizing

Calculations." In addition, as mentioned above, the licensee instituted a Calculation

Revision Notice (CRN) process that allows minor changes to calculations to be issued

and more easily tracked. This process was designed to reduce the number of amending

calculations to no more than nine in practice, based on a sample review, the inspectors

found no examples with more than two CRNs.

In spite of an apparently comprehensive effort to improve the calculation process, the

inspectors recognized a potential inconsistency in the way the licensee handled the

issue. The licensee's quality assurance team identified a concern related to the

preexistent backt0g in the DCIMS data base that had apparently been the cause of

several NRC-identified discrepancies. In addition, other persistent problems within the

calculation process were identified by the team. However, the licensee did not perform a

sample review of existing calculations to determine whether the DCIMS problem, in

addition to the other identified problems listed above, had created a situation where a

large number of calculations currently in effect may contain technical errors as a result of

these problems. The inspectors considered the lack of review of a sample of

calculations to represent a weakness in the licensee's overall response to the issue.

The inspectors reviewed the capability and use of the DCIMS database. It was a very

useful and powerful tool for searching for needed information concerning calculations,

evaluations, and reports.

_ _

,. . _ _ ._ .__

,

L

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24

Conclusips

A violation of 10 CFR Part 50, Appendix B, Criteilon lil, was identified conceming

discrepancies identified with four licensee calculations. The lack of a review of existing

calculations to determine if past problems identified by the licensee affected other

calculations of record was identified as a weakness.

E8.12 (CJosed) Insoection Followeo item 50-382/9720107: Licensee's Actions for Revising

the Modification Procedure to Address Field Routed Installations

Background

The NRC identified that the layout configuration of certain installed conduits was different

from that shown on the conduit and cable tray layout drawings. The NRC also identified

that there were no drawings or records of the seismic support details for the installed

conduit and cable tray supports. The licensee stated that this was due to conduits being

field routed.

The NRC noted that conduits attached to the wall of the emergency feedwater pump

room contained redundant cables supported on the same unistrut support. The licensee

stated that the supports were designed to withstand seismic design loads and were

passive. In addition, the licensee stated that redundant trains could be supported from a

common support from a seismic standpoint provided Appendix R and other criteria were

addressed.

The licensee's modification procedure discussed a constructability walkdown involving

the modification engineer, modification supervisor, craft supervisor, and quality engineer,

During discussions, the licensee stated that they would document the decisions arrived

at during the walkdown by including these in the form of work steps in the controlled work

package. The licensee stated that this would include conduit routing and seismic support

details that would be documented as sign-off steps. The need to revise the modification

procedure to address field routed installations was identified as Inspection Followup

Item 50 313/9720107.

Insoector Followuo

The inspectors reviewed the licensee's response to the inspection followup item

documented in Letter 1CAN099703, dated September 22,1997 (response letter to NRC

Inspection Report 50-313/97-201). The inspectors also interviewed licensee personnel.

The inspectors requested that the licensee supply details concerning how the Appendix

R requirements were met for the redundant cables supported on the same unistrut in the

emergency feedwater pump room.

The inspectors rewewed draft Calculation 85-E-0086-1, " Safe Shutdown Capability

Assessment," Revision 3, and found that the turbine driven emergency feedwater pump

cable and the motor driven emergency feedwater pump cable had a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire barrier

installed. In addition, the inspectors found that an NRC exemption was granted for

.- - . - - .

.

.

25-

configurations having less than 20 feet of separation. The inspectors reviewed licensee

Letters 1CAN068706, dated June 24,1987, and 1CAN048708, dated April 22,1987,

which requested the exemption for the cables. The exemption, granted in an NRC letter

dated October 26,1988, contained a stipulation that a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire wrap was necessary in

addition to a fire suppression system. During a walkdown, the inspectors determine)

that there was a fire suppression system in the pump room and that the cable was

wrapped with fire wrap.

The inspectors reviewed the revised Procedure 6030.005, " Control of Modification

Work," Revision 3, Change PC-1. The inspectors noted that the procedure was revised

to include instructions for the development of a single line drawing depleting the types of

supports used, the support spacing, and the conduit layout. The procedure required that

the draw!ag be included with the controlled work package.

E8.13 Ogied) Insoection Followuo item 50 313/97201-08: Licentee's Actions to Perform a

Revised Cable Pullina Calculation Usino Additional Desian Information Obtained from

the Cable Vendor

flitCMIQMDd

The NRC identified that Procedure / Work Plan 6030.112, " Installation of Raceway

Systems," used for the installation of raceways, included a table of conduit bend radil for

manufacturers' standard bends but not for long radius bends. The NRC asked the

licensee which cables required long radius bends as a result of the minimum bending

radius and maximum sidewall pressure restrictions during pulling. The licensee provided

a calculation that compared the cable minimum bending radil to the bend radil of the

conduit that was appropriate for a single cable. The NRC questioned the values for the

redil on the inside surface of the conduit bends as well as the accuracy of the cable

diameters and maximum cable bend radii. The licensee stated that they would obtain

updated cable vendor data to verify the conduit bend dimensions and would revise the

calculation accordingly. The NRC reviewed the preliminary calculation and noted that a

few largo diameter cables would require long radius bends. This item was opened

pending licensee actions to obtain additional design information from the cable vendor in

order to perform a revised cable pulling calculation,

insoector Followuo

The inspectors reviewed the licensee's response to the inspection followup item in

Letter 1CAN099703, dated September 22,1997, and interviewed licensee personnel.

The inspectors also requested a listing of condition reports involving damage to cables

that resulted from cable pulling; however, no such conditions were found.

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.

!

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26

The licensee stated that updated cable information would be obtained from cable '

vendors for approximately 10 percent of the cable codes listed in the cable minimum

bend radius review table. The licensee expected to receive this information by mid.

January,1998. The licensee also stated that the updated cable information obtained

would be used to revise the calculation for each of the applicable cable codes. The e

licens*a stated that the cable calculations would be revised by March 1998.

The inspectors determined that the licensee's corrective actions were adequate for the

current closure of this item based on the fact that the condition report search had not _

found any cables damaged by cable pulling and that the actions necessary to revise the

cable code calculations were being tracked and had acceptable completion dates.

E8.14 (Closed) Insoection Followun item 50-313/97201 11: Verification of Calculation Changes

Associated with DC Batteries

Backaround

The NRC identified that Calculation 92 E 0021-04," Battery DOS and 007 Recharge

Time," Revision 0, was not revised to show the reduced charging time resulting from a

change out of the Unit 1200 amp battery chargers with new 400 amp chargers. An

amending calculation,93 D 1010-04, " Recharge Time for D06 and D07 with Changes

per DCP 931010," Revision 0, calculated the revised charging time. This item was

opened pending incorporation of the amending calculation information into the parent

calculation,

insoector Followuo

The inspectors determined that this condition did not represent an actual prob'sm

because the battery recharge times were not required by license conditions and did not

affect battery capacity. In addition, as discussed in Section E8.11 of this report, thers

were no licensee procedures requiring incorporation of amending calculations into the

parent calculation within a specific time period. Based on discussions with the licensee,

the inspectors determined that information provided in amending calculations, but

missing from the parent calculation, should not result in a quality concern because

annotations provided in the parent calculation positively identified the existence of the

amending calculations.

