IR 05000201/2003031
| ML20128P376 | |
| Person / Time | |
|---|---|
| Site: | Waterford, West Valley Demonstration Project  |
| Issue date: | 05/23/1985 |
| From: | Constable G, Crossman W, Flippo T, Andrea Johnson, William Jones NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20128P342 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.D.2, TASK-1.G.1, TASK-2.B.1, TASK-2.E.3.1, TASK-2.E.4.2, TASK-TM 50-382-85-05, 50-382-85-5, NUDOCS 8506030518 | |
| Download: ML20128P376 (20) | |
Text
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APPENDIX A - ,- . 9,
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, U. 3. NUCLEAR REGULATORY COMMISSION ,
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., 't REGION IV
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NRC; Inspection Report: 50-382/85-05
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CP: CPPR-103 License: NPF-26 ;.
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-Licensee: Louisiana Power & Light Company (LP&L)
. 142 Delaronde Street [ , ,( '
f: ' New Orleans, Louisiana. 70174 s ,
: Facility'Narne: Waterford Steam Electric Station, Unit'3- t ' Inspection At's Taft, Louisiana ,
Inspection Conducted: February 1 through March 31, 1985
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. Inspectors: a@ M x; T /4 1EI5- ,
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G. L. Constable Da(e / '
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_ Senior' Resident Inspector d, d 50f s gs. ,g _ T. A. Flip'pb, Resident Inspector Date
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. R. Johnson, Reactor Inspector Date'
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iAssisting Personnels;
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; N k N ili Nuclear' Training. Instructo *
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J. J. Harrison,' Chief, Engineering Branch,' Region III , f' , s(paragraph 11)- ,,
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4 ,t , , , ,, D. Tomlinson,~ Reactor Inspector
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Approved:~ '/ 2 m - _ 3-
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W 4. Crossman, Chief, Project Section B, DAte '
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. Reactor Project Branch I c .: Inspection ' Summa ry - 'I -
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~ , Inspection Conducted February 1 throuah March 31, 1985 (Report 50-382/85-05) ,
LAreas Inspected: . Routine, announced inspection oft (1) startup test procedure i
' . review; (2). review of licensee'signific. tat construction deficiencies; (3),Three LMile Island(TMI) open items;((4) shift turnover review; (5) followup on ; , 3 previous'NRC Inspection findings;-(6) testiresults evaluation; (7) Phase III' '
s , .. test procedure' witnessing; (8) followup on allegations; and (9) quality w; - l assurance (QA) personnel qualifications. The inspection involved ' l 617l inspector-hours onsite by four NRC inspector ;
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, iResultst ~ Within the areas inspected, no violations or deviations were identifie .
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i DETAILS Pers'ons Contacted
. Principal Licensee-Employees *R. S. Leddick, Senior Vice President, Nuclear Operations ' *R. P. Parkhurst, Plant' Manager, Nuclear , =*T. F. Gerrets, Corporate,QA Manager *P. N. Dackes, Operations QA Hanager *L. F. Storz, Assistant Plant Manager, Operations and Maintenance S. A. Alleman, Assistant Plant Manager, Plant. Technical Staff 0. D. Hayes, Operations Superintendent *J. N. Woods, Plant Quality Manager *R. G. Pittman, Operations QA Audit Supervisor *G. E. Wuller, Onsite Licensing Coordinator *K. L.'Brewster, Onsite Licensing Engineer J. R. McGaha, Maintenance-Superintendent <
L. M. Meyers, Assistant Operations Superintendent
*Present at exit interview In addition to the above personnel, the NRC inspectors held discussions ' .with various operations, engineering, technical support, and administrative members of the licensee's staff.'
