IR 05000423/1997206: Difference between revisions

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{{Adams
{{Adams
| number = ML20202J819
| number = ML20248L910
| issue date = 12/05/1997
| issue date = 03/05/1998
| title = Insp Rept 50-423/97-206 on 970818-29 & 0908-19.Violations Noted.Major Areas Inspected:Licensees Configuration Mgt Plan & Emergency Core Cooling & Seal Injection Functions of Charging Sys
| title = Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-423/97-206.Actions Will Be Examined During Future Insp
| author name =  
| author name = Reynolds S
| author affiliation = NRC (Affiliation Not Assigned)
| author affiliation = NRC (Affiliation Not Assigned)
| addressee name =  
| addressee name = Bowling M
| addressee affiliation =  
| addressee affiliation = NORTHEAST NUCLEAR ENERGY CO.
| docket = 05000423
| docket = 05000423
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-423-97-206, NUDOCS 9712110251
| document report number = 50-423-97-206, NUDOCS 9803250013
| package number = ML20202J759
| title reference date = 02-02-1998
| document type = INSPECTION REPORT, NRC-GENERATED, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| document type = CORRESPONDENCE-LETTERS, OUTGOING CORRESPONDENCE
| page count = 51
| page count = 3
}}
}}


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=Text=
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March 5, 1998
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Report No.:  50 423/97 206 Docket No.:  50 423 License No.:  NPF-49    -
  * Mr. M. L. Bowling, Recovery Officer, Unit 2 l c/o Ms. Patricia Loftus, Director Regulatory Affairs for Millstone Station   RGQgED.
Licensee:  Northeast Nuclear Energy Company Facility:  Millstone Unit 3 3 Location:  Millstone Nuclear Power Stathn 156 Rope Ferry Road Waterford, Connecticut 06385 Dates:  August 18-29 and September 8-19,1997 Inspectors:  James Luehman, ICAVP, Leader, Sub-Team 3A Special Projects Office Stephen Tingen, Special Projects Office Donald Prevatte, Mechanical Engineer, Contractor *
Robert Quirk, l&C Engineer, Contractor *
Omar Mazzoni, Electrical Engineer, Contractor *
Victor Ferrarini, Mechanical Engineer, Contractor *
    * Contractors from Parameter, In Approved by:   Peter Kottay, ICAVP, Leader, Team 3 Special Projects Office Office o' duelear Reactor Regulation Enclosure 2 9712110251 971205 PDR ADOCK 05000423 G    PDR
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._ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ - _ _ _ _ _ _ _ . _   _ . _  .. _ _ __
Northeast Nuclear Energy Company P.O. Box 128   19?S 193 p fil 2: 08 Waterford, CT 06385 PUBLlc DOCUMCIT ROOH SUBJECT: INSPECTION REPORT NO. 50-423/97-206
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1. *    TABLE OF CONTENTS EXE CUTIVE S UMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1.0
;  I1.1nt rod uction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 System Description and Safety Function . . . . . . . . . . . . . . . . . . . . . . . . 1 inspection Scope and Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 a
2.0 i
Mechanical .......................................................
'    Charging System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 Recirculation Spray System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 Main Steam System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Conclusion s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12. . . . . . . . . . .
j  3.0
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E lect rical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Onsite ac Source - Emergency Diesel Generstor . . . . . . . . . . . . . . . 12
, Medium Voltage Distribution and Medium Voltage Switchgear . . . . . . 13
: Vdc Batteries . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
; Degraded grid protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
: Conclusion s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .,16. . .
          . . . . instrumentation and Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . 16 Transfer from Normal Chemical and Volume Control System Lineup to
,
High-Head Safety injection and Containment Sump Recirculation . . . , .16 Refueling Water Service Tank . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . 17
. Calculation Content and Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 i
4.4
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Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
; Structures and Supports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
, System Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
; Piping Stress Calculations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 i Thermal Expansion of Structural Steel . . . . . . . . . . . . . . . . . . . . . . . . . 22 Conclusion s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 Ope rations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 Charging and Safety injection System Operating Procedures . . . . . . . 23 ,
,
Charging and Safety injection System Annunciator    !
Response Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 :
. Operator Training . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6 l l    Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 1
> l M ainte na nce . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7 ; Charging and Safety injection System Maintenance Procedures . . . . . 27  ' Maintenance-Related intormation Notice (IN) Rwiew . . . . . . . . . . . . . 27 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 1 Surveilla nce . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 8 Technical Specification Surveillance . . . . . . . . . . . . . . . . . . . . . . . . . . 28
; inservice Pump and Valve Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30
: Con::!usions ...........................................31


i i          . i l
==Dear Mr. Bowling:==
l
This responds to your February 2,1998, letter, regarding the Notice of Violation that was transmitted with the December 5,1997, inspection report.


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Thank you for irforming us of the corrective and preventive actions documented in Attachment 2 of your letter. These actions will be examined during a future inspection of your license program.


. . _ . _ _ _ . . _ . . . _ . _ . . . _ . _ . _ _ . - . _ _ _ _ . _ . _ . _ _ . _ _ _ _ _ _ _ . _ _ . _ _ _ . . _ _ _
Your cooperation with us is appreciated.
.
' Entrance and Exit Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 Appendix A _ List of Apparent Violations, Unresolved items, and Inspector Followup Items . A 1 Appendix B Entrance & Exit Meeting Attendees . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-1 Appendix C List of Documents Reviewed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-1 Appendix D List of Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D-1
          ,
,


i
Sincerely,
.
/s/
I4 EXECUTIVE SUMMARY Millstone Nuclear Power Station inspection Report 50423/g7 206 During the periods from August 18 2g, and September 8 ig,1997, a team from the Nuclear Regulatory Commission's (NRC) Special Projects Office, Office of Nuclear Reactor Regulation, conducted a Safety System Functional inspection (SSF1) at Millstone Nuclear Power Station Unit 3. The inspection was done as a part of the NRC's oversight of the independent Corrective Action Verification Program (ICAVP). The purpose of this inspection was to assess the effectiveness of the licensee's Configuration Management Pieri(CMP)
J.Nakoski for Steven A. Reynolds, Chief ICAVP Oversight Branch Special Projects Office Docket No. 50-423 cc: See next page Distribution:
identifying ersat sf nonconformance with the plant's design and licensing bases by inspecting a sysbm that was not in the scope of the ICAVP contractor's reviww.- The team's review focused on tne emngency core coollag araf sealinjection functions of the charging syste The team identified apparent violations ami other findings that indicated that the licensee's configuration management program efforts undertaken, in ps,1, in respon6e to the findings of the igg 6 NRC "Specialinspection of Engineering and Ucensing Activities at Millstone Nuclear Power Station"(Report 50 423/g6 201), failed to identify some potentially significant licensing and design basis issue The team's principal findings were identified in the areas of operations, surveillance, and maintenance. The team found that the air in certain piping in the initially dry recirculation spray sye' m (RSS) could be swapt into the suctions of the charging and safety injection pumps du
PUBLIC/ Docket File SPO R/F J. Andersen M. Callahan, OCA R. Correia B. McCabe, OEDO S. Dembek      \d G. Imbro      \
  'ha cold leg recirculation phase of a loss-of coolant accident (LOCA). The pctential for air ing these pumps was recognized by Northeast Nuclear Energy Company (NNECO) prior to inwl plar.t startup. However, that review failed to completely identify and analyze the extent of the problem. Given the somewhat unique configuration of the RSS (an initially dry system subsequently providing not positive suction head for two water filled systems) the team was concemed not only that the licensee had initially failed to fully identify this issue but, even after the issuance of NRC Information Notice (IN) 88-23, " Potential for Gas Bindirg of High Pressure Safety injection Pumps," and its four supplements, this issue still went unidentifie The licensee's program for minimizing leakage optside the cortainment (a requirement ta Technical Specification (TS) 6.8 A) did not address inter system leakage that could result in radioactive water leaking into places where radioactivity could be vented to the atmosphere,
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such as the refueling water storage tank (RWST) following a LOCA. A minor leakage through such valves could result in unacceptable control room and offsita radiation doses in the event of an accident. The failure to test the vates in question also is indicative cf a weakness in the licensee's American Society of Mechanical Engineers (ASME) Code Section XI valve testing Progra The valve lineup procedure for verifying the TS requirement that all charging pump injection !
flow path valves, not locked or otherwise secured in position, are in their correct positions every 31 days was found to be inadequate. For this gocific case, the licensee was able to subsequently show, by use of other valve lineup procedures and reliance on valve position alarms, that it was unlikely that any of the valves not included in the designated lineup procedure could in fact remain out of positio .
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l A concem in the structural design area was the licensee's fklure to account for the post-
{ accident temperature rise inside containment on rigidly restrained structural steel. In an elevated temperature environment the steel will tend to expand more rapidly than the concrete
 
' around it. Excessive expansion of the steel could result in damage to the concrete structures '
as well as consequential damage to nearby equipment. At the end of the inspection the
] licensee was evaluating the potential effects of this isrus.
 
.
! In the area of mechanical design, the removal of certain mair; steam valves from the electrical j
equipment qualification program was identifed as an issue. Originally, the main steam block
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valves in question were part of the environmental qualification program. Subsequentl :
i questions relative to the actual quslification of the valves aros6 eAd the licensee performed an ;
i analysis that justifed the removal of the valg a from the program. The team concluded that !
based on the available documentation, the removal of the valves from the program was not
      '
properlyjustifie ;
The lack of adequate justification for the qualification of the seal material used on certain RSS containment isolation valves was an unresolved item. At the end of the inspection the licensee had not provided the team with sufficient information to support itt position that the material was qualifed for the radiation expasure it would receive during a design basis LOC In the areas of instrumentation and control and electrical design the team identified weaknesses in calculation control, control of documentation of assumptions supporting calculations, and inadequacies in the consideration of instrument uncertaintie Overall, the team found the material condition of the charging system and associated support systems to be good. The significant engineering calculation work ongoing, the outstanding testing of a major orifice modification to the charging system, and the outstanding issues the team identifed on the RSS (which supports the charging system in the recirculation phase of an .
accident), precluded the team from verifying that the system would perform its inten: led function ,
under all design conditions. However, with all modifications completed, and test results founc to be acceptable, the system will perform its intended functio The team identifed more findings than would be expected following an effort such as the licensee's CMP, Specifically, issues such as the effects of air entrainment in RSS piping, inadequate inter-system leakage monitoring of valves, and an nample of a valve lineup procedure that failed to meet TS regalroments, were expected to be identifed through the licensee's CMP efforts. Additionally, at least some of t".a less significant procedural and FSAR inconsisten:les noted by the inspection team, should also have been identifed and correctad by the licensee's CM The strengths identified during the inspection included an aggressive maintenance program, knowledgeable system engineers, and a document control organization that appears to have a good understanding of the programmatic challenges that remain in improving the unit's calculation control program, il
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4 l- Introduction l
i
}i On August 18 2g and September 81g, a team from the U.S. Nuclear Regulatory Commission's ;
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(NRC) Special Projects Office, Office of Nuclear Reactor Regulation, conducted a Safety i System Function Inspection (SSFl) at Millstone Nuclear Power Station. The inspection was l done as part of the NRC's oversight of the independent Corrective Action Verification Program
;
(ICAVP) at Millstone. The purpose of this inspection was to provide the NRC with additional l insights to evaluate the effectiveness of the licensee's Configuration Management Plan (CMP).
 
i          ,
The system selected for the inspection was outside the scope of the ICAVP contractor's review but was a system for which CMP was complete !
. Background The safety and seal injection functions of the charging system were selected for this inspection L from among the 88 Groups 1 and 2 systems covered by Title 10 of the Code of Federal
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Regulations, (10 CFR) Section 50.65, also known as the Maintenance Rule. All 88 Groups 1  i j
and 2 systems were evaluated oy the licensee's CMP. The specific criteria that were i considered in choosing the charging system included the relatively high number of other risk
! significant systems it interfaces with, the number of modifications and changes made to the
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system since plant licensing, and the relatively high risk significance the system has when considering accident scenario *
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j System Descrintion and Safetv Furdan    1
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' During normal operations, the charging system has numerous functions including providing
' makeup water to the reactor, seal injection cooling for the reactor coolant pumps (RCP), and boric acid for rea:tivity control. During and after an accdont, the charging system provides
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RCP seal cooling, the high pressure injection of borated water from the RWST during the injection phase, and one of the pathways for recirculation of contahment sump water during
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" later stages of an accident, it is this accident mitigation group of functions that was the focus of the inspection.
 
i AnaDection Score and Methode:esv
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3 The team used SSFI, inspection Module g3801, as its inspection approach Becauce the i
purpose of the inspection was to assess the effectiveness of the licensee's CMP, the emphasis
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of the inspection was system design, including modifications, and the appropriate translation of
. the design and licensing bau ' . ' operations and surveillance activities.
 
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; 1.3.1 Millstone Unit 3 Charging System inspection Boundaries
; Since the mechanical, electrical, control, and structural aspects of charging system interface with other plant systems, the inspection included some portions of interfacing systems. The
!
team identified the charging system inspection boundaries as follows:
e Seal Injection - From the charging system to containment isolation valve CHS*MV8100 and relief valve CHS*RV-8121. (RCP seal 2 and 3 leakoff piping and the No. 3 seal supply piping, and relief valve discharge piping to the pressurized relief tank (PRT) were
 
not within the scope of this inspection).
 
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      --,,ww-.-,, m,, ...,,+.yw,., -..,v.,
 
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Emergency core cooling system (ECCS) Injection and Recirculation The three 1      c%rgina pumps; piping and valves bounded by the volume control tank (VCT) outlet (d.OV CMS *LCV 1128): charging isolation valve CHS*MV4105; charging pump mini-
!      ficw line relief valves CHS*RS4510 A and B; charging pump suction recirculation
;
supply valves SIL*MV-8804A and B; charging s/ stem cold leg injection connections to l
reactor coolant system (RCS) loops 14; charging pump miniflow isolation valve
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CHS*MV4110; boric acid connections to the charging pump suction piping; the seal l i
water heat exchanger connection to the charging pump suction piping; the charging
{'      system header drain and test lines bounded by SlH*CV-8842 and 8843; and the
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hydrostatic test pump discharge piping connection to the charging system piping.
 
j            ,
;    e Suport Systems Charging pump cooling (CCE) (bounded by the CCE expansion tank
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and heat exchanger service water (SW) inlet / outlet valves SWP*V438, V848, V31A j      and V32A); ventilation supply / exhaust for the charging pump rooms (bounded by
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dampers HVR* MOD 40A, B, C1, C2 and HVR* MOD 4M. I, C1, and C2); ventilation to i      the auxiliary building, switchgear, and engineered safegw.<ds featwo motor control -
:
centers (MCCs) for electrical components within the scope of the inspection; electrical j-            ,
components within the system boundaries back to the MCCs including emergency j
diesel generator (EDG) loading; and instrumentation within the system boundaries back
            ;
j      to the initiating device or component, e
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' Mechanical i
, Charaina Svatem i
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,    2.1.1 Scope of Review
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l    The mechanical engineering review was aimed primarily at verifying that the charging system
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'    was capable of secomplishing its design functions, meeting all applicable codes, standards, regulatory requirements, and good engineering practices, meeting all licensing basis  ,
;    commitments; operating consistent with its design basis, and being tested in a manner to
,    accurately reflect its condition and its ability to perform its intended functio !
;
The interfacing and supporting systems for the charging system were reviewed to ensure that  i i    they could provide the necessary support for the charging system to perform its designed j
functions. The interfacing and supporting systems included the service water (SW), sealing and j    lubrication, instrument air, ventilation, reactor coolant, and recirculation spray.
 
4 During the mechanical and engineering review, special attention was focused on venfying the
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system and supporting systems could meet the minimum performance requirements; reviewing i    the systems' single failure design integrity; verifying fulfillment of the Final Safety Analysis
;    ' Report (FSAR) and TS commitments; and verifying adequate piping design temperatures and
[    pressures.
 
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4    2.1.2 Findings
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    (a)  Inadequate charging pump performance testing acceptance criteria jo verify meeting FSAR smallleak safe shutdown statement
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FSAR Section 6.3.3.2, * Loss of Reactor Coolant from Small Ruptured Pipes, from Cracks in large Pipes, or from the Ejection of a Control Rod which Actuate the Emergency Core Cooling System," provides the following discussion of the charoks system's safety function and cepobility for small reactor coolant system (RCS) breaks, " Ruptures of small cross sections l cause expulsion of the coolant at a rate which can be accommodated by the charging pumps l which would maintain an operational water level in the pressurizer, permitting the operator to j
' execute an orderly shutdown. A makeup flow rate from one charging pump is adequate to sustain pressurizer level at 2,250 pela [2,235 peig) for a break through a 0.375-inch diameter hole. This break results in a loss of approximately 17.5-Itvoec [127 ppm)."
 