The inspectors verified that amending Calculation 93 D 1010-04 had been incorporated

into Calculation 92 E 002104. The inspectors observed that the revised recharge times

were based on the updated 400 amp recharge capacity.

I '

d

m_ _ - - _ _ . _ . . , ..,,_,_--m =--. - . , _ , _ , - ---,--r-- -- - m. -

-__. .__ . .. . ._ _

.

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27

E8.15 (Closed) Insoection Followuo item 50 313/97201 12: Review of Current Desion BM!s for

Bvoassina Emergency Diesel Generator (EDG) Protective Trio

Background

For emergency starts of the Unit 1 EDGs, the licensee did not bypass those protective

functions delineated by Regulatory Guide 1.9, " Selection, Design, and Qualification of

Diesel-Generator Units Used as Standby Onsite Electrical Power Systems at Nuclear

Power Plants," to be bypassed under accident conditions. This was because the

licensee was committed to Safety Guide 9, " Selection of Diesel Generator Set Capacity

for Standby Power Supplies," which does not specify protective functions to be

bypassed. The concern was that the licensee's setup could result in a higher probability

of EDG loss during accident conditions. This item was left open pending review by the

NRC program office,

insoector Fougsp

Based on discussions with the NRC program office, the inspectors determined that the

licensee's adherence to Regulatory Guide 1.9 was satisfactory.

E8.16 (Ooen) U.Q[esolved item 50-313/97201 14: Lack of Testina Unit 1 Molded Case Circuit

Breakers

Background

The Updated Final Safety Analysis Report identified in Section 8.3 that the Class 1E

electrical distribution system was designed to meet the requirements of IEEE Standard

3081971, *lEEE Standard Criteria for Class 1E Electric Systems for Nuclear Power

Generating Stations ' Section 6.3 of this standard required periodic testing at scheduled

intervals to demonstrate that components that were not exercised during normal

operation of the station were operable. The standard further stated,'The specific tests

and the frequency at which they are performed depend upon the specific components

installed, their function, their environment, and the fact that they are in the maintenance

program of the unit (s). Illustrative examples of tests are given in Table 2." Table 2 of the

standard identified switchgear and one of the tests identified for illustrative purposes for

that type of component was an overcurrent trip test.

In contrast to these requirements, the licensee did not have a program to perform

periodic overcurrent trip testing of its Unit 1, Class 1E, molded-case circuit breakers.

The licensee developed a position paper that concluded that IEEE 308-1971 was not

applicable to molded-case circuit breakers. This conclusion was based on an

interpretation that Table 2 of the standard identified four tests for 600 V and below

switchgear: operation test, mechanical inspection, overhaul, and overcurrent trip test.

_ __ _

_ - . - - .

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ______________ _ ___

i

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28-

Due to the physical construction of molded caso circuit breakers, mechanicalinspection

and circuit breaker overhaul were not possible. Therefore, the licensee concluded that

the perindic tests in Table 2 were not applicoble to molded-case circuit breakers but were

rather intended for load center switchgear.

The inspectors considered the licensee's interpretation of Table 2 of IEEE 3081971 to

be in error, based on the following points:

  • Table 2 identified illustrative periodic tests for Class 1E components.

.

The identified tests were not an exhaustive or restrictive lists nor was Table 2 i

intended to exclude components for which it was not physically possible to l

perform all of the identified tests.

)

  • The user of the standard was expected to evaluate which of the identified tests

(or others) were appropriate for each component.

As a result of its operating experience assessments for NRC's Information Notices 92 51,

  • Misapplication and inadequate Testing of Molded-Case Circuit Breakers" and 93-64,
  • Periodic Testing and Preventive Maintenance of Molded-Case Circuit Breakers," the

licent,ee initiated a practice of performing electrical testing of molded case circuit

breakers in the receipt inspection process and performing mechanical exercising of the

breakers on a periodic basis. However, they did not implement the practice of periodic  ;

electrical testing. Information Notice 93-64 referred to Electrical Power Research

Institute Report NP 7410," Breaker Maintenance." Volume 3 of this document," Molded

Case Circuit Breakers," recommended electrical testing of molded case circuit breakers

on a 4 to 6 year periodicity.

Prior to the inspection documented in NRC Inspection Report 50-313/97 201, the

licensee performed a self assessment that identified that periodic electrical testing of

molded-case circuit breakers was not performed. The licensee established a task group

to determine if a testing program, replacement program, or a combination of both should

be adopted.

Insoector Followuo

The inspectors conducted interviews with personnel and reviewed the licensee's position

paper on molded-case circuit breaker testing and plans for developing a periodic testing

program for molded-case circuit breakers.

The licensee identified that they performed several activities related to safety related

molded-case circuit breakers to venfy that they will function under design basis

conditions and that the breakers are maintained accordhg to vendor requirements.

,

  • F

>

2g.

These activities included: testing Unit 2 electrical penetration breakers in accordance

with the Technical Specifications, tet, ting all new breakers prior to installation, preventive

maintenance inspection and cleaning of breakers and panels, manual cycling of some

breakers, and performing periodic thermography inspections of 480 V molded case

circuit breakers. The licensee stated that these testing activities have historically

resulted in very low failure rates.

The licensee informed the inspectors that they had undertaken a project to develop a

periodic electrical testing program for molded case circuit breakers. Specifically, the

licensee stated that a program would be developed to test the Unit 1 containment

electrical penetration breakers and the program would be implemented in the refueling

outage scheduled for the spring of 1998. At the time of the inspection, the licensee had

identified a sample population of 64 molded-case circuit breakers for inclusion in the

program, of which, approximately 12 wcre scheduled for testing in the next refueling

outage. Additionally, the licensee stated that it would categorize the remaining

safety related Unit i breakers by risk significance and develop and begin implementation

of a testing program by the end of 1998. With respect to Unit 2, the licensee identified

that safety related molded-case circuit breakers that were not included in the Technical-

Specification-required testing program would be evaluated for risk significance and a

testing program would be developed and implemented by the Unit 2 refueling outage

scheduled in 1999.

Based upon the previous record of satisfactory Technical Specification-required electrical

testing experience of Unit 2 molded-case circuit breakers and other maintenance and

testing activities the licensee has implemented, the inspectors did not consider the

licensee's lack of a periodic testing program for Unit 1 molded-case circuit breakers an

immediate safety concern. However, the licensee's practices will be reviewed further

with the NRC's program office to determine if the licensee's reference to IEEE 308-1971

in its Updated Final Safety Analysis Report imposed a regulatory requirement to

implement an electrical testing program for molded-case circuit breakers. This

unresolved item was left open pending review of this issue by the NRC program office.

E8.17 (Closed) Unresolved item 50-313/97202-15: Lack of includino all Safetv-Related Fuses

in Proce&dC

BacharWad

The NRC identified that the licensee had not adequately resolved previous concerns

re;ated to incorrect electrical fuse installations. The licensee had established a like for-

like replacement policy for installed fuses but had physically verified only 4 percent of the

fuses in Unit 1 to be correct in all aspects. The fuse control procedure required an

engineering evaluation only for blown fuses or when replacement fuses were different

from those installed. All other 3pplications involved a like for like replacement without an

engineering review. Under these conditions, an existing incorrect fuse would be

replaced with a like fuse. The NRC considered the like-for like fuse replacement policy

to be inappropriate considering the small percentage of fuses validated.