) Plant Status The U.S. Nuclear Regulatory Commission issued Facility Operating License NPF-38 together with Technical Specifications and Environmental Protection Plan to LP&L for the Waterford Steam Electric Station, Unit 3 on March 16, 1985. License NPF-38 authorizes operation of the Waterford Steam Electric Station, Unit 3, at core power levels not to exceed 3390 megawatts ~ thermal (100 percent power), and supersedes License NPF-26, issued on December 18, 198 , At,the end of the inspection period the-licensee was in the process of conducting startup testing at the 20 percent power leve . Startup Test Procedure Review The NRC inspectors reviewed the startup test procedures for performing power ascension testing of the plant. The procedures were reviewed for technical content, compliance with the Final Safety Analysis Report (FSAR), and ccepliance with licensee's administrative procedures. The startup test procedures reviewed are listed below: SIT-TP-716 Core Performance Record
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@ ;7 Ta.] :i (Closed)'SCD-37, Temperature Detectors - (RTDs) Failure
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j Z" ' . o On July 27l 1981, Combustion Engineering,'Inc.;(C-E). notified'EBASCO
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result in's nonconservative thermal margin / low pressure.(TN/LP) trip , i f 'setpoint and could permit ~possible operation in excesssof departure
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'- from nucleate boiling (DNB)311mits. The cause'of the erroneous RfD ..' '
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; signals was due to resistance changes in the~RTD. circuitry as a ,,
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result of, corrosion degradation of the RTD leads at the thermal block-m- w c . connection in the head of the RTD assembly. The corrosion was-M'J, Jc 'l
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metals in the RTD leads and the thermal block i
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A total ~of 33 Rosemount RTDs were utilized at Waterford 3. 0f these
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33',16 were safety-related R' ids" installed in steam generators hot and ! n ' cold ~1ess which provide' input to the plant protection 1 system., Eight j
?<; - of these 16,Rosemount RTDs were replaced with Weed RTDs which.left
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," 8. dual element Rosemount RTDs which were required'to be' ;
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inspector reviewed the licensee's u n - -
'o 2- environmentally corrective action and' sealed.; supporting The NRC, documentation for resolution oflthis y +
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> deficiency. This issue is: considered appropriately resolve '
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j ;.i . h g ' b '. ' -(Clos [ed)' SCE-80,- Unsatisfactory Strokina of EW Pump Turbine Steam l 9 m- p < Supply Shut Off Valves .
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lDuring bot' functional testing (HFT), automatic operation of & i' g (. ",ivalves'2MS-V611A and 2MS-V612B were found to be unsatisfactory.
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.s Stroking of;the valves was not smooth, and excessive force was needed
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,The licensee's corrective action was to replace the pneumatic n W */ operators for.the valves 2MS-V611A and 2MS-V612B with D-C motors.- g ,
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4 support the minimum emergency feedwater (EFW) control "lo-lo"
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, setpoint analyses. This item is close '
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. No' v'iola'tions or deviations were identifie .
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3' 'Following the cool down after HFT, it was discovered that the letdown b jccontainment isolation. valve (CVC-103) wa's stuck in the open positio '
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. seat'which reduced the~ clearance between the seat and the gate, Sq $4 *thereby causing the gate'to stick ope '
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L'C. A The-licensee's corrective action was to repair the valve with new ' - k.; si
' ' w.6 N ' O [ seats which are sized for maximum interference fit in the seat pocke ~
TLd~ f inforder to eliminate'the previously experienced seat float.- The,,' .' .. ',T y;g 3 m qg-- : closure'ofithe valve has'since been satisfactorily tested in the' cold
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' approximately 1.5 seconds. FSAR commitment for closure of valve
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; 0n March 3,'1985, at 0105 hours,' when the plant"was at normal l operating' temperature and pressure, the valve was, stroked once. 'The
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- closure time'was 1.5 ' seconds which is within'the allowable limit .
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;The NRC inspectors held discussions with_the licensee concerning the
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: operation of this valve and the licensee coassitted to the following
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a0 actions in a letter dated March 8, 1985,5 to Mr. Robert D. Martin.
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"The valve will be stroked in 100*F decrements during the next two
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~ cooldowns starting at about .100*F below normal operating temperature >
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. , cand pressure.' The surveillance conducted during cooldown will be in X accordance with' procedure OP-10-001. 'If the valve fails to close -
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'during testing, Technical Specification 3/4.6.3 will be~ evoked;and .
