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{{Adams
#REDIRECT [[IR 05000498/1997024]]
| number = ML20199J559
| issue date = 11/21/1997
| title = Insp Repts 50-498/97-24 & 50-499/97-24 on 970915-1112.No Violations Noted.Major Areas Inspected:Electrical,Mechanical & Instrumentation & Control Calculations to Assess Scope of Design Control Weakness
| author name =
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
| addressee name =
| addressee affiliation =
| docket = 05000498, 05000499
| license number =
| contact person =
| document report number = 50-498-97-24, 50-499-97-24, EA-97-523, NUDOCS 9711280169
| package number = ML20199J541
| document type = INSPECTION REPORT, NRC-GENERATED, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 48
}}
See also: [[see also::IR 05000498/1997024]]
 
=Text=
{{#Wiki_filter:,
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'O
                                        ENCLOSURE
                      U.S. NUCLEAR REGULATORY COMMISSION
                                        REGION IV
    Docket Nos.:    50 498: 50 499
    License Nos.:    NPF 76; NPF 80
    Report No.:      50 498/97 24
                    50-499/97 24
    EA No.:          97 523
    Licensee:        Houston Lighting & Power Company
    Facility:        South Texas Project Electric Generating Station, Units 1 and 2
    Location:        FM 521 - 8 miles west of Wadsworth
                    Wadsworth, Texas
    Dates:          September 15 through November 12,1997
  : Team Leader:    L. Smith, Senior Reactor inspector
                    Division of Reactor 3afety
    inspectors:    T. Alexion, Project Manager
                    Of tice of NRR, PD IV 1
                    D. Acker, Senior Project inspector
                    Division of Reactor Projects
                    P. Gage Senior Reactor inspector
                    - Division of Reactor Safety
                    W. Wagner, Senior Reactor Inspector
                    Division of Reactor Safety
  . Approved Ry:    Arthur T. HowellIll, Director
                    Division of Reactor Safety
    ATTACHMENT: - Supplemental information
                                              i
              kl  9e
 
.                                  -                                  .-                  .
  .
  .
                                                TABLE OF CONTENI.S
    E X E C UTIV E S U M M A RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iil
    Ill . E ng ine er ing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
            E2    Engineering Support of Facilities and Equipment .................. 1
                    E2.1    10 CFR 50.59 Implementation . . . . . . . . . . . . . . . . . . . . . . . . . . 1
            E3    Engineering Procedures and Documentation (37550) . . . . . . . . . . . . . . . 3
                    E3.1    Program for Control of Accuracy of Measuring and Test Equipment
                            Used in Surveillance Procedur es . . . . . . . . . . . . . . . . . . . . . . . . . 3
                    E3.2 Battery Service Surveillance Test Acceptance Criteria . . . . . . . . . 6
            E7    Quality Assurance in Engineering Activities . . . . . . . . . . . . . . . . . . . . . . 7
                    E7.1 Control of Calculation Amendments . . . . . . . . . . . . . . . . . . . . . . 7
                    E7.2 Effect of Design and Licensing Dasis Changes on Calculations . . . 10
                    E7.3 Licensing basis Change Effects on the Updated Final Safety Analysis
                            Report (U F S A R) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
                    E7.4 Safety Evaluation Not Performed for Calculation Revision .....                                          15
            E8    Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
                    E8.1    (Closed) Inspection Followup Item 50 498; 499/9009-01: review of
                            the licensee's self assessment and followup of the actions taken to
                            evaluste the apparent discrepancies in the setpoint program . . . . 16
                    E8.2 (Open) Unresolved item 50 498;.499/9716-01: Review of the safety
                            impact of the licensee's decision to defer the planned setpoint
                            program improvements and review of the overall calculation control
                            p r og r a m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 0
                    E8.3 (Closed) Licensee Event Report 50-498; 499/97-06: Inappropriate
                            surveillance procedure monitoring parameters. .....,.......29
                                                                ii
-
                                                                                                                                    .
 
  .
  .
                                          EXECUTIVE SUMMARY
                      South Texas Project Electric Generating Station, Units 1 and 2-
                                NRC Inspection Report 50 498; 499/97 24
      This team inspection evaluated the capability of the licensee's design engineering
      organization to develop an analytical basis for the design of the f acility. This inspection
      assessed control of the design and licensing basis, as well as, corrective actions for
      previously identified weaknesses related to the development of scaling and setpoint
      documents. The team reviewed electrical, mechanical and instrumentation and control
      calculations to assess the scope of design control weaknesses. The team also reviewed
    - 10 CFR 50.59 safety evaluations and 10 CFR 50.59 applicability screening evaluations.
      The inspection covered a 5 week period with two of those weeks conducted onsite. .
      The team found that the unreviewed safety question evaluations and screenings were
      generally logical and of good quality. The team found that the licensee had not promptly
      addressed self identified weaknesses related to the development of setpoint calculations.
      In addition, the team identified calculation errors and associated apparent violations related
      to design control, control of the licensing basis and corrective action. However, no
      examples of inoperable equipment were identified. -The licensee initiated and planned
      corrective actions for the program weaknesses identified by the team.
      Engineering
      *      In general, the unreviewed safety question evaluations reviewed were
              comprehensive and well researched. Of the screenings which the team reviewed,
              none were found that required an unreviewed safety quest.on evaluation
              (Section E2.1).
'
      *      Measuring and test equipment accuracy assumptions from the process
              instrument uncertainty calculations were not being correctly translated into
              the process instrument calibration procedures. This f ailure was identified as an
              apparent violation of 10 CFR Part 50, Appendix B, Criterion lli (Section E3.1).
      *      Loads were being added and subtracted from the vital batteries without
              documenting an evaluation of the impact of the changes on the battery service
              surveillance test acceptance criteria. However, the team reviewed the critical
              loading period and determined that the acceptance criteria was bounding
              (Section E3.2).
      *-      Amendments to calculations were not being adequately controlled. Three
              calculations had greater than 100 amendments. Eleven calculations had 15 to-
              99 amendments and 62 calculations had 6 to 14 amendments. The failure to assure
              that design changes were subject to design control measures commensurate with
l-          . those applied to the original design was identified as an apparent violation of
            -10 CFR Part 50, Appendix B, Criterion ill (Section E7.1).
L
                                                      iii
                                                                                                    )
 
.
.
  *  Design basis calculations were not always consistent with the physical design of the
    rfant. Calculations were not being systematically evaluated or revised, when their
    associated design inputs were changed. The f ailure to assure that design changes
    were subject to design control measures commensurate with those applied to the
    original design was identified as an apparent violation of 10 CFR Part 50,
    Appendix B, Criterion 111 (Section E7.2).
  *  The team identified two examples where the licensee did not accurately update the
    Updated Final Safety Analysis Report (Section E7.3).
  *  Af ter discussions with the team, the licensee identified an example of a
    programmatic weakness. Prior to the inspectio1, the licensee did not have a
    procedural requirement to evaluate calculation changes for impact on the Updated
    Final Safety Analysis Report. As a result, they failed to perform a safety evaluation
    for changes in refueling watei storage tank volumes and uncertainty analysis
    assumptions described in the Updated Final Safety Analysis Report. This failure was
    identified as an apparent violation of 10 CFR 50.59 (Section E7.4).
  *  The licensee did not recognize that setpoint calculation deficiencies identified in
      1992 and 1995 were conditions adverse to quality. As a result they did not take
      Ofective corrective action until prompted by the NRC. This f ailure was identified as
      ea Aparent violation of 10 CFR Part 50, Appendix B, Criterion XVI (Section E8.1).
  *  Dased on a sample review, the team concluded that the licensee's setpoint guidance
      procedure was technically adequate. The team reviewed the basis for 20 setpoints,
      which had been identified as potentially having no basis. The team determined that
      3 of 20 setpoints still had a weak calculation basis. The team concluded that a
      random sampling of existing safety-related plant process values indicated that
      appropriate uncertainty calculations had not always been accomplished through the
      end of 1996. The team selected 6 safety related process values from the Updated
      Final Safety Analysis Report and found that calculations associated with
      4 of 6 process values were incorrect and required revision. The unresolved item
      related to the review of the safety impact of the licensee's decision to defer planned
      setpoint program improvements was lef t open for additional review. The NRC plans
      to evaluate the results of calculation revisions, performed to address team concerns,
      which were not complete at the conclusion of the inspection (Section E8.2l.
  *  The licensee identified that they had not ensured that instrument measurement
      uncertainty related to reactor coolant system average temperature was adequately
      accounted for in operating logs. This licensee identified f ailure to assure the
      design basis was correctly translated into procedures was identified as an apparent
      violation of 10 CFR Part 50, Appendix B, Criterion ill (Section E8.3).
                                              iv
                                                                                            )
 
.
.
  * The NRC subsequently identified that wher, the licensee accounted for the
    instrument uncertainty related to reactor coolant system average temperature in the
    operating logs, they used vendor information to revise a surveillance procedure
    without voiding the conflicting site calculation and without understanding the
    technical basis for the reduction in uncertainty. The NRC identified failure to assure
    that design changes, including field changes, were subject to design control
    measures comrnensurate with those applied to the original design was identified as
    an apparent violation of 10 CFR Part 50, Appendix B, Criterion 111 (Section E8.3).
                                            v
                                                                                          j
 
  .
  .
                                              Beoort Details
    insoection Obiectives (37001 and 37550)
    This inspection was performed to continue tha review of the adequacy of the licensee's
    design control program related to instrument wtpoint criteria, initiated in NRC Inspection
    Report 50 498:-499/97-16. Design engineering management had previously deferred
    planned corrective actions to correct design control program problems and calculation
    deficiencies involving instrument setpoint criteria. This inspection included further reviews
    of the implications stemming from the licensee % decision to defer corrective actions for
    setpoint calculation problems, and the effectiveness of the licensee's corrective actions for
    identified setpoint calculation problems. In addition, this inspection rev6ewed design control
    measurer associated with the mechanical and electrical discipline to determine the scope of
    design control weaknesses. This inspection also included an evaluation of the licensee's
    control of changes to the f acility pursuant to 10 CFR 50.59, " Changes, Tests and
    Experiments."
                                              111. Enginnathig
    E2      Engineering Support of Facilities and Equipment (37001)
,
    E 2.1    10 CFR 50.59 Imolementation
      a.    luncction Sconc
            The team reviewed nine unreviewed safety question evaluations (USOEs) and five
              10 CFR 50.59 applicability screenings and subsequent nuclear safety evaluations for
            temporary modifications, permanent modifications, engineering change notices,
            engineering analyses, and procedure changes.
      b.    Observations and Findinas
            The team found that the USOEs and screenings were generally logical and of good
            quality. Summary statements or assumptions were either supported with additional
            information within the USOE, or supported by additionalinformation that was
            promptly provided by the licensee upon request, in general, the USOEs were
            comprehensive and well researched in that they provided the appropriate
            background information and considered the appropriate accidents and licensing
            basis. None of the screenings reviewed were found to require a USOE. The team
            found two USOE's, which warranted detailed discussion: one USOE that could have
            used more clarification or detail and one USOE involved a smallincrease in
            probability of equipment malfunction, that was found acceptable.
                                                      1
                                                                                                  )
 
  .
  .
    .USOE 95-0028 Reactor Vessel Head Stud Tensioners
    This USOE addressed a change to the Updated Final Safety Analysis Report (UFSAR)
    *o more generically describe the reactor vessel head stud tensic,aers and to remove
      .
    statements that were incorrect or misleading. This change did not affect the
    function of the stud tensioners or the way the tensioner operated. The USOE also
    stated that the change would not affect the operation of the tensioner in a manner
    that was not accounted for in " previous analyses."
    The team asked what was meant by the previous analyses. The licensee responded
    that the previous analyses referred to was the stress / fatigue analyses, which
    assigned usage f actors to cover normal operating conditions, trans!ent expectations,
    and the number of expected vessel head detensioning and retensioni.'q cycles during
    refueling outages. The licensee further indicated that in their view, no elaboration in
    the USOE was needed on what was meant by previous analyses because both the
    originator and the qualified reviewer of the USOE understood the meaning of the
    term " previous analyses."
    The team considered the USOE would have been more complete and would have
    better stood alone, if the previous analyses had been clearly identified.
    USOE 96 0016 Defeat of Main Turbine Generator Overspeed Trip
    This USOE addressed a temporary modification to defeat the electrical overspeed
    trip for the main turbine generator. Apparently, the electrical overspeed trip of the
    main turbine generator became inoperable, and troubleshooting of this circuit
    involved a certain measure of risk that possibly could have caused a main turbine
    generator trip. Since the f ailure mechanism was unknown at the time, the licensee
    performed a temporary modification to defeat actuation of the overspeed trip.
    The licensee's USOE stated that the overspeed trip was part of the mitigation for
    - turbine missile generation, and that the overall probability of turbine missile
    generation from destructive overspeed would not be significantly impacted.
    Therefore, the overall probability of turbine missile generation from all scenarios was
    not significantly impacted and was within the safety analysis report limit. The USOE
    also stated that disabling the electrical overspeed trip would not increase the
    probability of missile generation beyond the threshold of E-4 events per year.
    The team then asked the licensee for clarification on the statement that the
    probability of turbine missile generation would not be significantly impacted. The
    licensee then provided a supplement to the USOE (that was inadvertently not
    provided earlier) that was done because the licensee's Nuclear Safety Review Board
    had also asked for clarification of the meaning of "there is little impact to the overall
    probability of missile generation." The supplement stated that setting the probability
    of the electrical overspeed trip failure to "one" resulted in no change to the third
    significant digit of the probability of destructive overspeed (one of three types of
l
    overspeed), and that the overall probability for missile generation was not
                                              2
                                                                                              0
 