_ _

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.

,

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Insoector Followup

In response to the NRC finding, the licensee proposed to add the following DC

distribution equipment to the fuse control program by June 30,1998: D01, D02, D15,

D16, and D17 for Unit i and 2001,2002,2D26,2D27,2D41,2D42,2D51, and 2D52 for

Unit 2. Also, as a long term corrective action, the licensee stated that whenever a design

change was r ade to a safety related circuit protected by a fuse, a review of tha fuse list

would be performed by engineering to verify, correct, or add fuses to the procedure as

necessary.

>

The licensee reviewed their condition report database for fuse-related discrepancies

occurring since January 1,1993. Seven condition reports during this period addressed

wrong size or type of fuse. Only one of these condition reports affected Unit 1, and this

, fuse was installed in a nonsafety related circuit. One was in a nonsafety circuit common

to both units and five affected Unit 2. Of these five, only three affected safety related

circuits. None affected circuit operability. Based on these facts, the licensee concluded

that the fuse control program was being effectively managed. The inspectors determined

that it was unlikely that a pervasive fuse control problem existed and, therefore, did not

consider the matter to represent a safety concern.

E8.18 (Closed) Unresolved item 50-313/97201-16: FSAR Disneoancies

Backaround

The NRC identified five discrepancies in the license's FSAR, mostly involving

, Table 7-11 A, "R.G.1.97 Post Accident Monitoring Variables." The discrepancies ~

involved inconsistencies between the table and design documents concerning instrument

ranges and display configurations. One additional discrepancy concerned a note in

FSAR Section 14.2.2.1.3.1 stating that both steam generators would be isolated in the

, event of a steamline break; whereas, in actuality, only one steam generator would be

isolated. i

Insoector Followun

! Prior to the identification of these discrepancies the licensee was in the midst of a

comprehensive FSAR review for both units. This program was described to the NRC in

Letter CNRO 97/00010 "EOl Licensing Basis Assessment and UFSAR Review

Initiatives," dated May 7,1997. The licensee stated that this effort would resolve the

, generic concerns iaised by this item since this review effort would identify the same

types of errors discovered by the NRC. The specific errors were to be reviewed and

_

corrected using the ticensing document change process. These corrections were to be ,

made effective in the 1998 submittal of the ANO Unit 1 FSAR.

The inspectors reviewed Letter CNRO-97/00010 and noted that it referred to a generic

review of the Unit 1 and Unit 2 FSAR for accuracy. The inspectors reviewed a document

entitled, " Safety Analysis Report Upgrade Project," that explained the scope of this effort,

currently scheduled for completion in July 1998. Based on review of this material and

,

,, ,-,s.v ., ,_r.- --

..,y, ,-- -y , . w, - - - ,, -, -- - - , - __. , , ,,--- ---,-

.

.

-31-

discussions with licensee personnel, the inspectors concluded that the project as

pirnned would resolve the generic concerns related to this issue. At the time of this

inswa ine licensee was approximately 50 percent complete with this project and had

identified 895 potential discrepancies. Of these,391 vea strictly editorial, leaving 504

items having potential significance. However, the licensee stated that many items may

subsequently be deterrnined not to be FSAR errors since the reviewers have been

instructed to docuinent everything that might be a discrepancy.

The inspectors reviewed licensing Information Requests L97-0213, L97 0326, and

L97 0310 and confirmed that the specific errors identified by the NRC were scheduled to

be corrected in the next FSAR update.

The inspectors determined that the licensee's efforts were sufficient to correct the

specific and generic concerns associated with this inspection finding.

10 CFR 50.71(e) states that the updated FSAR shall be revised to include the effects of

all changes made to the facility or procedures as described in the FSAR. The licensee's

failure to properly update the FSAR regarding the five discrepancies described above

was considered a violation of 10 CFR 50.71(e). However, because of the favorable

progress being made by the licensee's FSAR review team and the fact that a description

of this project has been placed on the docket, and the NRC's view that the discrepancies

identified by the NRC likely woule. have been identified by the licensee's program, the

NRC is exercising discretion in accordance with Section Vll.B.3 of the Enforcement

Policy and is not taking formal enforcement action for this finding.

E8.19 (Closed) Unresolvnd item 50-313/97201-17: Ineffective Resolution of Previous

Corrective Actions

Backaround

During a plant walkdown, the NRC noted that two in-parallel check valves located in the

steam supply piping to the turbine-driven emergency feedwater pump (P7A) were

continually chattering because of steam leakage through downstream solenoid valves.

The NRC's concern was that wear from this mechanical action could eventually result in

failure of the check valves to perform their reverse seating function,

insoector FollQEuD

The Wspectors were informed that leakage through the solenoid valves had been a

problem at ANO since at least 1988 when CR 1-88 0015 identified this problem. The

licensee had performed maint3 nance on the valves on several occasions, but the

leakage problem persisted. At some point, the licensee determined that the small

amount of leakage was not detrimental to the system, since the check valves were not

experiencing abnormal wear and the heating of the downstream piping was considered

beneficial in reducing the potential for condensate-induced turbine overspeed events.

1 Then, the licensee deferred any plan for modification. However, in October 1996,

the licensee identified accelerated wearing of elastomeric surfaces on one of the

.

.

32-

two solenoid valves in the EFW steam lines. As documented in Condition

Report CR 196 0569, electricians found the O-ring on the operator of Solenoid

Valve SV 2663 to be hard and brittle following its installation on July 14,1992. The

licensee's equipment qualification program stated that this 0 ring was qualified for

60 years of service but did not consider seat leakage to be a normal condition for the

establishment of qualified life. The licensee determined that the solenoid valve remained

operable with the damaged 0-ring. Based on the O ring failure, the qualified life of the

O rings was being revised to be from 2 to 8 years depending on the service condition.

While the licensee considered a modification to the solenoid valves, they were also

considering installation of motor-operated globe valves and relecating the valveb closer

to the turbine.

The check valves performed a safety related function by isolating the piping from a

faulted steam generator. The licensee stated that one of these valves would be

disassembled and inspected during the next refueling outage in the Spring of 1998. In

February 1995, the other check valve was inspected and found structurally sound with ,

slight indications of scoring. During past tests of the check valve's reverse seating

capability, the results have been acceptable.

The inspectors were informed that the check valves had chattered since their installation

in 1986, even before the solenoid valves began to leak, because of the presence of a

drain / strainer orifice that creates a constant steam demand of 1600 pounds-mass per

hour. Since the check valves require iOOO pounds-mass per hour to remain constantly

open, the solenoid valve leakage couid actually reduce the chattering if it were of

sufficient magnitude. However, the current chatter rate of the check valves was 10 taps

per second, indicating that the solenoid leakage was well less than 3400 pounds mass

per hour.

The inspectors determined that the licensee had adequately ensured that EFW system

operabililty had not been compromised by the solenoid leakage problem. The inspectors

recognized that the check valve chattering was partNily independent of the leakage

problem and that the leakage had been observed tc to beneficial for preventing

condensation in the EFW turbine steam supply line Gese facts mitigated any safety

concerns re' < ed to this condition.