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, the; plant personnel will, respond accordingly.- ,
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n .A~1ong term solution will be sought to relieve,the plant from4 this '
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O"- , valve cycling exercise; should cycling prove'to be the necessary ! W[ > ,
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- No violations,or deviations were identifie '
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,# Three Mile Island (TMI) Open Items
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'd o P - J IThe' licensee l has installed ' the reactor [ coolant gas venUng system '
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] (RCGVS)L ,,This system allows:for the remote: venting of noncondensable qq Mr - t gases,ithrough either the reactor vessel head vent or pressurizer -
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W gyt i ' Lg steam , space vent,1which may collect in the RCS following certain post , 3/ y = - , g y (accident ~ conditions. :The RCGVS is described in' detail;in FSAR,. ,f ,9 y -Section 5'4.15,'"Reastor Coolant Gas VentjSystem'." The NRC inspector'
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Q'f S %' , h lf u = , : 4 reviewed the design basis for the RCGVS.and' determined that the <
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'nsdesign t-meets"the requirements.of NUREG-073 ,
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.;,jt i ., , /In addition, the.NRC inspector reviewed Emergency Operating s
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, Procedure OP-902-002f Revision 0, ' Loss of Coolant Accident Recovery ,
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J' D F Procedure,'.' and OP-902-003, Revision 0, " Loss of Forced Flow 'G' WM i
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Recovery Procedure." The'NRC. inspector determined that th ~
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- procedures provide adequate information and guidance to the operator jfr
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for. initiating and terminating RCGVS' usage. This item is considered
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y N* (Closed), NUREG 0737 (TMI Item I.G.1) Trainina Durina Low Power
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. The licensee-has. revised FSAR Sections 14.2.12.3.25 and.14.2.12.3.34
" Lto' reflect that natural circulation testing and training will be:
#,, , - conducted in Mode 3.under conditions of actual: decay heat-removal
_ ' following a reactor trip from 80% of rated thermal' power, level. - This
. test ~ is to be . conducted in. conjunction:with the ' loss of ' flow tes . '
The'NRC staff evaluated the above proposal-in NUREG-0737
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-Supplement 5,<"SER Related to the Operation of Waterford SES, Unit;
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N ," dated' June 1982. The NRC staff concluded that the testing
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~ and training objectives stated'in-the NRC staff's position.can be' ,
0- =i-readily accomplished during post-80 percent power trip condition ,
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The NRC. inspectors wi11' verify that,the licensee' meets the NRC s .
' staff's objectives for training,during' low-power testing during -
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,n . Mode 3' natural circulation operation. . This item is considered '
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a,yp, c .: [ Closed). NUREG-0737 (TMI Item II.E.3.1) Emeraency Power Supply for . HiO s '
'Pressuriser Heaters s ,,
s 'F' q The licensee has completed work on the pressurizer heater electrical y NJ supply which provides the capability to supply from either the' ' 4 nj 9't~'<~u .offsite power source or the emergency power source (when ~offsite'
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7, acg . > Epower is not available). A redundant group of pressurizer ,l' . a; , ' , proportional heaters,'each with a!capacityfof 150 kW, may receive
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,'The proportional, heaters-are powered from _the 480 V nonsafety
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383-S. -These safety-related busesare' designed to trip when the '
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receive-alloss of;offsite power (LOOP) or safety injection actuation
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heaters are~ protected by safety Class 1E circuit breakers. The 'N c nonsafety AJ 480 V switchgear breakers can be reclosed manually from N ';f e v x ' CP-1.once the emergency diesels"are on line, as long as a SIAS signal
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y , " Degraded Electrical Distribution Recovery Procedure," Section E,- ' Cl x ' provides the necessary guidance to direct the operator on when and
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, emergency buses. ~ In addition, LPE's Surveillance Procedure
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OP-903-28, Revision 0, "Pressuriser Heater Emergency Power Suppiy '
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Functional'. Test,"_provides instructionsLto demonstrate the ~
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T' . f + y operability of the pressuriser heater. emergency power. supply, as rrequired by'LPE Technical Specificat. ions. This item is considered
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: closed, n- n : ,,;.' ..