                        .  .            _                  . _-.      .
.
.
      statistically changed by deletion of the overspeed trip. While not statistically
      significant, the licensee agreed that intentional defeat of the overspeed trip device
      slightly increased the probability of generating a turbine missile. However, the
      licensee noted that af ter considering the overspeed trip defeat, the probability of
      generating a turbine missile was still much less than the criteria approved by the
      NRC staff, E 4 events per year.
      To verify the licensee's statement, the team reviewed a May 18,1993, Safety
      Evaluation by NRC's Office of Nuclear Reactor Regulation, related to the turbine
      maintenance program. The Safety Evaluation concluded that the licensee's turbine
      missile program was acceptable because the licensee's turbine missile generation
      probability satisfied the staff's requirements of E-4 per year for favorable orientation.
      The team also reviewed NUREG-1606, " Proposed Regulatory Guidance Related to
      implementation of 10 CFR 50.59 " NUREG 1606 stated that " Changes that might
      increase the probability of (turbine missile] generation from the existing level.to a
      level that is still below the specified criteria would not create a new type of
      accident, or increase the probability of an accident previously evaluated."
      Based on the above, the team found that deletion of the electrical overspeed trip
      was not an unreviewed safety question because the licensee's turbine missile
      generation probability continued to satisfy the staff's specified criteria of E-4 events
      per year. The team agreed with the licensee's USOE.
              _
  c.  Conclusions
      The unreviewed safety question evaluations reviewed were comprehensive and well
        researched. With one exception, the team found the background documentation
        contained in the nuclear safety evaluations to be well developed and complete. In
        general, the safety evaluations provided appropriately detailed bases for reaching
        conclusions regarding changes, tests, and experiments. All conclusions appeared to
        be logically supported and did not represent any unreviewed safety questions. Of
        the screenings which the team reviewed, none were found that required an
        unreviewed safety question evaluation.
  E3    Engineering Procedures and Documentation (37550)
  E3.1  &ngram for Control of Accuraev of Meas.Urina and Test Eauioment Used in
        Surveillance Procaduins
  a,    lasoection Scops
        The team performed a limited review of instrument setpoint program documents and
        maintenance and surveillance procedures to determine if measuring and test
        equipment (M&TE) accuracy assumptions made in the process instrument setpoint
        uncertainty analysis were being implemented in the field.
                                                3
                                                                                                J
 
.
.
  b. Observations and Findin25
    The team reviewed safety related setpoint calculations performed by the reactor
    vendor, Westinghouse. The team found that the Westinghouse calculations for
    safety-related instrument setpoints assumed that M&TE used would be four times
    more accurate than the installed instrumentation, except for detectors, where the
    M&TE was assumed to be as accurate as the detector. The team reviewed the
    licensee's surveillance program and determined that Plant Surveillance
    Procedure OPGP03 ZE 0005, " Procedure Preparation," Revision 12, had a checklist
    item for ensuring that the testing methodology supported setpoint assumptions.
    The team reviewed four procedures associated with setpoint criteria provided
    by Westinghouse and determined that two of the four procedures specified
    M&TE accuracies which were less accurate than was assumed by Westinghouse.
    Surveillance Procedures OPSP05 RC-0417, "RCS Flow Transmitter Calibration,"
    Revision 0, and OPSP05 MS 0514L, " Main Steam Pressure Loop Calibration,"
    Revision 1, both required a voltmeter accuracy of 0.15 percent, when an accuracy
    of 0.015 percent was assumed in the associated Westinghouse calculations.
      10 CFR Part 50, Appendix 0, Criterion Ill, requires that the design basis be correctly
    translated into procedures. The team considered that f ailure of Surveillance
    Procedures OPSP05 RC 0417, Revision 0, and OPSP05 MS-0514L, Revision 1, to
    incorporate the voltmeter accuracies assumed in the associate i Westinghouse
    uncertainty calculations was an example of an apparent design control violation
    re!ated to translation of M&TE accuracy requirements (50-498; 499/9724-01).
    The team discussed M&TE accuracy with the licensee, who noted that the actual
    instruments, which had been used to perform the procedures did have an accuracy
    of 0.015 percent. Th.s team noted, however, that the procedures listed test
    equipment or equal, so that an voltmeter with an accuracy of 0.15 percent could be
    used to perform the surveillance. Tha licensee stated that they would correct the
    procedures.
    The team reviewed two surveillance procedures associated with setpoint
    calculations performed by the licensee and determined that although test equipment,
    or equal, was specified, there was no listing of required accuracies. The team
    reviewed Procedures OPMPOS ZE-0034, "Caliteation of ITE-27 Relays," Revision 3,
    and OPSP06 PK-0005, "4.16kV Class 1E Degraded Voltage Relay Channel
    Calibration /TADOT Channel 1", Revision 4. The team was unable to determine how
    the licensee maintained control of M&TE accuracy when the procedures did not
    specify any accuracy requirements.
                                            4
                                                                                            j
 
  . - . .- --            .,      ._ - -
..
.
              The team requested that the licensee provide the accuracies of the test equipment
              specified in Procedure OPSP06 PK-0005 Procedure OPSP06 PK-0005 directed that
              the degraded voltage relays be tested using a test set monitor, either a EPOCH lit or
              an EPOCH 30. The licensee provided ir. formation which indicated that the accuracy
              of the voltmeter on the EPOCH 30 was plus or minus 1.0 percent of reading.
              The team reviewed the calculation for the degraded voltage relay trip setpoint,
              Calculation EC5052, " Degraded and Undervoltage Protection," Revision 3. The
              team found that this calculation required a test voltmeter accuracy of plus or
              minus 0.4 percent, in addition, as discussed in Section E8.2, the team considered
              the uncertainty analysis performed in this calculation to be in error. After
              considering the use of test equipment with an accuracy of 1.0 percent of reading
              and the calculation errors discussed in Section E8.2, the team determined that
              Procedute OPSP06 PK 0005 and Calculation EC5052 were inadequate to
              demonstrate that the degraded voltage relay trip setpoint met its design basis of
              ensuring adequate voltage to safety related equipment.
              The licensee reviewed this information, interviewed personnel, and concluded that
              all recent data for degraded voltage relay trip setpoints had been taken with a
              multimeter with an accuracy of 0.5 percent or better. The team noted that
                Procedure OPSP06-PK 0005 listed a multimeter in the list of test equipment, but at
              no time during the setting of the degraded voltage relay trio setpoint did the
              procedure specify that the multimeter be connected or used for voltage readings.
                As noted in Section E8.2, the licensee perforrned a revised uncertainty analysis for
                the relay circuit. This revised analysis was based on M&TE accuracy of .5 percent
                or better. Af ter considering the corrected uncertainty analysis and the licensee's
                use of more accurate test equipment, the team agreed with the licensee's
                conclusion that the equipment was operable.
                10 CFR Part bO, Appendix B, Criterion ill, requires that the design basis be
                correctly translated into procedures. The team considered that failure of
                Procedure OPSP06-PK-0005 to incorporate the M&TE accuracy requirements
                required in Calculation EC5052 was a second example of an apparent design
                control violation related to translation of M&TE accuracy requirements
                (50 498;-499/9724-01).
                The licensee stated that this issue was programmatit in that the required eccuracy
                i_nformation was not readily accessible to field personnel. The licensee stated that
                they initiated a number of actions to ensure that calculation assemptions would be
                included in field instructions. Included in this review was Condition Report 97 238,
                Action 104, to ensure electrical maintenanc6 and surveillance procedures
                implemented calculation requirements.
                                                          5
                                                                                                    0
 
                                                  __
  .
  e
    c.    Conclunops
          The team concluded that M&TE accuracy assumptions were not being adequately
          translated into surveillance procedures. For the identified examples, the licensee
          determined that appropriate measuring and test equipment had been used.
          However, the f ailure to translate M&TE accuracy requirements into surveillance        ,
                                                                                                  '
          procedures was an apparent violation of 10 CFR Part 50, Appendix B, Criterion ill,
          " Design Control." The licensee stated that this issue was programmatic in that the
          required accuracy information was not readily accessible to field personnel. The
          licensee planned to upgrade their program.
                                                                                                  l
    E3.2 Batterv Service Surveillance Test Acceotence Criteria
      a.  Scnns
          The team reviewed the battery load calculation, the Updated Final Saf t.ty Analysis
                                                                                                  l
          Report direct current load table and the battery service surveillance test to determine
          if the acceptance criteria was bounding.
      b.  Observations and Findinas
          The team noted that Calculation EC5008, " Class 1E Battery, Battery Charger, and
          inverter Sizing," Revision 10, had 10 amendments. The licensee stated this
          calculation was the calculation of record for determining plant electricalloading for
          the Class 1E 125Vdc system. The team found that Calculation EC5008 had been
          updated on November 27,1995 and appeared to be adequate for specific design
          basis accidents.
          The tearn noted that the des;ga change packages, included in the 10 amendments,
          did not identify the battery service surveillance test procedure for appropriate review
          within the design change process. The team found, as an example, that design
          change package 96-3056-34 installed a backup power supply for a damper actuator
          in the fuel handling building, which would be powered from the Class 1E 125Vdc
          station batteries. As indicated on the 10 CFR 50.59 screening form for the design
          change package, a review of the Updated Final Safety Analysis Report section 8.3
          and Calculation EC5008 were performed. The team noted that no references were
          made regarding updates to assure that affected surveillance procedures used to
          verify the operability of the Class 1E 125Vdc system were reviewed or updated.
          The team determined the licensee's program for design control was weak in this
          area, because a review of affected battery service surveiFance tests was not
          performed.
          The team selected the critical battery loading period, zero to one minute on the
          Channel I battery, for additional review. The team concluded that, af ter
          consideration of all of the amendments (some reduced the loading), the battery
          service surveillance test acceptance criteria was bounding.
                                                    6
l
,
 
  .-  . -    - - - . - - - - - -                          ----          --
    .
.    .
            c.      ' conclusions
                      The licensee's program for design change control related to battery testing was
                      weak. Loads were being added and subtracted from the vital batteries without
                      documenting an evaluation of the impact of the changes on the battery service
                      surveillance test acceptance criteria. However, the team reviewed the critical
                      battery loading period and determined that the acceptance criteria was bounding.
          E7          Quality Assurance in Engineering Activities (37550)
          E7,1        Control of Calculation Amendments
            a.        Insoection Sqogg
                      As part of the review of the licensee's calculation control program, the team
                      selected for review a calculation with a large number of amendments. -The team
                      reviewed the adequacy of the control of amendments to Calculation EC5002,
                      " Auxiliary Power System Load Study," Revision 4, dated October 30,1988.
            b.        Observations and Findinas
                      Calculation EC5002 was a computer aided study which determined plant total
                      connected electricalloads and expected loads for startup, full power. loss of offsite
                      power, and loss of coolant conditions. The licensee stated that this calculation vvas
                      the calculation of record for determining plant electricalloading.
                      The team noted that Calculation EC5002 had approximately 190 open and 30
                    - pending amendments. The licensee later determined there were 214 open
                      amendments. The licensee's definition of amendments included a number of
                      document types for accomplishing design changes. The licensee's design control
                      program allowed amendments to be attached to affected documents, without
                      changing the original document, or updating the list of effective pages. However,
                      the licensee required personnel making an amendrnrst to consider, and incorporate
                      previous amendments, if the previous and new amendments affected the same part
                      of the document.
                      The team reviewed all of the open amendments to Calculation EC5002 and
                      determined that, for the most part, there was no attempt by personnel making
                      amendments to EC5002, to incorporate previous amendments or there was no
                      attempt by personnel to determine the overall affect on plant loading as the number
                      of amendments increased. In addition, most of the amendments addressed only
                    . total connected load and not operational conditions, and many of the amendments
                    - only described the changes on a cover sheet, without making any changes to the -
                      actual calculation sheet
                                                                  7
                                                                                                            !
 