E8.20 (Closed) Unresolved item 50-313/97201-18: Lack of Seismic Sucoort of OTSG Pressure

Transmitters

(Closed) Unresolved item 50-313/97201-19: Lack of Desian Basis for Sucoort of Steam

Generator Instrument Sensina Lines

Backaround

The NRC observed that a clip designed to restrain the isolation valve and sensing line for

Steam Generator Pressure Transmitter PT 2667B was missing. The concern was that

the line could fail during a seismic event, potentially affecting the operability of the steam

_ _ _ _ _ _ _ _ _ .

.

.

33-

generator pressure controlloop. The NRC also observed that the corresponding clips on

the three other redundant pressure instrument channels in the same vicinity were loose.

The licensee initiated Condition Report CR 197-0058 to determine the operability status

and potential generic concerns. Before the end of the inspection, the licensee reported

to the NRC that the as found configuration was operable and that a walkdown of other

instrument tubing restraints indicated that the mounting clip problem was not widespread.

In the Unit 1 EFW pump room, the NRC found a 10-foot unsupported span of instrument

tubing for a pressure tranomitter (PT/PI 2811). This is a safety related sensing line that

monitors the discharge pressure of the turbine driven emergency feedwater pump. A

m. mum unsupported span of 2.5 feet was generally specified in the seismic drawing

details. The licensee issued Condition Report CR 107-0074 and, subsequently,

determined that the as-found configuration was operable. The licensee determined that

a generic issue existed and elevated CR 197-0074 to a significant condition, requiring a

root cause determination. The licensee performed a detailed walkdown of the EFW,

Decay Heat / Low Pressure injection, and High Pressure injection systems and identified

the following conditions;

Several installations were undocumented or unanalyzed, such as tubing supports

exceeding the 30-inch unsupported span and use of nonstandard instrument

supports.

Lack of design bases or criteria for determining instrument line slopes.

Drawing discrepancies between drawings and the as built condition.

Missing or incorrect tags on instrument valves.

Based on these findings, the licensee expanded the scope to include other nystems. For

example, Condition Report CR-197 0087 was later opened to document insidequately

supported emergency diesel generator instrument lines. In all,37 instrument

installations were found in Unit 1 that did not conform to either seismic or instrument

installation requirements as defined in the standard drawing details. Design engineering

deiermined that none of the 37 discrepancies created an inoperable condition.

Insoector Followuo

The inspectors reviewed Root Cause Analysis Report CR 197-0074," Instrument Tubing

not Mounted in Accordance with Seismic Design Requirements," dated March 24,1997.

The licensee identified the following root causes for this event:

(1) Inadeouate_trainina: Personnel involved in the installation and maintenance of

instrument tubing were not adequately trained or sensitized to seismic and

configurations requirements.

_ _ _ _ - _ _ _ _

. _ . . . ._. - - -- .. . .. -

.

.

34-

(2) Work oractics Personnelinstalling and maintaining instrument tubing either did

not use required configuration documentation during work activities or did not

follow these documents correctly.

(3) Desian confiaura@n and analygls: Design documentation and prints were

inadequate or not easily accessible.

The licensee walked down 40 instrument installations in Unit 2 and found that the quality

of the installations was much better than in Unit 1. Only four of the installations

inspected had notable deficiencies involving damaged mounting clips or brackets

resulting from removal and reinstallation to support work on nearby equipment.

The licensee performed the following corrective actions: (1) the individual discrepancies

were repaired as necessary, (2) a letter was issued on April 4,1997, to site groups

reiterating instrument tubing requirements, and (3) design engineering reconfirmed the

initial operability assessments fiom the tubing walkdowns. Additionallong term actions

included the following: (1) develop a scope for additional walkdowns, (2) determine

whether existing seismic tubing mounting requirements can be relaxed, (3) evaluate

enhancements to design drawings associated with tubing, (4) evaluate the work planning

processes that involve instrument tubing or the removal of tubing for interference

purposes, and (5) provide training on seismic and configuration requirements for

instrument tubing.

The inspectors discussed the status of the proposed long term actions discussed above

with the licensee. A calculation was in the final stages of review that redefined the

acceptable support span lengths for various configurations. Once this calculation was

complete, the licensee was planning to perform additional walkdowns of those safety-

related components in Unit 1 that were potentially susceptible to instrument tubing

discrepancies. This effort was to be completed by May 1998. The licensee was also

planning to provide personnel training and to clarify drawings that contained confusing

tubing support information.

l The inspectors determined that the licensee's completed and proposed corrective

actions (each of which were assigned completion due dates) were sufficient to address

the issues raised by this item and considered the overall response to be strong,

particularly considering that none of the identified discrepancies had been evaluated to

affect operability.

10 CFR Part 50, Appendix B, Criterion 111, " Design Control," states that the design control

measures shall provide for verifying or checking the adequacy of the design.

The licensee's failure to provide measures to ensure the correct translation of seismic

design requirements to instrument tubing installations and the maintenance of this

configuration during plant operations in Unit 1 was identified as a third example of a

violation of 10 CFR Part 50, Appendix B, Criterion ill (50-313/9721-02).

!

1 -. _-. _ _

.

i

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Conclusions

The licensee's failure to provide measures to ensure the correct translation of seismic

design requirements to instrument tubing installations and the maintenance of this

configuration during plant operations in Unit 1 was identified as a violation.

E8.21 (Gosed) Unresolygd item 50-313/9720120: Inadeauate Work Plan for the Controlof

Post-maintenance Testina

Backomund

The NRC identified a misaligned limit switch on Valve CV-4804, rector building vent.

The misalignment caused the actuator cam to drag on the limit switch arm toward the

end of the valve stroke instead of riding on the roller bearing. Repeated operation in this

configuration could have resulted in a faulty valve position indication. Although this

particular application was not safety related, there existed a concern that other safety-

related applications may have a similar misalignment problem.

The licensee recognized that the post modification plan and job order for this valve

(00954697) did not include a post maintenance test to verify proper operation of the limit

switch cam assembly. The licensee reopened this job order and accompanying work

plan with the intent to add steps to reconfigure and test the assembly. One other valve in

Unit 1 (CV-1845, quench tank sample valve) had a similar mounting design, and a similar

revision was scheduled for its work plan. The licensee performed a review for other

similar limit switch configurations but did not identify any other of these configurations in

either Units 1 or 2.

Insoector Followup

The inspectors reviewed the work plan revisions discussed above and observed that

steps were added to verify that the limit switch rollers were in proper contact with the

cams. The licensee considered that this verification in the work plan coupled with the

existing post maintenance test that ver;fies smooth motion and proper valve position light

sequencing would be sufficient to ensure that the condition would not recur (i.e., no

revision to the post modification plan was considered necessary). The inspectors

observed that the post maintenance testing problem,in this instance, did not affect

safety related equipment and determined that the licensee had adequately addressed

the specific and generic implications of this finding.

E8.22 (Closed) Unresolved item 50-313/972121: Inadeauate 50.59 Review Associated v@

the Removal of the Unit 1 BWST Vacuum Breaker and Followuo Corrective Action

,

Backaround

This item hvolved an inadequate 10 CFR 50.59 safety evaluation performed by the

licensee for a temporary covering placed over the borated water storage tank vacuum

relief valve flange. The evaluation was inadequate, because it did not consider the air

_

'

.. .. . .. ..

.

. . .