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' J(ChosAd)iNNREG-0737-(TMIItemII.E.1.1)AuxiliaryFeedwaterSys' tem *i
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' t, The licensee reevaluated the EW systes using event tree and'f'ault g
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- a vt . treeilogic techniques for potential EW sfailure under various loss of '
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, usin,feedwater transients. The results of the evaluation, stated in >
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FSAR Appendix 10.4;9.B. were reviewed by the NRC inspectors'and M '
@~ 'n determined to adequately address failures,which could result from '
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, 1 human: error, common 'causes, single point vulnerability, testing, and '
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1LPE's. staff also compared the EW design w'ith the requirements of.. p' j
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; Standard Review Plan (SRP) 10.49 and Branch Technical Position Dml ' * ; g
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(BTP)'ASB'10.1. . The.NRC inspector verified that'tt.e Licensee's- .
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- V_ evaluation, stated in FSAR Appendix 10.3.9A;ms Tabla 10.3.9Aa14 is'in j;
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V' a b [ An: analysis of the Waterford 3'EW against:the NRC's requirements for ,
i. . % EW flow to the ' steam ' generators to ensure adequate removal of core
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p" h if ? fdecayheatis'documentedinFSARAppendix1 10.4.9A as Table 10.4.9A- t
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the licensee has . adequately considered ' plant transient and accident P- , conditions:which could affect removal of' reactor decay hea .'
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requirements of NUREG 0737_ for the evaluatioi; of the EW system. .In :
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addition, theJ11censee has met the NRC's short and long term '
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recommendations is reflected'in the Waterford 3 Technical- r D, # ce*. Specifications and LP&L't;0perating Procedure OP-9-003, " Emergency"b '
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.Feedwater," Revision 3. This item is' considered closed.
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A WC' ' 4 consoles in the control" room,. technical support center (TSC), and
, l emergency operations facility;(EOF) as required by NUREG-073 '
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' Nowever, the computer at Waterford 3,' . specifica11y CPU-3, . cannot
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: execute both the emergency response (including SPDS) and NSSS
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(including COLSS)' software concurrently. LP&L committed, in their letter.W3P85-0432,<to-install an additional redundant CPU by ' '
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j-i : June 1935 to allow for simultaneous operation of NSSS and emergency *
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operators (NPO) will be able to' initiate the SPDS program from the ' F
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control room'during Modes 6 through 3 and have it available within ,,
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% , .< , t .The NRC frispectors' observed that the NPO was able to load the SPDS ;
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^ ,! 8PDS program had to be reinitialized from the computer room. The :
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-licensee has agreed to maintain a computer operator on shift until an f
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While' reviewing the. operability of the SPDS program, the NRC . ' .
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inspectors noted that"several< displayed parameters were incorrect,- s .
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t , . including steam' generators and pressurizer level. Discussions with the. licensee revealed _that wo'rk is presently in progress to verify
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4 r implemented the'SPDS software as' required by NUREG-0737,'Supplenent 1
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.using'the proposed! interim solution.s The,NRC inspecto'rs will monitora '.
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'the testing of CPU,4 and will' verify that the installation of the '
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.. ':, ' . . , > < .f.- (Closed)F NUREG-0737 (THI Ites'II.E.42) Containment Isolation
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as Table 6.3-32. This section'of the SRP' requires that there be.3 , i p? ( 5 l , , S
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isolation." The NRC inspector reviewed the licensee's evaluation and .