    - .--    - -          - - -        - - - _- -                ---          -.        .-.-.- -.-
  ,
  ,
              A few of the specific problems discovered by the team were as follows:
              *        Design Change EC 62, dated October 16,1991, made load changes for
                        Buses 3E171EMCE1C2 and B4 without incorporating changes made by
                        Design Change EC 49, approved February 26,1990, causing the loss of-
                        power loads to be incorrect.
              *        Design Change EC 62, dated October 16,1991, made load changes for
                        Bus 3E151ESGOE1 A without incorporating changes made by Design
                        Change EC 32, dated August 30,1989. Design Change EC 32 had increased                I
                        load on the bus by 83 kilowatts (kw), which was not recognized by EC 62,
                        thus, the totalload shown by EC 62 was 83 kw low.                                    I
                                                                                                            l
                                                                                                            '
              *        Design Change PCF 176712A approved June 19,1995, made load changes
                        to the total connected load of Bus 8E171EMOC1F2 without incorporating
                        changes made by Design Change PFC 211205A, approved April 27, 1994.-
                *      Design Change MDCN 90037 04, issued December 18,1995,added
                        technical support center diesel control circuits to the full power loads, but not  '
                        to the total connected loads or loss of power loads.
                *      Design Change ECN 88 L 00100, dated June 6,1989, added a 37 kw load
                        by cover sheet without showing the specific changes to Calculation EC5002.
                Based on this review, the team was unable to determine that the licensee had              .
                                                                                                            '
                adequate design control for the electrical distribution system. The team discussed
                this issue with the licensee. They stated that all of the amendments to EC5002
                were minor, and that other documents, such as design drawings, were available to
                show that the present design configuration was adequate, During this discussion,
                the team determined that the licensee had not purchased the computer program
                used by the architect engineer to perform the original design. As a result, they
                could not easily recalculate total connected load and had not maintained a
                calculation commensurate with the original study. The licensee stated that they had
                recently purchased new sof tware and were in the process of inputing electrical
                distribution system design and operational data into a new computer aided study.
                Because the team was concerned with the adequacy of the electrical distribution
                system, the team reviewed two calculations specifically performed to ensure
                adequate electrical distribution support for safety related equipment, Calculation
                EC5092, " Class 1E Standby Diesel Generator Loading Analysis," Revision 0 and
e                Calculation EC5008, " Class 1E Battery / Battery Charger / Inverter Sizing," Revision
                  10. The team determined that the licensee. had updated these calculations and, in
            ..
                general, the calculations indicated adequate electrical distribution system support for
                safety related equipment under accident conditions which required operation of the
                emergency diesel generators and safety related batteries.
                                                                                                          !
                                                        8
          m
                                                                                                          j
 
.
.
  Even though Calculations EC5008 and EC5092 had been updated more recently and
  appeared to be adequate for specific design basis accidents, the team was unable to
  conclude that the licensee had maintained adequate control of the design of the
  electrical distribution system without a full understanding of the results of the
  auxiliary power system load study that was covered by Calculation EC5002.
  The licensee initiated Condition Report 9714608 to revise Calculation EC5002.
  Af ter 20 person days of research, the licensee was able to demonstrate to the team
  that the total effects of all the amendments to Calculation EC5002 were well within
  the capabikties of the affected buses. In response to the team's findings, the
  licensee performed a self assessment regarding the generic aspects of the Isrga
  number of open amendments posted against specific calculations. The licensee
  identified that two other calculations had greater than 100 amendments, that
  11 other calculations had between 15 and 99 amendments, and that 62 calculations
  had between 6 and 14 amendments. The licensee evaluated the immediate safety
  significance of these design control measures. They noted that in several cases
  there were alternetive methods for controlling the design. For example, the licensee
  noted that plant changes associated with cables and raceways were also controlled
  by a computer aided circuit and raceway program. The licensee indicated that they
  had carefully maintained this program, to ensure that plant changes had not
  overloaded cables or raceways.
  The licenseo stated that review of Calculation EC5002 and other calculations with
  amendments, led them to conclude that the present practice for amendment control
  did not meet engineering management's expectations for control of chlculations.
  The licensee stated that they planned to limit most calculations to five amendments,
  and planned to revise all calculations with more than five amendments to
  incorporate all of the amendments. For example, the licensee stated that they
  would revise the cable and raceway calculations discussed above to ensure that the
  design had been adequately maintained by use of the computer aided cable and
  raceway program. The licensee noted that a relatively small percentage of the total
  population of approximately 6000 calculations was affected.
  The team noted that prior to March,1994 amendments were controlled by
  Procedure OEP 3.070, " Preparation of Engineering Calculations." During this period
  Calculation EC5002, described above, had over 50 active amendments. Rather than
  solving the problem, the licensee relaxed the requirements and, af ter March 1994,
  effectively allowed an unlimited number of amendments to exist before a calculation
  was revised.
    10 CFR Part 50, Appendix 8, Criterion Ill, requires that design changes, including
  field changes, shall be subject to design control measures commensurate with those
    applied to the original design. The team noted that the South Texas Operations
    Quahty Assurance Plan, Chapter 6.0, " Design and Modification Control,"
    Section 5.2.3 stated that design analyses shall be sufficiently detailed as to
    purpose, method, assumptions, design input, references, units, and status
    (preliminary or final) such that a technicelly qualified pe son can review and
                                            9
                                                                                        d
 
  . .                                        .      _        .    -                      -      - - -
*                                                                                                        i
.
                                                                                                          l
            understand the analyses and verify the adequacy of the results without recourse to
            the originator. The team concluded that design changes to Calculation EC5002                  j
            were not subject to design control measures commensurate with those applied to
            the original design. As a result a technically qualified person could not review and    .
            understand the analyses and verify the adequacy of the results. The team
            concluded the f ailure to update Calculation EC5002 commensurate with the original
            design was an apparent design control violation related to amendment control                  !
                                                                                                          '
            (50 498; 499/972A 02h
      c.  CDDdu11gHS                                                                                    I
                                                                                                          l
            Amendments to calculations were not being adequately controlled. Three                        l
            calculations had greater than 100 amendments,11 other calculations had between                )
            15 and 99 amendments, *nd 62 calculations had between six and 14 amendments.
            The team reviewed one c. 4e in detail and found amendments were not being
            integrated to develop a coherent analytical basis for the auxiliary power system load
            study. The f ailure to assure that design changes were subject to design control
            measures commensurate with those applied to the original design appeared to be a
            violation of 10 CFR Part 50, Appendix B, Criterion Ill.
      E7.2 Effect of Desian and Licensing Basis Chances on Calculations
      a.  Inspaction Scops
            The team reviewed electrical and mechanical calculations to assess the adequacy of
            design control measures. The team reviewed these calculations to determine
            whether the impact of design and licensing basis changes had been fully considered.
      b.  Observations and Findings
            Diesel Generator Transient Loading
            During the revicw, the team observed that Condition Report 95-10936 identified
            potential problems with the transient loading of the emergency diesel generatcis.
            Based on this condition report, the team reviewed the emergency diesel generator
            transient loading calculation.
            The licensee indicated that NEl Peebles Electric Products, Inc., Study
            Order T 1031, " Transient Voltage Response of the Diesel Generator Units, Trains A,
            B, and C to Postulated Emergency Loading," Revision 0, and associated changes
            provided the technical analysis that supported the conclusion that the emergency
            diesel generators would perform satisf actorily during a loading sequence upon
            loss-of-offsite power or upon a loss of-offsite power with a loss-of-cooling accident.
                                                    10
                        ,m    y                                                -m
 
                                                _      __
.
                                                                                        ;
.
  The last change to this study that was provided to the team was made in 1989.          I
  The team observed that many changes had been made to the loading of the                i
  emergency diesel generators since the study had been done, as indicated in
  Calculation EC5092, " Class 1E Standby Diesel Generator Loading Analysis,"
  Revision O.
  The team found that there was conflicting and incorrect information within the
  transient study. For example, inputs listed on the computer modeling did not match    j
  the loading inputs developed in the study and changes had been made to the study
  that changed base information without explanation. The team also found that the
  study contained conflicting load sequencing information, and that the study did not    ;
  match existing configurations because numerous changes had been made to loads          ;
  and load sequencing since the study was last updated in 1989.                          !
  Despite the numerous problems, the team determined that all the loading and            1
                                                                                        '
  loading sequences listed as input information for the computer modeling were
  conservative with respect to the current load and load soquencing calculated in        i
  Calculation EC5092. Current total loads and individualli ad block additions were
  appropriately less than sssumed in the study. The licens se stated that they planned
  to update the emergency diesel generator transient loadhg analysis using newly
  purchased computer programs.
  10 CFR Part 50, Appendix B, Criterion ill, requires that design changes, including
  field changes, shall be subject to design control measures commensurate with those
  applied to the original design. South Texas Opetations Quality Assurance Plan,
  Chapter 6.0, Section 5.1, states that measures shall be esiablished to document
  selection of design inputs. Changes to specified design inputs, including
  identification of their source, shall be identified and documented. While the team
  did not have a safety concern in this case, the team concluded that the licensee had
  changed the design innuts to Study Order T 1031, Revision 0, without identifying
  and documenting that the inputs had been revised and evaluating the effect. As a
  result, the current design was not being controlled commensurate with the original
  design. The failure to identify and evaluate the load changes against the study to
  confirm it was still bounding was one example of an apparent design control
  violation related to identifying and documenting design input changes
  (50 498;-499/9724-03).
  Change in Auxiliary Feedwater System Requirements
  The team found that the licensee made several changes to the dynamic
  requirements for the auxiliary feedwater system which affected various calculations.
  On May 27,1994, the NRC staff issued Technical Specification Amendment 61 for
  Unit 1 and Amendmei.t 50 for Unit 2 to reduce the system flow requirements from
  550 gallons per minute (gpm) to 500 gpm. On May 2,1995, the NRC staff issued
  Technical Specification Amendment 78 for Unit 1 and Amendment 67 for Unit 2 to
  change the Technical Specification main steam safety valve tolerance from i 1 %
  to 13% This had the effect of increasing the required pump discharge pressure.
                                          11
                                                                                      J
 
                      __    _                      .    ._              _
,
.
  At initial licensing the licensee installed new pump impellers to improve system
  performance and generated new pump curves. The licensee also had modified the
  piping configuration and developed a more detailed estimate of system resistance.
  The team reviewed current calculations to determine if they had been updated to
  reflect these changes. The team found that the following five calculations had not      ,
  been revised or evaluated when the system parameter used as a design input was
  updated.
  *        'the piping configuration was changed and the change was not identified and
          documented in Calculation MC5004, "AFW Suction Line Sizing and Pump
          Available NPSH," Revision 2 dated August 22,1995. Specifically, the
          available net positive suction head in Calculation MC5004 was calculated
          based on 100 feet of 6 inch piping. The latest isometric drawings showed
          approximately 23 feet of 6 inch pipe and 109 feet of 8 inch pipe.
  *        The design pressure for the turbine driven auxiliary feedwater discharge
          pipmg was changed and the change was not identified and documented in
          Calculation MC5060, " Auxiliary Feedwater Line Sizing," Revision 1, dated
          Der ember 6,1985. Calculation MC5060 was based on Revision 1 of
          Ce station MC5001, " AFW Pump D!scharge Pressure," and did not include
          the crianges in Revision 4 of Calculation MC5051, issued March 27,1992.
  *        The syn tem reristant e model was updated to reflect current piping
            configurations and the change was not identified and documented in
          Calculation MC6864, "AFW Pump Runout Flow," Revision 2, dated June 7,
            1989. Calculation MC5864 was based on Colculation MC5861, " Auxiliary
            Feedwater (AFW) Pump Design TDH and Flowrate," Revision 1, dated
          January 15,1987, and did not include the changes in Revision 3 of
            Calculation MC5801 issued July 14,1997.
  *        Minimum system flow requirements, maximum system pressure
          requirements, and system resistance were updated and the changes were not
            identified or documented in Calculation MC5056, " Auxiliary Feedwater (AFW)
            Control Valve Sizing: AFW System Resistance," Revision 12, dated
          January 3,1986. Calculation MC5056 used these parameters as inputs and
            was not updated when the t,ystem requirements changed.
  *        Maximum system pressure, the auxiliary feedwater pump curve and the
            system resistance were updated and the changes were not identified or
            documented in Calculation MC5924, " Auxiliary Feedwater Regulating
            Valves Anticipated Cycles," Revision 1 dated April 16,1987.
            Calculation MC5924 used these parameters as inputs and was not updated
            when system requirements changed.
  10 CFR Part 50, Appendix 8, Criterion lil, requires that design changes, including
  field changes, shall be subject to design control measures comn,ensurate with those
  applied to the original design. South Texas Operations Quality Assurance Plan,
                                            12
                                                                                        -)
 
.
.
    Chapter 6.0, Section 5.1, states that measures shall be ests.olished to document
    sclection of design inputs. Changes to specified desig*1 inputs, including
    identification of their source, shall be identified and documented. The team
    concluded that the licensee chariged the design inputs to the five noted calculations
    without identifying and documenting that the inputs had been reviced. As a result
    the current design basis ca culations were not consistent with the physical design of
    the plant. The five failures to identify and evaluate the impact of auxiliary
    feedwater system changes as described above represented five additional;xamples
    of an apparent design control violation related to identifying and docummiting design
    input charges (50 498; 499/9724 03).
    Calculation impact Review Sheet
    The licensee agreed that calculations were not always being updated to match the
    current plant configuration. Several of the examples involved the output of one
    calculation being used as an input to a second calculation. When the first
    calculation was revised, it changed the design input to the second calculation.
    However, the second calculatinn was not revised To address this programmatic
    weakness, the licensee revised their calculation control program te 'nclude an impact
    review sheet. The licensee planned to use the impact review to identify and
    document the impact of calculation changes on procedures, the Updated Final
      Safety Analysis P ort, and other calculations,
  c. CDDclusions
      The licenseo changed the design inputs to Study Order T 1031, " Transient Voltage
      Response of the Diesel Generator Units, Trains A, B, and C to Postulated Emergency
      Loading," Revision 0 without identifying and documenting that the design inputs had
      been revised. As a result the current design was not being controlled commensurate
    with the original design. While the licensee's study to demonstrate adequate
      emergency diesel generator voltage, in response to accident loads, was out of date
      and contained errors, it bounded the current loading calculations and was not
      considered to be a safety concern. The team identified five similar examples of
      inadequate design change control for the auxiliary feedwater system. The licensee
      revised key system parameters without identifying and documenting the design
      input changes to five calculations, lne team concluded that the licensee was not
      systematically ensuring all calculations affected by design and licensing basis
      changes were reviewed and updated as needed to be consistent with the physical
      design of the plant. The team concluded these f ailures to identify and document
      design input changes resulted in a f ailure to assure that design changes were subject
      to design control mr asures commensurate with those applied to the original design.
      This appeared to be a violation of 10 CFR Part 50, Appendix B, Criterion Ill.
                                              13
                  -      -        -.
                                                                                            j
 