!
According to TS 3.4.6.2.e, " Reactor Coolant System Operational Leakage," the total flow a charging pump would have to deliver would be 127 gallons por minute (gpm) for the leak, plus 40 gpm to the RCP seals, plus at least 60 ppm through the minimum flow bypass line (the required minimum flow per SP 3604A (B & C).1, Rev g, dated July 28,1997, " Charging Pump A (B & C) Operational Readiness Test") for a total of 227 gpm. The required developed head, neglecting elevation differences, which would be relatively minor, would be the RCS pressure of 2,235 psig plus the charging system resistanc TS 4.5.2.f.1, " Emergency Core Cooling Systems Surveillance Requirements," establishes the charging pump surveillance testing performance limit of a 2,411 paid developed head at recirculation flow conditions, and surveillance procedure SP 3604A (B & C) 1, Rev. g, dated July 28,1997 " Charging Pump A (B & C) Operational Readiness Test," was intended to test this requirement. Both of these were intended to assure that the pumps' minimum performance was above the " Accident Analysis Minimum Acceptable Curves" contained in procedure EM 31121, Rev. 6, dated December 10, igg 6, "lST Pump Operational Readiness Evaluation,"
which were derived from the Westinghouse accident analyse The team's review of these documents revealed that, even when the charging system resistancs was neglected, the point representing the above quoted FSAR statement (i.e,227 gpm at 2,235 paid, when plotted on the charging pump minimum performance curve, was approximately 217 poig above the curve, and approximately 173 psig above the curve that could be projected from the TS minimum allowable point. Although a calculation that provided the charging system resistance along the normal flow path was not located, the accident injection flow path would be'similar. The calculation for that flow path indicated a system
  .
resistance of about 104 pol under these flow condiuons. The acceptance criteria specified in TS 4.5.2.f.1 were found, however, to be adequate to demonstrate the charging pumps' ability to meet the accident analyses performance requirements, which were considerably lower than the FSAR statemen The team reviewed the latest actual pump performance test data. The "A" pump had the best performance, and the actual data from a May ig,1997, test, indicated that with virtually any system resistance taken into account, and without accounting for instrument uncertainty, it would not have met the FSAR statement of 127 gpm at 2250 psi The pump performance specified by the FSAR statement does not appear to be required to comply with any regulatory requirement and, as discussed above, pump performance appeared adequate to meet accident analysis and ASME Code testing requirements. Therefore, the safety significance of this finding is low. Nevertheless, the plant (as-built) does not meet the FSA _  .
      ..
 
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.      4 The deficiency above represents one instance in which the Millstone Unit 3 FSAR was not maintained up-to-date or did not reflect the actual plant configuration or operating practice. As such, it constitutes a violation of 10 CFR 50.71(e). (VlO 423/g7 206-01)
  (b) Incomplete relief valve replacement item evaluation Relief valve 3CHS*RV811g was rcplaced with a different valve by Replacement item Evaluation (RIE) No. PSE MP3E 94-101. This relief valve provided overpressure protection for the letdown piping downstream of the pressure control valve that is downstream of the letdown heat exchanger. It was replaced because the original va!te leaked, and it was no longer available from the vendor in its original design configuratio One of the design features of the replacement valve that is different from the original was the sealing gasket material. The original coating gasket material was asbestos and the replacement was Buna N rubber. The original asbestos material was qualified for the life of the plant; however, the Buna N material was qualified for 13 years at expected operating
,  condition Nuclear Group Procedure (NGP) 6.12, Rev.1, " Evaluation of Replacement items,"is the
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licensee's procedure that govems such replacements, and it required that RIES includes a Preventative Maintenance Management System (PMMS) update to identify material with finite lives that would require periodic replacement, (i.e., the Buna N gasket in this valve), it also requires a bill of materials (BOM) update. The RIE contained neither update and consequently neither change required was performed by the licensee, in response to this finding, the licensee initiated Condition Report (CR) M3-g7-3105. Because this valve provides overpressure protection for the piping and has no direct accident mitigation function, its failure potential as a result of seal deterioration has little safety significanc The failure to update the PMMS and the BOM, as required, is considered one example of a violation of Technical Specification 6.8.1, failure to follow procedurer. (VIO 423/g7 206-02)
  (c) System relief valves' discharge path does not meet the ASME code ASME Boiler and Pressure Vessel Code,1971, Section lil, " Nuclear Power Plant Components,"
Article NC-7000 requires that " Vessels, piping, valves, and pumps which are pressure-containing components of the systems within the scope of Subsection NC thall be protected from pressure and (coincident) temperature that are in excess of the desip conditions."
 
Relief valve 3CHS*RV811g was intended to provide this mtection for the letdown piping and components downstream of pressure control valve 30!;S*PCV131. This valve's pressure was set at 300 psig per the design pressure requireme;its this piping. Relief valve 3CHS*RV8123 was intended to protect the RCP sealloakofi piping and components upstream of the seal water heat exchanger. Its set pressure of 150 psig is also per the piping's design pressure requirement. The dischn,ge piping for both of these relief valves combines into a common line connected to the VCT. This line and the VCT are also protected by relief vt tve 3CHS*RV8120 set at 85 psig. (The VCT normally operates at approximately 15 psig.) 3CHS*RV811g has a 4-inch discharge, and 3CHS*RV8123 has a 3-inch discharge; the common discharge line into the VCT necked down from 4 inches to 3 inche _ _ _ _ _  --    -
_ - -
__
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(
D. Mcdonald P. McKee D. Screnci, OPA      [0g DOCUMENT NAME:P: reply.JI Ta receive a copy of this document, indicate in the box "C" copy w/o attach /enci"E" copy wfattach/enci"N" no copy OFFICE  ICAVP 1 Q ICAVP / C:lCAVP , C NAME  JLuehmanh PKo  SReynoldM h#
,     5    l i
f DATE  3/[/98  198 3/[/98 I  OFFICIAL RECORD COPY 9003250013 900305 -
Article NC 7512 of the ASME Code," Pressure Drop Considerations," states,"In determining the set pressures and capacities required to comply with these rules, full accont shall be taken of the pressure drop on both inlet and discharge sides of the pressure relief device at full flow
PDR ADOCK 05000423 G  PDR
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condition. In addition, back pressure arising from discharge to closed storage or dissipation systems or from discharge of other pressure-relieving devices through common discharge piping, shall be taken into account." Article NC 3677.3, *Discherp9 Piping from Pressure-Relieving Safety Devices," subparagraph (d), states, "It is recommended that individual (relief valve) discharge lines be used but if two or more reliefs are combined, the disearge piping shall be designed with sufficient flow area so as not to affect opc, ration of the relief devices and ,
in no case shall the area of such common lines be less than the combined area e' alllin discharging into it."
 
Contrary to Article NC-7512, the back pressure that relief valves 3CHS*RV8115 and 3CHS*RV8123 would discharge to the VCT pressure could be as high as 85 ps4 Therefore, when considering a 10 percent accumulation, the relieving pressures in the lines being protected could be as high as 415 psig and 250 poig, well above the ASME Code allowable for maximum relieving pressures of 330 psig and 165 psig, respectively. Contrary to Article NC-3677.3, the area of the common 3-inch discharge line for thess two valves is significantly less ;
than (approximately 1/3 of) the combined area of the two discharge lines.
 
-
In response to the common discharge line flow area concems the licensee s%d that the ASME Code requirement related to multiple relief valves that protect the sar, + vnponent for a common event, and that these valves were designed to lift because C f unrelat ant ,
Therefore, it is not assumed that the relief valves lift concurrently. The licenset ~. no analysis to support its position that no single event could challenge both valves simultaneously. Because the common discharge piping necked down to 3 inches, thereby not even providing full flow area for relief valve 3CHS*RV811g which has a 4-inch discharge area, the licensee's position
;- does not adequately address the team's piping area concem and fails to address the back -
pressure concem.


'
f
Cumpliance with the ASME Code aside, the safety significence of this finding appears to be minimalt Although the design may have allowed the piping to be overpressurized,it is not likely that a failure would occu Failure to meet the code required size for the charging system relief valve discharge piping is an apparent violation of 10 CFR 50.55a. (VIO 423/g7 206 03)


  (d) Non-control of prohibited materials in plant While ingsetion the charging system seal injection piping to RCP "A," the team observed a yellow plastic high-pressure (HP) tape applied to the piping. Chloride and fluoride containing materials, such as this tape, that comes in contact with stainless steel piping, in combination with elevated temperaturas and stresses,' can cause cracking and stress corrosion of the pipin Procedure CC1, Rev. 2, dated April 14,1997, " Control of Chemical Consumabl6 Products,"
Northeast Nuclear Energy Company Millstone Nuclear Power Station Unit 2 cc:
recognizes this fact. In a note preceding Step 1.6.2, "The restriction on chemical consumable >
Lillian M. Cuoco, Esquire Mr. F. C. Rothen Senior Nuclear Counsel  Vice President -Work Services Northeast Utilities Service Company Northeast Utilities Service Company P. O. Box 270  P. O. Box 128 H:rtford, CT 06141-0270 Waterford, CT 06385 Mr. John Buckingham  Ernest C. Hadley, Esquire Department of Public Utility Control 1040 B Main Street Electric Unit  P.O. Box 549 10 Liberty Square  West Wareham, MA 02576 New Britain, CT 06051 Mr. John Streeter Mr. Kevin T. A. McCarthy, Director Vice President- Nuclear Oversight Monitoring and Radiation Division Northeast Utilities Service Company Department of Environmental Protection P. O. Box 128  )
product use on corrosion resistant metal surfaces of reactor plant systems in this procedure are
79 Elm Street  Waterford, CT 06385 H:rtford, CT 06106-5127 Mr. David Amerine Regional Administrator, Region i Vice President - Nuclear Engineering U.S. Nuclear Regulatory Commission and Support 475 Allendale Road  Northeast Utilities Service Company King of Prussia, PA 19406 P. O. Box 128 Waterford, CT 06385 First Selectmen Town of Waterford  Mr. Allan Johanson, Assistant Director Hall of Records  Office of Policy and Management 200 Boston Post Road  Policy Development and Planning Waterford, CT 06385  Division 450 Capitol Avenue - MS# 52ERN Mr. Wayne D. Lanning  P. O. Box 341441 Deputy Director of Inspections Hartford, CT 06134-1441 Special Projects Office 475 Allendale Road  Mr. M. H. Brothers King of Prussia, PA 19406-1415 Vice President - Operatior s Northeast Nuclear Energy C9mpany Charles Brinkman, Manager P.O. Box 128 Washington Nuclear Operations Waterford, CT 06385 ABB Combustion Engineering    I 12300 Twinbrook Pkwy, Suite 330 Mr. J. A. Price  l Rockville, MD 20852 Unit Director- Millstone Unit 2 Northeast Nuclear Energy Company  I Senior Resident inspector P.O. Box 128 Millstone Nuclear Power Station Waterford, CT 06385 cla U.S. Nuclear Regulatory Commission    i t P.O. Box 513 Nirntic, CT 06357
,
      - ______-__-_-_
to prevent induction of corrosion of nickel alloys by ingredients or contaminants of chemical
;
products used on the metal surface when the metal surface is exposed to high system temperature and pressure," and in Step 1.6.2, "lE Job involves application of chemical
_
_ _ . _ _ . ___ _ __._ _ ._ _ _ _ _


_ . _ _ _   _ . _ - _ . . ._ _ ._ - _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _  _
i .
NorNeast Nuclear Energy Company            Millstone Nuclear Power Station Unit 2 cc:
,
, .      6
'
consumable product to a corrosion resistant, pressure retaining metal surface of a primary or
'
secondary system, PERFORM the following: ENSURE product is labeled "A" for Chemical Product Use Category DE ENSURE compliance with restrictions Chemical Consumable
'  Restricted Use Permit in AWO." Attachment 2 " Controlled Chemical Consumable Prodects,"
listed " nuclear grade tapen" as a type of chemical consumable product to which the administrative controls of the procedure applies. Additional direction was provided by a


memorandum from the Chemical Control Coordinator, to " Distribution," dated July 13,1994, states that "these tapes [ yellow HP tape and silver duct tape) are not to be used on    i primary / secondary systems."
Mr. B. D. Kenyon Chief Nucler Omcer- Millstone Northeast Nuclear Energy Company P.O. Box 128 Waterford, CT 06385 Citizens Regulatory Commission ATTN: Ms. Susan Perry Luxton 180 Great Neck Road Waterford,CT 06385
 
i Subsequently, the licensee issued condition report ME g7 2748 to document and evaluate this    ,
condition, and Trouble Report (TR) No. 22M3075644, to clean and inspect the affected are ,
The tape appears to be newly installed ( l.a., since the last plant shutdown, and, therefore, would not have been exposed to high temperatures, and the piping would not have been exposed to high stresses). Therefore, all the necessary conditions to initiate cracking or stress
,
i corrosion did not yet exis t The use of tape on primary systems is considered an example of a violation of Technical Specification 6.8.1, failure to follow procedures. (VIO 423/g7 206-02)
  (e)  incomplete nut to-stud thread engagement
'  Specification SP-ME 570, Rev. 3, dated December 21,1995, " Field Fabrication and Erection of Piping and Supports," page 18 3, requires, "All bolted or stud connections shall have full thread engagement with the nuts." Common Maintenance Procedure C-MP 715A, Rev. O, dated February 25,1994, " General Practices for Threaded Fasteners," requires in Section 4.13. " Final
;
Inspection and Documentation [of bolted joints)," VERIFY nuts have full engagement on studs -    ;
or bolts.
 
'
During the walkdown of the charging system, the team discovered two instances where flanged piping joints did not have full thread, stud-to-nut engagement. The first was on RCP "A" at the seal water injection line pump connection flange where two of the four studs were not fully engaged. The second was on the sealinjection line connection for RCP "D"where one of the four studs was not fully engaged. The last time these joints had been assembled was when the RCPs were replaced in 199 '
Subsequently, the licensee inspected these joints and found that in every case full engagement was lacking by less than one thread, and in both cases, the nuts were of heavy grade.- Their
;
engineering evaluation determined that since the load on bolted joints is carried by the first few threads, the structural integrity of the joints was not degraded. Therefore, the team agreed that there was little safety significance Of this findin The lack of adequate stud-to nut engagement is another example of a violation of Technical Speci9 cation 6 8.1, failure to follow procedures. (VIO 423/97-206-02)
l  (f)  Single failure of charging system check valves
, -  - The team identified two instances where a single component check valve of the ECCS flow j  paths of the charging system was not completely redundant One of these valves was check
            .
a.,..-, v-~. , - - -- . *- - -- , --nn-,---,--.www.- - .--------w--< .-------n - ~~ ~ m ,,,,w,-~-, --- n , .-- - - - - , - w,rv v-- - - , -. ,,w -, -
 
_ _ _ _ __  _ _ _ _ _ __  _ _ _ _ _ _ _ _ _
-
i
:
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,    7 valve 3CHS V261 in the single common suction line from the RWST and the other check valve 3CHS V5 in the single common injection line. During the injection phase of a LOCA, the active failure of either of these valves to open would incapacitate the system, and during the recirculation phase, the passive failure of 3CHS V5 could also totally block system flow (V261 would not be in the flow path during the recirculation phase). This design appears to be
; contrary to that discussed in the FSA Section 3.1 of the FSAR, "Conformance With NRC General Design Criteria," states that "the
.
!
!
Applicants conclude that MP3 fully satisfies and complies with the [ single failure criteria} GDC."
l Deborah Katz, President Citizens Awareness Network P. O. Box 83 l  Shelburne Falls, MA 03170 The Honorable Terry Concannon Co-Chair Nuclear Energy Advisory Council Room 4035 Legislative Office Building Capitol Avenue H:rtford, CT 06106 Mr. Evan W. Woollacott Co-Chair Nuclear Energy Advisory Council 128 Terry's Plain Road Simsbury, CT 06070 Little Harbor Consultants, Inc.
 
Section 3.1.1.2, " Definition of Terms Used in Single Failure Criteria," defined an active failure as
"one in which mechanical movement must occur to complete the component's intended
: function. An active component failure is failure of the component to complete its intended
 
'
function upon demand." One of the examples of an active failure provided in FSAR was "the failu.a of a check valve to move to its correct position." "The ECCS is designed to accept a i single active failure following the incident without loss of its protective function." Under the heading of " Passive Failure Criteria"It states, " Adequate redundancy of check valves is provided to tolerate failure of a check valve during the long-term as a passive component."
 