. determined that the containment isolation system design complies with the requirements of SRP 6. , In addition, the licensee reviewed each of the systems that impact .,
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. containment isolation to determine if they are essential.or
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- 4 / nonessential based on whether the system is necessary to:
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.(1) maintain the integrity of the reactor coolant boundary; (2) shut the reactor down and meintain.in a safe condition; and (3) prevent or
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l mitigate the consequences of accidents which could result in the
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potential for offsite exposure. The'NRC inspector verified that each nonessential system, as described in FSAR Table 6.2-32, is maintained
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in a closed position or will close on receiving an isolation signa These valves, once closed by an isolation signal, can only be reopened by deliberate operator actions following resetting of the containment isolation signa The containment isolation high pressure trip set point for I initializing containment isolation of nonessential penetrations has . been established below that permitted by the Technical Specifications l limits. The licensee used' explicit set point methodology to account for individual instrument uncertainty such as instrument loop error, set point variance, and instrument drift. This method ensures that the trip setpoint is sufficiently below the allowable value to prevent high containment pressure without initiating a containment isolation signal, i
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In addition to the containment high pressure isolation, the containment purge system is designed to automatically isolate on a containment purge isolation signal (high radiation). Operability of the containment purge isolation valves has been analyzed to ensure I closing against the most severe design basis accident. The analysis showed that the containment purge isolation valves are capable.of closing under any accident conditions when Ifmited to an opening of l 52'. Modifications to limit valve openings have been complete l The NRC inspector determined that the licensee has fulfilled the ! requirements of NUREG-0737 for containment isolation dependabilit I LP&L bas also impicmented Operating Procedure Op-903-075,
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" Containment Purge Valve Isolation System Operability Check,"
Revision 3, and off Normal Operating Procedure Op-901-020, "High Airborne Activity in Containment," Revision 2, to ensure the
, containment isolation system operates as designed. This item is
. considered close "
; . N violations or deviations were identifie .
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The NRC inspectors reviewed the plant procedure for shift turnover (OP-100-07). The NRC inspectors reviewed the procedure to assure that the checklist provided for the oncoming control room operators included the following items, Assurance that critical plant ~ parameters are within allowable limit Assurance of availability and proper alignment of all systems
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essential to the prevention and mitigation of operational transients and accidents by a check of the control consol Identification of system and components that are in a degraded mode of operation permitted by the Technical Specification ' It.is the NRC inspector's observation that a new checklist or an upgrade of the old checklist is warranted to assure that the oncoming shift supervisor has adequate information regarding the availabillty and proper alignment of all systems essential to the prevention and mit.igation of an accident. This item was discussed with the licensee and they have agreed to revise the procedure to address this area of concern. This is considered an open item (50-382/8505-02).
No violations or deviations were identifie ' Followup on Previous NRC Inspection Findings
, ' *(Closed), Open Item 50-382/84-07 Construction Appraisal Team (CAT) Finding ,
6.2 and 6.3 on Masonry Walls - i .
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) The NRC inspector reviewed the documentation associated with the rework o , o masonry wall S-24 and verified the wall was constructed as per desig T - This item is considered close ,
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N)oviolationslordeviationswereidentifie ' Test Results Evaluation ,
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The NRC inspectors reviewed initial fuel load and Phase III test results to verifylthat (1) all changes, including deletions to the test program, had been reviewed for conformance to the requirements established in the
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FSAR and Regulatory Guide 1.68; (2) deficiencies had been adequately addressed and corrective action completed; (3) the licensee correctly analyzed the test data and verified it met the established acceptance
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criteria; and (4) the startup organization as well as the Plant operating Review Cocaittee (PORC) had reviewed and accepted the test results. The following test packages were reviewed:
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-11-SIT-TP-400 Initial Fuel Load SIT-TP-502- Post Core Reactor Coolant System Flow and Coastdown Measurements SIT-TP-503 Rod Drop Measurements SIT-TP-505
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Pressurizer Effectiveness SIT-TP-506 , Leak Rate Test SIT-TP-508 Reactor Coolant System Heat Loss The NRC inspectors determined that each of the above test packages was properly reviewed by the licensee and met the applicable acceptance
.' . criteria.The following observations were made while reviewing the test result SIT-TP-502 - The reactor coolant flow coastdown measurements, '
following a planned simultaneous trip of all four reactor coolant