  _  . - _ _ _ _ _                    _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ .. _ _ _ _ _ _ __ _ _ _ _ .
                                                                                                                  r
    *
                                                                                                                  l
                                                                                                                  !
    .
                    E7.3 Licensina basis Chance Effects on the Undated Final Safety Analysis Reggs                i
                          LUFSAR)
                    , s.  'Insnaction Scone
                                                                                                                  i
                          During the review of electrical and mechanical calculations discussed in.
                          Section E7.2, the team evaluated the UFSAR to determine if it was consistent with      ;
                          the latest revision to the calculations.
                      b.  Observations and Findinos
                                                                                                                  '
                          The team found two examples where the UFSAR was not accurately updated during
                          the last revision as required by 10 CFR Part 50.71(e), " Maintenance of Record,        '
                          Making of Reports." Both examples involved incorporation of design information
                          which had been reviewed and approved by the NRC staff.                                  ;
                                                                                                                  i
                          *        The license revised the response to Oudstion 440.30N to say that the
                                    485,000 gallon limit was all usable. The team noted that the 485,000 gallon  ;
                                    limit also included an unusable portion which had been accounted for in      ;
                                    calculations and reviewed by the NRC staff.
                                                                                                                  '
                            *      The licensee did not comprehensively update UFSAR Section 7A.ll E.1.1,
                                    " Auxiliary Feedwater System Evaluation," when the auxiliary ieedwater
                                    system flow was reduced. They did not change the amount of time to
                                    specified to purge the volume between the water solid portions of the
                                    auxiliary feedwater/ main feedwater system and the steam generators
                                    following an auxiliary feedwater pump start. However, the licensee
                                    determined that these purge times were not an input to the safety analysis;
                                    therefore, there was no safety significance to this finding.
                            The licensee stated that the change notices, which initiated these updates were
                          prepared before the licensee increased their emphasis on UFSAR accuracy. The
                            licensee previously committed to a complete reverification of the UFSAR to confirm
                            its accuracy. The licensee stated that they had not yet performed the reverification
                            of these specific safety analysis report sections in question.                        i
                            This item is unresolved pending further NRC review of the 10 CFR 50.71(e) issues
                            (50 498; 499/9724-04).
                      c,    Conclusions
                            The team identified two examples where the licensee did not accurately update the
                            UFSAR.
                                                                                                                  ,
                                                                                                  14
e          ,
                                                                                                                O
 
                                            -  ._                _  -      -                .
.
.
  E7.4 Safetv Evaluation Not Performed for Calculation Revision                                    .
        The team noted that the licensee lacked a procedural requirement to perform a
        10 CFR 50.59 safety evaluation screening review of calculation revisions to identify
        those calculation changes which impacted the f acility as described in the UFSAR.
        The licensee subsequently identified an example of a calculation change which              ,
        changed the f acility as described in the safety analysis report without a 10 CFR
        50.59 safety evaluation.
        The licensee provided Calculation MC5037, " Determination / Validation of RWST
        Level Sotpoints," Revision 7, dated July 11,1995, and Condition Report 97 14434,
        dated September 17,1997 to the team. Condition Report 9714434 identified
        differences between Revision 7 of the calculation and UFSAR Section 6.3.1. The
        licensee identified discrepancies between nominal refueling water storage
        tank (RWST) volumes specified in the safety analysis report and the volumcs
        calculated in Calculation MC5037. The licensee stated that the differences were
        minor and that they wNid correct the UFSAR.
        The team reviewed the calculation and the UFSAR and noted that the discussion of
        instrument uncertain les was also different between the two documents, but agreed
        with the licensee that the differences appeared minor.
        However, the team noted that the more conservative instrument uncertainty                ,
        assumed in the safety analysis report provided additional margin against air ingestion
        into the suction of the emergency core cooling system pumps. Both the calculation
        and the UFSAR treat all of the water above the opening of the suction pipe as
        available water. The team was concerned that this assumption is not technically
        valid. This concern is discussed further in Section E8.2.
        10 CFR 50.59(b)(1) requires that the licensee maintain records of changes in the
        f acility made pursuant to this section, to the extent that these changes constitute
        changes in the f acility as described in the safety analysis report. Further, these
        records must include a written safety evaluation which provides the basis for the
        determination that the change does not involve an unreviewed safety question. The
        f ailure to perform a written safety evaluation for Calculation MC5037,
        "DeterminationNalidation of RWST Level Setpoints," Revision 7, dated July 11,
        1995, which changed the f acility as described on page 6.3-4 of the UFSAR is an
        apparent violation of 10 CFR 50.59(b)(1) (50-498; 499/9724 05).
        The licensee recognized the generic implications of this issue and revised their
        calculation program to require impact reviews for calculation revisions. These
        impact reviews will consider whether the f acility as described in the UFSAR was
        impacted by the calculation change.
                                                  15
                                                                                                )
 
.
.
  c.  Conclusions
      Af ter discussions with the team, the licensee identified an example of programmatic
      weakness in that there was no procedural requirement to evaluate calculation
      changes for impact on the UFSAR. As a result, they failed to perform a safety
      evaluation for changes in RWST volumes and uncertainty analysis assumptions.
      This was considered to be an apparent violation of 10 CFR 50.59.
  E8  Miscellaneous Engineering issues
  E8.1 (Clas.cd) Inspection Fol!nwuo item 50-498: 499/9609-01: review of the licenseo's
      self assessment and followup of the actions taken to evaluate the
      apparent discrepancies in the setpoint program,
  a.  Dankaround
      The licensee originally identified weaknesses in their setpoint calculation progiam
      during their engineering assurance assessment in May of 1992, initially they
      planned a corrective action program which involved various contractor assessments.
      One of the first assessments, completed in December of 1993, documented that
      while program improvements were required, immediate operability problems
      probably did not exist. However, later contractor assessments, submitted in April
      and Juno of 1995, identified 35 calculations which required major revision for
      reasons which indicated that the calculations did not meet the licenseo's design
      control program. Af ter prompting by the NRC, the licensee included these
      calculations in their corrective action system in December of 1996; however, they
      still were not identified as conditions adverse to quality. The licensee planned
      additional assessment in 1997.
      The NRC initiall > concluded in NRC Inspection Report 50-498; -499/96-09 that there
      was no regulatory basis for determining that a violation of NRC requirements had
      occurred. However, the NRC initiated an inspection followup item to review the
      results of the licensee's self assessments in more detail and to follow the actions
      taken to evaluate the apparent discrepancies in the setpoint program.
  b.  InsprcioLf.ollo.wup
      The team reviewed the following five engineering self assessment reports to
      evaluate the licensee's program for control of design basis calculations and confirm
      that design and licensing basis requirements were being met:
      *      Engineering Assurance Assessment 92 01, " Instrument Setpoint
              Methodology," dated May 27,1992:
      *      Hurst Consulting, incorporated, Task 1, " Independent Assessment of the
              South Texas Project Setpoint Control Program," dated January 21,1994;
                                                16
                                                                                            a
 
                                                                                            l
.
.
  *      Hurst Consulting, incorporated, Task 2, "STP Setpoint Program Assessment
          (Identification of Existing Setpoint Calculations and the detail of revision
          needed)," dated April 18,1995; and,
  *      Hurst Const i ting, incorporated, Task 3, " Identification of Setpoints with and
          without Calculations," dated June 2,1995.
  *      Engineering Self Assessment of Setpoint Calculations, Methodology and
          Control, Condition Report 96 16020, dated February 21,1P97.
                                                                                            '
  The team found two of the engineering self assessments were performed by the
  licensee and three by a third party engineering firm to evaluate the plant
  instrumentation setpoint calculations, methodologies and programs. These self
  assessments were performed from January 1992 through January 1997.
  Failure to Promptly Correct Deficient Setpoint Analyses identified in 1992
  Engineering Assurance Assessment 'J2 01 was conducted from January 30 through
  April 28,1992, in this assessment, the licensee identified that it was necessary to
  revise several of their channel statistical allowance calculations because the
  instrument uncertainties associated with the Veritrak transmitters were in excess of
  those assumed in the ssfety analysis. This resulted in non-conservative technical
  specifications, which the licensee corrected in Amendments 61 and 50 to License
  Nos. NPF 76 and NPF 80, respectively. The assessment also included the
  observation that there was an error in the RWST recirculation swap over setpoint
  analysis.
  Since several of these items were significant, the assessment included
  programmatic observations as well for a total of 14 observations.
  Observation 92-01 13, classified as Level11 (potential plant safety impact),
  addressed programrnath concerns regarding setpoint calculations. In
  Observation 92-01 13, the licensee documented that s'ome of the technical
  assumptions, contents, scopes, and results of the plant design basis calculations
  were not consistent with the physical design of the p! ant.
  As a result of NRC questioning, during NRC Inspection Report 50-498;-499/96-09,
  the licensee conducted an assessment in January,1997, of the completion of the
  corrective action for the 14 observations. The licensee found that they had not
  implemented all of the corrective actions for Observation 92-01-13. Specifically,
  the design engineering department plan of action to conduct a " review and repair"
  of the design basis setpoint calculations was not implemented. The licensee
  initiated Condition Report 97-1526 to investigate why the Design Engineering
  Department had not implemented this plan of action,
  in Condition Report 97 1526, the licensee noted that the actual " review and repair"
  of instrument setpoint calculations was to be addressed by an approved business
  plan program element, H9 CALC, " Plant Setpoint Program." in accordance with the
                                          17
                                                                                          ]
 
.
,
  Operations Quality Assurance Plan, this response was accepted by quality assurance
  as documented in Station Problem Report 92 0584. However, the licensee
  subsequently decided not to perform a 100% review and repair program, because
  they believed the remaining discrepancies with the calculations were non-safety
  significant. Instead, according to Condition Report 97 1526, engineering
  management elected to utilize the business plan program element funds, to develop
  the compute *ized scaling / calibration data sheet program. This change in work
  priority was supported by operations and raaintenance management. Therefore,
  Condition Report 971526 concluded that the cause was " resources used to
  complete higher priority work."
  The team noted that the business plan item was not viewed as a condition adverse
  to quality in accordance with 10 CFR Part 50, Appendix B, Criterion XVI. As a
  result, licensee management believed that they were only deferring planned program
  enhancements.
  The team found that South Texas Operations Quality Assurance Plan specified
  acceptable design control measures for setpoint analyses and calculations performed
  to confirm compliance with design limits. Chapter 6.0, Section 5.2.3, stated that
  design analyses shall be sufficiently detailed as to purpose, method, assumptions,
  design input, references, units and status (preliminary or final) such that a
  technically qualified person can review and understand the analyses and verify the
  adequacy of the results without recourse to the originator, Section 5.5 stated that
  measures shall be established to control the approval, issuance and changes of
  design documents to prevent the inadvertent use of superseded design information.
  Design documents included setpoints with tolerances and design limits. Section 5.7
  stated that errors and deficiencies found in approved design documents, including
  design methods, that could adversely affect the quality related structures, systems,
  or components shall be documented and action taken to correct and prevent the
  recurrence of deficiencies.
  Chapter 13, " Deficiency Control," which applied to deficiencies discovered in
  activities under the scope of the Operations Quality '.ssurance Plan (including design
  control), stated that procedures shall be developed for the control of activities which
  do not conform to established requirements. These procedures shall provide for the
  identification and documentation of deficient conditions, resolution and/or
  disposition, documentation of the corrective action taken, and actions to be taken to
  assure timely corrective action on deficiencies.
  The team determined that calculations which form the basis for specifying setpoints
  and tolerances or form the basis for confirming compliance with design limits that
  did not meet the criteria of the Operations Quality Assurance Plan Sections 5.2.3,
  and 5.5 were deficient and were, by definition, conditions adverse to quality, which
  required prompt identification and correction.
                                          18
                                                                                          1
 
f
  ,
  .
    10 CFR Part 50, Appendix B, Criterion XVI, includes the requirement to promptly
    identify and correct conditions adverse to quality including deficiencies.
    Considering, that the licensee had already identified that programmatic deficiencies
    of the type described in Observation 92-01 13 could adversely affect quality re%ed
    components, the team determined that cancelin9 the corrective action for
    Observation 92 01 13 was one example an apparent violatic.7 of 10 CFR Part 50,
    Appendix B, Criterion XVI (50 498; 499/9724 06).
    Failure to Promptly Correct Deficient Setpoint Analyses identified in 1995
    As a part of their long term efforts to address the setpoint analysis issue, the license
    hired contractors (Hurst) to perform an assessment of the currant state of their
    setpoint calculations. In 1995, the licensee received information from this
    assessment that 35 calculations required major revisions for recsons that would
    make them deficient with respect to design control standards included in the
    Operations Quality Assurance Plan. As discussed in the 1995 Hurst assessment
    reports, the 35 calculations required major revision because they were deficient due
    to invalid / unverified assumptions, inconsistent format, non retrievable references,
    references with no revision numbers, no methodology specified, no instrument
    uncertainties considered, no calibration or process effects considered and/or
    outdated references. In December of 1996, following questioning by the NRC,
    engineers documented the issues raised in the Hurst assessment reports as a
    condition not adverse to quality in Condition Report 96-16020 because they
    determined that the setpoint calculation discrepancies were not safety significant.
    In addition, ever, af ter the issue was raised in NRC Inspection
    Report 50 498' 499/96 09, the licensee did not perform an operability
    evaluation of all the calculations listed in the 1995 assessment as having major
    problems. Operability was questioned again in May of 1997, during NRC
    Inspection 50 498; 499/9716 and the licenseo performed and operability
    evaluation at that time. As a result, the team determined that the licensee had no
    basis for defet.,ng corrective actions to review existing setpoint calculations in
      1995.
    The team found that the licensee's basis for concluding that the setpoint calculation
      discrepancies were proven to be non safety significant was documented in their
      Letter ST HL AE-5705 to the NRC, dated July 22,1997. Letter ST HL-AE-5705
      was written in response to NRC Inspection Report 50-498; -499/9716 regarding
      setpoint calculations issues. Regarding the safety significance of the setpoint
      calculations, this letter stated that "Although not documented at the time,
      Engincering evaluated the reports and concluded that the operability of safety
      related components was not af fected by these calculation problems."
      During this inspection the team asked the licensee to provide any information related
      to actions the licensee had actually taken to reach the conclusion that operability
      problems did not exist in 1995. Licensee personnel stated that the vendor indicated
      in 199; that no operability issues existed and that licensee personnel believed the
                                              19
                                                                                              -)
 
