ML20132D686: Difference between revisions

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{{Adams
#REDIRECT [[IR 05000361/1996017]]
| number = ML20132D686
| issue date = 12/16/1996
| title = Insp Repts 50-361/96-17 & 50-362/96-17 on 961118-22. Violations Noted.Major Areas Inspected:Engineering
| author name =
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
| addressee name =
| addressee affiliation =
| docket = 05000361, 05000362
| license number =
| contact person =
| document report number = 50-361-96-17, 50-362-96-17, NUDOCS 9612200069
| package number = ML20132D671
| document type = INSPECTION REPORT, NRC-GENERATED, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 17
}}
See also: [[see also::IR 05000361/1996017]]
 
=Text=
{{#Wiki_filter:. .. . . . _ . . .__ -          .  . _ . . . _    _ _ . . . _ . _ . .    ___ - _ . . . . _ _ _ _
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  ~
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                                                        ENCLOSURE
                                                                                                          '
l                                    U.S. NUCLEAR REGULATORY COMMISSION
                                                        REGION IV
                                                                                                          .
          Docket Nos.:            50 361                                                                l
                                  50-362                                                                i
            License Nos.:          NPF-10
                                  NPF-15
          Report No.:            50-361/96-17                                                          >
                                  50-362/96-17
          Licensee:              Southern California Edison Co.
l          Facility:              San Onofre Nuclear Generating Station, Units 2 and 3
          Location:              5000 S. Pacific Coast Hwy.                                            i
                                  San Clemente, California
                                                                                                          ,
            Dates:                November 18-22,1996                                                    >
            Inspectors:            M. F. Runyan, Reactor Inspector
                                  M. B. Fields, Project Manager
                                  J. J. Russell, Resident inspector
                                                                                                          '
            Approved By:          C. A. VanDenburgh, Chief, Engineering Branch
                                  Division of Reactor Safety                                              ;
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                                                                                                          l
            Attachment:          Supplemental Information
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(
                                          '
    9612200069 961216
    PDR      ADOCK 05000361
    0                          PDR
                                                                      . - . _ _ .
                                      _  _
 
                                                                                                                                    1
                                                                                                                                    !
.
.                                                                                                                                    1
                                                              2-
                                              TABLE OF CONTENTS
  EX E C UTI V E S U M M A R Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
  R e p o rt D e t a il s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
  111. E ng i n e e ri n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
        El    Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
                E1.1    Reactor Head Vent Line Loss of Coolant Accident Limiter
                        Orifice Valve S31201MU995 Normally Locked Closed, Found                                                    ,
                                                                                                                                      '
                        Open - Unit 3 .....................................                                                      5
                                                                                                                                    I
        E.8    Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10                          1
                E8.1 (Open) Inspection Followup Item 50-361:362/9526-02:                                                            i
                        Failure to Perform 10 CFR 50.59 Evaluation . . . . . . . . . . . . . . . 10                                l
                E8.2 (Closed) Licensee Event Report 95-04: Inoperable Fire Control
                        System......................................... 12
                E8.3 (Closed) Inspecticn Followup Item 50-361:362/9526-03:
                        Licensee Event Report 60-361:362/95-16): Heating,
                        Ventilation, and Air Conditioning /High Energy Line Break
                        I nt e r a ctio ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
  V. M a n a g e m e nt M e e ting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
                                                                                                                                    l
                                                                                                                                    l
        X1    Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . .                    ..........              15 1
                                                                                                                                    1
                                                                                                                                      1
                                                                                                                                      ,
                                                                                                                                    I
                                                                                                                                    I
                                                                                                                                    .
 