This section also states, "Two trains of pumps, heat exchangers, and flow paths are provided for redundancy." Although the licensee initiated CR M3-97-2140 on July 11,1997, to address single failure issues, that CR originally addressed only passive failures but was later expanded 4 to generally address active failures as wel In response to this finding, the licensee maintained that only power operated valves were required to be considered for active failure. The licensee implied that the check velves at issue 3 were "especially qualified for service" and, therefore, exempt from active component failure per FSAR Section 3.1.1.3, " Application of Single Active Failure Criterion." An "especially qualified"
; valve was a valve with some special feature or operating condition that would effectively preclude active failure. This section cited the safety injection accumulator check valves as examples of "especially qualified" check valves, and it stated, the post-LOCA high, differential, ,
pressure that would ensure their opening. However, this reason was not applicable to the charging check valves since they would not necessarily experience high, differential, pressure - 1 for all accident conditions, particularly, for small break LOCA where the d#ferential pressure would be relatively low. No other special feature or conddion was identified. Finally, the licensee maintained that it was Westinghouse's position that single failure of check valves did not have to be considered and that the FSAR statements were unintentional and incorrec Correcting the inconsistency between the plant as-built and the initial design by deleting the l FSAR statements, as suggested by the licensee's CR needs to be evaluated. This is identified 1 as an Unresolved item. (URI 423/97-206-04)
(g) Other concoms      l (1) - The following are examples of minor problems with calculations and procedures where the overall reliability of the documents was not affected. In some instances, multiple !
calculations existed that addressed the same subject with no apparent hierarchy or l indication of obsolete or invalid calculations being supersede '
,
Calculation P(R)-0983, *NPSH Evaluation for ECCS Pumps RHS, SlH, CHS - Maximum Safeguards," Rev. O, dated April 23,198 _ . _ _ .  .-_ _ -
 
_ - _ . _ _ - _ _ - - - _ _ _ _ _
.
.
 
Calculation 294, "NPSH Available for ECCS Pumps," Rev. 4, dated September 9,198 Calculation 3-ENG-181, " Determination of Available NPSH to Charging Pumps During Gravity Boration," Rev. O, dated November 14,199 Calculation P(R) 0982, Rev. O, "ECCS Pressure Drop Calculations Based on Westinghouse Piping Resistance Criteria."
 
Calculation P(R) 0982, "ECCS Pressure Drop Calculations Based on Westinghouse Piping Resistance Criteria," Rev. Csiculation RFS-P 1515, " Performance of Modified 4 Loop ECCS," Rev.1, dated January 17,1973.
 
l l
C,alculation 12179 US(B)-311, "RSS Branch Flow Analysis with Degraded Pump Curve,"
Rev. O, Change 1, dated April 18,1997.
 
i i    Calculation 12179-US(B)-245, " Branch Flow Rate Analysis for Safety injection and Containment Recirculation System," Rev. O, Change 3, April 18,199 Calculation UR(B) 393-0, " Lifetime Radiation Dose to 3RSS*MOV20A, B, C, and D,"
Rev. O, dated April 30,198 Calculation UR(B)-400, " Gamma and Beta Equipment Qualification Doses for Normal Operetttns, Depressurized LOCA and Pressurized LOCA." Rev. Although the licensee had identified, on a generic basis, that there was inadequate control of multiple calculations on the same subject, all of the calculations identified above except for the first two relating to charging pump not positive suction head (NPSH) had not been previously identified as specific examples by the license (2) The team identified errors in the following calculat'ons and procedure (a) Calculation SE/FSE-C-NEU 0154 Rev. O, Change 1, dated June 3,1997," Millstone Unit 3 ECCS Evaluation - Future Tech Spec Change Basis."
 
Assumption 1 of the calculation regarding hydraulic resistances was in error, if thm predicted resistances were 30 percent higher than actual, reducing them back to the actual resistance values would require a global toduction factor of 76.9 p,:. cent, not 70 percent. Using a reduction factor of 70 percent would reduce the resistances to 91 percent of the actual values, which would produce predicted flow values higher than actual and be non-conservative with respect to the 10 CFR 50.43 accident anaF/ se The safety significance of this finding is minimal. The flow model was used to predict
    . flow orifice sizing and the throttle valve positions associated with a recent modification where flow orifices were installed to eliminate excessive throttle valve erosion arid to prevent system runout during the recirculation phase of an accident when the charging pumps were being boosted by the RSS pumps. The team verified that during testing of this modification, the correct throttle positions would be set to provide the required
        - ''
        ,,:%.
 
. _ _._ _ _ ___ . . _ _ _ _ . _ _ _ . _ _ _ _ _ . . _ _ . _ _ . _ _ . _ _ _ _ _ _
.
e l
system resistances on the basis of observed flows, and if necessary, orifice plates  !
would be resized. The flow model would subsequently be adjusted to conform to the observed system flow coniition (b) Procedure EN 311121, "lST Pump Operational Readiness Evaluation," Rev. 6, dated December 10,1996. (Provides instructions to establish the baseline reference values io analyze data from all pumps tested in the pla t * a part of the ASME Section XI testir,0 program.)
 
Step 4.1.5.s provided conversion factors if the medium being pumped was seawate This conversion factor was incorrect because pressure in feet of water is a standard term of measurement, expressed in feet of standard water (i.e., fresh water at 68'F),
regardless of the density or temperature of the medium being pumpe Attachment 4, page 22, the TS reference for the containment recirculation pumps curve should have also included TS 4.5.2. Attachment 4, page 23, the last pump referenced should have been 3CHS*P3C and not 3CHS*P3 Attachment 4, page 24, the pumps identified on this curve were the charging pumps; they should have been the safety injection pump Attachment 4, page 25, the correct TS reference is 4.5.2.f.3 not 4.5.2. Attachment 4, page 26, the correct TS reference is 4.6.2,1.b not 4.6.2. Allowance was not made for the 10 gpm imbalance that could occur between the injection lines as required by the Westinghouse accident analysis. This would raise the acceptance criteria from a 452 gpm to a 462 gpm. The team reviewed the most recent test results for the "A" and "C" pumps to ascertain the potential effect of this finding on the current plant condition. They were found to be well above the minimum TS requirement, it was, therefore, considered unlikely that with considsration of the instrument uncertainty, they would have failed to meet this requiremen The licensee stated that the above minor issues will be evaluated and corrective actions implemente .2 Recirculation Sorav system 2.2.1 Scope of Review Verify that the recirculation spray system (RSS), a support system for the charging system, could perform the required support functions of providing a suction source and pressure boost to the charging pumps during the recirculation phase of an acciden _-.  . _ , .-.-    _ . - . - .
 
_ __ _ _ -
_ .. _ _ _    _._______
.
i
 
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k'
j  2.2.2 Findings        t The team reviewed licensee event report (LER) 89-012, " Containment Lookage in Excess of j
Limits Due to Valve Leakage," dated July 15,198g, which documents the failure of an 10 CFR j
i Part 50, Appendix J, Type C, test of the *A* RSS pump containment sump suction, containment isolation valve,3RSS-MOV-23A. This valve was a 12-inch butterfly vdvo with a " vulcanized l  rubber seat" that had separated from its mounting surface, and as a rmit, its leakage rate i
i  exceeded the total allowable for all containment valve .
Corrective actions included the removal of the valve from the system and shipping it to Pratt, j  the manufacturer, for overhaul and a post-maintenance bcal leak rate testing (LLRT) after
;
reinstallation. The team determined from the information provided in the LER, that the valve
! ,
seat material was not changed. However, the team noted that the valve seating material used
 
may not be correct for this application in that the licensee could not demonstrate that its was l  adequate for the environment to which it would be subjected, i
l The LER described the seat material as " vulcanized rubber," a generic term applicable to a
          ,
myriad of elastomer materials. Per Calculation UR(B) 393-0, " Lifetime Radiation dose to 3RSS*MOV20A, B, C, and D," Rev. O, dated April 30,1984, the accident radiation dose to these seats was 2.2x10' reds, Another calculation, UR(B)-400, " Gamma and Beta Equipment Qualification Doses for Normal Operations, Depressurized LOCA and Pressurized LOCA," Re O, predicted the accident dose at 2.4x10' reds. Neither calculation accour ted for the 40-year lifetime dose to the valves. (It should be noted that the accident exposure would be concurrent with the accident design temperatures, maximum 280'F and exposure to borated water.)
 
These valves would be required to withstand the 40-year lifetime dose plus the accident radiation dose, temperature, and chemical environment and still be capable of performing their containment isolation functio The team determined that the valve seats were made of ethylene propylene torpolymer (EPT).
 
Per Parker 0-Ring Handbook, ORD 5700, page A2-12., no elastomer compounds, including ethylene propylene compounds, should be considered for exposures above 10' rods at room    ,
temperature conditions because of their loss of' memory,' and, hence, loss of their sealing capability. For higher temperatures, the document stated that the combination of radiation and temperature would further degrade the material A Parker Seal Company test report that was provided to the team by the licensee showed that the best EPT compounds, S604 70 and E740 70, had compression sets of 20.0 percent and 28.6 percent at 10' rods respectively, and were the only compounds that could be recommended for service at 10' rods or slightly above for room temperature applications. The Parker Seal test report also indicated that other EPT compounds had a relatively low radiation    ,
resistanca. To assure thn; an EPT was qualified, the particular compound would need to be know A Pratt Valve Company letter claimed that the material was qualified to 10' reds. However, the backup material provided with this letter addressed the material properties of tensile strength, elongation, and hardness but failed to address the compression set. Compression set is particularly important for the seals in question because, unlike static O-ring seals trapped in a seal groove that will fill the space it is trapped in, even if it completely disintegrates, these seals are not confined and the deteriorating material could rapidly fall to provide adequate sealing f-
    ,n- --, ,. , .- . . - - - - - - - - ,.,r,.m.- m .-- --- - - -e-e ,- -- . , v.. n--n v c e-
 
_ _ _ _ _ . _ . _ _ - -- __. _ _- _ _- _ .- - _ _ _ _ _ -  _ _ _ . _ _ _
,
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capability. This backup information also showed some very dramatic degradations in
        )
elongation and hardness at 10' rods with no indications of what were the maximum allowable degradations. The Pratt Valve Company vendor documents did not identify tim specific compounds of the EPT, therefere, the team could not determine the qualification and acceptability for these valve seats.
 
'
A significant contributor to the post LOCA offsite and control room doses is liquid leakage from systems outside of containment. For a single passive failure of a pressure retaining component in the RSS, such as a pump seal, the system's containment isolation valves are intended to )
l limit this leakage to the values used in the dose analyses. If, however, these valves fall to seal '
because of inadequate design, then the leakage would not be limited, and the offsite and control room doses could exceed the 10 CFR Part 100 and general design criterion / criteria (GDC) ig limits, respectivel The team concluded that the RSS pump suction containment isolation valves,3 RSS MO'!-
23A, B, C, and D, had not been demonstrated to be able to perform their design basis function of primary containment isolation under all required conditions. This finding questions the qualification of other similar containment isolation valves. The licensee was asked to provide a list of these valve The adequacy of the qualification of the RSS sump isolation valves remains unresolved pending NRC review of the licensee's further analysis related to the qualification of the valve (URI 423/g7 206-05) Main Steam System 2.3.1 Scope of Review The review of the main steam system was not within the originally intended scope of this ;
inspection. However, during the initial document retrieval phase of the inspection, the team was provided a calculation list containing Calculation 313, " Temperature Response of a Motor to the Peak Main Steam Line Break (MSLB) Transient Wnhin the Main Steam Valve Block
- (MSVB)," Rev. 2, dated August ig,1993. Because it was the intent of the team to review equipment qualification, this calculation was requested and evaluate .3.2 Finding Main steam system atmospheric dump valves 3 MSS *PV20A D were provided on asch of the main steam lines to allow for automatic relief of overpressure conditions by dumping steam to the atmosphere. In parallel with the atmospheric dump valws, remote manually operated bypass dump valves 3 MSS *MOV74A D were also installed to allow operators to dump steam in a controlled manner during plant cooldown when the nonsafety-related main condenser was not available. Upstream of both valves, on a common line, normally open remote manually
  ..
actuated isolation valves 3MM*MOV18A-D were designed to meet single failure criteria by isolating the steam line following a failure of the atmospheric or remote operated dump valves thus preventing uncontrolled plant cooldown during or following a plant transient. Both valves are electric motor-operated valve _  _ _ _
 
_ _ _ _ _ _ _ _ _.. _ . _ _      _ . _ . _ _ _ _ _ _ _
i 4            *
 
;
  -          ,
;
j .
:        12 In 1994, the licensee performed a modification under DCN NME S-009194, " Electrical
-
Equipment Qualification Master List Deletion of 3 MSS *MOV18A,-188,18C, 180 (Limitorque Operators)," dated February 25,1994, that removed the electric motors associated with the isolation valves from the environmental qualification (EQ) list. The licensee's two justifications
:
for this modification wsre that (1) these valves performed no safety related function and (2) that
{    single failure was not required to be considered for components in the penetratien area.
 
l    Subsequent to this modification, thermal insulation blankets that protected the valve motors
;    from the steam line break environment were removed.
 
;
 
10 CFR 50.49, Environmental qualification of electrical equipment important to safety for
;
nuclear plants, requires that equipment important to safety be included in a program for i '
qualifying electrical equipment, included in this requirement are safety and nonsafety related electrical equipment whose failure under postulated environmental conditions could prevent the j
accomplishment of safety functions such as controlled cooldown. The team concluded that j    given the function performed by the isolation valves and their potential exposure to the harsh j
environment of an accident, the valve operators were required to be qualified and incorporated into the EQ program. Failure to maintain the isolation valves' valve operators in the environmental qualification program is a violation of 10 CFR 50.49 requirements. (VIO 423/97-
;    206-06)
!
; Conclusions i
!    Among the more significant findings in the mechanical area was the removal of the motor-l    operated isolation valves for the main steam atmospheric dump and bypass valves from the environmental qualification program. The failure to demonstrate the qualification of the seat i    material for the RSS containment isolation valves is also a potentially significant issue. The l    control of calculations is a programmatic issue, which the licensee, as well as the team, l    identified as a concem. The team noted the document control organization's initiatives to j    address this issue.
 
!
l ElaGitiGal i
            *
i
, Onsite ac Source - Emeroencv Diesel Generator
;
j    3. Scope of Review
!
The team reviewed from the emergency diesel generators (EDG) A to 4.16 kv bus 34C, down j    to the 4.16 kV bus 34C and connections, to charging pump 3A.
 
i
 
3.1.2 Findings i
.
'
An intentional time delay of 0.2 seconds was introduced in the differential protection to allow the clearing of ground faults by the neutral breaker (see calculation 12179GM-60-03.421CB, dated
-
June 1,1984). The timer (device 162G) setpoint calculation did not include any evidence to
    . substantiate the selection of the 0.2 second time delay. The team was concemed about the i    actions introduced to delay the differential protection. This would have the potential to result in j    heavier damage to the generator in the case of a full, three phase, shor1-circuit, fault in the j    windings. In view of the fact that one of the principal advantages of the differential protection is F    its inherent ability for fast detection of faults, the team was concemed that there was no
,
I
__ ____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _  _ _ _ _ - _ _ _ _ _ _ . _ _ . _ , _ . . . _ _ . - _ - . , - - - , _ _ _ . . - . - - .
 