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pumps (RCP), was less conservative than that assumed in FSAR Section 15.3.2.1. The RCP flow coastdown curve is used by C-E to establish the maximum thermal margin for the COLSS and setpoints. To ensure that the COLSS is conservative and the plant operated within the analyzed operating parameters, the COLSS penalty factor EPOL1 has been changed from -4.7175 to -7.000. This new COLSS penalty factor will remain in effect until RCP flow coastdown can be reanalyzed during SIT-TP-727, "80% Loss of Flow - Natural Circulation." SIT-TP-505 - The flow settings on the pressurizer continuous spray valves (RC-302A and RC-302B) could not be adjusted to maintain the pressurizer spray line temperature 25'F to 30*F colder (actual 50*F) , than the average RCS cold leg temperature at> steady state conditions as required by the acceptance criteri C-E, in letter C-CE-9390, dated February 8, 1985, stated that the maximum allowable temperature differential for the spray nozzle could be increased to a value of' 85* The NFT data for the pressurizer spray line temperature indicated readings of approximately 522'F with <four RCPs operating and both cos.tinuous spray bypass valves fully open. The pressurizer spray line low temperature ' alarm is presently set at 525'F which is resulting in an almost constant alarm conditio C-E, in their
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letter C-CE-9390, recommended resetting the alarm to 520'F to void
. the constant alarm during steady' state operating conditions. In addition,' Technical Specification 5.7, " Component Cyclic or Transient Limits,". Table 5.7.1,.for the pressurizer spray nozzle should be revised to account for the reduced netpoint.' The present Technical Specification limit calls for a differential temperature of.130*F between'the pressurizer water (approximate 1y'653*F at 2250 psia) and the spray water temperature, with less than four RCPs running. The licensee is presently reviewing a change'to allow the limit to be revised to 140*F, to avoid the present' required calculation of usage ,
factors for virtually every operation of the pressurizer spray
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'. N., 4 O ' Phase'~III Test Procedure Witnessingi ,
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? 's e '.. SIT-TP-500 Determination of Auxilisry Spray. Flow Split [ , -
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Determination .
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1ae .The person,el conducting the test verei cognizant of. the test <
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/# ;. t acceptance criteria, precautions,'and prerequisites prio.r toL ' . s 3 :beginning the tes :
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.Three allegers",7all requesting anonymity,;werg interviewed by two1NRC
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representative 4 on Decentse 18,'1984.s They: expressed; concern with
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1 specific relding practices' employed at'Waterford 3 during constructioni '
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These con,: erns and the NRC findings are stated: al~
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. . Reactor surge line restraints were originally welded and then
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~The:allegers st'tqd a that the reason given for this' change was that . W'
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jThe[NRC'inspectorreviewed~the-documentationfortheserestraints.and *
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7 :S J ' conducted interviews'with cognizant engineering and NDE personnell ~ * ~ : ~" e This documentation review and subsequentiinterviews revealed that, in 4 y y i
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<some cases', thetinspection1 requirements were' changed in.tSe' original ,
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4, gg ,' design;of.the surge line restraints, all of:the welds were" designated g e
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fMi -Normally this'would:be accomplished using radiographye In cases ~ Ji G1 f %
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smay_'be examined using an approved ultrasonic, technique, .Because of
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Cn V, fthe complexity;of these restraints and severe? access limitations, an-j; f ; ~ . . , yengineering' decision was made to, radiograph!all: accessible areas of 6- 8: '
? E theseifull-penetration' welds and perform anfultrasonic examination of . b? the, remaining areas. Because of the' difficulties encountered in ' ',
f _ yJ ~ . "using two;methodsiof volumetric inspection on parts;of the same weld S( - <!j' _
, ; joint, a subsequent' engineering analysis was, performed.and:it was-1 ' decided that certain ' welds in these supports ha'd;been over designed '
=and could be. changed:to partial penetration welds without
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S A compromising the intended function of the structures. The acceptance '
,t' f" criteria for_ partial penetration welds is-based upon' the performance ^' - -
of a surface examination by either the liquid penetrant or magnetic -
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particle < method. 'No " blanket" change was made for the weld design
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a: changelof these restraints. A provision was included'in th engineering analysis for each of the welds in this category that each
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- . of the welds'in this category must'be documented in a Field Change (
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, Request (FCR) and evaluated individually'by engineering prior-to this i
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, -"down grading." The NRC inspector reviewed four FCRs relating to-
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these restraints and ascertained that-each had been initiated,
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,' evaluated, and dispositioned in accordance'with the appropriate
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1 (; - procedures. . ~ In some cases the requirements formwelds were changed
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each weld under consideration was clearly stated on.each FCR and each
4 ' wa~s) signed by thefdesign engineers performing the. evaluations.