      -  .  -        .  . .. --            -.-        - - - . _ - -      _ . .    -      . - . - -
    .                                                                                                  ,
  .
                                                                                                      ,
I
    *
                                                                                                      ;
                                                                                                      ,
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                                                        3-                                            '
l                                                                                                      1
.
(                                            EXECUTIVE SUMMARY
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l                            San Onofre Nuclear Generating Station, Units 2 and 3
l                              NRC Inspection Report 50-361/96-17:50-362/96-17
i
        Enaineerina
        *      The failure to properly align orifice gate valve 3MU995 in the reactor coolant gas
              vent system prior to Unit 3 refueling outage startup in September 1995 was              ,
              identified as a violation of Technical Specification 6.8.1 (Section E1.1).
                                                                                                        l
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        *
              During the period of time that Unit 3 was operated with valve 3MU995 improperly
              open (September 1995 to September 1996), the probability of a small break loss of
              coolant accident was slightly increased. However, the reactor coolant gas vent          i
              system remained capable of performing its design function. Consequently, the
              safety significance of the mispositioning was low (Section E1.1).
        *      The mispositioning of valve 3MU995 revealed a weakness in the licensee's locked
              valve program, where a single procedural error resulted in the failure to properly
              align a valve important to safety. The licensee had initiated corrective actions that
              should correct this weakness (Section E1.1).
        *      Upon discovery of the mispositioning of valve 3MU995, the licensee did not
l              perform a comprehensive review of documented valve lineups to ensure that no
              other similar valve mispositionings had occurred. The licensee performed this
              review at the inspectors' request. The inspectors considered the licensee's initial
              response, which included only a check of one corresponding valve lineup in Unit 2,
              to be weak (Section E1.1).
        *      The inspectors considered Licensee Controlled Specification 3.4.102, which
              addresses component configuration, test, and surveillance requirements for the
              reactor coolant gas vent system, to be weak in not addressing the position
              sensitivity of valve 3MU995 (and 2MU995)(Section E1.1).
        *      The licensee missed an opportunity to earlier identify the mispositioning of
              valve 3MU995 from valve lineup information that was available during the entire          ,
;              year the valve was mispositioned (Section E1 1).                                        l
        *      A previous violation concerning valve 3MU995 was reviewed but left open because
              the licensee had not adequately updated the Updated Final Safety Analysis Report
              to describe the actual configuration of the reactor coolant gas vent system and had
              not addressed a weakness (identified by the NRC at the time the violation was
,              issued) in the 10 CFR 50.59 screening process that appeared to be a contributing
l              cause to the violation (Section E8.1).
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                                                                                                    .
 
                                                                                            1
  *
                                                                                            l
                                                                                            l
  *
l                                                                                          1
1                                            -4-
                                                                                            i
!                                                                                          I
                                                                                            1
    * Licensee efforts to identify and correct a high energy line break interaction through ,
      ventilation ductwork to other spaces housing environmentally sensitive equipment      1
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      was indicative of strong, proactive engineering (Section E8.3).
                                                                                            1
                                                                                            l
                                                                                            l
                                                                                            l
                                                                                            l
                                                                                            l
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                                                                                            I
 
  ,__  _ -                -                  ~~    - - - . _ _ _ _    -_    . _ .      -. -        _. . -
      .
      .
                                                                    -5-
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i                                                      Report Details
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            Summarv of Plant Status
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            Unit 2 and Unit 3 were operated at 100 power during the inspection.                              ;
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                                                      ll1. Enaineerina
            E1    Conduct of Engineering
                                                                                                              ,
            E1.1  Reactor Head Vent Line Loss of Coolant Accident Limiter Orifice
                  Valve S31201MU995 Normally Locked Closed. Found Ooen - Unit 3(37550)
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            a.    Insoection Scope
                  On September 27,1996, auring preparations for draining the Unit 3 reactor coolant
                  system to midloop, the licensee discovered that valve S31201MU995(3MU995)
                  was open. This valve, a manual orifice gate valve located in the reactor coolant
                  vent gas system, is designed to limit the flow rate during reactor head venting
                  operations to less than the capacity of a single charging pump. An orifice is drilled
                  into the seat of the valve, such that when the valve is closed, a flow diameter of
                  0.18 inches is provided. When the valve is open, a larger 0.62-inch diameter
                  opening is provided. This valve was required by procedure to be locked closed
                  during Modes 1 through 4. The valve had been open since the last refueling outage
                  in September 1995. Thus, the plant was operated with valve 3MU995 out of its
                  normal position for approximately 1 year.
                  The inspectors reviewed the circumstances involved in the mispositioning of
                  valve MU995 and the safety significance of the valve being left open for 1 year of          ;
                                                                                                              '
                  reactor operations.
                                                                                                              l
                  in addition, the inspectors reviewed a previous violation concerning valve 3MU995
                    (361:362/9526-02). This violation identified the licensee's failure to perform a
                  safety evaluation in accordance with 10 CFR 50.59 or to update the final safety
                    analysis report during installation of a field change that replaced an orifice plate
                    originally installed in the gas vent system with valve 3MU995 (an identical change        i
                    was made to Unit 2).
t
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                                                                                    . - .
 