- - - - - - -  . - -.- . _ . - - . - _ .    - . - - - . -. __
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i          1
.      13
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objective evidence that an analysis was performed to provide justification for allowing the differential protection to trip in a longer time than that of its intended desig On the basis of input information provided by the licensee, the delay in the differential protection actuation would only occur for the case of fault not involving ground ( l.a., phase to-phase or i  three phase faults). For a fault involving ground, the neutral breator would be tripped without i  delay, thereby, ensuring that the ground fault was eliminated, and, thereby, continuing the operation of the generator, However, for phase to-phase faults and for three-phase faults, the
          !
delay would be in effect, which would subject the machine winding to the potential destructive  !
thermal and mechanical effects of very high and prolonged short-circuit currents. Because the
'  other train would be available to provide required emergency power, the team considered this an equipment protection issue rather than a safety issue. The licensee initiated an evaluation to determine the acceptability of the of the timer estling of 0.2 second ;
t
- Medium Voltano Distribution and Medium Voltage Switchamar 3.2.1 Scope of Review Medium voltage distribution and moolum voltage switchgear was evaluated from the 4,16 kv bus 34C through 480 V substation 32R,480 V MCCs 3A2 and MCC 3A1, to charging pump and ventilation system valves and motor .2.2 Findings (a) Surge protection of medium voltage equipment
          .
One Line Diagram 12179 EE 1C, indicated that surge arresters were provided on'the 345 kv side of the RSST, The team requested an evaluation of the protection afforded by these arresters for surges transmitted through the transformer windings by capacitive and inductive effects. The licensee performed a study in response to the team's question. The evaluation
 
considered as acceptance criterion the impulse withstand capability of the medium voltage  t switchgear and not the impulse withstand capability of the class 1E motors connected to the bus. The impulse withstand of the motors would tend to be quite lower than that of the switchgear, due to the limited insulation and geometric distance between the winding tum Therefore, use of the impulse withstand capability of the switchgear as acceptance criterion was judged to be nonconservative because of the potential common mode failure constituted by a single surge traveling down the transformer and causing simultaneous damage to redundant class 1E motors of the two safety-related train The licensee indicated that the evaluation of the surge protection will continue. Adequate surge protection for 1E motors is considered an unresolved item pending the licensee's evaluation of the plant's surge protection. (URI 423/g7-206-07)
  (b) Prpwn of cables against short circuits in the end load devices The team was unable to find any evidence in Specification SP-M3 EE 269." Electrical Decign Criteria", Rev.1, dated July 23,1997, that cable insulation had to be designed to survive the short circuit currents flowing to faults in downstream parts of the distribution system. The team t
,,.n , . - , .,--,r-,-r -, , .
n,-n,..  -. , -
      - ,-, ,-.----..,-n , - . - . . , - - - , ,--,.c , - --- n, ,,. . - . - - - - - - .  -. - - - . - . - . - -  _ - - . - - -
<.
.
',      14
 
found this issue significant because the plant committed to adherence to IEEE Standard 30 Paragraph 5.2.1(6) of that standard, " Protective Devices" states " Protective devices shall be i
:
provided to limit the degradation of the Class lE power systems," implying that only the part of
  - the system where the fault has occurred should be expected to become damaged. In cases of j
cables feeding loads, IEEE 308 dictates that the cable should not be damaged for faults
 
;
occurring at the load device. Faults at the terminals of load devices have a higher probability than faults in the cable itself. Therefore, the team requested information on how many other
!  safety related cables could sustain damage in case of faults at the terminals of the load
 
devices. Upon funher investigation, the team found that a number of the safety related 4 kv
:        !
;
cables for Millstone Unit 3 could be damaged by faults at the load devices. The safety <
significance of this issue is mitigated by the fact that the consequences of an assumed
{
'
electrical fault would affect only one of the two redundant electrical divisions.
 
,
}  The licensee indicated that the short circuit current levels were in the process of being re-
{  evaluated, and that this issue was a start up issue. (Ref. CR M3 g7 2358, dated February 25, i  1997). The evaluation of the qualification of unit cables to withstand short circuits is an
}  Inspection Followup item pending the licensee's fullidentification of the problem and NRC j  review of the licensee's evaluation. (IFl 423/g7 206-08)
,
i Vdc Batteries
;  3.3.1 Scope of Review i
,
'  The system was evaluated from the 480 V MCC 3A2, down to the battery charger 301A 2, to
.
125 Vdc battery 1, down to 125 Vdc Panel 38YS*PNL 1, to 38YS*PNL 1V, to fuse distribution panel 38YS*PNL17F and to 125 de loads associated with systems inspected, also from the i  480 V MCC 3A2, down to inverter INV 3, and the 120 Vac loads, and from the 125 Vdc
{  distribution panel 301 A-2, down to inverter INV 3, and the 120 Vac loads.
 
!  3.3.2 Findings
.
  (a) Calculation of battery capacity I
The battery capacity Calculation BAT-SYST-1240E3, Rev.1, did not consider the condition of plant operation involving safety injection (SI). Calculation BAT-SYST-1240E3, Rev.1, i
paragraph 6.1, " Scenario Development," page 21, contains conclusions from quoted
;_  paragraphs of the FSAR that "the design basis event (DBE) is a Loss of Power (LOP) with loss
{  of battery charger." A further statement is made to the effect that Si does not have to be i-  considered as a design condition. "SI, concurrent with a LOP and loss of battery charger output, i  is not part of the design basis for the two hour battery discharge." While it may be acceptable to disregard Si conditions for the battery 2-hour rating, it is not acceptable to do so for the 1 minute rating of the battery (Ref. Criterion 17, and FSAR, paragraph. 8.3.2.1). This is because the 1-minute rating would apply under a loss of-offsite power (LOOP) and Si conditions for the period of time until the EDG can be started and connected to the bus, and is able to provide power to the battery charger.
 
.__ _ _ __  ._ Appendix A
}
!
List of Apparent Violations, Unresolved items, and inspector Followup items l This report categorizes the inspection findings as apparent violations (VIO), apparent violations i
being considered for escalated enforcement (EEI), unresolved items (URis) or inspector followup items (IFI)in accordance with Chapter 610 of the NRC Inspection Manual. An apparent violation is a matter about which the Commission has concluded there is enough information to conclude a violation of a legally binding requirement has occurred. The violation is classified as apparent until the NRC assigns a severity level and the licensee is given the appropriate chance to respond to the NRC's determinations. A URI is a matter about which the Commission requires more information to determine whether the issue in question is acceptable or constitutes a deviation, nonconformance, or violation. The NRC may issue enforcement action resulting from l-its review of the identified URis. - An IFl is a matter for which additional information is needed that was not available during the inspection.
 
L ltem Number Finding Type Section  Title 423/97 206 01 VIO 2.1.2(a) Volation for failure to property update FSAR in 6.1.2(c) ecoordance with 10 CFR 50.71 423/97 206-02 VIO 2.1.2(b) Violation for failure to follow procedures in -
2.1.2(d) accordance with TS 6. .1.2(e)
423/97 206-03 VIO 2.1.2(c) Volation for failure to comply with ASME Code
_
5. of record as required by 10 CFR 50.55a 423/97-206-04 URI 2.1.2(f) Untosolved item to resolve check valve single failure requirements 423/97-206-05 URI 2.2.2.(2) Unresolved item concoming quellfication of RSS isolation valve seats 423/97 206-06 VIO 2. Violation of 10 CFR 50.49 for failure to have certain main steam valves in the environmental qualification program  I-423/97 206-07 URI 3.3.2(a) Unresolved item concoming the adrwacy of the unit's voltage surge protection 423/97 206-08 IFl 3.3.2(b) Inspection followup Item concoming cable short circu'.t qualification 423/97 206-09 URI 3.5.2(a) Unresolved item concoming 1-hour battery connection 423/97-206-10 IFl 3.5.2(b) Inspection followup Item to verify labeling of hydrogen analyses indicators A1
_
 
. - . . - . . - ---..-.. - - -.  -.-  - - . -
        --
,
,
.      15 The licensee indicated that the Sl conditions were enveloped by the analysis conducted and
.
'  that a clarification would be provided to the calculations. This is an unresolved item pending NRC revt " of the licensee's clarification of the battery calculation 'URI 423,sf 206 Og)      l
 
  (b) Lack of labeling on battery room hydrogen concentration indicators The hydrogen analyzers instrumentation did not indicate the units of measurement nor the identification of the battery in which they were associated. The lack of this information could result in noncompliance with Station Procedure SP 3712 NA, " Battery Surveillance Testing,"
precaution 3.1.5, which states " Read hydrogen meter prior to entry into any battery room; maximum concentration for entry is 2 percent hydrogen."
 
The licensee agreed that adequate labeling would be required, and initiated the corrective measures by Station Procedure OAg, Attachmt nt 2, " Label Request Form," to develop a label
;
for each Hydrogen Analyzer to indicate its ID, the unit of measurement and its associated battery. This is an Inspection Followup item, (IFl 423/g7 20610) Degraded grid protection 3.4.1 Scope of Review l
l  Calculations, assumptions, and protective device settings associated with the protection of plant l  equipment from the effects of low voltage on the offsite grid were evaluated.
 
l  3.4.2 Findings (a) Time Delay for Undervoltage Protection The team reviewed Calculation NL-040, and found insufficient substantiation for the selection of a long,1.8-second delay, in the output of the undervoltage protection, which, as defined,
  - operates below 70 percent voltage. There was no evidence that the selection of the time delay had been examined in light of the related consequences of allowing plant operation at zero "
voltage. If the undervoltage condition were such as to cause total voltage collapse (a level of zero volts), all 480 V magnetically held contractors would drop out, and all running 4.16 kv and 480 V motors fed from circuit breakers would slow down, with many stalling due to the collapse in input power. The drop out of contactors would cause associated safety related loads to be disconnected. In the unlikely event that voltage restoration occurs within 1.8 seconds prior to the under voltage relay completing its protective function, simultaneous motor reaccelerating and restarting would occur, which may cause over currents sufficier e. to produce undesired tripping of protective relays. The team did not find justifbation for the assumption that a 1.8-second delay would be acceptable in the case of total voltage wilapse. - The licensee stated that the delay setting of 1.6-second will be evaluated, This 6 an unresolved item pending NRC -
review of the licensee's evaluation of the 1.8-second delay. (URI 423/g7 20611)
        .
-.- ---  - . _ . - . - _ . -  _ _ _ - . - . __- .--
 
_ _ . _ _ _ _ . _  _ . _ _ - . _ _ _ _ . _ -
      .____
'
!'
I
.
16 Conclusions
:
,
There were no violations identified. Calculation inconsistencies and lack of adequate document
!  bases for calculations, setpoints and conclusions will be addressed by the licensee.
 
.
        ,
j instrumentation and Control    '
        ! Transfer from Normal Chemical and Volume Control Svatem Linnuo to l-ushe+u l j
Safetv iniection and Containment Sumn RMrMa'lan l
        '
4.1,1 Scope of Review I  Evaluate instrumentation and controls associated with the transfer from normal CCS lineups to j
high head safety injection and from high head safety injection to high-head containment ,
l  recirculation were evaluated. Documentation reviewed included TSs, P&lDs, elementary wiring j diagrams, logic diagrams, test loop diagrams, instrument uncertainty calculations, hydraulic
;
calculations, calibration procedures, calibration results, and normal and emergency operating
!
procedures. Equipment walkdowns were conducted in the main control room (MCR), auxiliary j
shutdown panels (ASP), charging room cubicles, charging pump cooling heat exchanger area, and along the charging and high-head safet. 'njection flow path .1.2 Findinas
:
  (a) Improper description of valve logic
 
,
Valves 3CHS*8804A and 3CHS*8804B open during emergency core cooling system (ECCS)
i  suction transfer from the refueling water storage tank (RWST) to the containment sump. The control logic for the valves is very similar. However, FSAR Table 6.3.3 " Motor Operated i  Valves in the ECCS System," had incorrect logic information for these valves. This was ,
,  previously identified by the licensee and was addressed under FSAR Change Request '
 
  (FSARCR) g7-MP3 323. The team noted that the interlock description for valves 3CHS*8804A and 3CHS*8804B were different. 3CHS*8804A was correct, but 3CHS*8804B was still incorror' The draft FSARCR stated 3CHS*8804B cannot be opened unless residual heat '
removkl (RHR) inlet isolation valve 3SlH'8702A, or 3SlH'8702B, or 3SlH'8702C is fully osa The correct interlock description should be that 3CHS*88048 cannot be opened unless 3SlH'8702A, or 3SlH*87028, or 3SlH'8702C is fully stu The falk a to property correct a deficiency in the FSAR is an example of a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action. (VIO 423/g7 206-12)
  (b) Inadequate Valve Test Procedure Procedure SP 3604A.5, Rev.10, " Chemical and Volume Control System Valve Operability Test," includes testing of altemate charging path valves 3CHS*190A and 3CHS*1908. These valves should only be tested during cold shutdown conditions when normal charging is isolated, but this prerequisite was not included in the procedure. The operations procedures (ops)
group initiated a Station Administration Procedures Group Report Feedback Form to add this prerequisite to the procedur '
.
*    17 The failure to correctly specify when to test CHS*190 A and B is an example of a violation of Technical Specification 6.8.1, inadequate procedures. (VIO 423/97 20613)
(c) Inadequate bypassed and inoperable status indication FSAR discrepancies in Section 7.3.1.1.5 with bypassed and inoperable status indication equipment was being corrected under FSARCR 97-MP3101, However, the approved FSARCR did not identify that the Si pumps cooling pump being inoperable caused the SI pump itself to te inoperable. CR M3 97 3134 was written to correct the FSA The failure to property correct a deficiency in the FSAR is another example of the apparent violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action. (VIO 423/97-20612)
(d) Improperty labeled auxiliary shutdown panel puchbuttons Auxiliary shutdown panel (ASP) pushbuttons for charging pump suction valves 3CHS * 1128 and 112D were labeled "OPEN/ AUTO" and "CLOSE/ AUTO," and the control switch for Si pump 3SlH.P1 A was labeled "STOP-AUTO-START." These Train A devices do not have an automatic function from the ASP because of safe shutdown concerns, although their corresponding B train components in the ASP and both trains in the MCR do have automatic functions. Resolution to this labeling problem is being tracked under CR MS 97 316 (e) Improper RWST draw-down assumption Calculation 12179-US(B)-295 Rev. 5, with Change Notice 1, "RWST Draw-down Rates and
_
Switch Over Levels," demonstrates that 20 minutes of ECCS flow is available prior to the manual transfer from cold-leg injection to cold-leg recirculation modes. The maximum charging pump flow assumMI in the this calculation and the FSAR was 820 gpm. Analysis for the Vantage 5 fuel in Westinghouse Letter No. FSSE/CWBS-1200 "MP3 FSAR Input Data,"
indicates charging flow could be 876 gpm. Therefore, the worst-case drawdown time would be marginally less than that indicated in the calculation. This is addressed in CR M3 97 295 .2 Refuelina Water Service Tank 4. Scope of Review The team reviewed instrumentation and controls associated with the refueling water service tank (RWST) SQSS*TK1 including elementary wiring diagrams, logic diagrams, test loop diagrams, isometric drawings, instrument setpoint calculations, calculations instrument uncertainty for hydraulic calculations, calibration procedures, calibration results, and emergency operating procedures. Walkdown and verification of safety-related level transmitters and switches were also conducte .2.2 Findings:
TS 3.5.4.a requires at lea:t 1,166,000 gallons in the RWST which corresponds to approximately 57 and 59 feet of water in the tank. An overflow pipe limits the tank volume to 1,186,110
. gallons, resulting in 20,110 gallon TS control ban .. _ -. -
  .-  .-. -
      - - - -  _
l'
!            ! :
            ,
Surveillance Test Procedure 3604C.1 is used to verify compliance with TS 3.5.4.a. that    '
            .
requires the RWST water level to be maintained above the minimum level of 1,166,000 gallon '
The procedure specifies the use of the wide range instruments for establishing RWST levels.
 
!
A document identified as WCAP 14353, " Westinghouse Setpoint Methodology for Indication, Control, and Protection Systems for Millstone Nuclear Power Station Unit 3 24-Month Fuel    ;
;
Cycle Evaluation," Rev. O, Tables 3109 and 3110, indicates that the wide-range RWST-level
  - instruments have an uncertainty of 5.1 percent of span for the analog corttrol room indicators,  ,
:
and a 4.2 percent of span for the computer points corresponding to RWST volumes of 61,557 and 50,594 gallons, respectively. Since the margin of instrument error is greater than the specified control band, surveil;ance test procedure 3604C.1 cannot verify the compliance with i
TS 3.5. '
l  The failure of ST 3604C.1 to ensure the TS requirement is another example of a violation of Technical Specification 6.6.1, inadequate procedures. (VIO 423/97-206-13) The team verified that existing control room annunciators which are actuated by a more accurate narrow range    ,
;
instrument could be used by control room operators to obtain accurate RWST losl.
 
! cabl=Han Content and Coggj 4.3.1 inspection Scope:
The team reviewed several calculations as well as procedures describing how instrument uncertainty calculations are performed.
 