______ ____________

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36-

flow restricting effects of a mesh screen that was an integral part of the temporary

covering. The following is a summary of key events:

  • On December 4,1996, while Unit 1 was at full power, the iicensee removed the

8-inch diameter vacuum relief valve from the borated water storage tank for a

surveillance test. When the valve was removed, the licensee installed a white

plastic bag over the 8 inch diameter open tank flange to act as a foreign material

exclusion (FME) cover. Two slits, each approximately 8 inches long, were cut in

the plastic cover to allow air flow. This work was done using Job Order

00952943. The job order stated that if suction for the tank was necessary, a

supplemental vent path was provided by a 4 inch tank overflow line that was

always open during plant operations.

  • On December 5,1996, the system engineer observed that the tank flange was

covered with what he believed to be a solid FME cover and initiated Condition

Report 106 0663 to document the potentialinoperability of the borated water

storage tank. Subsequently, the system engineer determined that the tank was

operable since the FME cover contained two slits that would permit tank pressure

and vacuum relief. The plastic bag was installed for approximately 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The licensee subsequently ter*ved the plastic bag and installed a blank flange

on the tank flange with a 1 inch gap between the flange and tank flange to permit

pressure and vacuum relief. Around the 1 inch opening, a mesh screen was

placed to prevent foreign material introduction into the tank.

  • Condition Report 196-0663 also documented the discovery of a through wall

crack on the inlet flange of the removed vacuum relief valve and assessed the

operability of the tank. The operability assessment did not address tank

operability while the plastic bag was installed as an FME cover. The assessment

only addressed the flange and 1 inch gap and stated that the blank flange with

the 1 inch air gap installed over the tank flange would provide sufficient air

movement to vent the tank during rapid drain downs. At this time, any air flow

restriction from the installed mesh screen was not considered,

.

  • On December 12,1996, the gap dimension was changed from 1 to 3-inches

under Job Order 00952943. This was done to increase the venting capacity

based on engineering judgement that the 1 inch gap may have been marginal.

  • On January 22,1997, Significant Condition Report 197-0019 was initiated and

identified that a safety evaluation was not performed for se plastic bag installed

over the tank flange for approximately 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. In addition, the condition report

identified that the procedure for testing the valve (Procedure 1306.034, " Testing

of Unit 1 Pressure Vacuum Relief Valve PSV 1617, PSV 2423, and PSV 1412,"

Revision 2) did not allow removal of the vacuum relief valve during power

._ _ __ __ _ __- _ _ _ _.. _ __ _ _ _ _ _ _ _ . _ . . _ . _ _ _ _

  • -

.

.

37 -

. operations and that the safety evaluallon for the procedure did not include

l consideration for valve removal at powcr. The condition report included an _

>

operability assessment for the tank when the plastic t sg was installed on the tank

i flange. The licensee sta;ed that the tank was operable based on engineering

judgerrent that the plastic bag would break before an internal vacuum would

'

,

cause damage to the tank.

'

+ On February 3,1997, the licensee issued Condition Report CR 197-0031 to

address the previously made assumption that the 4-inch overflow line would, by

itself, provide sufficient vacuum relief. The conclusion of this effort was that,

< contrary to prior judgement, the 4 inch line was not large enough to prevent the

formation of an internal vacuum in the tank during rapid draindown. .

  • On February 11,1997, the licensee prepared Temporary Alteration 971-001 that

documented the removal of the vacuum relief valve and the installation of the

i blank flange with a 3 inch gap over the tank flange. The temporary alteration was

prepared approximately 2 months after the actualinstallation. Temporary

alterations were not issued for the previous two temporary coverings (plastic bag

and 1 inch gap flange connection). Temporary Alteration 971001, which >

contained the first 10 CFR 50.59 evaluation performed in association with these .

events, concluded that the borated water storage tank was operable with the  ;

3 inch gap blank flange installation. j

'

  • On March 4,1997, in response to concerns from the NRC, Condition Report 1

,

96-0663 was amended to incluc's a discussion of the air flow blockage caused by

the mesh screen over the 1 inch gap. After considering the screen restriction, the -

licensee concluded that the borated water storage tank was still operable under

this configuration.

,

insoector FolloWMD

The inspectors reviewed the documents referenced above and had discussions with

licensee personnel, in review of the facts, the inspectors identified several concerns in

addition to those addressed in the previous inspection, as discussed below. <

Procedure issu_es

'

The inspectors found, in agreement with Significant Condition Report CR-197 0019, that

'

the licensee's use of Procedure 1306.034 during power operations was improper. This

procedure was originally written for use during refueling outage conditions as indicated  ;

by paragraph 4.3.1 of the procedure, which stated that inspection of the valve would be

performed each refueling outage. The 10 CFR 50.59 evaluation for the procedure did

4

not address performing maintenance or testing of the vacuum relief valve while the plant

was at power. Also, the procedure did not address temporary provWons for vacuum

relief, which would be an essential element for online maintenance of the vacuum relief

valve. Although one of the corrective actions from the condition report was for the

,

,--v--e - - ,,,,, y ,-e , , , . ,-, ~ . ~ , - , ., , - ,,-r , - , , - , , - - , ,yw --e , ,e~ n.- -,--a, n- -rm.,.

.

.

38-

licensee to revise Procedure 1300.034 to allow valve removal and maintenance of the

VMye while at power, this revision had not been corr,pleted at the time of this inspection.

10 CFR Pr.rt 50, Appendix B, C6terion V, require', thst activities affecting quality.shall be

oresenbod by docum6ntsd procedures of a type appropriate to the circumstances and

shall be accomplished !,1 accordance with these procedures. The licensee's failure to

comply with the plant condition prerequisites of Procedure 1306.034 was identified as an

apparchtTiolent.,n (50 341iW21-03).

Desian Controlinna

The inspectors determined that the licensee had performed temporary changes to the

borated water storage tank without the use of proper design control measures. The

licensee removed the tank vacuum relief valve and sequentially installed three temporary

covering configurations, each of which constituted a change to a safety related

con'oonent, without use of the temporary alteration procedure (a temporary alteration

wat eventually prepared for t,1e third configuration, but this was issut j approximately 2

months after installation).

The inspectors reviewed Procedure 1000.028," Control of Temporary Alterations,"

Revision 20. Attachment i to this procedure listed the criteria for requiring a temporary

alteration, which included a situation where a design change was made to safety-related

equipment without the removal of the equipment from service. Using this criteria, the -

inspectors concluded that, because the borated water storage tank was not taken out of

service for these evolutions, a temporary alteration was required. The use of a

temporary alteration was intended to ensure that plant changes were subjected to design

control measures commensurate with the original design.

10 CFR Part 50, Appendix B, Criterion til, requires those design changes, including field

changes, shall be subject to design control measures commensurate with those applied

to the original design. The licensee's failure to provide design control measures

<

commensurate with initial design for three temporary changes made to the borated water

storage tank vacuum relief valve was identified as an apparent violation

(50-313/9721 04).

Operability issues

The plastic bag was attached to the tank flange in a manner such that the 8-inch

'

ventilation slits may not have been pulled into the opening during an event (e.g., a safety

injection event due to a large nreak loss of coolant accident that would cause a rapid

tank draindow). Therefore, the venting function of the line could have been defeated.