                              _ -                -              .    .-        .  _
  .
  .
A
                                                6-
    b. Observations and Findinas
        Failure to Follow Procedire
.
        The inspectors interviewed operators and operations' supervision, reviewed copies
        of Procedure SO23-31.4, Attachment 3, Temporary Change Notices 15-1 and
        15-2, "RCS Post Fill Valve Alignment;" NRC Inspection Report 50-361;362/95-26
,      with licensee responses; Updated Final Safety Analysis Report Section 9.3.7;
        Licensee Controlled Specification 3.4.102;and NUREG 0737, " Clarification of TMI
        Action Plan Requirements." The inspectors also reviewed Procedure SO23-0-17,
        Temporary Change Notice 10-41, " Locking of Safety-Related Valves and Breakers,"
        and reactor coolant system Piping and instrumentation Drawing 40111. The
        inspector al.so reviewed a briefing paper submitted by the licensee on November 18,
        1996, and Licensee Event Report 50-362/96-005," Reactor Head Vent Valve
        Mispositioned."
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4
        During the Unit 3 Cycle 8 refueling outage in September 1995, following core
,      reload, the reactor coolant system had been filled solid, then drained to
        approximately 50 percent pressurizer level for integrated leak rate testing and
        integrated engineered safety features testing. At that time, the licensee entered
i      reactor coolant system fill Procedure SO23-3-1.4, Temporary Change Notice 15-2,
        Procedure Modification Permit 1, " Filling and Venting the RCS." Attachment 3 of
        this procedure included a check that valve 3MU995 was locked closed. During the
        fill and vent, this valve is normally opened to expedite the evolution, then closed
        upon completion of the evolution. However, because vent rigs were to remain          i
        attached to the reactor coolant system, the procedure was modified to temporarily
        defer Attachment 3 and, thus, to permit, among other changes, valve 3MU995 to
        remain open. The procedure modification permit process was used, which included
4      final approval by the operations manager. The attachment, if performed, checked
        reactor coolant system vents closed. The reactor coolant system was then taken
        solid a second time and a bubble was drawn. The second time the fill procedure
        was used a different senior reactor operator initiated Attachment 3, but wrote
        "N/A" for various valves in the attachment that he thought would have been
        checked by the previous performance of the attachment. Those checks had been
        deferred as explained above, and had not been performed.
        During an outage in September 1996, valve 3MU995 was found in an open position
        during routine work. The licensee immediately verified that Attachment 3 had been
        performed for Unit 2, and, thus, had confidence that valve 2MU995 (same valve in
<
        Unit 2) was in the required closed position. The operations manager stated that the
        only valve found out of position was valve 3MU995, as the other vent valves in the
        attachment had been positioned correctly when the vent rigs were removed in
        accordance with the procedure for removing the individual vent rigs.
 
  c                                                                                      I
                                                                                          :
                                                                                          I
                                                                                          ,
  '
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                                                                                          ,
                                              -7-                                          l
                                                                                          1
                                                                                          !
                                                                                          l
    The alignment performance guidelines for Procedure SO23-31.4, Attachment 3,
                                                                                          '
    Temporary Change Notice 15-2, "RCS Post Fill Valve Alignment," allowed for            l
    unbracketed steps, including the check of valve 3MU995, to be omitted if
'
                                                                                          '
    verification was made that they had been performed during a previous fill. The
'
    inspectors interviewed the senior reactor operator who directed that the              1
    unbracketed steps not be performed. He stated that he could not remember what        i
    he had checked for verification. The inspector considered this a violation of the    1
    alignment performance guidelines of this procedure, since the procedure directed      l
    bracketed steps to be performed, unless they were verified as having been              I
    previously completed. Specifically, the alignment performance guidelines listed
    below Step 1.3, Attachment 3, Temporary Change Notice 15-2, stated, "All
    unbracketed steps will be performed now . . . unbracketed steps may be marked
    N/A after verifying completion during the previous fill." The senior reactor operator
    had marked "N/A" but had not verified completion of this step during the previous
'
    fill.
                                                                                          l
    Technical Specification 6.8.1, in effect in September 1995, required that written    l
    procedures be established, implemented, and maintained covering activities
<
    recommended in Regulatory Guide 1.33, Reudon 2, Appendix A.
    Procedure SO 23-3-1.4 was included in the supe of this regulatory guide. The
    failure to correctly follow Procedure SO23-31.4 was identified as a violation
    (violation 50-361:362/9617-01). This violation, though generally meeting the
    criteria for a noncited violation, has been cited for the following two reasons:
    (1)      The licensee missed an opportunity to earlier identify the mispositioning of
              valve 3MU995. Valve lineup information was available indicating the
              mispositioned status of this valve for the entire year it was mispositioned.
'
    (2)      The licensee did not perform an immediate generic review of locked valve
              program valve lineups to ascertain whether other valves may have been
              similarly mispositioned (only the corresponding Unit 2 valve was checked).
              This is discussed in greater depth under " Programmatic Weakness" below.
    Safety Conseauences
!
    The licensee had installed valve 3MU995, an orifice disc gate valve, during the
,
    Cycle 8 refueling outage in September 1995. Previously an orifice plate had been in
    this position. The orifice plate and the closed orifice gate valve both limited the
    vent flow rate to the capacity of a single charging pump. With the orifice gate
      valve open, the flow rate would have exceeded the entire normal charging capacity.
      As a consequence of this fact, operating Unit 3 with valve 3MU995 open
      effectively added approximately 171 feet of ASME Class 11 piping to the reactor
      coolant system pressure boundary. In accordance with the Updated Final Safety
i    Analysis Report, Section 3.2 (the Q-List), the reactor coolant system was required
      to be Class I piping. The piping downstream of valve 3MU995 was rated for
      reactor coolant system pressure and temperature, and was normally pressurized to
 