!  -
4.3.2. Findings        .
  (a) - Errors in RWST setpoint calculation
 
There were several minor errors in RWST setpoint calculations dating from the original design period as well as in new calculations. Setpoint calculation SP 3QSS 1, Rev._1, establishes the  '
voltage level associated with the RWST high temperature alarm. The desired setpoint is 49'F, but because of an error in the calculation, the actual specifed setpoint is 3.2 volts which    >
equates to 48'F. The associated Loop Calibration Report indicates that the instrument was calibrated to the desired setpoint of 49'F, and % the incorrect value in the calculatio Therefore, there are two errors: (1) the calculadon is wrong, and (2) the plant procedures co not
;
agree with the calculation. This resulted in action request (AR) 97021209 being written to revise the calculation. Additionally, setpoint calculations SP-3QSS 5, Rev.1, and SP 3QSS-6,
;
  "3QSS TS38 RWST Temperature," Rev. O, establish setpoints for RWST temperature contro They incorrectly refer to valves 30DS-SOV26A and 268. Only one valve 3CDS SOV26 exist CR M3-97 2797 was initiated to resolve the proble (b) Errors in the RWST level interlock channel calibration    .
Calculation 3451803-1232 E3, Rev. O, "RWST Level Interlock Channel Calibration,"is a new -
calculation; it was performed using a spreadsheet and verified using hand calculations. It i
provides values used to calibrate the RWST low-low and tank empty level switches. The team i  identifed the following concoms:
'
            ,
J- ~ - . . - . . - ~ . . . - _ , . . - _ _ . _ . . _ - _ _ , _ _ . _ , _ . . . , . , _ . . , _ . , . , . . , , _.,,__._,.w.ww-...
_
          .m ..m.,._. --
 
._  . _  -__ _ _ _ _ . _ . __ _ _ _ _ . . _ _    _ l-
)l i
.
 
j  e The use of an incorrect instrument uncertainty value for tank empty level switches  ,
j  3QSS*LS56A D was a result of inadequately justified assumptions for boron
;
concentration, temperature, and tank empty level switch drift values . Equations used in i
!
the numerical calculations were not included in the calculation used in the Microsoft Excel @ spreadsheet which was not maintained as a QA document.
 
;
;  e The use of incorrect instrument drift values could result in the instruments being j  inaccurate and that not being noted during their associated surveillance test procedures.
 
!
!
Condition Report MS 97 3169 was written to address some of these problems.
:  The followup of 4.3.2a and b is identified as an inspector Followup item (IFl 423/97 20614).
 
;
- Conclusions
          :
!  The team identified a number of issues regarding instrument uncertainty Adequate
!'
consideration of instrument Uncertainties tivas raised on Unit 2 in adversa condition report (ACR) 3577. The resolution of that ACR should be broadened in order to address the concems j  raised in this area as well as those in the mechanical portion of this inspection.
 
l- Structures and Supports
.
          . System Modifications I
          '
l  5.1.1 Scope of Review i
;
System modifications including the modification that was required to install the new restrictive  -
          ,
;
orifices 3SlH*R038,039,041, were evaluated. These orifices were installed to protect the hot  .
;  and cold leg injectiren valves from erosio .
c
!  5.1.2 Findings i-          '
i Calculation Change Notice (CCN) No. 6, was performed to justify the installation of restrictive orifices 3SIH'R038,039, and 041. The installation of these orifices is accomplished by
;
removing a 14 inch section of a 1.5-inch schedule 160 pipe and installing a restrictive onfice l  with an outside diameter of 3.0 inches and an inside diameter of 0.3 inches.
 
.
The piping calculation that was performed to substantiate the acceptebility of CCN No. 6 to l  Calculation 12179-NP(B)-X100700, Rev.1, did not include the stress indices that are required  ;
. for ASME Class 1 piping enalysis in accordance with NS 3600 of the ASME Cod <
The existing calculation revision (through CCN 5) had no restrictive orifices. CCN No. 6 i  considered only the effects that the increased weight / mass (of the new orifices) would have on 5'
the system response, stresses, and pipe support loadings on adjacent supports. However, the 4:  calculation failed to provide consideration for the different stress indices as required by NS-t 3600. Since the indices for the restrictive orifices are higher than for straight pipe, the resulting stresses and interactions would also be higher. Therefore, the calculation should have provided considerations for these higher indices,    i
.
l A
- - , - - - - - - - -~,    -..-., ,~.-.---o..-w --,-.w.,-,,,r,,w--v---.-we-,,,.,
 
;-... - .- _ - . . - - - . . _ . _ - . - - - - _ . _ _ - _ . . - . - . _
 
.
 
!'
The safety significance of this finding appears to be minimal. The restrictive orifices were
        :
)  installed in a section of ASME Code Class 1 piping which experiences relatively mild transients. !
The resulting stresses, although higher, will still be less than the allowable stress limits. The '
;
>
failure to consider the stress indices required by the ASME Code is another example of a violation of 10 CFR 50.55a. (VIO 423/g7 206-03)    l
: Pining Strama Calculations
;  5.2.1 Scope of Review Review analysis procedures and guidelines that were used to perform the piping and pipe support analysis include M14g - Specification for Piping Engineering and Design ASME lil;
,
:
Code Calculation 1,2, 3; ANSI B31.1, Class 4; and Stone & Webster Engineering Corporation, i
North East Technical Memorandum (NETM) NETM-30 Procedure for Preparation, Review, Approval and Control of Power, Hydraulic, and Engineering Division Stress Data Packages and  '
l  Engineering Mechanics Division Document Large bore piping including suction piping from RWST to charging pumps was reviewed. This
!
Section included yard piping (above and below ground), piping runs in engineered safety -  -
feature (ESF) building, pipe tunnel, and auxiliary building Piping analysis was reviewed including different amplified response spectra, buried piping, and outside pipir,3 Additional large bore piping from charging pump discharge to the RCS injection nozzle for both the injection and recirculation modes were reviewed. Features reviewed included the containment penetration anchor and the ASME Code Class 1 and 2 parts of the charging pump discharge.
 
.
The following features were also evaluated:
!
e      >
small bore piping stress analysis providing cooling to the charging pumps bearings e  charging pump calcu'ations for pump foundation and nozzle loading i
e  approximately 15 pipe support calculations (both large and small bore) -
e  containment unsleeved pipir g penetration calculation for charging pump discharge i
e  RHR line rupture restraints and associated space frame  .
e  steel load reverification program and analysis for containment annulus pine rack framing 5.2.2 Findings      ,
(a) Response spectra analysis
*
_
_
The team identified a potential source of nonconservative analysis. The procedures require that only the response from frequencies under 50-Hz or from the first 50 modes (whichever
; comes first) need to be considered. In the case of a rigid system (i.e., significant frequencies .
;
L l
u__.,____ _ _ _ _  ___ - - _ - - - - -  -
 
<
          )
.
I j      21 I
greater than 50-Hz) or a very flexible system (i.e., many more significant modes than 50 Hz)
{  potential exists that not all or most of the mass will be considered in the analysis.
 
;
1  The FSAR states: "All significant dynamic modes of responses under seismic excitation with t
frequencies less than 50M: or modes less than 50, whichever is reached first, are included in j
the dynamic analysis described in FSAR Section 3.78.3.8." The concem is that these cutoff j
frequencies do not ensure that all or most of the system mass is accounted for in the seismic
]  analysis. The three areas of particular interest are: (1) stiff systems where signifcant parts of 1  the response are above the 50 Hz cutoff, (2) relatively large and soft systems where there are
;
j numerous modes less than 50 Hz such that the 50 Hz modes cutoff would not include most o the system mass, and (3) bng piping runs with an axial restraint (i.e., long rigid runs greater j  than 50 Hz). For these types of systems the pipe support loadings may be under predicted.
 
!
The licensee inoicated that they would review the calculation and evaluate the team's finding,
!          i
          :
!  This is an unresolved item pending licensee analysis of pipe support loadin (URI 423/g7 20015)
  (b) Masses not considered in analyzing containment annulus pipe rack structures  l l  The team identified a concem that the entire mass of the attached piping was not considered in i
the analysis of the rack structure ~ The rack structure was considered rigid without any
!  Justification 'or the assumption. Therefore, the rigid range accelerations were used in the rack -
!  calculation. The actual frequency of the structure could be signifmantly less than assumed
!
when all the attached pipe support mass is added to the mass of the structure. A lower frequency would result in a higher strucWal amplification of the acceleratio :
I
*
The annulus pipe rock calculation demonstrates that the rock is structurally adequate to withstand the revised pipe support lesdings generated during the igg 6, Piping Reconciliation l  Effort because of an e;evated temperature of the SI system and other piping systems attached
!
  - to the rack structure. The main function of the pipe rack structure is to provide a structure to l  - support piping, conduit and ducts in the annulus area. Therefore, the most significant loadings
-
on the structure are the forces and moments as a result of supports for piping, conduit, and
:  ducts. A review of the calculation revealed that the structure was accelerated using the rigid
;.
  . range accelerations (i.e., the structure was assumed rigid). The calculation provided no  ;
!~  justifying calculation or explanation as to why the structure could be considered rigid. This  '
'
structure, unlike a normal pipe support, is a very large frame structure with numerous attached j  concentrated masses (i.e., supported piping). It appears to be prudent to consider the potential amplification of the structuidl system when performing the structural verification calculation.
 
,
The inclusion of attached masses ava the resulting higher structural amplification should be  ,
  - considered as they will increase the total loading and stresses for the rack structure should be  t I  considered.-
t
!
The safety significance of this finding for the specific area appears to be minimal since the
.!
          ,
^
  - calculated stress for the annulus pipe rack indicates that a relatively large margin exists using the current loadings. The application of the larger amplified structural response is not expseted i  to result in stresses or loadings that exceed the current allowable limit.
 
B
-,.-m_ , - -.. ,.-.m4 . - , , a.-,m,.~, _ ., , , ,. , - _-..m , ,,..-,, , , ,,_e,-.y , ,_ ., .-.m,,.,,-.... -ya,m- my m.-, ,v.., ,
 
__ - - . - _ - - _. - _ - - - _ - _ _ - - - - -
.
.
 
This is a violation of 10 CFR Part 50, Appendix B, Criterion lil, Design Contro (VIO 423/97 20616) Thermal Exoansion of Structural Steel 5.3.1 Scope of Review Evaluate the effects of the thermal expansion of steel structures that were not considered fot
,
the elevated conts'nment temperature following an accident. Post accident temperature of the containment can be as high as 280 7 causing the structural steel to expand. That expansion will cause thermalloadings if the structure is configured such that the thermal expansion will be resiste .3.2 Findings The team identified at least one structure (pipe wh!p structuralin the steam generator cubicle area) that has the potential to generate significant loadings as a result of the restraint of the structural steel growt The structural steel reviewed has been essentially installed to provide support for pipe whip restraints. The main members consist of W14x343 members which are rigidly framed between the crane rail wall and the inner cubicle wall. The concem is the effect of the differential in thermal expansion betwecn the concrete and steel when the containment heats up to 280 The linear coefficient of expansion for steel and concrete are about the same (6.6 x 104 for steel vs 5.5 x 104 for concrete). Therefore, under steady state conditions, there will be negligible stress or loading because of differential thermal expansion. However, under transient conditions, the steel will absorb heat at a significantly higher rate than the much more massive concrete structure resulting in stresses as a result of end constrain The thermal growth will be resisted by the relatively rigid walls. The maximum generated force exerted on the steel and wall will be approximately 4,000 Kips. This force will be relieved by any clearances that may be available because of bolted joints (after slippage), and the available structural flexibility of the wall and steel structure. The resu4ing loadings will need to be checked against the wall / structural capacity to determine the severity of this concem. In addition, consideration must be given to assessing the effects of these loadings in conjunction with other loadings that may potentially be concurrent with or follow this even The safety significance of this finding is indeterminate at this time for this structure or other similar structures. This item is considered untenolved piwing licensee identification of all structures where this may be a problem and NRC review of the licensee's conclusion (URI 423/97-20617) Conclusions The principal finding of the team in this area was the inadequate consideration of the effects of accident temperatures on rigidly restrained stee .
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23    '
i Operations l Charging and Safety Inlae' inn Svatem Doeratina Praea&en
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6.1.1 Scope of Review      =
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Evaluate the charging and safety injection system emergency, abnormal and normal operatirr '
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procedures, TSs, Technical Requirements Manual (TRM), FSAR description, and accident analysis. Observe operators perform surveillance procedures,  i
.        ;
j  6.1.2 Findings (a) Inadequate Technical Specification 3. '
Millstone Unit 3 TS 3.4.3 states, in part, "The pressurizer shall be operable with a water volume
'  of less than or equal to g2 percent (1656 cubic feet)." The bases is that TS states, in part,
  "The limit is consistent with initial SAR assumptions." However, FSAR Figure 15.5 2, which
'  documents the pressurizer level transient resulting from an inadvertent ECCS injection at full powe starts at approximately 1,250 cubic feet (the upper level of the normal pressurizer
:
contr, band).
 
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On the basis of the transient from which that figure was derived, FSAR Section 15.5.1.3, '
;
concludes that follcwing an inadvertent ECCS actuation, "the pressurizer will not reach a water-l  solid condition with or without power operated relief valve (PORV) actuation, prior to 10 minutes from event initiation." (It is assumed that no operator action will be taken for 10 minutes after
'
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initiation as the operator goes through the appropriate procedures to determine whether the :
ECCS actuation was a valid or invalid signal, in which case injection should be terminated.)
 
,
,  in reviewing this transient, which would result from charging pump injection, the team identifed
;  an apparent inconsistency between the TS and the conclusion of the FSAR. If the transient
'
described in the FSAR was initiated at the highest TS allowable pressurizer level rather than at
,
the top of the normal control band, it appears likely the acceptance criteria of not reaching a solid water condition within 10 minutes could not be achieved. When this issue was raised with the licensee, the team was informed that this issue had been previously identified in CR M3 g7-322 In dispositioning the CR, the licensee concluded on the basis of information from Westinghouse, that the wording of the TS was never meant to imply that the limiting condition for operation established the initial condition for accident analysis and that Westinghouse had submitted a change request to the standard Westinghouse TS (NUREG 1431) to delete the statement oiscussed above from the TS bases. However, the team noted that it was not just the TS bases that would appear to imply the use of the higher pressurizer levelin the analysi FSAR section 15.0.7 states, in part, '' control system action is considered only if that action results in more severe accident conditions. No credit is taken for control system operation if that operation mitigates the results of an accident."
 
Even assuming a failure of the pressurizer level control system, it is extremely unlikely that the pressurizer would ever be operated at or near the TS pressurizer high-level limit for any
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significant length of time. The pressurizer high-level annunciator alarm at 70 percent and the operator response to the corresponding procedure would assure that under alllikely conditions i en inadvertent ECCS actuation would occur at or below a pressurizer level which would ensure
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the FSAR scoeptance criteria would be met. Nevertheless, given the language of the FSAR, Section 15.0.7, the licensee's CR conclusions appear to be inconsistent with the FSAR's description of the accident analysis mothe#0gy.
 
,
The licensee's proposed change to the TS to osolve the apparent inconsistency with the FSAR
! will be evaluated by the NRC. This is an inspec or Followup item. (IFl 423/97 206-18)
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(b) Technical Specification 3.7,14, " Area Temperature Monitoring," is inadequate
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During the inspection of the ventilsetion that supports operation of the charging and charging
'
cooling pumps, the team noted that the licensee has previously identified a number of problems with calculations and procedures associated with the system. For example, the licensee identified that the charging pump area temperature alarms were set at 11g 1 (which is above the TS limit of 110 T) and that design calculations show that temperatures in the those areas could reach 111 to 113 T, under normal and accident conditions respectively, when the design temperature was 104 After reviewing the TS and the licensee's responses as to why the various design, environmental qualification, and operational temperature limits are consistent (with the above noted exceptions), the team noted the followin First, the APPLICABILITY statement of the TS is improper. The TS is only applicable when the equipment is required to be operable as ststed in the TS. However, the basis of the TS is environmental qualification and not operability. Therefore, the TS should be applicable at all times given that elevated temperatures in the rooms can affect equipment qualification whether or not it is required to be operable, Second, the ACTION statement a. of TS 3.7.14 is inadequate. The ACTION statement only requires recording the cumulative amount of time by which the temperature exceeds the limit if that is less than 20 Y and for less than 8 hour Because the maximum abnormal excursion (MNE) temperature (a one time, 6-hour, maximum temperature the c,quipment is postulated to experience before an accident) for the charging pump room, for example, is 120 Y (while the TS limit for that room is 110 T), the ACTION statement fails to ensure thet the charging pumps will be maintained at or below required environmental temperatures or that such exposures will be sufficiently limited during the lifo of the equipment. Additionally, the licensee informed the team that the temperature value used in the TS is equal to that of the MNE. If the TS temperatures are set exactly et MNE, it is unclear how the 2.2 Y instrument error discussed in the TS is accounted for in the limit. Finally, the rationale for having a TS to control plant area high temperatures for EQ purposes and only the controlling operability-related low temperature for the charging pump rooms in the TRM is unclea The safety significance of these findings is low. Even with the improper setting of the charging area temperature alarms there are a number of other alarms and indicators that would alert the operator before teniperatures in the area would be a problem even under elevated ambient temperature conditions. With regard to the low temperatures presently controlled by the TRM, the action required by the TRM and the definition of system operability contained in the TS should adequately address this issue while the problems related to the TS are being resolve _ _.________ _ _ _    -._____ . _ _ _ . _  . _ _ _ _ _ _ _
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i The licensee's corrective actions to resolve the inconsistencies in temperature settings is an inspector Followup item. (IFl 423/g7 2061g)
i    (c) FSAR description of auxiliary building ventliation inconsistent with design i
i FSAR Chapter g.4.3.1, itern 12, states that air flow in the auxiliary building shall be maintained
;
from the least contaminated to the more contaminated spaces. The team noted during the LOCA recirculation mode of operation that the charging pump cubicles would potentially be the i,    most contaminated areas in the auxiliary building, in the ventilation system's winter alignment,
,
the auxiliary building ventilation system draws air from the charging pump cubicles and distributes a part of the air to other areas in the auxiliary building in lieu of the charcoal filters.
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At the ered of the inspection period, the licensee issued C3 M3 g7-03161 to resolve the FSAR
{    discrepancy associated with the flow of contaminated air in the r.uxiliary building.
:
The inconsistency between the FSAR and the closign of the charging pump ventilation is
!    another example of a violation of 10 CFR 50.71 (e). (VIO 423/g7 206 01)
!
, Charaina and Safety Iniection System Annunciator Remnanne Pr== hem i
6.2.1 Scope of Review
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Evaluate applicable emergency operations procedures (EOPs), annunciation response
;    procedures, and ops.
 
i 6.2.2 Findings
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    (a) Containment recirculation cooler SW flow annunciator response inadequate l.
 