.
.
        vendor would have indicated operability concerns in the 1995 report, if the vendor
        had operability concerns. The team determined that the licensee's July 22,1997,
        letter was potentially misleading and the reliance by licensee personnel on the
        vendor was not commensurate with an acceptable operability review.
        The team determined that the licensee missed a second opportunity to identify
        design deficiencies within the setpoint calculation program as conditions adverse to
        quality requiring prompt corrective action.10 CFR Part 50, Appendix B,
        Criterion XVI, states that measures shall ae established to assure that conditions
        adverse to quality such as deficiencies are promptly identified and corrected. The
        f ailure to promptly identify the 35 deficient setpoint calculations requiring major
        revision as conditions adverse to quality requiring prompt corrective action is the
        second example of an apparent violation of 10 CFR Part 50, Appendix B,
        Criterion XVI (50-498; 499/9724 06).
        Current Status of Planned Corrective Actions Related to Setpoint Deficiencies
        The team found that on March 19,1997, ir. Condition Report 97-5666 the licensee
        initiated an action to revise affected setpoint calculations to ensure compliance with
        the newly developed programmatic controls. In addition, on August 5,1997, the
        licensee committed to a full upgrade of their setpoint analysis by December 1998.
        The licensee stated that they expect to complete this upgrade in conjunction with
        their ef fort to transition to improved technical specifications,
    c.  Conclusions
        Setpoint calculation deficiencies, which were identified by the licensee in 1992 and
        1995, were not promptly corrected. The licensee f ailed to identify these
        deficiencies as conditions adverse to quality or to take eff ective corrective action
        until prompted by the NRC. These failures were apparent violations of
        10 CFR Part 50, Appendix B, Criterion XVI.
  E8.2 10 cent Unresolved item 50-498: 499/9716-01: Review of the safety impact of the
        licensee's decision to defer the planned setpoint program improvements and review
        of the overall calculation control program,
  a.  Backaround
        The NRC reviewed the adequacy of the licensee's design control program related to
        instrument setpoint criteria. The NRC also reviewed three setpoint calculations and
        identified the following deficiencies:
        *      The low-level alarm for the essential cooling water pond, which is the safety-
                related ultimate heat sink, had been disabl3d by a design changt but the
                calculation was not voided.
                                                  20
                                                                                                ]
 
.
4
    *      The diesel generator fuel oil storage tank level setting calculation did not
            include any uncertainty for changes in fuel specific gravity, differences in
            temperatures, height measurements, instrument loop, and test instruments.
            The NRC was .:oncerned that the alarm setpoint might not ensure that
            operators were alerted to a low-level condition prior to exceeding the
            technical specification limit. The licensee stated that an alternative local
            indicator was actually used to confirm technical specification compliance.
    *      The residual heat removal system low pressure alarm setting calculatio'1
            included a 6 psig margin for instrument error, but the calcuiation did not
            provide any technical basis for this 6 psig margin. Based on the span of the
            detector, the NRC calculated that the actualinstrument uncertainties would
            be greater than the 6 psig established by the calculation for instrument error.
            A licensee representative stated that, due to spurious low pressure alarms,
            they were in the process of revising the calculation and planned to reset the
            alarms. The licensee representative stated that a preliminary calculation
            established an instrument loop uncertainty of 15 psig.
    The NRC noted that even though the instrument setpoints provided the operators
    with important information on the status of safety related equipment, they were not
    setpoints that were required to be established in accordance with Regulatory
    Guide 1.105. The NRC concluded there was a need to conduct further review of
    the safety implications of the management decision to not correct identified design
    control deficiencies,
  b. lDioector Followun
    The team reviewed the licensee's setpoint program for compliance with licensing
    commitments. The team clso reviewed a sampling of instrument setpoint
    documents associated with two general areas: setpoints previously identified by a
    contractor as having no identified basis, and setpoints selected from safety related
    UFSAR described operations.
    Setpoint Criteria
    The licensee's guidance for determining setpoints was contained in
    Procedure SZ120ZO1028, " Design Criteria for Instrument Loop Uncertainty
    and Setpoint Methodology," Revision O.
    The UFSAR, Table 3.121, stated that South Texas conformed to the intent of
    Regulatory Guide 1.105, " Instrument Setpoints for Safety Related Systems."
    Regulatory Guide 1.105, Revision 2, endorsed instrument Society of America
    Standard ISA S67.041982, "Setpoints for Nuclear Safety Relatea instrumentation
    Used in Nuclear Power Plants," for ensuring that instrument setpoints remain within
    technical specification limits. The Instrument Society of America Standard was
    updated in 1994.
                                            21
                                                                                            ._}
 
    .o
    c-
        Based on a sample review, the team considered that Procedure SZ120ZO1028
        implemented the Instrument Society of ' America Standard.
        Setpoints With No Basis
        LA contractor, Hurst Consulting Inc., reviewed a number of safety related instrument    a
        'setpoints and identified a number of setpoints that did not identify a calculation to
        document the setpoint basis. The results of this review were contained in
        Letter ST 5W-HS 090237 of June 2,' 1995. =
                                                _
                                                                                                  ,
                                                                                                  !
        The team selected 20 of these setpoints for followup review and requested the
        licensee to provide any information they currently had related to the setpoint bases.
        The licensee determined that the basis for several of the setpoints was developed in
        site calculations. The licensee determined that some of the setpoints were provided
        by Westinghouse as part of their standard design. The team limited followup to the
        Wastinghouse provided setpoints to reviewing documentation which indicated that
        Westinghouse considered the licensee's setpoints conservative. After reviewing the
        documentation provided by the licensee, the team determined that three of the
          setpoints had a weak calcu'ation basis.
        Ranciot Coolant Pumo Seal inlection Flow
        To ensure adequate reactor coolant pump seat lubrication and cooling is maintained,
          adequate sealinjection flow must be maintained. The licensee determined the
          minimum and maximum reector coolant pump seal injection flow from the reactor
          coolant pump technicalinstruction manual. The team determined that the pump
.        technicalinstruction manual provided a valid basis for acceptable seal flow rates.      l
  '
          However, the team r,uted that the licersee had not accounted for instrument
          uncertainties, when determining the minimum acceptable flow value. The technical
          manual specified a minimum seat injection flow of 6 gpm for each reactor coolant
          pump seal, with a designated maximum limit of 20 gpm. Thr technicalinstruction
          manual stated that the normal operating values were between 8 and 13 gpm per
          pump. Emergency operating procedures specified a setpoint range vvhich included a
          lower minimum value of 6 gpm and a maximum normal value of 13 gpm as an
          acceptable range for seal injection flow. With a lower limit of 6 gpm for indicated
          flow, instrument uncertainties could place the actual pump seal flow outside the
          specifications of the technicalinstruction manual,
                                                                                                  ,
          Volume Control Tank (VCT) Level
          To ensure that suctico to the charging pumps was not lost, upon loss of inventory in.
          the VCT, the licensee had an alarm and automatic swap-ovei of the supply to the
          charging pumps from the VCT_to the RWST at three percent levelin the VCT. This          '
          swap over ensured that the charging pumps always had a source of process fluid
          during normal plant operation and protected them against loss of net positive
          suctiori head and consequent cavitation.
                                                  22
                                                                                                j
 
[
  e.
  .-
    The licensee indicated that the besis for the three percent swap-over was conta;aed
    in Design Change PCF 308010A, dated September 2,1993. l'CF 308010A
    indicated that the setpoint had been one percent, but testing in 1993 indicated that
    had the VCT been drained, no swap-over would have occurred. With the VCT
    drained below the associated level instrument taps, two level indicators read 1.6
    and 3.1 percent, so no swap-over would have occurred. Based on this test, the
    licensee raised the' swap over and alarm setpoints to three percent, however, no
    calculation wat performed to ensure that the three percent setpoint was adequate
    to account for instrument uncertainties.
    The team questioned the adequacy of the three percent setpoint, given that during
    testing one of the two associated ecators read greater than three percent with the
    VCT drained. The team asked f 1 - ior information on the VCT and associated
    swap-over circuits, in order to cac J. ate a uncertainty using the licensee's setpoint
    guldence, Procedure 5Z12Z01028. The licenseo provided the team with the
    requested information, along with a preliminary calculation which indicated that the
    instrument uncertainty associated with the three percent setpoint was 2.7 percent.
    The team revie'wed this calculation and determined that the calculation was
    generally performed using the licensee's setpoint guidance.
    The tean; observed that within the 2.7 percent, the licensee allowed 1.0 percent
    uncertainty for process measurement errors and treated the uncertainty as a
    random, inde. pendent error. Process uncertainty covers non-instrument related
    uncertainty. The licensee stated that the 1.0 percent process uncertainty was a
    conservatism added based on judgement and not on review of actual potential
    process uncertainties.
    The team determined that two process uncertainties existed, changes in fluid
    density due to temperature changes and differences in the elevation between the
    instrument taps and the associated sensors. The team noted that
    Ce:culatica JC5280, " BAT (Boric Acid Tankl Level instrument Uncertainties /
    Setpoirit: " Revision 0, included process uncertainty due to fluid density changes as
    a bias but the revised preliminary VCT calculation did not. The licensee provided the
    team with sensor elevation information which indicated that the sonsors were not
    located at exactly the elevation c' the instrument taps, therefore, the team
    considered that this uncertainty also needed to be included in the revised
    calculation.
    The team considered that the specific uncertainty due to fluid density changes and
    sensor locations needed to be included in the calculation, and were required to be
    considered as biases, not random, independent uncertainties. Because bias errors
    must be directly added to the statistical formula for random, independent
    uncertMnties (normally a square root sum of the squares method) moving the
      1.0 pei,.ent process uncertainty from a random, independent error to a bias would
    increase the instrument uncertainty from 2.7 percent to 3.5 percent.
  '
                                            23
                              - -                                                          )
 
0
.
  In addition, the team observed that capillary tubes were installed but not included in
  the licensee's uncertainty calculation. Procedure 5Z12Z01028 stated that capillary
  tubes may affect the accuracy of the sensors and that the vendor should be
  contacted to provide assistance in determining the effect. The team asked the
  licensee for information relating to the accuracy of Lapillary tubes. The licensee
  stated that there was no information available on site on the capillary tubes, but
  they believed the capillary tubes had no effect on the sensors used for VCT level
  determination.
  The team discussed the VCT level swap over setpoint with the licensee. The
  licensee stated that they believed the actual process uncertainties were less than
  1.0 percent, and that the calculation provided to the team was preliminary, The
  licensee stated that a more formal calculation would be performed as part of actions
  on Condition Report 97 15708. The licensee stated that if the uncertainties were
  determined to be greater than 3.0 percent, a setpoint change would be considered.
  This issue will remain open pending an NRC review of the final calculation to
  confirm the licensee has successfully assured that the automatic VCT/RWST swap-
  over would be reliable and that air-binding of the centrifugal charging pumps would
  not occur.
  Reactor Coolant System Low Pressure Residual Heat Removal Valve InteriorA
  The reactor coolant system low pressure residual heat removal setpoint was a
  permissive interlock, which allowed operations personnel to open residual heat
  removalisolation valves at a decreasing reactor coolant system pressure of
  332 psig. The licensee indicated that the basis for the 332 psig isolation valve open
  permissive interlock setpoint was contained in Design Change ECN 88E344A, dated
  September 11,1988.
  The team found that ECN 88E344A discussed a changa to the permist.ive interlock
  to account for instrument uncertainty. TW design change indicated that
  Westinghouse agreed to a temporary setpoint of 332 psig, but that efforts should be
  instituted to revise the Technical Specifications and return the setpoint to 350 psig.
  The team asked the licensee what actions had been taken to resolve this issue in the
  current and pending improved technical specifications. The licensee reviewed
  - associated records and informed the team that no action had been taken. The team
  asked the licensee if continued operation with a temporary setpoint was acceptable.
  The licensee initiated Condition Report 97-14670 to evaluate why no apparent
  action had been taken to date and whether the pending improved technical
  specifications should be revised. The licensee did not consider this to be an
  operability concern.
                                          24
 