This would have left the 4-inch overflow line as the only venting source for the tank,

which, according to the licensee's calculations, was not adequately sized to prevent a

vacuum from forming in the tank during a rapid tank draindown. Consequently, the

inspectors questioned the operability of the borated water storage tank during the 30

-__-

.

.

t 39

hours that the plastic bag was attached to the vent flange. Because of a lack of

redundancy, an inoperable borated water storage tank would render the entire

emergency core cooling system inoperable. The licensee performed several tests and

calculations in an attempt to support the original engineering judgement that the tank

was operable in this configuration. Following the onsite inspection period; the licensee

provided the following evaluations and calculations for NRC review;

from Don Phillips to Charles Tyrone, dated November 25,1997.

  • Calculation 97 E-02100 01, "BWST Venting Capability Through 4 inch Overflow,"

Revision 0, December 3,1997.

  • Calculation 97 E 021101, "BWST Level Analysis," Revision 1, December 4,

1997.

.

Calculation 97 E-0209-01," Evaluation of BWST Tank T 3 for Potential Vacuum

Load in Support of Operability Determination associated with CR-196-0663,"

Revision 0, December 8,1997.

The licensee concluded that borated water storage tank was operable during the period

of time the plastic bag was attached to the vacuum relief valve flanga. However, the

analyses clearly demonstrated that the borated water storage tank was in a degraded I

condition and could have plastically deformed above the water level in the tank in the

event of a large or medium break loss of coolant accident. This deformation, according

to the licensee's evaluation, would not have prevented the operation of the emergency

core cooling system as designed. Also, the effect of the internal vacuum on the tank

level instrumentation could have caused an early transfer of emergency core cooling

water system suction from the borated water storage tank to the reactor building sump.

This early transfer would have reduced the net positive suction head margins for the

emergency core cooling system pumps. The licensee determined that this would not

have rendered any of these pumps inoperable.

'

10_CFR 50 59 Issues

The inspectors determined that the licer,see had failed to evaluate whether temporary

changes made to the borated water storage tank had created an unreviewed safety

- question.

10 CFR 60.59, " Changes, Tests and Experiments," permits the licensee to make

changes to the facility and to procedures as described in the Safety Analysis Report

without prior Commission approval provided the change does not involve a change to the

Technical Specification and does not involve an unreviewed safety question. A

!

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.

-4 0-

proposed change, test, or experiment shall be deemed to involve an unreviewed safety

question if a possibility for an accident or malfunction of a different type than any

evaluated previously in the safety analysis report may be created. The licensee shall

maintain records of changes in the facility, and these records must include a written

safety evaluation that provid6s the bases for the determination that the change does not

I involve an unreviewed safety question.

The inspectors identified the following two examples where the licensee made changes

to the borated water atorage tarik vacuum relief line without adequately eva!uating

whether an unreview6d safety question existed:

  • The licensee reinoved the existing vacuum relief valve and covered the tank

flange with a plaatic bag taped to the tank. In the absence of a safety evaluation,

the licensee used engineering judgement to consider the tank operable under this

configuration. However, when questioned, the licensee had to perform extensive

testing and analysis beicte concluding that the original engineering assumption

was correct and that the tank was operable. The installation of the plastic bag

appeared to create the possibility for an accident or a malfunction of a different

type than any evaluated previously in the Safety Analysis Report and may have

created an unreviewed safety question.

  • The licensee installed a blind flange separated by a 1 inch gap from the tank

flange and installed a foreign material exclusion screen over the gap. The

unreviewed safety question evaluation of this configuration failed to consider the

air flow restriction created by the foreign material exclusion screen.

The failure to properly determine whether an unreviewed safety question existed as

t

required by 10 CFR 50.59 was identified as an apparent violation (50-313/9721-05).

Conclusion

Three apparent violations were identified associated with temporary coverings placed

over the borated water storage tank vacuum relief valve flange, when this valve was

removed on December 4,1996, for testing and maintenance. The apparent violations

involved the failure to perform proper safety evaluations as described in 10 CFR 50.b9,

the failure to perform the vacuum relief valve testing under the plant conditions as

prescribed in the test procedure in accordance with the requirements of 10 CFR Part 50,

Appendix B, Criterion V, and the failure to utilize the correct design control processes

(temporary alterations) for the temporary cover configurations placed over the valve

flange in accordance with 10 CFR Part 50, Appendix B, Criterion Ill.

,. .. . . . .. .

.

..

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-o

-_. . _. - . . _ - . ---- - - -

0

0

41-

E8.23 (Cloigd) Unresglyed item 50-313/9720122: Revised Calculations Associated with

BWST Vonexing and Pumo NPSH

Background

The licensee prepared an engineering report,93 R 1002-01, "ANO-1 BWST Outlet

Vortex Suppressor," dated February 5,1993, to examine the potential for vortexing in the

BWST resulting in a loss of net positive suction head (NPSH) for the emergency core

cooling pumps. The conclusion of the report was that the ECCS pumps would have

sufficient NPSH. Within this report, the licensee identified that it had failed to consider

the effect of instrument error in the vortexing calculation. This meant that the actual level

in the tank at the time of switchover from the BWST injection mode to the recirculation

mode may have been less than the nominal value, bringing the hydrodynamic system

closer to a state of cavitation and air entrainment than had been originally assumed. The

licensee issued Condition Report CR 197-0039 to investigate this condition. Within this

report, the licensee concluded that the ECCS system would have remained operable.

The licensee concluded that only the high pressure injection pumps would be susceptible

to an NPSH problem but only at low tank levels near the switchover point to the

recirculation mode. At this time in the accident, the high pressure injection (HPI) pumps

could be shut down because the low pressure injection pumps alone could provide

adequate core cooling. Within CR 197-0039, the licensee also was to complete

Revision 1 to Engineering Report 93 R 1002 01 and to determine whether other

calculations may have been affected by a similar failure to consider instrument error.

Insoector Followup

The inspectors reviewed the documents listed above and discussed the issue with

licensee engineers. The inspectors determin. d that this instrument error was only

reflected in the vortexing calculation and did not affect actualinstrument readings.

10 CFR Part 50, Criterion lil, " Design Control," states that the design control measures

, shall provide for verifying or checking the adequacy of design, such as by the

l performance of design reviews.

The licensee's failure to check properly the adequacy of Engineering

Report 93 R 1002-01 was identified as a fourth example of a violation of 10 CFR Part 50,

Appendix B, Criterion lli (50 313/972102).

Conclusioni

The licensee's failure to check the adequacy of an engineering report properly that failed

to account for instrument error in a calculation of vortexing in the borated water storage

tank was identified as a violation.

1

_ _ -.

.