  .
i
  .
l                                          -8-
    2250 psia because of the presence of the orifice. The valve, when closed, had
l  provided a code break between these two classes of piping, because it was sized to
    limit flow to the capacity of one charging pump. Consequently, a break
    downstream of the valve would have resulted in a leak and not a loss-of-coolant
    accident, regardless of the size of the break. With the valve open, the 0.18
    square-inch orifice diameter was replaced with a O.62 square-inch opening. At
    2250 psia, a 0.62 square-inch opening would have allowed flow greater than the
    capacity of three charging pumps, but still within the total makeup capacity of the
    emergency core cooling system pumps. The resulting small break loss-of-coolant
    accident would have been bounded by the existing small break analysis.
    The head vent system was constructed to Section lll of the 1974 ASME code, with
    Schedule 160 376 stainless steel. The licensee informed the inspectors that the
    only difference for Class I and ll system piping less than 1 inch, in terms of
    allowable materials, fabrication, design, construction, and testing, as the 1974
    code, was that the welds for Class I piping were both radiographed and
    dye-penetrant tested as postweld-required nondestructive examination, while the
    welds for Class li systems were only radiographed. The inspectors found that
    because approximately 171 feet of piping had been added to the reactor coolant
    system pressure boundary, and because this piping was not as completely
    examined as the Class I piping already in place, the probability of a small break
    loss-of-coolant accident had increased. However, the inspectors also found that,
    while certain surf ace defects may only be detected by surface examination
    depending on their orientation to the radiographic beam, radiography still provided a  i
    reasonable assurance that significant defects in the weld would have been detected.    !
    The inspectors determined that the increase in the probability of a small break      )
    loss-of-coolant accident was low, because the additional piping was constructed
    and designed to Class I standards and was normally pressurized to reactor coolant
    system pressure.
    The inspectors also determined that the reactor coolant gas vent system remained
    capable of performing its design function, despite the increased flow rate that would
    have been present without the flow restricting orifice in service. This observation
    was based on conversations with the licensee and a review of a hydraulic analysis    ,
    performed by the licensee at the inspectors' request. The inspectors determined      !
    that the solenoid-operated valves, located downstream of the orificed valve and
    providing isolation from the pressurizer relief tank and containment atmosphere,
    were capable of opening and closing against the postulated differential pressures
    and flows. Based on information provided by the licensee, water hammer and
    stresses created by the additional flow rate were within acceptable tolerances for
    the system. The operators may have lost pressure control when the system was
    utilized, due to the pressure drop from the increased flow, but control would
    probably have been quickly recovered following the expected level recovery
    provided by the automatic response of the emergency core cooling system. By
    procedure, the operator response would have been to stop venting and regain
    subcooling in the reactor coolant system. Natural circulation may have been
 