                _ __                        _ _ _ . _      __. .                              __
                                                                                                  '
  .. _
                                                                                                      !
  *                                                                                                j
        The team ob' served that the technical basis for this setpoint. as discussed in the-          I
        licensee's Residual Heat Removal System Design Basis Document, allowed a range                ;
        of pressure which would make a setpoint of 332 technically acceptable, based on
        an informel calculation of instrument uncertainties. Therefore, the team agreed with
        the licensee that there was no immediate operability concern.
        Safety Related Process Values Selected from Final Safety Analysis Report
        The team selected six safety-related process values described in the UFSAR and
        asked the licensee for the calculation basis for these valuer,, to determine the
        adequacy of supporting calculations. The team determined that two of the values
        chosen had an adequate design basis which considered appropriate instrument and              '
        process uncertainties. The remaining four values are discussed below.
        RWST Usable Volume and Level Instrument Setnoints
        The team reviewed uncertainties associated with two RWST requirements: having
        350,000 gallons available to support emergency core cooling and having sufficient
        volume to supply the ECCS pumps, during swap-over of pump suction from the
        RWST to the emergency sump, approximately 11,100 gallons.
        The licensee Provided the basis for both of these volumes in Calculation MC5037,
        "Determinati3n/ Validation of RWST Level Setpoints," Revision 7, dated July 11,
        1995. Calculation MC5037 indicated that maintaining the level above the Technical
        Specification limit of 458,000 gallons would ensure that 350,000 gallons would be
        available for emergency core cooling. The licensee calculated the available volume
        by subtracting the volume remaining in the tank, when swap-over to the
      - containment sump is initiated, from the total available water maintained in
        accordance with the technical specification limit. Calculation MC5037 determined
        the volume available to allow the valves to swap-over by subtracting the water
        below the bottom of the suction piping from water remaining when swap-over is
        initiated, in both of these calculations, the licensee considered instrument
        uncertainty.
        The team reviewed Calculation MC5037 and observed that Revision 7, dated
        July 11,1995, determined that instrument uncertainty was 4.38 percent. The team
        determined that the uncertainty calculation was completely invalid, in that it
        assumed that most sensors uncertainties did not exist, and incorrectly calculated the
        remaining uncertainties. The team discussed Calculation MC5037 with the licensee.
!      The licensee agreed that the instrument uncertainty calculation was incorrect and
        documented this problem on Condition Report 97-14434.
        The licensee performed a preliminary' uncertainty calculation and determined that the
        actual uncertainty was 4.09 percant. The licensee considered that this preliminary
        calculation bounded the incorrect calculation. However, the team reviewed the
                                                      25
L
,
                                                                                        _    ..
                                                                                                  N
 
                                ._
.
.
  preliminary calcu' tion and observed that the licensee had included a random,
  independent 1.0 percent process uncertainty. The team questioned this uncertainty
  for the reasons discussed above for VCT level uncertainties. The licensee stated
  that they planned to perform a more detailed uncertainty calculation as part of
  Condition Report 97 14434.
  Tnis issue will remain open pending an NRC review of the final uncertainty
  calculation to determine whether or not the licensee has adequately accounted for
  level instrument uncertainties associated with maintaining an adequate usable
  volume in the RWST.
  When determining net positive suction head available, the team observed that the
  licensee took credit for the volume of water from the low-low alarm / swap-over poir't
  of 11 percent all the way down to the bottom of the downward facing suction
  piping. The licensee calculated the maximum flow rate to be 22,200 gpm, which
  was also specified in the Updated Final Safety Analysis Report. The team
  considered that, at the flow rates discussed in the Updated Final Safety Analysis
  Report, vortexing would cause air ingestion and loss of net positive suction head to
  ECCS pumps before the bottom of the 24 inch diameter suction piping was reached.
  The downward facing suction intake was located one foot from the bottr.m of the
  tank, with the centerline of the p:pe exiting the tank at the three foot lesel. The
  licensee stated that they did not expect that the RWST level would actuelly drop to
  the bottom of the suction piping. However, they had not calculated the minimum
  expected level. The team calculated that, based on instrument uncertainties, the
  low low alarm / swap-over start point, and 11,100 gallon allowance for swap-over
  valve operations, the required levelin the tank would be approximately 2.75 feet
  above the bottom of the tank, which is below the center line of the suction intake
  pipe.
  The team asked the licensee for a calculation or test which indicated that air would
  not be ingested into the ECCS pump suction piping due to vortexing in the RWST.
  The licensee provided the team with a drawing which showed that a deck grating
  vortex suppressor was installed around the suction inlet inside the RWST Drawings
  for this vortex suppressor indicate that it was approximately four feet,10 inches
  long; six feet wide; and two feet high. The licensee stated that, based on a review
  of criteria provided in Regulatory Guide 1.82, " Water Sources for Long-Term
  Recirculation Cooling Following a Loss-of-Coolant Accident," Revision 2, they
  concluded that the design of the vortex suppressor in the RWST would ensure that
  vortexing would not cause air ingestion above the team's calculated 2.75 feet.
  However, the licensee had no calculation, test or analysis N support this conclusion.
  The team observed that the vortex suppressor in the RWST did not meet the design
  criteria discussed in Regulatory Guicio 1.82 as acceptable to limit air ingestion.
  Subsequent to the onsite inspection the licensee provided an additional informal
  calculation to support their view that the installed configuration was operable and
  that unacceptable pump air ingestion would not occur. The team noted that the
                                          26
 
    - .--            -                  .  .        .    .-          -    -  -.        .  .- - -.
  .-
                                                                                                        I
                                                                                                        l
  .
            minimum expected level of the RWST was not clearly defined in either the
          - calculation or the UFSAR. Therefore it was difficult to accurately assess the
            licensee's application of Regulatory Guide 1.82 test data. The licensee agreed that
            they had not clearly documented their design basis. They planned to revise                  1
            Calculation MC5037, and perform an evaluation to correct and clarify their design
            and licensing basis for the minimum RWST level and maximum expected flow at                i
            swap-over, and any other assumptions necessary_to assess air ingestion potential.
                                                                                                        )
                                                                                                        '
          - This item remains open pending additional NRC review of the finalized calculation,
            the safety evaluation for the change in UFSAR design assumptions, and the
            application of Regulatory Guide 1,82 data to the vortex suppressor configuration.
            This inspection will be performed to confirm the licensee has successfully assured
            that the automatic RWST/ containment sump swap-over would be reliable and that
            air binding of the ECCS pumps would not occur.
            AFST Usable Volume and LeveLinstrument Setnoints
            The team reviewed the instrument Jncertainty calculations associated with the
            AFST and noted that Calculation MC6082, " Misc. AFST Losses," Revision 2,
            determined that instrument uncertainty was 2.87 percent of instrument span, which
            was equivalent to approximately 15,500 gallons. When the team requested the
            uncertainty calculation for review, the licensee provided a new preliminary
            calculation which indicated that instrument uncertainty was equivalent to
            20,500 gallons. This potentially reduced the available design margin for the AFST
            by 5000 gallons. The licensee issued Condition Report 97-15906 to validate the
            preliminny calculation and update the AFST calculation, as necessary. The team
            reviewtJ the licensee's preliminary calculation and considered the calculation
            reasonable, although a detailed validation was not accomplished.
            During discussions with the licensee, the team determined that this potential
            nonconservatism was identified several weeks ago. The item was being tracked as
            an action that was added to Condition Report 97 15807 when a contractor first
            identified the potential nonconservatism. The team noted that the condition report
            was not identified as a condition adverse to quality requiring prompt corrective
            action.
.
            The team also noted that the calculation of record only included a design margin of
            approximately 6500 gallons between the estimateo worst case required volume of
            auxiliary feedwater and the amount of usable auxiliary feedwater maintained by the
            technical specification limit. Similar to the discuseions in Section E7.1, the team
            determined that a potential 5000 gallon reduction in design margin should have been
            viewed as a condition adverse to quality.
                                                    27
                                    _ _.      .                . . . .  ._
 
                                    . _ .  .              ..-      .-
                                                                                        '
  .
  .-
        ' This issue will remain open pending an NRC review of the final uncertainty
          calculation to determine whether or not the licensee has adequately accounted for
          levelinstrument uncertainties associated maintaining an adequate usable volume in
          the AFST.
          To estimate the amount of auxiliary feedwater needed to achieve residual heat-
          removal initiation conditions, the licensee calculated tiie losses expected to occur
        - during system operation. The team reviewed Calculation MC6082 and determined
          that with one exception the calculation conservatively modeled losses associated
          with each case considered. The exception was that the licensee had modeled the
          reserve required to prevent vortex formation without considering all flows. The
          licensee had not included flows associated with postulated pipe breaks and valve
          f ailures. -The licensee agreed and evaluated the discrepancy and determined that
          even af ter including the additional flow the specified vortex reserve was sufficient.
        .They planned to revise their calculation to correctly determine the necessary vortex
          reserve.
          This issue will remain open pending an NRC review of the final loss calculation to
          determine whether or not the licensee has adequately accounted for vortex reserve
          associated maintaining an adequate usable volume in the AFST.
          Dearaded Grid Voltana Relav Setooint                                                    I
          The team reviewed Calculation EC5052, " Degraded and Undervoltage Protection,"
          Revision 3, dated August 7,1997, which determined the adequacy of the degraded
          grid relay trip setpoint to ensure that the grid source would trip before Class 1E
          equipment became inoperable, due tn low voltage. The team observed that
          Revision 3 changed the calculation of instrument uncertainties for +he relay. The
          team considered that this uncertainty calculation was inadequate because it did
          not include any uncertainty for calibration accuracy or follow the licensee's
          dpproved setpoint program guidelines. The licensee agreed that the uncertainty
          calculation within Calculation EC5052 was incorrect. The licensee issued Condition
          Report 97-15658 to correct the calculation. The licensee performed a preliminary
          uncertainty calculation and concluded that the actual uncertainty was less than that
          indicated in Calculation EC5052.- The team reviewed the preliminary calculation and
,          agreed.
          This issue remains open pending further NRC review of the finalized uncertainty
          calculation.
          Comoonent Coolina Water (CCW) Surae Tank Volume
.,
          During review of calculations associated with the component cooling water system
          the team identified that Calculation MC6007, "CCW Surge Tk Vol & Lev,"
            Revision 1, dated January 15,1987, and Calculation ZC6019, "lLUE and
                                              .
            Determination of Trip Setpoint(s) for CCW Surge Tank Level NNS isolation,"
            Revision 1, d' ated March 24,1986, calculated the volume of the surge tank by
.                                                  28
                                                                                                  i
      ._                                                                                          $
 
l
    .
    .
            different methods and indicated different volumes. The licensee indicated both
            calculations were active, but had different purposes. The team discussed the two
            calculations with the licensee. The licensee concluded that Calculation MC6007
            was correct and that Calculation'ZC6019 was wrong by about 400 gallons. The
            licensee issued Condition Report 97-14448 to correct Calculation ZC6019. The
            licensee concluded that there were large margins for setpoints determined in
            Calculation ZC6019 and that the error did not affect tank operability. The team
            reviewed the calculations and agreed with the licensee's conclusions.
      c.    Conclusions
            Based on a sample review, the team concluded that the licensee's setpoint guidance
            procedure was technically adequate.
            Based on a review of twenty setpoints which had been identified as potentially
            having no basis, the team concluded three of the setpoints had a weak calculation
            basis.
            Based on a random sampling of existing safety-related plant process values, the
7
            team concluded that appropriate uncertainty calculatiens had not always been
"
            accomplished through the end of 1996. The calculations for four of six process
            values were incorrect and required revision. The team also determined that the
            analytical basis for preventing air ingestion into the ECCS pumps from the RWST,
            during swap-over to containment sump, was not adequate.
  1
            The unresolved item remains open pending completion of the new followup
              inspections described above.
      E8.3 (Closedi Licensee Event Reoort 50-49 1 499/97-06: Inappropriate =.urveillance
              procedure monitoring parameters,
      a.    Backaround
              On May 7,1997, the licensen identified that the operator log surveillance procedure
              for reactor coolant system riterage temperature did not take into account i'istrument
              measurement uncertainties to ensure that actual reacto; Mr+ system
              temperature was less than the safety analysis limit. The safety analysis simit of
              598*F was developed in support of the Vantage SH fuel upgrade, submitted in
              license amendment application ST-HL AE-4364 dated May 27,1993. This analytical
              limit was incorporated into the Technical Specifications on May 27,1994,
      b.      Insoector Followuo
                                                                                                    {
              The team reviewed the basis for Technical Specification Surveillance
                Requirement 4.2.5.1 and the licensee's calculation for establishing the
                measurement uncertainty associated with wide range reactor coolant system
                average temperature, Calculation ZC7002, " Analysis of uncertainties for the DNB
                                                          29.
                                    _ - - - _ - - - _ - -                                          )
 