!
During review of the annunciator response procedures, the team identified that OP 3353.MB1C,
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    ' Main Board 1C Annunciator Response,' Rev.1, Change 5,1 1B 'CTMT Recire CLR SW FLOW Hl/LO' was inaccurate.
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For Train A, Step 6.2.2 requires that if there is a SW rupture downstream of 3 SWP*MOV57A i
and 3 SWP'MOV57C refer to OP 3308 "High Pressure Safety injection and STOP 3 SlH' PIA, i    safety inje:: tion pump." For Train B, Step 6.4.2 requires similar action but conditions the
;    response with "lE Safety injection is no.t actuated."
 
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Given that the equipment in question would only be in operation during an occident or during surveillance, the Train A direction to stop a safety injection pump under M conditions is improper, in that, it directs safety equipment to be secured during an accident. The
;
inconsistency in the annunciator response procedure for the SW flow to the recirculation spray heat exchanger is another example of a vio!ation of Technical Specification 6.8.1, inadequate j    procedures. (VIO 423/g7-206-13)
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    (b) Reactor plant compoWte cooling water exchanger SW flow annunciator response inadequate d
 
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The alarm noted that setpoint of the low service water flow to the reactor plant component cooling water system (RPCCW) heat exchanger is set at 6200 gpm. The present calculations i  show inat 7125 gpm are needed to achieve safety grade cold shutdown. The licensee indicated that the setpoint (which is also the service water flow rate to be restored according to ,
l  abnormal operating procedure (AOP) 3561, " Loss of Reactor Plant Cooling Water," Step 2.b) is not based on the service water flow needs for safety- grade cold shutdown. The AOP valve is based on service water flow needs during accident conditions, with some consideration for i  setting the alarm low enough to preclude spurious alarms. Verification of adequate service
;
water flow to accomplish safety-grade cold shutdown is provided by a weekly test performed on
[  the service water side of the RPCCW heat exchangers which has an acceptance limit of 7200 gpm.
 
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While having the annunciator and AOP values set on the basis of the safety required accident flow rate ard a weekly test to verify flow set on the basis of safety-grade cold shutdown
]  requirements appeart, reasonable, there is one weakness in the procedural implementatior; of this arrangeinent. Should the annunciator actuate or the AOP need to be entered because of
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low service water flow, there is no specific guidance in the restoration steps of the AOP or in the
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annunciator response proceduie that directs the operator to subsequently verify flow that I
safety-gra.de cold shutdown has also been reestablished. As implemented, the annunciator
,
could alarm, flow could increase just enough to clear it, and the unit could remain in a condition l  without adequate flow to achieve safety-grade cold shutdown until the next surveillance (up to
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i  one week).
 
l  The inconsistency in the required service water flow between the AOP, the annunciator, and the
:  weekly surveillance test is another example of a violation of Technical Specification 6.8.1,
)  inadequate procedures. (V!O 423/g7-206.13)- Ooerator Trainino 6.3.1 Scope of Review Evaluate the charging and safety injection system operator training and design change training modules associated with charging pump cubicle and charging system valva disk modification .3.2 Findings The te,am found that the training nodules adequately addressed system modifications and operation .4 Conclusions The team found the system engineers assigned to the charging system and other interfacing systems to be strength. The assigned engineers had a thorough understanding of their assigned systems and exhibited questioning attitude .
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27 Maintenance Charoina and Safetv inime+1an Svatem Me!ntenance Preceduram 7.1.1 Scope of Revh Evaluate the charging pump and charging pump cooling pump vendor technical manuals,  :
witness maintenance performed on the charging pump cubicles ventilation duct, and conducted -
an irxlepth system walkdown. Review preventive maintenance, work requests, condmon reports, design changes, and maintenance procedures for the following components:
o Charging pumps 3CHS*P3A,B,C e Charging pump cooling pumps 3CCE*P1 A,B e Charging pump di::::harge check valves 3CHS*V46,47,48 e Charging pump recirculation valve 3CHS*MV8110 e Charging pump suction valves 3CHS*LCV1128.C.D E e Charging pump 6se.,harge valves 3CHS*MV8801A,B e Charging pump cooling heat exchangers 3CCE E1A,B
      !
7.1.2 Findings Work Order M3 96-05462 indicated that Teflon tape was used to seal the rssistance temperature detector thermowells on the charging pump 3CHS*P3B gear box. The use of Teflon tape was confirmed during the walkdown of Pump 3CHS*P38.- As a result of this finCng, the licensee issued CR M3-97 2907 that stated that Teflon is not compatible with the radiation dose rates expected in the charging pump cubicles during the recirculation phase of a loss-of-coolant accident. CR M3-96-0067 had previoustj identified the use of Teflon tape on components in the containment. Corrective action required that maintenance personnel be notified that the use of Teflon tape was not acceptable and that Teflon tape be remove from the components inside the containment. Corrects s action for CR M3-96-0067 did not address the -
identification of Teflon tape on components Iccated outside of the containment:
The use of Teflon tape on components outside containment is another example of an apparent violation of 10 CFR Part 50, Apper4ix B, Criterion XVI, for inadequate corrective actio (VIO 423/97-206-12) Maintenance-Related Information Notice (IN) Revb; 7.2.1. Scope of Rev'mw Review licensee's evaluations for the following NRC in e 92-61: Loss of High-Head Safety injection e
93-13: Undetected Modification of Flow Characteristics in the High-Pressure Safety injection System    !
e 93-42: Failure of Anti-Rotation Keys in Motor-Operated Valve.t. Manufactured by '
Velan
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e 94 5g: Accelerated Dealloying of Cast Aluminum Brone Valves Caused by Microbiologically induced Corrosion e
94-78: Recent Failures of Charging / Safety injer . , safts 7.2.2. Findings The team found that the ins were property avaluated and correctivw mc . e implemented by the maintenance program when appropriat .3 Conclusions Ovvall, based on this inspection, the team found maintenance to be strength as a number of longstanding charging system maintenance problems appear to have been corrected through aggressive actio .0 .Sptveillance Technical Soecification Survalliance 8.1.1. Scope of Review Review the surveillance procedures that implement charging / safety injection TS Surveillance Requirements 4.5.2.b,4.5.2.e.1,4.5.2.e.2.a,4.5.2J.1,4.5.2.g.2, and 4.5.2.h.
 
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8.1.2 Findings -
i (a) Air in RSS piping during cold-leg recirculation During the review of surveillance procedures that vent the charging system ECCS piping, the team found that sections of RSS piping upstream of valves 3RSS*MV8837A,3RSS*MV88378, 3RSS*8838A and 3RSS*88388 are maintained in a dry layup condition. Each of these four separate sections of piping are approximately 14 to 17 feet in length and 8 inches in diamete During the ECCS recirculation mode of operation, two of the four sections of dry piping would be aligned to the suction of the charging and safety injection pumps. The team had two concems with this confguration. First, there is no test or analysis that verifies that injecting the air in the piping to the suction of the pumps will not unacceptably degrade pump or system performanc Second, the effects of potential water hammer, as water is injected into this dry piping with its significant bends and elevation differences, had not been analyze Licensee evaluation of IN 88-23, Potential for Gas Binding of High Pressure Safety injection Pumps, including Supplements 1 through 4, did not identify that there was sir in the RSS pump d:scharge and charging and high-pressure safety injection pumps' suction piping. Although the iN and supplements did not specifically address the licensee's configuration, the potential for air and the adverse effect of air in ECCS pumps suction piping was thoroughly discussed. Given that the RSS system configuration is unique, in that it actuates from a dry condition, a thorough review of the system on the basis of the IN should have identified this issu Subsequent to the inspection, the licensee informed the team that on August 22,1985, a Design Deficiency Report (DDR) had been initiated that discussed the potential "or air in the
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recirculation spray system being injected into the suction of the charging and safety injection pumps. After reviewing DDR 641, the team concluded that the DDR fomed on another potential air problem (which has subsequenty been resolved) and failed to include the piping of
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conco The failure of the licensee to property evaluate and correct the potential for the injection of air i  into the charging and safety injection systems is an apparent violation of 10 CFR 50,
,  Appendix B, Criterion XVI. (eel 423/97-206-20)
  (b) Inadequate venting of RSS piping
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TS 4.5.2.b.1 requires that in Modes 1,2,3, and 4, ECCS piping, with the excephon of the RSS pump, heat exchanger and associated piping, be verified to be full of water by venting the ECCS pump casings and accessible discharge piping high points every 31 days. The team found that the charging pump discharge piping was vented in accordance with this TS requirement, i  However, one section of RSS pump discharge piping thet supplies the suction to the charging i
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and high-pressure safety injection pumps during the recirculation mode of operation was not vented in accordance with this TS requirement. The purpose of procedure SP3610A.3, "RHR System Vent and Valve Lineup Verification," Rev. 3, is to perform this TS requirement, and the procedure did not include vent valve 3SIL*V992. The licensee documented this procedural deficiency in CR M3-97-288 The failure to vent the RSS piping is an apparent violation of TS 4.5.2.b.1. -(VIO 423/97-206-21)
  (c) landequate procedure for verification of charging lineup i
TS 4.5.2.b.2 requires that Modes 1,2,3, and 4, the position of each valve in the ECCS system
>
flow path that is not locked in position be verified every 31 days. The team reviewed procedure SP 3608.4, "High Pressure Safety System Vent and Valve Lineup Verification," Rev. 4. The purpose of this procedure is to perform TS 4.5.2.b.2 requirements for valves in the charging system. The team found that the procedure was deficient because it did not include all charging system valves in the ECCS flow path. The procedure did not verify the position of valves 3CHS*MV8438A, 3CHS*MV84388, 3CHS*MV8438C, 3CHS*MV8468A, 3CHS*MV84688, 3CHS*LCV112D, 3CHS'LCV112E, 3CHS*V49, 3CHS*V50, 3CHS*VS1, 3CHS*V706, 3CHS*V44 and 3CHS*V707. These valves are not locked in position. The licensee stated that the position of these valves was reviewed during the monthly performance of procedure SP3604.C2,
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Boration Flow Path Verification, Rev. 4. Although the requirement to verify the position of valves is not specifically addressed in SP 3604.C2, operators would take the appropriate corrective a  actions if valves were found to be out of their required position. During the inspection, the licensee issued CR M3-97-2832 because the Surveillance Program did not correctly identify the procedure that accomplished TS 4.5.2.b.2 Surveillance Requirement The inedequate valve lineup procedure for verifying the charging system valve lineep every 31 4-days is an example of a violation of Technical Specification 6.8.1, inadequate procedure (VIO 423/97-206-13)
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30 innervice Pumn and Valve Testina 8.2.1 Scope of Review Review the procedures that test the following pumps and valves to verify that they were tested in accordance with the ASME Boiler and Pressure Vessel Code, Section XI, o Charging pumps 3CHS*P3A,B,C e Charging pump cooling pumps 3CCE*P1 A,B e Charging pump discharge check valves 3CHS*V46,47,48 e Charging pump injection header check valve 3SlH*V5 e Charging pump suction valves 3CHS*LCV1128,C, e Charging pump suction check valve 3CHS*V261 e Charging pump injection valves 3SlH*MV8801 A,B e
Charging pump cooling pump discharge check valves 3CCE*V13A,B e Cold-leg iniection check valves 3RCS*V2g,70,106,145 e
Charging pump mintflow valves 3CHS*MV8511 A,B and 3CHS*MV8512A,B 8.2.2 Findings Valves 3CHS*LCV112D,3CHS*LCV112E,3CHS*V261,3CHS*MV8511A,3CHS*MV85118,
!
3CHS*MV8512A and 3CHS*MV8512B are not classified as ASME Boiler and Pressure Vessel Code, Section XI, IWV 3420, Category A, valves, and, therefore, not tested for seat leakag These valves are required to be classified as Category A valves and tested for seat leakage because during the recirculation mode of operation, seat leakage past these valves would accumulate in the RWST causing an increase in the radiation dose in the control room and at the site boundary. TS 4.0.5 requires that purr ps and valves be tested in accordance with Section XI of the ASME Boiist and Pressure Vessel Cod TS 6.8.4.a requires that a program be established to reduce leakage from systems outside containment that could contain highly radioactive fluids during a serious transient or acciden The program is required to include integrated leak test requirements for the charging part of the chemical and volume control system. To meet TS 6.8.4.a requirements, the valves above are required to be periodicelly checked for seat leakag Further, the NRC had previously issued IN 91-56, " Potential Radioactive Leakage to Tank Vented to Atmosphere," that related directly to tHs issue. Because the licensee did not utilize FSAR leakage rate assumptions in Da analysis of the IN, an opportunity to identify this programmatic deficiency was misse Calculation 746P(R), "ECCS Leakage Outside Containment," dated May 12,1997, stated that during ECCS operation, the estimated leakage rate of fluid from the containment to areas outside of the containment is approximately 1 gallon per hour (gph). FSAR Table 15.6-g,
" Assumptions Used For the Radiological Consequences of a LOCA Analysis," states that the maximum post LOCA equipment leakage rate is assumed to be two times the maximum operation leakage rate for a totalleakage rate of approximately 2 gph. The licensee's evaluation of IN 91 56 stated that a leakage rate of 50 gpm (3000 gph) to the RWST from the ECCSs for a 24-hour period war acceptable. This leakage significantly exceeded FSAR leakage rates. At i
 