                                                      ..
  .
  .
                                            9
.
    momentarily lost, but would have been reestablished when venting was stopped.
    Based on the above facts, the inspectors determined that the reactor coolant gas
    vent system remained operable with valve 3MU995 in the open position.
    Based on the above, the inspectors concluded that the safety consequence of
.  operating Unit 3 with valve 3MU995 open was low. The probability of a small
3
    break loss-of-coolant accident increased slightly but the consequences were
    bounded by the current accident analysis. TM reactor coolant gas vent system was
    still capable of performing its design function.                                      j
    Proarammatic Weakness
.
    The licensee's corrective actions taken and planned to prevent recurrence of this
    incident consisted of immediately verifying that the Unit 2 valve 2MU995 was
    closed, generating a controlling document to stipulate procedure flow during outage    j
'
    recovery (reviewing integrated procedures and identifying where important decisions    l
    are made and should have management review), having a final systems alignment          1
    checklist of important valves prior to containment closecut (which will include        )
    valve MU995), reviewing the event with all operators, and appropriate disciplinary
    action.                                                                                ,
                                                                                            1
    In contrast to the licensee's assertion that multiple errors caused the mispositioning  i
    of valve 3MU995, the inspectors found that a single operator error caused this          l
    event. The error was the failure to follow the procedure cited above and discussed
    in the Notice of Violation. The inspectors noted that several other barriers had
    failed, but did not consider them Os programmatically required. The operations
    manager stated that it was his expectation that the operators who used the
'
    procedure modification permit process (leaving valve 3MU995 open) for the original
l  performance of the fill procedure should have communicated this change better,
    that the shift turnover process should have communicated this change, and that the
    reviewer of the procedure performance in which the valve was not closed should
    have determined that the "N/A" was inappropriate. While acknowledging that one
    or more of these could have prevented the mispositioning, the inspectors found that
    these expectations were not written into the procedure and, hence, did not
    represent clear programmatic requi ements. Since the mispositioning had only one
    clear procedural barrier, the inspector questioned how the licensee was assured of
,
    the correct positioning of the remaining locked, safety-related valves in both units.
    The operations manager stated that a locked valve had not been found
    mispositioned in approximately 10 years. The inspectors acknowledged this, but
    also noted that no effort had been made to verify that similar decision points had
    been made correctly, accomplished by checking attachments performed at the end
    of the most recent outages for both units. The licensee acknowledged and
    responded to this concern. At the end of the inspection period, the licensee was in
    the process of verifying correct performance of the Unit 3 procedures performed
    during recovery from the most recent Unit 3 outage. Unit 2 was scheduled to
    enter a refueling outage approximately one week following the exit meeting, during
 
.
                                              -10-
      which the position of locked valves would be physically verified. The licensee
      prioritized the review of attachments on Unit 3 and did not have time before outage
      entry to perform a similar review on Unit 2.
                                                                                            l
      The inspectors determined that the corrective ections proposed by the licensee, as
      discussed above, would most likely prevent recurrence of this event. These actions    J
      would appear to correct the weakness in the locked valve program that permitted
      valve 3MU995 to be mispositioned as a result of a single procedural error.            j
                                                                                -
                                                                                            l
  c.  Conclusion                                                                            i
                                                                                            l
      A violation was issued for failing to follow procedure. The safety consequence of
      leaving valve 3MU995 open was low, because the probability of a small break
      loss-of-coolant accident was only slightly increased, and because the reactor
      coolant gas vent system remained operable. However, the licensee's corrective
      actions were not thorough in that it did not immediately begin to verify other
      safety-related locked valves (other than the identical valve in Unit 2) were in the
      correct position, even though, in this instance, only one clear procedural error
      precipitated the valve mispositioning. (The inspectors recognized that the licensee
      continued to characterize the event as a multiple-error occurrence thereby,
      explaining, somewhat, the lack of more decisive action). The inspectors also noted
      that the licensee missed an opportunity to earlier identify the mispositioning event.
      in light of these concerns, the NRC determined that the violation, despite generally
      meeting the criteria for a noncited violation, should be cited.
  E.8  Miscellaneous Engineering issues (92903)
  E8.1 (Open) Insoection Followuo item 50-361:362/9526-02: Failure to Perform
        10 CFR 50.59 Evaluation
      Backaround
      The licensee replaced the reactor coolant gas vent system flow limiting orifice,
      described and depicted in the Updated Final Safety Analysis Report with an orifice
        gate valve (3MU995) without performing a 10 CFR 50.59 safety evaluation and
        without changing the Updated Final Safety Analysis Report. These two omissions
        resulted in a Severity Level IV violation. NRC inspection Report 50-361;362/95-26
        also identified a weakness in Procedure SO123-XXIV-10.21," Field Change Notice
        and Field Interim Design Change Notice," Revision 5, regarding the lack of a written
        basis justifying whether a plant change required a 10 CFR 50.59 safety evaluation.
        The licensee responded to this violation by letter dated February 20,1996, which
        provided the results of the 10 CFR 50.59 safety evaluation, and a commitment to
        update Updated Final Safety Analysis Report Figure 9.3-15 (reactor coolant gas
        vent system).
 