          ~ .ex-+, .a =              +      e 4 . . + . . . . 1-mn -
                                                                      +=s - an...              ..m - ,am, y      s .an  -.a..r. u __ - - .. . .  a  .4
      . .
      y
      :.                                      [
                                Related Parameter .- Tavr ." Revision 0, issued on July 7,1987. The team observed
                                that the technical spec wation basis in the license amendment annotated that the'
                                temperature limit was palytical, and that the measured value needs adjustment to
                            '
                                account for measurement uncertainties before comparison with the required limit.
                                The team noted that 10 CFR Part 50, Appendix B, Criterion lil, " Design Control,"
                                                                                                              ~
                            L requires that measures shall be established to assure that apolicable regulatory
:                              requirements are correctly translated into procedures. The team determined that'                                                      ;
                                the licensee had identified an apparent design control violation, which lasted from                                                    l
                                May 27,1994, to May 7,1997, related to translation of measurement uncertainties                                                      .!
                                (50-498; 499/9724-07).                                                                                                                I
                              The team found that Calculation ZC7002, had annotated a 3.5'F uncertainty for
                                average temperature of the reactor coolant system, which when subtracted from the -
                                safety analysis limit of 598 F resulted in an upper indicated value of 594.5'F. The-
                                team noted that the licensee stated in Licensee Event Report 50-498/97-06 that a                                                      ,
                                temporary operationallimit of 593*F was established for e"erage reactor coolant
                              temperature. The team determined this value was conse- itive
*
                              The team reviewed current surveillance procedure, Procedure UPSP03-ZQ-0028,
                                " Operator Logs," Revision 33, dated August 21,1997, associated with setpoint
                              calculations for departure from nucleate boiling parameters to assess the long term
                              corrective actions, associated with Licensee Event Report 50 498/97-06. The team
                              fourvd that the operator logs used to perform the surveillanca for ensuring that
                            -average coolant temperature is within the safety analysis indicated an upper limit'of
                                595'F, which did not aopear to be conservative.
                              The team questioned the licensee's implementation of a less conservative
                              uncertainty in the surveillance procedure compared with the information provided in
                          - the design calculation. Plant personnelinformed the team that vendor information
                              regarding departure from nucleate boiling parameters was provided in
,
                              Letter ST-WN HS-97 0018, dated June 30,1997. The team observed that the
                              uncertainty for average coolant temperature as stated in the June 30,1997, letter
                              was 2.1'F, which implied an upper limit for indicated average temperature of
                          ' 595.9'F, which was_ consistent with the techrncal specification surveillance limit
                              indicated on the operator logs. Plant personnelinformed the team that Condition
                              Report 97-14628 was subsequently issued to void Calculation ZC7002.
                          ~ The team questioned the basis for selecting the less conservative value supplied via
                          = a letter compared with the design calculation information. Plant personnel were
,
                              unable to provide the basis for selecting the less conservative value. After NRC
                              questioning, the licensee requested a copy of the vendor calculation and evaluated
                              the differences between the two calculations. The licensee was able to
                                        _
, .
                          -
                              demonstrate that the information provided in Letter ST-WN-HS-97-0018 was
i'                        . correct.
4
                                                                                  30
              1
1
    A. ,                _                                              _          - . . . _ . _            -  .,  -  -                  -      ig..    , . .~ _
 
        - -. .                              ~    -            .-      . -          ..  -        -      ...
      .
    e
      9
                                                      ~
                                                                  ~
                      The team was concerned that a sclection of uncertainty parameters regarding a -
                      parameter significant for monitoring departure from nucleate boiling was made
                      without performing a technical review or evaluation to. understand the basis of
                    . vendor supplied information that was in conflict with ir. formation contained in a site
                      design calculation.
                    ~ 10 CFR Part 50; Appendix B, Criterion 111, reo"hc;; that design changes, including-
                      field changes, shall be subject to design control measures commensurate with those
                      applied to the original design. South Texas Operations Quality A ssurance Plan,
                      Chapter 6.O! Section 5.5, states that measures shall be established to control the
                      approval, issuance and changes of design documents to prevent the inadvertent use
                      of superseded design information.
                      Section 1.13 of Procedure OEP-3.07Q "Prepaiation of Engineering Calculations,"
                                              ~
                      Revision 4, states, in part. " Design calculations which are no longer required to
                      support design activities are to be voided..."
                      The team determined that use of vendor information to revise a surveillance
-                    procedure without voiding the conflicting site calculation and without understanding
                      the technical basis for the reduction in uncertainty was an apparent design control
                      violation related to use of vendor informat;on and control of superseded documents
                      (50 498;-499/9724-08). .
                c.    Conclusions
                      The licensee identified an apparent design cont.ol violation related to failure to
                      translate instrument measurement uncertainty into surveillance procedures. The
                      NRC subsequently identified and apparent design control violation related to the
                      f ailure to adequately review vendor information and failure to control superseded
                      documents. These were apparent violations of 10 CFR Part 50, Appendix B,
                      Criterion 111.
                                                  V. Management Meetinas
              X1 Exit Meeting Summary
              The team met with the management of South Texas Project Electric Generating Station on
              October 20,1997 to conduct an exit interview. During the exit, the licensee provided
              additionalinformation, including their basis for requesting reconsideration of two proposed
              violations. Following the exit, the team considered the licensee's comments, performed
          - additional research and concluded that the licensee position was correct. The NRC
              reviewed the remaining proposed enforcement actions and determined a predecisional
              escalated enforcement nieeting was warranted. A subsequent exit interview was
          - conducted by telephone on November 12,1997, to discuss the change in characterization
              of the findings.
I                                                              31
i
h -.
I
i
'
          - ,                    _                                          _ . _ . .
 
          ..                      .    .-      . _ . .        .    ._ _ . . -  . . _ _ _ . . _ . . _ . _ . _ _ --    _ --
      : ..
      .,p'
      ,
              --
      1,1
,                      - Durh.g the inspection the_ licensee provided several proprietary documents to.the team. . For
                ' - the most parti these documents'had previously been submitted to the agency and they .
                        '
                                                                                                                            ;
          E-            -fw'ere ac'cepted by the NRC staff as being proprietary documentsi All proprietary
                        - documents were returned to'the licensee. .
                        '
                                                                    .
                          .
                    '
                    .-'
t
7'
                      a
        ,
    .
                                                                            32.
I .,
                                                          ..
                                                                                                  l,
 
                                                  ________ _ _ ___
  m
  *
i
    4
                                                  ATTACHMENT 1
                                        SUPPLEMENTAL INFORMATION
                                    PARTIAL LIST OF PERSONS CONTACTED
4
          Licensee -
        ' C. Albury, Superves;9r, Nucleer Fuel and Analysis
          M. Campbell, Engineer, Plant Auxiliary Section, DED
        1 J. Carbone, SGRP inc. Project Engineer -
          T. Cloninger, Vice President, Nuclear Engineering                                      .
                                                                                                  !
          J. Cottam, Plant Auxiliary Supervisor, DED
        .W. Cottle, Executive Vice President
        : D,' Gore, Supervisor, Nuclear Fuel and Analysis
        : J. Groth, Vice President, Nuclear Generation
          W. Harrison, Senior Consulting Engineer
          S. Head, Sr. Consulting Engineer, Licensing
        - B. Humble, Plant Auxiliary Supervisor, SED
            M. Kanavos, Manager, Mechanical / Civil Engineering
            A. Kent, Manager, Electrical and Instrumentation and Control Systems
            R. Kersey, Engineer, N&SSS Section, DED
            T Koser, Licensing Engineer
            D. Leazar, Manager, Nuclear Fuel and Analysis
            M. McBurnett, Licensing Manager
            C. Pham, Engineer, Balance of Plant Section, DED
            D. Rencurrel, Manager, Electrical /l&C, DED
          : Vi Starks, Design Engineer
            S. Thomes, Manager, Design Engineering Department
            B. Wellborn, Supervisor, l&C Design
          'NBC
          - D. Loveless, Senior Resident inspector
                                          INSPECTION PROCEDURES USED
          37550              Engineering
          37001              =10 CFR 50.59 Safety Evaluations
                                    . ITEMS OPENED. CLOSED. AND DISCUSSED
      ,
                                                                                    1
                                                                    _ _ _ _ _ - _ _  _ _ _ - - _
 
                            .  -    . - ---. - ~                  -  -          .-    -    -
  ,
  .
    ~.
    e
      Ooened
      50-498;-499/9724 01  APV  M&TE accuracy assumptions included in the_ process -
                                  instrument uncertainty calculations were apparently not
                                  translated to the process instrument calibration surveillance
                                  procedures as required by 10 CFR Part 50, Criterion lli
                                  (Section E3.1).
      50-498;-499/9724-02  APV  Excessive uw -f amendments resulted in desicn changes
                                  which were apparently not subject to design ce,atrol
                                  measures commensurate with those applied to the original
                                  design as required by 10 CFR Part 50, Criterion 111
                                  (Section E7.1).
      50-498;-499/9724-03  APV  Calculations were not being updated or evaluated when the
                                  design inputs were changed. As a result the current design
                                  calculations were apparently not being controlled-
                                  commensurate with the original design as required by 10
                                  CFR Part 50, Criterion lli (Section E7.2).
      50-498; 499/9724-04  URI  Two UFSAR Inaccuracies were identified: 1) inaccurate
                                  description of usable AFST volume and 2) inaccurate
                                  description of purge volume (Section E7,3).
      50-498, 499/9724-05  APV  Lack of procedural requirement to evaluate calculation
                                  charges for impact on the UFSAR results in an apparent
                                  f ailure to perform a safety evaluation required by
                                  10 CFR 50.59 (Section 7.4).
      50-498; 499/9724-06  APV  Apparent failure to promptly identify and correct setpoint
                                  calculation deficiencies identified in 1992 and 1995 as
                                  required by 10 CFR Part 50, Appendix B, Criterion XVI
                                  (Section E8.1).
'
      50-498;-499/9724-07  APV  Licensee identified failure to translate instrument
                                  uncertainty for t,y into operations surveillance procedure is
                                  an apparent violation of 10 CFR Part 50, Appendix B,
                                  Criterion 111 (Section E8.3).
      50-498; 499/9724-08' APV    Failure to evaluate a design change related to instrument
                                  uncertainty for t,y and f ailure to void the superseded calc
                                  is an apparent violation of 10 CFR Part 50, Appendix B,-
                                  Criterion ill (Section E8.3).
      Closed
                                                  2
                                                    .
 
                                                                              ..        -  . _.. .
            ,
                                                                                  -
                                                                                                              1
  .A-
.;' 4
  .                                                              .
                                                                                                              l
        50 498;.-499/9609-01          IFl      - Review of licensee's self assessment and followup of
                                                  actions taken to evaluate apparent discrepancies in the-
                                                  setpoint program.-
        ,50-498;-499/97 06
                  _
                                      lLER-      Inappropriate surveillance procedure monitoring
                                                  parameters.
          Discussed
          50-498;-499/9716 01          URI        Review of the safety impact of the licensee's decision to
                                                  defer planned setpoint program improvements.              {
                                            LIST OF ACRONYMS USED
      'AFST        auxiliary feedwater storage tank
                                                                                                            &
        AFW          auxiliary feedwater
        CCW-        component cooling water
        cfm          cubic feet per minute
        CR          condition report
        CVCS        chemical volume and control system
        DBD          design basis document
      -DNB          departure from nucleate boiling
        ECCS        - emergency core cooling system
        EOP-        emergency operating procedures
        gpm          gallons per minute
        HELB-        high energy line break
        HVAC      . heating ventilation and air conditioning
        ILUE        instrument loop uncertainty evaluation
        LOCA      - loss of coolant acciderit
        LOOP        loss of offsite power
        M&TE        measuring and test equipment
        NNS          nonnuclear system
                                                            3
                            .
                                          9
                                              -                                  -- -
                                                                                                , , -
 
        ,                  -    _                      _
                                                                        _ . -    _
                                                                                      ,,            ._
                #
                      . , .                                                                            %
    ;,              a                                                                                  i
    g .-.
      <                                                                                                +
            NPSH:              inet positive suction head
            psig :              pounds per square inch gage:
            OPDS;            7 quality parameter display system
            RCS              . reactor coolant system                                                  ,
            RHR                  residual heat removal system -
          -RSB,                -Reactor Systems Branch
          . RWSTE              .siuehng water storage tank
            PWR-                pressurized water reactor
            SER.                Safety Evaluation Report
            SR-                ' surveillance requirement
            (.STPEGS -            So'uth Texas Project Electric Generating Station
            -TDH                  total developed head
            TS                  technical specification
            -TADOT:-            trip actuating device operational test
              UFSAR                Updated Final Safety Analysis Report
            :USQE                  unreviewed safety question evaluation
              VCT.                volume control tank
                                                          DOCUMENTS REVIEWED
                Unreviewed Safety Question Evaluations-
                Number                          Title                                    Revision
                USOE 95-0001                      UFSAR Section 6.2.6.1 and Table 6.2.6-  Revision 0
                                                  1 (Allow Flexibility in Selection of
                                                  Systems / Penetrations not to be Vented
                                                  During a Type A Test)
              EUSOE 95-0011                    ~ RCS Administrative Cooldown Limit      Revision O
                USOE 95-020-                      This Change will Limit the Accumulator  Revision O
                                                  Water Temperature to Less Than or
                                                  Equal to 90 Degrees Fahrenheit
  ; _.
                    ,
.
                                                                        4
  -
                                                                +-
 