.

l -4 2-

E8.24 [Gosed) Unresolved item 50 313/97202 23: InadtSuate 10 CFR 50.59 Evaluation

6110 dated With BWST Releastts

Background

This issue involved a problem witi 10 CFR 50.59 safety evaluation performed for

Temporary Alteration 97-1001, whi n is discussed in Section E8.23 of this report. In

assessing the potential post loss of coolant accident (LOCA) radioactive release through

the 3 inch gap foreign material exclusion (FME) covedag on the flange of the borated

water storage tank (BWST) vent, the licensee addressed only offsite exposure and did

not address control room exposure, which could ba considerably more significant. Also,

the safety evaluation did not refer to the licensee's response to information Notice 9156,

  • Potential Radioactivity Leakage to Tank Vented to Atmosphere "

Insoector Fe20Wup

The licensee revised the 10 CFR 50.59 evaluation for Temporary Alteration 971-001 to

include a discussion of control room dose resulting from potential backleakage to the

BWST. In the evaluation, the licensee stated that no accident consequence analysis

takes credit for holdup or filtering of radioactive releases to the BWST, and, therefore,

back leakage to the BWST is beyond the licensing basis. Based on recent test results of

the BWST supply and return valves, the licensee stated (in the revised safety evaluation)

that little or no backleakage would be expected to occur post LOCA. Consequently, the

licensee determined that there would be no change to the estimated control room

dosage as a result of the temporary alteration.

The licensee reviewed its response to information Notice 9156 and determined that the

conclusions reached as a result of Temporary Alteration 97-1001 were not affected by

this iirmation notice.

The inspectors considered the lack of reference to control room dose in the original

10 CFR 50.59 evaluation to be an oversight, but observed that two mitigating factors

existed. First, the original evaluation did not quantify a release rate based on the

assumption that the release would be negligible. Had the release been quantifiable, an

assessment of control room dose would have been necessary. Also, the fact ' hat no

credit was taken in the licensee's accident analysis for containment of BWST radioactive

leakage relegated the issue to a beyond-licensing basis status, The inspectors

considered the licensee's response to this issue to be adequato.

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IV. PlanLSunpart

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F2 Status of Fire Protection Facilities and Equipment

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a. Insoection Scooe (92903)

During a walkdown of the Unit i emergency feedwater system pump room, which was

performed in support of the review of various items discussed in Section E, the

inspectors inspected the condition of the electrical cable fire barrier material,

b. Qhstryations and Findings

The inspectors identified a deficiency in the fire wrap for Conduit EJ2029, which provided

protection for the control circuitry for the turbine-driven emergency feedwater pump. On

a vertical run of the conduit, a collar joining two sections of fire wrap had loosened and

slid down. This resulted in about a 2 inch portion of the conduit being exposed.

The inspectors reported the condition to the licensee and showed the deficiency to a fire

protection engineer and an operator. The licensee agreed that the as-found condition

was unacceptable and immediately declared the fire barrier inoperable. In accordance

with the requirements of its fire protection program, the licensee verified the operability of

smoke detection equipment and control room alarm capability and established an hourly

fire watch. Condition Report 197-0313 was initiated to document the item and Job

Request 928623 was initiated to repair the fire wrap.

The inspectors asked the licensee whether there was a program for inspecting fire

barriers and whether the fire wrap in the emergency feedwater pump room was included

within the scope of the program. The licendee provided the inspectors with

Procedure 1307.062, * Unit 1,1 Hour Cable Fire Wrap Inspection," Revision 0. The

inspectors reviewed the procedure, discussed fire wrap inspection techniques and the

process of dispositioning inspection observations with licensee personnel, and verified

that inspection of the pump room was included within the scope of the procedure. The

inspection of the sub; set ' ire wrap was last performed with satisfactory results in July

1997. The procedure was required to be performed every 6 months. At the time of the

inspection, no explanation was available for the cause of the deficiency. The licensee

determined that a noncompliance did not exist, because the surveillance interval had not

been exceeded and the damage could reasonably be assumed to have occurred

following the most recent inspection.

The inspectors considered the licensee's actons on this matter to have been

appropriate,

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c. Conclusions

The inspectors identified a fire barrier deficiency and the licensee took prompt

compensatory actions as required.

V. ManSment Meetings

XI Exit Meeting Summary

The inspectors presented the inspection resu'ts to members of licensee management by

telephone on January 6,1998. -The licensee acknowledged the findings presented.

During this meeting, the licensee stated that it will expand its definition of margin of

safety as it pertains to the performance of safety evaluations performed under

10 CFR 50.59. This issue is further discussed in Section E2.1 of this report. i

The inspectors asked the licensee whether any materials examined during the inspection  !

should be considered proprietary. The licensee sta..J that one document reviewed by

the inspectors contained proprietary information. The proprietary information from this

document was not discussed within this inspection report.

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ATTACHMEliT 1

SUPPLEMENTAL INFORMATION

PARTIAL LIST OF PERSONS CONTACTED

Licemen

G. Ashley, Licensing SupeMsor

E. Blackard, Design Engineer

M. Cooper, Licensing Specialist

G. Dobbs, Electrical and instrumentation and Control Supervisor

J. Howell, Design Engineer

R. Hutchinson, Vice President, Arkanshs Nuclear One

R. Lane, Director, Design Engineering

D. Mims, Licensing Director

D. Phillips, Nuclear Steam Supply Syatem Supervisor

J. Richardson, Design Engineer

C. Tyrone, Director of Design Engineering

T. Waldo, Supervisor, Modifications

C. Zimmerman, Plant Manager, Unit i

Nf1C

J. Melfi, Resident inspector

INSPECTIQfiERQCIQUEES USED

37001 10 CFR 50.59 Safety Evaluations

92903 Followup of Engineering issues

ITEMS OPENED. CLOSED. AND DISCUSSED

Ooened

50-313/9721-01 VIO Failure to incorporate EFW Flow Limits into Procedures

(Section E8.7)

50-313/9721-02 VIO Design Calculation Deficiencies (Sections E8.8, E8.11,

E8.20, and E8.23)

50-313/9721-03 APV Failure to Perform BWST Vacuum Relief Valve Testing

Procedure Under Proper Plant Conditions (Section EB.22)

50-313/9721 04 APV Failure to implement Temporary Alterations for BWST

Vacuum Relief Valve Removal (Section E8.22)

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50-313/9721 05 APV Failure to Perform 10 CFR 50.59 Evaluations for BWST

Vacuum Relief Valve Removal (Section E8.22)

Closed

50-313/9627-03 VIO Failure to Notify NRC Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of Declaration of NUE

50 313/96-009 LER Fire in Reactor Building During Heatup Resulted from

Cracked Weld in Oil Line

50-313; 368/96512-01013 EA inadequate Lube Oil Collection Systems for Reactor

Coolant Pumps

50 313/96512-01023, and EA Failure to identify and Take Prompt Action to Lube Oil

01033 Leakage

50-313/97201 01 IFl Licensee's Actions to Revise the Technical Specification

Bases for the Minimum Water Vo!ume in the Condensate

Storage Tank

50 313/97201 02 URI EFW Flow Rates Exceeding B&W Flow Rates

50 313/97201-03 URI EFW Piping Configuration Differences

50 313/97201-05 URI Evaluation of EFWP Pump Room Environment

50 313/97201 06 URI Drawing and Calculation Revisions

50-313/97201-07 IFl Modification Work Procedure Revisions

50 313/97201 08 IFl Add;tional Vendor infonnation Necessary for Revised

Caole Pulling Calculation

50-313/97201 11 IFl Verification of Calculation Changes Associated With DC -

Batteries

50 313/97201 12 IFl Review of Current Design Basis for Bypassing EDG

Protective Trip

50-313/97201 15 URI Lack of including all Safety Related Fuses in Procedure-

50 313/97201 16 URI FSAR Discrepancies

50-313/97201 17 URI Ineffective Resolution of Previous Corrective Actions

50 313/97201-18 URI . Lack of Seismic Support of OTSG Pressure Transmitters

50-313/97201 19 URI Lack of Design Basis for Support of OTSG Instrument

Sensing Lines

50 313/97201 20 URI Inadequate Work Plan for the Control of Post-

Maintenance Testing

.50 313/97201 21 URI Inadequate 50.59 Review Associated with BWST Breaker

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50-313/97201 22 URI Revised Calculations Associated with Vortexing and