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the end of the inspection period, the licensee initiated CR M3-97-3218 to reevaluate valve leakage to the RWST, CR M3 97 3218 listed the following valves that are not seat leak teste j l
3CHS*LCV1128,C,D,E 3SlH*MV8920 3SlH'V261 3SIL*V3 t 3CHS*MV8511A,B 3SlH*MV8814 3SIL*MV8812A,B 3SIL*V9 3CHS*MV8512A,B  3S!H*MV8813 3SlH*MV8806 3SlH*V11  3RHS*V43 The failure to leak check certain valves does not meet the requirements of 10 CFR 50.55a and is an apparent violation of TS 6.8.4. (eel 423/97-206-22) Conclusions Both the RSS air injection issue and the inadequate program for control of leakage outside containment are significant because of their potentialimpact un safety. Both issues could have been detected earlier if the licensee had property evaluated generic information it had availabl The inadequacy of the valve lineup procedure used to verify, in accordance with TS, the charging valve lineup every 31 days was indicative weaknesses in the licensed CM .0 Entrance and Exit Meetinas Upon arrival onsite, the team conducted an entrance meeting to formally brief the licensee on the scope and duration of the inspection. A partiallist of persons who attended the entrance is contained in Appendix After completing the on-site inspection, the team conducted an exit meeting with the licensee on September 24,1997, that was open to public observation. During the exit meeting the team leader presented the results of the inspection. A partiallist of persons who attended the exit meeting is also contained in Appendix B. The team leader also presented the findings and answered questions concoming the inspection at a meeting with the public at the Waterford town hall on the evening of September 24,199 __ . _ _ . _ . _ _ .  . _ _ _ _ _ . _ _ _ . . _ _ _ . _ -  - _ _ . _ _ _ _ _ _ _ - . . _
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423/97 206-11  URI 3.6.2(a) Inspection followup item on licensee to evaluate 1.8 second delay in undervohage
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protection 423/97-206 12  VIO 4.1.2(a) inadequate corrective action to correct 4.1.2(c) identified discrepancies in accordance with 10 7. CFR Part 50, Appendix B, Criterion XVI
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423/97-206 13  VIO 4.1.2(b) Failure to have adequate procedures in 4. accordance with TS 6. .2.2(a)
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6.2.2(b)
8.1.2(c)-
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423/97 206-14  IFl 4.3.2(b) Inspection followup item to correct calculation oirors in accordance with CR M3-97-3169 423/97-206-15  UFsi 5.2.2(a) Unresolved item concoming pipe support analysis
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423/97-206-16-  VIO 5.2.2(b) Violation of 10 CFR Part 50, Appendix B, Criterion lli
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423/97-206-17  URI 5. Unresolved item concoming the effects of i
thermal expansion on rigidly restrained steel 423/97-206-18  IFl 6.1.2(a) Inspection followup item concoming the adequacy of the licensee's proposed change to the bases of the pressurizerlevelTS 423/97-206-19  IFl 6.1.2(b) inspector followup item conwming the
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resolution of charging pump area temperature _
inconsistencies 423/07-206-20  eel 8.1.2(a) Apparent violation of 10 CFR Part 50 Criterion XVI for failure to identify and take corrective actions for airin the RSS piping 423/97-206-21  VIO d.1.1(b) Violation for failure to vent RSS piping in accordance with TS 4.5.2. /97-206-22  eel 8. Apparent violation of TG 6.8.4 and 10 CFR 50.55a for failure to leak check certain valves j      A-2
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Appendix B Entrance & Exit Meeting Attendees NAME  ORGANIZATION Neil Cams +  NU Chief Nuclear Officer Dave Goebel+  NU VP Nuclear Oversight Dave Amerine+  NU VP Nuclear Engineer and Support Evan Wocliacott*  NEAC Terry Concannon+  NEAC Richard Laudenat*  NU-lCAVP Barry Pinkowitz'  MP3-OPS Robert Andren*  MP3-Design Thomas McCarthy  MP3-ICAVP Lead Denny Hicks *  MP3 Director
! Joe Fugere  Mgr ICAVP I Raymond Necci  NU-CAMP Michael Brothers  MP3 Martin Bowling  MP2 Harry Miller *  MP2 Gil Olsen*  MP3 William Travers +  NRC/ Director, SPO Eugene Imbro+  NRC/ Deputy Director, ICAVP,, SPO Peter Kottay  NRC/ICAVP,SPO Steve Reynolds . NRC/ Chief, ICAVP SPO Jim Luehman  NRC Team Lcader, ICAVP,SPO Victor Ferrarinl*  NRC/ Contractor Don Prevatte*  NRC/ Contractor Stephen Tingen*  NRC/ICAVP SPO Omar Mazzonl*  NRC Contractor Harold Eichenholz  NRC/ICAVP,SPO Tony Ceme+  NRC Senior Resident inspector Beth Korona*  NRC Resident inspector
* Attended Entrance Meeting only
+ Attended Exit Meeting only B-1
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. _ _ - _ _ _ _ - _ _ - _ - _ - - - - - - - _ - - - - - - - - -
.
Appendix C List of Documents Reviewed PROCEDURES
] Abnormal Ooeratino Procedures
        :
-
AOP3555, Rev 7, Reactor Coolant Leak
-
AOP3560, Rev 3, Loss of Service Water
-
AOP3561, Rev 4, Loss of Reactor Plant Cooling Water
-
AOP3566, Rev 4, immediate Boration
] h, Emeroency Ooeratina Procedures
-
EOP35 E-1, Rev.14, Loss of Reactor Secondary Coolant
-
EOP35ES 1-1, Rev.12, Si Termination
-
EOP35ES13, Rev 6, Transfer to Cold-Leg Recirculation
-
EOP35E 3, Rev.13, Steam Generator Tube Rupture
-
EOP3503, Rev 11, Shutdown Outside Control Room
-
EOP3506, Rev 5, Loss of All Charging g, Ooeratina Procedures
-
OP3203, Rev 15, Plant Startup
-
OP3252 Rev 3, Change 1, Operatur Aids
-
OP3301D, Rev 11, Change 3. Reactor Coolant Pump Operation
-
OP3304A Rev 25, Charging and Letdown
-
OP3314J, Rev 4, Auxiliary Building Emergency Ventilation and Exhaust
-
OP3326, Rev 18, Change 11, Service Water System
-
-
OP3330A Rev 13, Change 1, Reactor Plant Component Cooling Water OP3330D Rev 5, Change 1, Charging Pump Cooling
-
-
OP33531s, Rev. O, Change 2, Instrument and Service Air Panel Annunciator Response OP3353.MB1 A-MB4C, Main Board Annut-iator Response
-
OP3353VP1A-C, Main Ventilation and Air Conditioning Annunciator Response Soecial Procedures
-
SP3601F.4, Rev. 9, RCS Pressure isolation Valve Test
-
SP3604A.1, Rev 9,7/28/97, Charging Pump A Operational Readiness Test
-
SP3604A.3, Rev. 8, Charging Pump C Operational Readiness Test
-
SP3604A.5, Rev 10,8/7/97, Chemical and Volume Control System Valve Operability Test
-
SP3608.4, Rev. 2. High Pressurs Safety injection Valve Lineup Verification
-
SP3608.6, Rev.11, Safety injection System Valve Operability Tes SP36128.4, Rev 12, Change 3,4/17/97, Containment Local Leak Rate Test Type C Penetrations
-
SP3626.13, Rev 15, Service Water heat Exchangers Fouling Determination C-1
      . _ _ . . .. ----.
 
__ ____ -_____ _ __ _ _ _
.
-
' SP3646A.17, Rev, 9, Train A ESF With LOP Test
-
SP3859,Rev 3, 8/29/96, RWST Boron Concentration Nuclear Group Proced_tgag NGP 6.12, Rev.1, Evaluation of Replacement items L Maintenance Procedures
-
MP3783 EA Rev 3, Component Cooling Pumps Motor Replacement for Fire Protection g, Miscattaneous Procedures
-
EN 311121, Rev 6,12/10/96, IST Pump Operational Readiness Evaluation
-
CP 807/2807/3807AA, Rev 0, 9/13/93, Boron Analysis 2. DRAWINGS
-
12179-EM 102A-F, Reactor Coolant System
-
12179-EM 103A, Rev 15,5/30/97, Resetor Coolant Pump Seals
-
12179-EM-104A-D, Chemical & Volume Control
-
12179-EM-105A, Rev 12,6/27/97, Charging Pump Sealing and Lubrication
-
12179 EM 112A, Rev 25,2/7/97, Low Pressure Safety injection
-
12179-EM-112C, Rev 16,2/7/97, Low Pressure Safety injection / Containment Recirculation
-
12179-EM 113A&B, High Pressure Safety injection
-
12179-EM-115A Rev 19 - Quench Spray & H Recombiner
-
12179-EM-121B Rev 14 - Reactor Plant Component Cooling Water
-
12179-EM-133B, Rev 34, 4/15/97, Service Water
-
12179 EM-148A&B, Reactor Plant Ventilation
-
12179-ESK-5CS Rev 11 - Charging Pump P3A
-
12179-ESK-5CT Rev 10 - Charging Pump P3B
-
-
12179-ESK-5CU Rev 11 - Charging Pump P3C (Train A Power)
12179-ESK 5CV Rev 9 - Charging Pump P3C (Train B Power)
-
12179-ESK 5DE Rev 16 - RHR Pump P1A
-
12179-ESK-5DF Rev 16 - RHR Pump P1B
-
12179 ESK-6PK Rev 13 - VCT Outlet 1128
-
12179-ESK 6PL Rev 13 - VCT Outlei 112C
-
12179-ESK-6PM Rev 14 - VCT Outlet 1120
-
12179-ESK-6PN Rev 11 - VCT Outlet 112E
-
12179-LSK-9-1C Rev 11 - Reactor Plant Component Cooling Water
-
12179-LSK-26-2,2C, 2.2D- VCT
-
12179-LSK-26-2.3A through 2-2.3L - Charging Pumps
-
12179-LSK-26-2.A Rev 9 Seal Water Isolation
-
12179-LSK-27-17A -27-17C - Safety injection Actuation
-
12179-LSK-27-7A Rev 12 C-2
.
.__
    - - - -  '
 
.. . - - - - . - ..-..-.- _ - - _ - - -  .-
      - . - - . _. --
.
, , CALCULATIONS
)  -
P(R) 746, Rev 0,3/10/82, ECCS System Leakage Outside Containment
  -
P(R) 0982, Rev 0, ECCS Pressure Drop Calculations Based on Westinghouse Piping
,
Resistance Criteria
  -
P(R)-0983, Rev 0,4/23/84, NPSH Evaluation for ECCS Pumps RHS, SlH, CHS -
Maximum Safeguards
  -
P(R)-1057, Rev 0,12/4/84, Process Requirements for Procurement of Suction Pipe Relief Valves for CHS*P3A, B, C
  -
P(R)-1183, Rev 0,10/3/96, Operating Pressures and Temperature. Nr the CHS System to be Used As input in the Stress Data Package
  -
i P(R) 1183, Rev 0,10/28/85, Operating Pressures and Temperatures for the CHS System to be Ueed in the Stress Data Package il
  -
UR(B)-382-0, Rev 0,4/6/84, To Determine if Various Valves in the ESF and Auxiliary Building Are Less Than 1.0 X 10' Reds
  -
UR(B)-393-0, Rev 0,4/30/84, Lifetime Radiation dose to 3RSS*MOV20A, B, C, and D
  -
,
UR(B)-398-0, Rev 0,4/30/84, Post LOCA Beta Dose for Equipment in Contact with Containment Sump Water
  -
UR(B)-400, Rev 0, Gamma and Beta Equipment Qualification Doses for Normal Operations, Depress 1rized LOCA and Pressurized LOCA
  -
230, Rev 1,4/18/80, Available NPSH for Varying Recire. Pump Flow Rate - Reference Elevation
  -
232, Rev 2,4/15/85,- Floor Sumps Water Supply as a Function of Floor Water Depth
  -
249, Rev 3,5/15/92, Determination of Maximum Water Level inside containment Following a LOCA
;-  -
294, Rev 4,9/9/85, NPSH Avanabis for ECCS Pumps
  -
313, Rev 2,8/1953, Temperature Response of a Motor to the Peak MSLB Transient Within the MSVB
  -
900P(B), Charging Pump and Component Cooling Pumps Areas Ventilation System
  -
1130P(b), Temporary Ventilation for CCP Pumps Area During Loss of Primary Ventilation due to Fire
.
-  -
86-317-742 GM, Rev 1,11/15/95, Required Service Water flow to the Charging Pump
-
Lube Oil Coolers - 3CCE*E1A&B
  -
90 069-1130 M3, Rev 0, Change 2, 5/22/97, Service Water System - Summary of Westinghouse Heat Exchanger Calculations
  -
,
93-LOE-342E3, Rev 1, 8/23/93, RCP 3RCS*P1B Seal Leakoff Range Flow Loop Uncertainty & High Alarm Setpoint
- -
94-ENG-01037-M3, Rev 0, 8/8/94, Charging Pump Coolers 3CCE*E1 A&B; Required i  Service Water Flow With a 78'F SW Temperature
'
  -
3-ENG-181, Rev 0,11/14/90, Determination of Available NPSH to Charging Pumps During Gravity Boration
: -
W3-517-409-RE, Rev 0,1/4/84, Attemate Means of Cooling the Safety injection and
;  Charging Pumps
  -
SP-3CHS-12, Rev 0,4/30/85,3CHS*RV8351 Containment Penetration Z62 Relief Valve
  -
FSE/SS-NEV-1924, Rev 0,5/21/93, SWS PEGISYS Model Uncertainty
 
i C-3
 
. . ._ - _ _ _ _ _ _ _ __ __ _ _ _ _ _ _ _ ._ _ __ _
^
,
        !
  -
,
SE/FSE-C-NEU-0154, Rev 0, Change 1,6/3/97, Millstone Unit 3 ECCS Evaluation -
Future Tech Srec Change Basis
  -
SP-3CHS-9, huv 0,4/17/85, Charging Pump Suction Line Overpressure Protection Relief Valves - 3CHS*RV8501A, B & C
 
-
NSP-193-CHS, Rev 0,12/30/86, CHS System Relief Vahre Setpoints
: -
'  SP-3SlH-6, Rev 0, 7/13/85, 3SlH*RV8870 Containment Penetration Z99 Overpressure Relief Valve 1 -
SP-3CHS 2, Rev 0,1/27/84, Pressure Set Point 3CHS-PS393A, B, C Charging Pump Aux. L.O. Pump Control i -
3CCE-TIC 37 A/B, Charging Pumps Coolers Outlet Water Temperature Controller, dated
.
5/20/82
-
12176-NP(F)-677-XD, Rev. 2 inc. CCN 1 thru 3, Small bore pipe stress analysis: CVCS Containment
. -
12176-NP(F)-743-XD, Rev. 2 inc. CCN 1 thru 2, Small bore pipe stress analysis: CVCS Containment
-
12179-US(B)-2.5, Rev 0, Change 3,4/18/97, Branch Flow Rate Analysis for Safety Injection and Containment Recirculation System
.
-
12179-US(B)-311, Rev 0, Change 1,4/18/97, RSS Branch Flow Analysis with Degraded
,  Pump Cume
-
;  12179-US(B)-342, Rev 1, Change 2, Recirculation Sprav Heat Exchanger UA's i
;
-
12179-NP(F)-730, Ree 4, Small Bore Pipe Stress Analysis Charging Pump Cooling, j  Aux. Bldg. PLI CP 410113 thru 118, 515 thru 519
-
;
12179-NP(F)-729 Rev. 3, Small Bore Pipe Stress Analysis: Charging Pumps Cooling
;
Piping Aux. Bldg Iso's CP-410100,111,112,120,511,514
-
12179-NP(F) 2017,-2029,-2030, Rev. 3, Comparison of Calculated Equipment Nozzle i  Loads with Allowable Values for Charging Pumps 3CHS*P3A, B, and C
-
l  12179-NM(S)-677, Rev.1, Charging Safety injection Pumps 3CHS*P3A, B, C Embedmont Loads
-
,
12179-NS(B)-120, Rev. O, Class 2, Unsleeved Penetration Calculation (part for the high pressure safety injection inside and outside containment)
-
,
12179 NM(B)-127-JAK, Rev. 2, Inc. CCN 1, Residual Heat Removal Line Pipe Rupture
{  Restraints and Spa::e Frame
-
12179-NP(B)-X10700, Rev.1, Inc. CCN 1 thru 6, Containment Structure Annulus Piping -
ASME Class 1 & 2
-
i 12179-NP(B)-X11004, Rev.1, inc. CCN 1 thru 4, High Pressure Safety injection (SlH)
piping AUX. BLDG Piping (ASME Ill Code Cl. 2 piping)
i -
12179-NP(F)Z-110-108, Rev. 2, Inc. CCN 1, Design of Pipe Support: 3-SIH-2-PSA108
-
12179-NP(F)Z-110-013, Rev. 5, ine CCN 1 thru 5, Design of Pipe Support: 3-SlH-2-PSR013
-
12179-NP(F)Z-110-014, Rev. 3, ine CCN 1, Design of Pipe Support: 3 SlH 2-PSRO14
-
12179-NP(F)Z-110-015, Rev. 3, Inc. CCN 1 Design of Pipe Support: 3-SlH-2-PSR015
-
12179-NP(F)Z-110-017, Rev. 3, Inc. CCN 1 thru 2, Design of Pipe Support: 3-SlH-2-PSRO17
-
12179-NP(F)Z-110-018, Rev. 2, Inc. CCN 1, Design of Pipe Support: 3-SlH-2-PSR018
-
12179-NP(F)Z-110 623, Rev. 2, Inc. CCN 1, Design of Pipe Support: 3-SlH-2-PSR623
-
12179-NP(F)Z-110-624, Rev.1, inc. CCN 1, Design of Pipe Support: 3-SlH-2-PSR624
-
12179-NP-SR-007. Rev. 3, P.E. Certified Stress Report for High Head Safety injection C-4
 
. _  _ _ _ _ . _ _ . _ - - __  _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ .
, .
.
    -
,
12179-NP(FF2707-XD, Rev. O inc. CCN 1 thru 2, Comparison of Calculated Equipment
 