  *
i
l
,
  e
                                            -11-
l
l
    Followuo
    The inspectors reviewed the 10 CFR 50.59 evaluation and Updated Final Safety
l  Analysis Report change for completeness and accuracy. In addition, the inspectors    ,
    reviewed Procedure SO123-XXIV-10.21 to determine if the weakness identified in
    NRC Inspection Report 50-361:362/95-26(discussed in the preceding paragraph)
    had been corrected.
                                                                                          t
    The inspectors concluded that the 10 CFR 50.59 safety evaluation addressing the
    modification to the reactor coolant gas vent system was acceptable. The safety
i  evaluation properly identified the systems and components affected by the change,
    the paremeters of the accident analysis affected by the change, and the potential
    effects of system or component failure. The evaluation contained acceptable
    responses to each of the seven Nuclear Safety Analysis Center-125 questions
    designed to identify an unreviewed safety question, as defined by
    10 CFR 50.59(a)(2). Sufficient detail was provided in the safety evaluation to allow
    an independent reviewer to conclude that the change did not result in an
    unreviewed safety question. The inspection team observed that the safety
    evaluation could have been stronger with regard to the administrative processes for
    controlling the position of the orifice valve. Because of the importance of the
    position of this valve, it would have been prudent to state clearly in the safety
    evaluation that the procedures to assure proper valve position needed to be highly
    reliable and single-failure proof.
                                                                                          :
    The inspectors noted that the 10 CFR 50.59 safety evaluation included an
    assumption that valve 3MU995 (and 2MU995) would always be closed in
    accordance with the valve lineup procedures contained within the locked valve
    program. The licensee credited the independent verification of valve position
    provided within the locked valve program to conclude that two independent errors
    would have to occur for the valve to be left open. Therefore, postulating the valve
    being left open would have exceeded single failure analysic requirements. The
    inspectors concluded that the licensee was correct in their interpretation of the
    requirements and that the safety evaluation's exclusion of mispositioning was
    acceptable. As discussed in Section E1.1 above, the inspectors concluded (during
    this inspection) that a single procedural error in combination with several examples
    of poor practice had caused the mispositioning. However, the inspectors concluded
    that the originators of the 10 CFR 50.59 evaluation were justifiably unaware of this
    weakness in the locked valve program and that, based on information available at
    the time, the single error assumption made in the 10 CFR 50.59 evaluation was
    acceptable.
    Regarding the Updated Final Safety Analysis Report update for the reactor coolant
    gas vent system, the inspectors concluded that while the change made to
l    Figure 9.3-15 was accurate, the Updated Final Safety Analysis Report change was
    not complete. The orifice in this system is referred to in several places in Updated
    Final Safety Analysis Report Section 9.3.7. A complete Updated Final Safety
 
    -  -                            -        .                -        -  -
!
  A
  4
  '
l
.
                                                      -12-
l
                                                                                                    ,
                                                                                                    ,
                                                                                                    l
              Analysis Report change would have modified Section 9.3,7 to make it clear that the
              reactor coolant gas vent system orifice was replaced with a gate valve that could      l
              act as a restricting orifice when closed, or could be opened to expedite fillirig and
              venting operations,                                                                    l
              The inspectors reviewed the specification defining the operability of the reactor      i
              coolant gas vent system. The current location of this specification is Licensee        )
              Controlled Specification 3.4.102. This specification did not reference                !
              valve 3MU995. The inspection team observed that a more complete description of        I
              the vent system would include this valve, and its proper position, during Modes 1,    '
              2,3, and 4. As discussed in Section E1.1 of this report, the reactor coolant gas
              vent system could be considered operable with valve 3MU995 open, but its
              conformance to design is dependent on this valve being closed.
              The inspectors reviewed Procedure SO123-XXIV-10.21 to determine if the
              weakness identified in NRC Inspection Report 50-361:362/95-26had been
              corrected by the licensee. The inspectors concluded that no changes had been
              made to the applicable portions of the procedure, and that documentation of the
              basis for determining if a plant change required a 50.59 safety evaluation was still  l
              absent from the procedure.                                                            '
              During review of this issue, the inspectors noted that the 10 CFR 50.59 screening      i
              process combined many impact evaluations into a single review effort, with only        '
              one set of initials provided for verification purposes. The procedure did not require
              the analyst to document which sections of the Updated Final Safety Analysis Report
              were reviewed for potential impact. The inspectors concluded that if the personnel
              responsible for evaluating the change to the reactor coolant gas vent system were
              also required to document which sections of the Updated Final Safety Analysis
              Report were reviewed during the design process, there would have been a
              significantly higher likelihood of recognizing the need for a 10 CFR 50.59 safety
              evaluation.
              The violation was left open because: (1) The Updated Final Safety Analysis Report
              update for the modification did not include an adequate description of the existence
              and function of the orifice gate valve (3MU995), and (2) the weakness identified in
              the 10 CFR 50.59 initial screening (no written justification of screening decisions)
              had not been corrected.
      E8.2 (Closed) Licensee Event Report 95-04: Inoperable Fire Control System
              Backaround
!
!
              During Surveillance testing, the licensee discovered that preaction
              valve SA2301MU469 was inoperable. The valve rel ease weight stuck upon
              electronic actuation and the valve did not open as required. The binding problem
l            was intermittent, evidenced by the fact that during repeated cycling the valve
 