    4
  .
              USOE 95-0028 -        (Modify UFSAR Description of Reactor                    Revision O '
                                    -Vessel Stud Tensioner
            - USOE 96-0004        - Mid Loop Flow lncrease from 1500 -
                                                      .
                                                                                            Revision 0
                                    ' gpm/RHR Tra'n to 3000 gpm/RHR Train
            J USOE 96 0014        - Plant Cooldown with Control Rods                      - Revision O .
                                      Partially Withdrawn
            . USOE 96-0016          Temporary Modification T2 96 4520-4,                    Revision 0
                                    - Defeat of the Electrical Overspeed Trip
                                      for the Main Turbine Generator
              USOE 96-0101          - Control Room Envelope HVAC -                          Revision 0
                                      Emergency Makeup Flow Control
                                      Damper B2HEFCV9585 Temporary
                                      Modification
              USOE 97-0008          UFSAR Change Notice 2142, Maximum                      Revision 0
                                      Boron Concentration in Modes 3,4 and
                                      5
                10 CFR 50.59 Applicability Screening Reviews
                Number                  Title                                                    Re. ision
                MOD 87-030              Replace the Transformers in Class-1E                      Revision O (approved
                                        Channel I and IV Battery Chargers                        03/31/94, but not yet
                                                                                                  implemented)
                DCP 9415610              Four Changes to the OPDS Processirig                    Revision 0
                                        Algorithms and Display leonics (P:,moval
                                        of Automatic Essential Chilled Water
                                        Flow Control to the Essential Chillers,                                        '
                                        Removal of trie Residual Hert Removal
                                      . Flow Interlock, Modification of the Cold
                                        Overpressurization Mitigation Curve, and
                                        Completion of Adverse Containment
                                        Condition Human Engineering
                                        Deficiency)
                DCP 9612225-7-          This Change is to Allow the Use of an                    Revision 0
                                        Alternate Battery (Power Battery
                                        Company P/N PRC636) in Emergency
                                        Lighting Fixture Model M 19
                                                                                  5-
..    . .. .    .  .
                      .                                _ - _ _ _ _ - _ - _ _ _ -    _ _ _  __.              _.      A
 
                          'n                                                            - --- - -              _      __
    .
. .
    *
l
                                                            Clarify Wording of UFSAR Section              Supplement 0
                                            DCP 97-9555 4
                                                            9.5.3.2.3 in Order to Remove Ambiguity
                                                            in Statement Describing Sealed Beam
                                                            Battery Pack Units as Seismically
                                                            Supported Only and Remove Possibility
                                                            of Interpretation as Seismically Qualified
                                                            Clarify References to "1/2T Compact          02/21/97
                                            CN 2134/97-646
                                                            Tension (CT) Fracture Mechanics Test
                                                            Specimens" in UFSAR Sections 5.3.1.6
                                                            and 5.3.2.1
                                              Calculations
                                                                    Title                            Revision
                                              Number
                                                                    Class 1E Standby Diesel          O
                                              EC5092
                                                                    Generator Loading Analysis
                                                                    Voltage Regulation              6
                                              EC5000
                                              EC5002
                                                                    Auxiliary Power System Load 4
                                                                    Study
                                                                      Class 1E Battery, Battery        10
                                              EC5008
                                                                      Charger, and Inverter Sizing
                                                                      Degraded and Undervoltage        3
                                              EC5052
                                                                      Protection
                                                                      Class 1E Battery Duty Cycle      0
                                              EC5068
                                                                      in Station Blackout
                                                                      BAT LevelInstrument              0
                                              JC5280
                                                                      Uncertainties /Setpoints
                                                                      AFW Suction Line Sizing and      2
                                                MC5004
                                                                      Pump Available NPSH
                                                                      Determination / Validation of    7
                                                MC5037
                                                                      RWST Level Setpoints
                                                MC5041                  Design and Operating            1
                                                                        Pressure and Temp of the
                                                                        Auxiliary Feedwater
                                                                        Containment Penetrations
                                                                        AFW Pump Discharge              4
                                                MC5051
                                                                        Pressure
                                                                                      6
      - - _ - - - _ - _ - _ - _ _ _ - _ _ .          -
 
                . - . . .
                          = . - ~ .          -    ~ .      -      .            . - -        . . - .    -          -          . _.
    a
    4
      .
              MC5056                              Auxiliary Feedwater (AFW)-          1                                                >
                                                  Control Valve Sizing -                                                              1
            ' MC5057 -                              Maximum and Minimum-                3 r.:nended .                                    l
                                                    Flow Requirements of                                                                ;
  -
                                                  Auxiliary Feedwater System                                                            i
              MC5060                              Auxiliary Feedwater Line            1
                                                                                                                                        )
                                                    Sizing                                                                              ;
            MC5861                                Auxiliary Feedwater (AFW)            0,1,2,3                                          I
                                                  Pump Design TDH and .
                                                  Flowrate
            MC5864                                AFW Pump Runout Flowrate            2
        .    MC5871                                Verification of AFW 10              1
                                                  Minute Unattended
                                                  Operation
            MC5924                                Auxiliary Feedwater                  1
                                                  Regulating Valves -
                                                  Anticipated Cycles
            MC6007                                CCW Surge Tk Vol & Lev              1
            MC6082                                Misc. AFST Losses                    2
            MC6092                                AFST Volume & Level                  2
                                                  Setpoints
"-          Study Order T-1031 (NEl              Transient Voltage Response          O
            Peebles Electric Products,            of the Diesel-Generator
            Inc)                                  Units, Trains A,- 8, & C to
                                                  Postulated Emergency
                                                  Loading
            ZC6014                                CVCS HELB Mitigation                1
                                                  Temperature Switches
            ZC6019                                ILUE and Determination of            1
                                                  Trip ~Setpoint(s) for CCW
                                                  Surge Tank Level NNS
                                                  isolation
            ZC7002-                                                                                    '
            Analysis of Uncertainties -          0'
            for the DNB related ~
,            parameter'- Tavg
                                                                      7
<-
t
          !
                          - , - ,    , , , - .  -
                                                              --, .
                                                                        - - . . -                      -    . , . . -
                                                                                                                        .-- - . -
 
,
        4 --
      3
      :.
                iRHR Closure Alarm-        -O
                ' Setpoint_- Surtion Valves
                  Drawings
                  Number ~                    Title -                      Revision
                . D 781247_ -                Stainless Steel Liners For =15
                                              Auxiliary Feedwater -
                                              Storage Tanks
                ;35520                        Auxiliary Feedwater Pump  ; superseded ,
                                              Curve
                  46306=                      Auxiliary Feedwater Pump    6/22/95 ~
                                              Curve
                  3G69PAF602-                Auxiliary Feedwater "AF"-    Sheet 15/R6-
                  SS141FOOO24                Piping and Instrument        2                ,
                                              Diagram Auxiliary
                                              Feedwater
                  SS199FOOO2O                Piping and Instrument        26                ,
                                              Diagram
                                              Condensate Storage
                  5Z121Z50001 Sh 8            Delta T and T average      O
                  SZ121Z50001 Sh 9            RCS Hot Leg Temp NR        O
                ,5R 17 9-Z-42410              Letdown isolation Valves    8
                                              Logic
    ,
                  9-E DJAA-01 #1              125VDC Class 1E              13
                                              Distribution SWBD E1 A11
                  9 E DJAB-01 #1              125VDC Class 1F              14
                                              Distribution SWBD E1D11
                  9-E-DJAC 01 #_1            125VDC Class 1E              13
                                              D stribution SWBD E1811-
r                9 E DJAD 01 #1              125VDC Class 1E              11
                                            . Distribution SWBD E1C11
                ' Modification Packages
                  Number                    . Title                      Revision
  -
  -                                                        8
            ;:
                                                    -          - - .  .    - . - . .  - --
 
r
  " $'      ,
f  .
        E C 6 2-:                      Electrical Calc.                0-
        EC 32 :                      - Calculation / Load Study      0
        EC 49:        -
                                      - Electrical Calc./ Load Study ' O-
      .PCF 308010A                  -Change the VCT to RWST          O
                                      Switch Over and Low Low-
                                      Alarm Set Point from 1% to.
                                        3% VCT Level-
        PCF 176712 -                  Add Lighting at the            O
                                      Chemical Feed Skid -
        PCF 211205                    Add Lighting Around the        O-
                                      . Steam Generator Startup .
                                      Feed Pump
        ECN 88 L-OO10G                . Provide Permanent Power        0-
                                      to Trailer
      - ECN 88E344A                  Change Unit 1 Spray            0
                                      Additive Tank Alarms                    -
        MDCN 9003704                  Provide Enclosure for Tech      O
                                      Support Center Diesel
                                      Generator - Unit 1
                                                                                  !
        Procedures and instructions -
      - Number                        Title                          Revision
        Procedure OPGPO5 ZA-          10 CFR 50.59 Evaluations      6
        0002
        Procedure OPGPO3-ZE-          Procedure Preparation-        12-
      :0005
        Procedure OPSPOS-MS-          Main Steam Pressure Loop      1
        0514LL                        Calibration
        Procedure OPSPOS RC-
                  '
                                        RCS Flow Transmitter-          0
        0417                          Calibration
                                                                                  i
                                                                                  '
                                                      9
 
(::
    -o
      %
      .-
          Procedure OPMPOS-ZE-    Calibration of ITE-27 Relays 3
          0034
          Procedure OPSP06-PK-    4.16KV Class 1E Degraded    4
          0005                    Voltage Relay Channel
                                  Calibration /TADOT-Channel
                                  1
          Procedure 5Z120ZO1028    Design Criteria for          O
                                  Instrument Loop
                                  Uncertainty and Setpoint
                                  Methodology
          Procedure OPSP03-SI-0020 Safety injection System      2
                                  Miscellaneous and Train
                                  1 A(2A) Valve Operability
                                  Test
          Procedure OEP 3.07Q      Preparation of Engineering  4
                                  Calculations
          Procedure OPSP03 ZQ-    Operator Logs                32 & 33
          0028
          Procedure EOPT-03.25    STPEGS EOP Technical        5
                                  Guidelines
          Procedure 52529ZB01024  DBD EOP Setpoints            O
          Procedure 5Z529Z901003  EOP Sctpoint Document        1
          Procedure 3ZO10ZO1027    Design Criteria for          2
                                  Instrument Scaling
                                  Methodology
          Procedure OPGPO4-ZA-    Piant Instrumentation        1
          0011                    Scaling Program
          Procedure OPMP08-Zl-    Generic Temperature          12
f.        0011                    Switch Calibration
          Procedure OPGPO3-ZM-    Installed Plant              13
          0016                    instrumentation Calibration
                                  Verification Program
          Procedure OPSP06-DJ-      125 Volt Class 1E Battery  2
          0004                    Service Surveillance Test
          Procedure OPSP06 DJ-      125 Volt Class 1E Battery  2
          0002                    Quarterly Surveillance Test
                                                10                                        ,
                                                                        . - - - _ _ _ - _ .
 
            _ _ _ _
    o
i
    ,
                    Procedure OPSP06-0J-                Battery Charger 8 Hour        5
                    0006                                Load Verification
                      Condition Reports (CR) and Nonconformance Evaluations
                      Number                              Title                        initiation Date
                      CR 9714434                          Comparison of Revision 7      09/16/97
                                                          of the RWST Determination
                                                          of Level Setpoints
                                                          Calculation to the UFSAR
                    -CR 97-15906                        Provide an Instrument          10/02/97
                                                          Uncertainty Calculation in
                                                          Support of AFST Level
                                                          Instrument Setpoints
                    'CR 9714448                          Resolve Discrepancy          09/16/97
                                                          Between Calculation
                                                          MC6007 and ZC6019
                      CR 97-14670                        investigate the Possible      09/18/97
                                                          Revision of TS SR 4.5.6.2A
                      CR 97-14608                          Evaluate impact of Large      09/24/97
                                                          Number of Amendments to
                                                          Electrical Calculation
                                                          EC5002
                      CR 97-15708                          Calculate of VCT
                                                          Instrument Uncertainties
                      CR 97 238 Action 104                Ensure that M&TE
                                                          Requirements are
                                                          Accounted for in
                                                          Procedures
                      CR 97-15658                        Correct instrument
                                                          Uncertainties in Calculation
                                                          EC5052
                      CR 9510936                          Potential for Simultaneous    09/21/95
                                                            Start of Containment Spray
                                                            Pump and Other EAB
                                                            Chillers During a LOOP with
                                                            LOCA
                                                                          11
                                                                                                        I
  ..  .. .-
                        _ _  _ _ _ _ _ _ _ _ _ _ _ - _ _
 
[=                                                                      <
                                                                                      j
    .
                                    ''
    j-
    -o              -
                                                                                        ;
                    CR 97714628l              Revise , Void, or Retire DNB 09/18/97
                                                                .
                                                                                        !
              .                            parameter calculation
                    CR 97 8349              incorrect limits for DNB -    05/07/97-
                                            parameters used in
                                            surveillance T/S '4,2.5.1
  -
                      \
                  - Miscellaneous
                    Number-                . Title                        - Revision
                    Design Basis Document    Auxiliary Feedwater System    Revision 2
                  - 5S149MB1016 :
                    Design Basis Document    Reactor Coolant System        Revision 2
                  - SR149MB1027
                    Design Basis Document -  Residual Heat Removal          Revision 3
                    SR169MB1021              System
                    Design Basis Document    Class 1E 125V DC System        O          <
                    4E529EB1111
                    License Amendment 63/52
                    License Amendment 73/62
                                            Technical Specifications
                                            Updated Final Safety          Revision 5
                                            Analysis Report
          u-
i
            e
                $
                                                                                        V
        D
                        .,
                                                          12
          Ty
      1
                                                              J
                                        --
}}

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