Nozzb Loads with their Allowable Values for: Charging Pump Cooling Pump 3CHS*P1 A
    -
    -
12179-NM(B)-534-lE, Rev.1 Refueling Water Storage Tank Design & Analysis 12179-SEO-C30.15, Rev.1, Refueling Water Storage Tank Foundation Evaluation for Outlet Loads-3QSS*TK1
    -
12179-NM(SF514-lE, Rev. O, Refueling Water Storage Tank Seismic Analysis
    -
12179-NP(BFX10717, Rev.1, Inc. CCN 1 thru 7, Pipe Stress Analysis: Containment Structure Annulas Piping ASME Class 1& 2
    -
12179-NP(BFX10730, Rev.1, Inc. CCN 1 thru 4, Containment Structure Annulus Piping SlH to Cold-Leg Loop 3 ASME Class 2 & S Piping
,    -
12179-NP(BFX10702, Rev.1, Inc. CCN 1 thru 9, Containment Structure Annulus Piping,
,
'
High Pressure Safety injection to Cold-Leg Loop 2 & 4
    -
12179-C24.1, -C24.23,2/14/78. Buried Piping-Conc. Encasement EC-24A,B and Service
,
Water Encasement at Col. Line 44.3 ESF building
    -
 
    .12179-P(RF1179, Rev.1, Documentation of Operating Temperature and Process used as input to SDP-SlH
;    -
03705-S52.31, Rev. O, Steel Load Re-verification Program Annulus Pipe Rack Framing
 
Analysis Containment Building 4 PLANT DESIGN CHANGE REQUESTS
.
    -
DCR M3-97008, Rev 0,3/18/97, Replacement of ASCO SOVs
    -
DCR M3-96077, Rev 0,3/24/97, ECCS Orifices and Throttle Valves
;
    -
DCR M3-97527, Rev 0,2/28/97, Replacement of ECCS Flow Measurement Orifices
'
    '3SlH'FE917, 918, and 921 with Similar Calibrated Onfices
    -
'
PDCR MP3 93-163, Rev 0,9/13/93, Modification to Swing Check Valve 3CHS-V165
    -
PDCR MP3-95-020, Rev 0,4/17/95,3 CMS *MV8507 A/B Disk Modification
    -
PDCR MP3-95-003, Rev 0, 3/2/95, RCP No 1 Seal injection Valves Replacement
    :
.
RIE # PSE-MP3E 94-101,2/1/95, Replacement of Relief Valve 3CHS*RV8119 I    -
i PDCR MP3-93-126, Rev 0,8/7/93, CHS & RCS Valve Yoke Bolts Replacement
    -
PDCR M3-91-067, Rev 0,4/4/91, Change Auto Makeup Setpoint from 120 gpm to 80 e    gpm-
-    -
-
MMOD M3-96606, Rev 0,7/17/97, Revision of Lube Oil on Charging Pump Sealing and Lubrication P&lD j
    -
PDCR MP3-89-095, 6/4/89, "C" Charging Pump Orifice Installation ADVERSE CONDITION REPORTS ACR 3577 (unit 2)
:
, SPECIFICATIONS
,
:    -
Engineering Specification SP-M3-EE-0333, Rev 0, Environmental Conditions for h    Equipment Qualification
    -
Specification 2362.200-164,1/10/85, Mechanical Equipment Environmental Qualification i LICENSING DOCUMENTS i
;    - FSAR Chapters 6,9, and 15
    -
FSAR Section 1.2.10, Engineered Safety Features
';
s C-5
_ _ _ _ _ _ _ _ _ _ - _ _ _ . _ _ _ _ _ _ . _  _. - _ _ _ _- _ _ _ _  _ _ _ _ _
 
    .-    -
.- -_- . - __ _  -
        )
-
        !
I
.
  -
FSAR Table 1.3-1, Design Comparison    !
  -
FSAR Section 1.8 Conformance to NRC Regulatory Guides
  -
FSAR Section 1.8N, NSSS conformance to NRC Regulatory Gukies
  -      ,
FSAR Section 3.1, Conformance with NRC General Design Criteria  l
  -
FSAR Section 3.2.4, Other Classification Systems
  -
FSAR Appendix 3B, Environmental Design Conditions
  -
Millstone 3 Technical Specifications
  -
FSAR Section 5.4.1.2.2, [RCP) Pump Assembly Description
  -
FSAR Section 5.4.1.3.10, Shaft Seal Leakage
  -
FSAR Section 5.4.1.3.11 Seal Discharge Piping
  -
FSAR Section 5.4.3, Reactor Coolant Piping
  -
FSAR Question Q480.3 and Answers
  -
Millstone 3 Technical Requirements Manual, Change 39 LICENSEE EVENT REPORTS (LERs)
  -
LER 89-12,7/5/89, Containment Leakage in Excess of Limits Due to Valve Leakage
  -
LER 89-022,10/25/89, Valve Stroke Time Testing in the Wrong Direction Due to Transcription Error
  -
LER 91-011,5/10/91, Both Trains of high Pressure Safety injection System inoperable Due to Relief Valve Leakage
  -
LER 92-O'F "2/3/92, Both Trains of High Pressure Safety injection Inoperable
  -
LER 93c . 6/21/93, inadequate Surveillance Testing of High Pressure Safety injection Check Valves
  -
LER 94 007,5/13/94, Violation of Engineered Safety Feature Response Time of Quench
,  Spray System
  -
LER 94-010, 8/24/94, Both Trains of Charging inoperabia Due to Procedural Deficiency
  -
LER 96-028, & 01,12/13/96, Potential Overcooling of Charging Pump Lube Oil System Due to Failure of Air-Operated Temperature Control Valves
  -
LER 96-029,9/29/96, Functional Deficiency in the Setting of the Emergency Core Cooling System Throttle Valve Positions
  -
LER 96-031,10/6/96, Potential Failure of Safety Related Control Valves Due to Failure of Non-Qualified Air Regulators
  -
LER 96-032,10/9/96, High Pressure Safety injection Relief Valve Piping Hydrostatic Test Non-Compliance with ASME Code Due to Personnel Error
  -
LER 96-044,12/4/96, Qualification of Containment Systems Following a Design Basis Accident
  -
LER 97-008, 2/22/97, Failure to Enter Technical Specification 3.0.3 Action Statement for MSIV C;osure
  -
LER 97-018, 3/7/97, Technical Specifcation Parameter Compliance OTHER DOCUMENTS / MANUALS REVIEWED
,
  -
Henry Pratt Company Drawing of Valve 3RSS-MOV23A, B, C, and D, S&W Dwg. N , Sh 110, Rev 1
  -
Lonergan "LCT" Series Relief Valves Vendor Manual, Document ID # OIM-186-1 A, /Rev 1G 8/2/96
  -
Charging / Safety injection Pumps Vendor Manual, Document ID # 25212-001-019A, Rev L,5/20/97
'
C-6 a
_ _ _ _ _ _ _ _ _ __
 
_ - -. . . - - . - - .. . . - .= - . .- .. -
      --- . _ . - .
.
-  -
'
 
Electrical Equipment Qualification Program Manual, Rev i
  -
Memo from: D. T. McDaniel to M. P. Pearson: Operability of CHS Pumps Without Fans !
,
3HVR*12/14 in Service in Mode 5, dated October 10,1992\
l  -
Letter from: W. J. Johnson, Westinghouse to R. C. Jones, NRC: Use of 2700*F PCT t Acceptance Limit in Non-LOCA Accidents
  -
PORC/SORC Meeting Minutes - March - August 1997
  -
Fire Protection Evaluation Report
  -
Technical Bulletin 79-6, Pacific Centrifugal Charging / Safety injection Pumps, 9/25/79
  -
Installation, Operation and Maintenance Manual No. 25212-001-019A, Charging / Safety injection Fumps, Rev. L
  -
Installation, Operation and Maintenance Manual, OIM-042,002A, Charging Pump Cooling Pump, Rev 2F
  -
.
Work Order M3-94-03094 dated 7/27/95, Replace Antirotation Key on 3CHS*MVB111 A
  -
Work Order M3-9319263 dated 8/30/93, Replace Antirotat5n Key on 3CHS*MV811D
:
+
E
 
.
      .
 
C-7
 
.- _ ... _ . . . _ . _ . . _ _ . . . . . . _  . . _ _ _ . . _ _ _ . . _ _ _ _ . _ _ _ _ - _ _ _ _ . _ _ _ _
.
          ?
. il        Appendix D  -
,
;
List of Acronyms
 
ACR(s) adverse condition resort (s)
    -AOP(s) ebnormaloperating procedure (s)
'
A action request '
ASME American Society of Mechanical Engineers
:    ASP auxiliary shutdown pane!
BOM bill of materials BTP Branch Technical Position i
.
CCE charging pump cooling -
l    CCN celculation change notice CF Code ofFederalRegulations i
CR(s) condition report (s)
DBE design basis event DDR design deficiency repost
.
. ECCS emergency core cooling system
!    EDG emergency diesel generator
 
eel escalated enforcement item
'
EEQ electrical equipment qualification EOP(s) emergency operation procedure (s)
'
. EMI/RFT/ electromagnetic / radio frequency interference / electrostatic discharge EP ethylene propylene EPRI Electrical Power Research Institute
,
'
EPT ethylene propylene terpolymer EQ equipment qualification
;    ESF engineered safety feature l    FSAR .
    - Final Safety Analysis Report i    FSARCR Final Safety Analysis Report Change Request
;    GDC general design criterion / criteria F    gph gallons per hour gpm gallons per minute
:
HP high-prescure
!
'
ICAVP Independent Corrective Action Verification Program IEEE- - Institute of Electrical and Electronics Engineers
'
 
IN(s) Information Notice (s)
j  iPE individua! plant examination
;  . lST inservice testing
!  LCO limiting condition for operation l'        D-1
  -
1, I          ;
          , ._ .  ,~  _ - . _ _ . _ _ . _ .
 
_- . .- - . . . - -. _. -.
t a
LER(s) licensee event report (s)
'
LLRT localleak rate testing LOCA loss-of-coolant accident LOOP loss of offshe power LOP loss of power MAE maximum abnormalexcursion MCC motor control center MCR main control room MNE maximum normal excursion MOV motor-operated valves
,
MSLB main steam line break MSVB main stessa valve break NGP(s) Nuclear Group Procedure (s)
NNECO Northeast Nuclear Energy Company NRC U.S. Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation
:
OP(s) operating procedure (s)
P&lD piping & instrumentation diagrams PMMS preventive maintenance management system PORV(s) power operated relief valve (s)
'
QA quality assurance
,
RCP(s) reactor coolant pump (s)
RCS reactor coolant system RG Regulatory Guide RHR residual heat removal RIE Replacement item Evaluation RPCCW Reactor plant component cooling water RPS reactor protection system
, RSS recirculation spray system RSST reserve station service transformer RWST refueling water storage tank SI safety injection SSFl safety system functionalinspection SW service water TR(s) trouble report (s)
 
TRM Technical Requirements Manual TS(s) technical specification (s)
URl(s) unresolved item (s)
: VCT volume control tank D-2
!


i i  1
Millstone -ITPOP Project Omco P. O. Box 0630 Niantic, CT 06357-0630 Mr. Daniel L. Curry Project Director l Parsons Power Group inc.
. Vdc volts, direct-current i


  .
2675 Morgantown Road l Reading, PA 19607 l
D-3
Mr. Don Schopfer Verification Team Manager Sargent & Lundy 55 E. Monroe Street              !
Chicago,IL 60603 i
_ _ _ _ _ _ _ _ _ _ . - _ _ _ - - _ _ _ _ _ _ _ _ . ______      _  _
}}
}}

Latest revision as of 20:36, 15 December 2021

Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-423/97-206.Actions Will Be Examined During Future Insp
ML20248L910
Person / Time
Site: Millstone Dominion icon.png
Issue date: 03/05/1998
From: Reynolds S
NRC (Affiliation Not Assigned)
To: Bowling M
NORTHEAST NUCLEAR ENERGY CO.
References
50-423-97-206, NUDOCS 9803250013
Download: ML20248L910 (3)
v • d • e


Text

l a

March 5, 1998

.~~~

fbL 1

  • Mr. M. L. Bowling, Recovery Officer, Unit 2 l c/o Ms. Patricia Loftus, Director Regulatory Affairs for Millstone Station RGQgED.

Northeast Nuclear Energy Company P.O. Box 128 19?S 193 p fil 2: 08 Waterford, CT 06385 PUBLlc DOCUMCIT ROOH SUBJECT: INSPECTION REPORT NO. 50-423/97-206

Dear Mr. Bowling:

This responds to your February 2,1998, letter, regarding the Notice of Violation that was transmitted with the December 5,1997, inspection report.

Thank you for irforming us of the corrective and preventive actions documented in Attachment 2 of your letter. These actions will be examined during a future inspection of your license program.

Your cooperation with us is appreciated.

Sincerely,

/s/

J.Nakoski for Steven A. Reynolds, Chief ICAVP Oversight Branch Special Projects Office Docket No. 50-423 cc: See next page Distribution:

PUBLIC/ Docket File SPO R/F J. Andersen M. Callahan, OCA R. Correia B. McCabe, OEDO S. Dembek \d G. Imbro \

\

D. Mcdonald P. McKee D. Screnci, OPA [0g DOCUMENT NAME:P: reply.JI Ta receive a copy of this document, indicate in the box "C" copy w/o attach /enci"E" copy wfattach/enci"N" no copy OFFICE ICAVP 1 Q ICAVP / C:lCAVP , C NAME JLuehmanh PKo SReynoldM h#

f DATE 3/[/98 198 3/[/98 I OFFICIAL RECORD COPY 9003250013 900305 -

PDR ADOCK 05000423 G PDR

f

Northeast Nuclear Energy Company Millstone Nuclear Power Station Unit 2 cc:

Lillian M. Cuoco, Esquire Mr. F. C. Rothen Senior Nuclear Counsel Vice President -Work Services Northeast Utilities Service Company Northeast Utilities Service Company P. O. Box 270 P. O. Box 128 H:rtford, CT 06141-0270 Waterford, CT 06385 Mr. John Buckingham Ernest C. Hadley, Esquire Department of Public Utility Control 1040 B Main Street Electric Unit P.O. Box 549 10 Liberty Square West Wareham, MA 02576 New Britain, CT 06051 Mr. John Streeter Mr. Kevin T. A. McCarthy, Director Vice President- Nuclear Oversight Monitoring and Radiation Division Northeast Utilities Service Company Department of Environmental Protection P. O. Box 128 )

79 Elm Street Waterford, CT 06385 H:rtford, CT 06106-5127 Mr. David Amerine Regional Administrator, Region i Vice President - Nuclear Engineering U.S. Nuclear Regulatory Commission and Support 475 Allendale Road Northeast Utilities Service Company King of Prussia, PA 19406 P. O. Box 128 Waterford, CT 06385 First Selectmen Town of Waterford Mr. Allan Johanson, Assistant Director Hall of Records Office of Policy and Management 200 Boston Post Road Policy Development and Planning Waterford, CT 06385 Division 450 Capitol Avenue - MS# 52ERN Mr. Wayne D. Lanning P. O. Box 341441 Deputy Director of Inspections Hartford, CT 06134-1441 Special Projects Office 475 Allendale Road Mr. M. H. Brothers King of Prussia, PA 19406-1415 Vice President - Operatior s Northeast Nuclear Energy C9mpany Charles Brinkman, Manager P.O. Box 128 Washington Nuclear Operations Waterford, CT 06385 ABB Combustion Engineering I 12300 Twinbrook Pkwy, Suite 330 Mr. J. A. Price l Rockville, MD 20852 Unit Director- Millstone Unit 2 Northeast Nuclear Energy Company I Senior Resident inspector P.O. Box 128 Millstone Nuclear Power Station Waterford, CT 06385 cla U.S. Nuclear Regulatory Commission i t P.O. Box 513 Nirntic, CT 06357

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NorNeast Nuclear Energy Company Millstone Nuclear Power Station Unit 2 cc:

Mr. B. D. Kenyon Chief Nucler Omcer- Millstone Northeast Nuclear Energy Company P.O. Box 128 Waterford, CT 06385 Citizens Regulatory Commission ATTN: Ms. Susan Perry Luxton 180 Great Neck Road Waterford,CT 06385

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l Deborah Katz, President Citizens Awareness Network P. O. Box 83 l Shelburne Falls, MA 03170 The Honorable Terry Concannon Co-Chair Nuclear Energy Advisory Council Room 4035 Legislative Office Building Capitol Avenue H:rtford, CT 06106 Mr. Evan W. Woollacott Co-Chair Nuclear Energy Advisory Council 128 Terry's Plain Road Simsbury, CT 06070 Little Harbor Consultants, Inc.

Millstone -ITPOP Project Omco P. O. Box 0630 Niantic, CT 06357-0630 Mr. Daniel L. Curry Project Director l Parsons Power Group inc.

2675 Morgantown Road l Reading, PA 19607 l

Mr. Don Schopfer Verification Team Manager Sargent & Lundy 55 E. Monroe Street  !

Chicago,IL 60603 i

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