    ._ _                                        _-
  ,
  :
  .
                                                        -13-
                occasionally opened. The licensee concluded that the valve failed to open because
                of incorrect assembly by the manufacturer and a mispositioning of the weight
                switch. Both problems were necessary for the valve to fail. The licensee
                determined that the valve had been inoperable for an extended period and that the
                Technical Specification requirement to set a fire watch within 1 hour of failure had
                not been met.
,
                This valve must open to charge the sprinkler system in the Unit 2 emergency diesel
*
                generator (2G002) room. Had the valve failed to open, the sprinkler system would
                not have been available, and the licensee would have had to rely on the remote fire
                alarm and manual fire fighting efforts to mitigate the consequences of a fire.
                The licensee concluded that the valve became inoperable on September 20,1994.
                The problem was discovered on November 14,1994, at which time a fire watch
'
                was established and maintained until the valve was repaired and tested
                satisfactorily.
                Followun
4
                The inspectors reviewed documentation and discussed this event with the licensee
                and concluded that: (1) the licensee had promptly acted to set a fire watch and to
                repair the valve, (2) that the failure of the valve was the result of a coincidental
                combination of two remote causal factors, (3) that no prior information existed to
                suspect a generic problem necessitating testing of preaction valves beyond the
                nominal surveillance schedule, and (4) that the licensee's preventive maintenance      l
l
                and testing program was not at fault for not preventing the failure and remained        l
                appropriate for future assurance of operability. The inspectors did not consider this
                event to constitute a violation because the licensee acted within one hour              ]
                (Technical Specification limit) of the time of discovery to set a fire watch. Based on  ;
*
                these considerations, the inspectors considered this item closed.                      I
                                                                                                        l
"
          E8.3 (Closed) Insoection Followuo item 50-361:362/9526-03: Licensee Event
                Report 50-361:362/95-16): Heating, Ventilation, and Air Conditioning /High Energy
                Line Break Interactions
                Backaround
                The licensee determined that the original architect / engineer review of the plant's    I
                response to high energy line breaks had failed to identify the potential for steam
                from a line break to migrate from environmentally harsh areas to environmentally
                mild areas through ventilation ducts. The concern was that nonenvironmental-
                qualified safety-related components in the assumed mild environment spaces could
                be exposed to high temperature steam and fail to function as designed. The
                licensee had identified this issue as part of a larger program designed to evaluate
                                                                                                        l
                the adequacy of all types of barriers in the plant. The issue pertained to both units.
                                                                                                        l
 
    .. . __. _ ._      _ _        _.      _ _ _ -._. .            _ _ _ _      __    _  _ _ _ _ _ _ _ _ _ _
  a
  :.                                                                                                              '
  .
                                                            -14-
1
l
l                  The licensee performed an extensive review to identify spaces vulnerable to this
                  mechanism and to evaluate the potential effect on safety-related components within
                  those spaces. The following four areas of concern were identified:
                  (1)      Control Buildina/ Safety Eauipment Buildina
                            Temperature switches were installed to stop ventilation fans in the event
i                          high temperatures are detected. Some pipe supports were strengthened to
l                          lessen the likelihood of pipe whip causing transmission of steam between              ;
l
                            spaces. All actions were complete for these areas.
                  (2)      Penetration Buildina
                            A 2-inch steamline was cut and capped. No other corrective actions were                l
                            planned.
                  (3)      Radwaste Buildina
                            Support modifications were planned. The engineering work was completed
                            but no field implementation had occurred to date. The concern with this
                            area was potentialloss of the swing charging pump. For interim operability,            I
                            the licensee credited alternate charging sources, including high pressure
                            coolant injection.
                  (4)      Doahouse (steam aenerator blowdown valve area)
                            The concern with this space was that a ventilation fan in the main steam
                            isolation valve area could bring steam into the doghouse through the
                            ductwork. This could adversely affect the remote position indication of the
                            steam generator blowdown valves. The licensee developed a field change to
                            replace the limit switches on these valves with qualified switches. The
                            switches were on order. Other equipment in the doghouse was potentially
                            affected but was located lower in elevation and out of the way of the
                            expected steam intrusion. For interim operability, the licensee credited
                            thermal stratification of the doghouse, which would prevent sensitive
                            equipment from exposure concerns.
                  Followuo
                  The inspectors discussed this issue with the licensee and reviewed Licensee Event
l                  Report 95-16 and Nonconformance Report 951100064. Based on this review,
l                  inspectors concluded that the licensee had performed commendably, first, to
l
                  identify this issue, and, second, to take aggressive corrective actions. These
;                  actions were indicative of a strong, proactive approach to plant safety.
.
!
l
l
 