Pump NPSH

50 313/97201-23 URI Inadoquate 50.59 Evaluation Associated with BWST

Releases

Discussed

50-313/97201-04 URI Inadequate Piping Pressure and Temperature

Specifications

50-313; URI Consideration of Multiple Hot Short Actuations

-368/9623-01

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50-313/97201 14 URI Lack of Testing of Unit 1 Molded Case Circuit Breakers

LIST OF ACRONYMS USED

A/E Architect / Engineer

BWST Borated Water Storage Tank

CR Condition Report

CRN Calculation Revision Notice

DC Direct Current

DCD Design Configuration Documentation

DCIMS Design Configuration Information Management System

EDG Emergency Diesel Generator

EFW Emergency Feedwater

FME Fore;gn Material Exclusion

FSAR Final Sa'ety Analysis Report

MOV Motor-operated valve

NPSH Net Positive Suction Head

QAT Quality Assurance Team

SIMS Safety Information Management System

ULD Upper Level Document

WBGT Wet Bulb Globe Temperature

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AIIACHMENT 2

LIST OF DOCUMENTS REVIEWED FOR ITEM 50 313/97-201-06

CALCULATIONS

Lalculation A-86," Items Discussed with Daleas R.S. of CE MSLB, ECCS and LOCA,"

Revision O.

Calculation Quality Action Team Final Reput, September,1996.

Calculation 32-1229968-00, " Final Draft, ANO AFW Capacity - Phase 1".

Calculation 32 5000450-G0, "ANO AFW Capacity, Phase 2," Revision 0.

Calculation 80-D-1083A-02, "EFIC DC Valve Torque Calculation Under Reduced Voltage

Condition," Revision 1.

Calculation 80-D-2085-01, " Correlation of Teletector to Activity Release," Revision 0.

Calculation 80-D-2085-02, " Provide Correlation Between MR/hr Reading Detector Activity Steam

Line Result Table Method," Revision 0

Calculation 85-D-1042-03, "AMSAC Instr Loop Errors & Setpoints," Revish15.

Calculation 85-E-0053-20, " Combustible Loading Calc for Fire Area G," Revision 29.

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Calculation 86-E-0002-01, "EDG 1 & 2 Load Study," Revision 8.

- Calculation 86-E-0020-01, "lE Station Battery 2D11 Duty Cycle'& Sizing Calc," Revision 7.

Calculation 87 E-0026-09, "ErM Pump Room Temperature Profile," Revision 0.

Calculation 88 E-0086-01, "lE Bulletin 88-04 Review for P7A and P78 Minimum Flow

Evaluation," Revision 0.

Calculation 89-E-0029-02, " Cal Factor Calc ANO-2 MS Line Rad Monitor," Revision 0.

Calculation 89-E-OiO2-01, " Terminal Voltage Calc for DC MOV," Revision 8.

Calculation 89-E-0144-01, "EDG Loading for 2A3 & 2A4 Bus," Revision 3.

Calculation 90-E-0116-01, "ANO-2 EOP Setpoint Document," Re.wision 4.

Calculation 91 E-0090-06, " Heat Load Determination for Post Accident for Rooms 95 98 99100

104109110 & 149 for Post Accident Cooling," Revision 2.

Calculation 92-E-0021-01, " Emergency Duty Cycle and Battery Sizing Calculations," Revision 4

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Calculation 92-E-0077-04, " Unit 1 EFW System Pump Performance Requirements,' Revision 0,

- Calculation 93-D-1008-05, 'MOV Sizing Calc for CV-2620,* Revision O.

Calculation 93-R-1029-30,'ANO-1 SRR for Auxiliary Building HVAC System ABH V,'

Revision O.

Calculation 94-D-5033-01, * Removal of Local Fire Trouble Horns from D21 Breaker 34,"

Revision 0.

Calculation 94-E-006101," Tech Spec Allowable Outage Time (AOT) Utilizing Attemate AC ,

EDG," Revision O.

Calculation 95-D-1004-03, 'DCP 95-1004 Load Addition to RC1," Revision O.

Calculation 95-D-1011-01,"EDG Load Add for DCP 95-1011," Revision 1.

Calculation 95-D 7064-01, " Vital AC Panels & Instrument AC Panels," Revision 0.

Calculation 95-D-7072-01,"D01 Battery Loading Amendment," Revision O.

Calculation 96-E-0065-01, " Unit i Safety Survivability Eval Under Stall Thrust / Torque,"

Revision 1.

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Calculation 96-R 0002-02, " Generic Letter 89-10 Program Valves Limit Switch Setpoint,"

Revision 1

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CONDITION REPORTS

CR-C-96-0060, re Calculation Room Control and Discrepancies, March 22,1997.

- CR-C-97-0053, re Calculation 92-D-2001-49 Approval, February 11,1997.

CR-C-97-0058, re Upper Level Document Discrepancies, February 13,1997.

CR-C-97-0059, re Design Configuration Documentation Discrepancies, February 13,1997.

CR-1-96-0184, re Seismic Qualification Calculations, May 23,1996.

CR-1-97-0017, re Accuracy Analysis Calculations for EFW Flow Transmitters, January 21,1997.

CR 2-97-0304, re Discrepancy in Instrument Setpoint Calculations 2LT-5659-1 and 2LT-5668-2,

May 28,1997

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DESIGN CHANGES

Design Change Package 94-1003, "Feedwater (MFW Venturi Replacement)," Revision 2

Design Change Package 95-2001, " Control Room Emergency Chiller Installation - Air

Conditioning Modification," Revision 1

ENGINEERING STANDARDS

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Engineering Standard GES-39, " Calculation Numbering Schemes and Codes," Revision 2

LICENSING INFORMATION REQUESTS

LIR L97-0035, ' Action items from ANO 50.54(f) Design Bases Assessment: Maintenance of

Calculations [1.1,1.2),",

LIR L97-0037, " Action items from ANO 50.54(f) Design Bases Assessment: DCIMS

Improvements [1.6],",

LIR L97-0201,"NRC AE Design Basis inspection," July 31,1997

LIR L97 0202,'NRC AE Design Basis inspection," July 31,1997.

LIR L97-0289, " Response to IR 97-202 AE Design Basis inspection," September 28,1997

LIR L97-0290, " Response to IR 97-201 AE Design Basis Inspection," September 26,1997

PROCEDURES-

Procedure OP-5010-015," Engineering Calculations," Revision 1 PC1,

Procedure OP-5010-019, " Computer Software Control," Revision 1 PC1

MISCELLANEOUS

Memorandum ANO-97-00225, M. Stroud to ANO Engineering Personnel," Reminder of Design

Considerations for Possible Impact to Battery and EDG Lording Calculations," March 3,1997

Design Configuration Documentation Discrepancies 2M7-07-01,02,03,07,08,09,12,14.

Design Configuration Documentation Discrepancies RBS 02-05,06

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