                                                          . . - _ . _ . . _ . . _ _ -_ _ ._ _ _ . _ _ . - . , _
o
.'
.
                                          -15-
      The inspectors were confident that the licensee was on schedule to complete all
      proposed corrective actions. Interim operability bases appeared reasonable.
      Accordingly, this issue was considered closed.
                                V. Manaaement Meetinas
  X1 Exit Meeting Summary
      The inspectors presented the inspection results to members of licensee management
      at the conclusion of the inspection on November 22,1996 and during a conference
      call conducted on December 16,1996. The licensee acknowledged the findings
      presented. The licensee did not identify as proprietary any of the information
      presented to the inspectors during the inspection.
                                                                                                                l
                                                                                                                ,
                                                                                                                1
                                                                                                                l
                                                                                                                l
 
  . . . . . . _  _ _ . . . . . _ _ . , _ __ _ . . _ _ . _ . _ _ _ _ _ _ _ _ _ _ . . _ . . _ . _ . . _ _ . _ . _ . . _ _ . _ . . _ _
    m
    a
    *
                                                                                                                                    .
                                                                                                                                    "
      .
                                                                                                            ATTACHMENT
                                                                            PARTIAL LIST OF PERSONS CONTACTED                        l
                Licensee
                D. Axline, Licensing Engineer                                                                                        -
                D. Brieg, Manager, Station Technical
                R. Clark, Manager, Quality Engineering and Fuels
                C. Coker, Supervisor, Design Engineering                                                                            ,
                G. Gibson, Manager, Compliance                                                                                      t
                J. Hedrick, Manager, NEDO-Plant Engineering
                R. Krieger, Vice President, Nuclear Generation
                D. Nunn, Vice President, Engineering and Technical Services
                J. Rainsberry, Plant Licensing Manager                                                                                i
                S. Root, Supervisor, NEDO
                H. Smith, Compliance
                K. Stagle, Manager, Nuclear Oversight
                                                                                                                                    -l
                R. Waldo, Manager, Operations                                                                                        '
                M. Wharton, Manager, Engineering Design
                C. Williams, Supervisor, Compliance
                T. Yackle, Manager, Safety Review Committee
                Other Oraanizations
                                                                                                                                      I
                W. Peabody, Consultant, Peabody Associates
                                                                                                                                    >
                NRC                                                                                                                  '
                J. Sloan, Senior Resident inspector
                                                                            LIST OF INSPECTION PROCEDURES USED
                IP 37550                                          Engineering
                IP 37551'                                          Onsite Engineering
                IP 92903                                          Followup- Engineering
l                                                              LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
l
                Opened
!
I.              50-361:362/9617-01 VIO Failure to Follow Procedure (Section E1.1)
:
!
i
                                                                                      ,
 
                                                                          - -
I
j c                                                                                        ,
l o                                                                                        l
l
  .
  '
                                                                                            1
                                                                                            '
    Closed
    50-361:362/95-04    LER Inoperable Fire Control System (Section E8.2)
    50-361;362/9526-03 IFl Heating, Ventilation, and Air Conditioning /High Energy Line    l
                              Break Interactions (Section E8.3)
    50-361;362/95-16    LER Heating, Ventilation, and Air Conditioning /High Energy Line
                              Break Interactions Interactions (Section E8.3)
    Discussed
    50-361;362/9526-02 VIO Failure to Perform Safety Evaluation
                              (Section E8.1)
i
                                                                                          /
}}

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