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<accordance with code:and no improprieties were noted.~'The NRC \'
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e -- Pb?" ,Theyallegers stated that a piping restraint in Cell 2B, located ,y . 1?4- " inside,the containment building,-had cracked. The restraint locate gg , { p % at'approximately the 15-foots level, was repaired unsuccessfull Mseveral-times 1before the design'was changed. Their concerns are with h^ f6, ;
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cause.of the original cracking.and the adequacy of the design" ~ ", W ,7
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< changes.y'They expressed
? & fconcrete'behind the embedded plate to which'the restraint is mounted;,.
concern with.the possible damage done to'the- '
# 9 7 : Their concern is that the preheating 'of this plate prior' to welding
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i restraint.reworkiperformedintlie'areacitedb[theallegersinan ; 7~" effortito" identify the specific structure referred t By comparing
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ithat the subject structure wasta "C".stop' as.shown on drawing G-696 w I' 7 : S06;R-3', : sheet 6.- CCR-AS-3625 states that' repeated cracking iof one'
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#j' , 1of th'e: cracking:and"a means to prevent recurrence. Thefengineering' M m
u _ , f analysis determined thatLlocked_-in' weld stresses and' rigidity caused
;g * a [ by thefrestraint geometry;were responsible for.the repeated cracking
_ y; ' @G 'in1the finalLweldi.9A design. change 1for this[ restraint wa's proposed
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f v UM y [and~is:shownfon.thelFCR.(Thechangefnot(onlyalteredthe1 restraint n iconfigurationibut' changed 'some. of the ; welds from full-penetration to F ,% '
. 1 partial-penetlation welds. 'Thii changinglof theftype of. weld being -
{used' automatically changed the. inspection requirements from a-
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a~ so , ' 4/* . Evolumetri'c examisation to*a surface examination? r (The signatures'oni _
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p and.that the FCR was properly dispositioned. iThe restraint was
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. , . cracking.'1 Concerning the heat-induced spalling of;the concrete
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Lbehind;the plate,1the NRC-inspectortinterviewed,several civil and-structural yngineers ,wholagreed;that the weld preheat applied to an
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Tembedl plate offthisisizecand' thickness, could not transmit sufficient-
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'Lheat?toithe concrete to causefeven minor'spalling1behind the pl' at ~
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,1 1 being-tested.and certified by the licensee. The allegers feel that-Lf ~
welder certification program.
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/The licensee hss' opted;to qualify, certify, and maintain weldsrs9at 2 .
the Waterford.3 plant rather than contract these servicescthrough?
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V established afwelder training 'and certification center onsite. All Lwelder, testingD is; performed under'the< direction _of the'LP&L assistant '
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W94$ u ' mechanical 1superinten' dent with the' accept / reject res'ponsibility NM
. ,
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. .assignedtothe:LP&L'plantiquality' group.yTheNRC;inspectorreviewed .
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f12 randomly selected' welder qualification recordsLand conducted , t WW#' ., ,t laterviews.with three representatives of the' plant quality' group.-.# '
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gWgMpy w , 'g" aiinspector that the qualification'and. certification of~welderstwas s
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. ;being conducted in accordance with Procedure MM-1-052, Revision 2,'
: and'.the ASME' CodeF ;The records generated for each welder ties ~t
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S.8 1' Ng findicated that, as a minimum, the material identification, welder ~ ]yv , 5: ,, % identification;f est; t coupon fiti-up, and test coupon final weld 'y' .w < ; - ,
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N group. Acceptance :or rejection of a1 test coupon is based solely on 1:n O *(.g g s 3'ithe resultstof radiography or physical bend- test as observed by the ,
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_4The:subjectofthe$1' legation liswhethertheilicensee_isprovid'ing ' r/ - "'f 96 . iadequate. monitoring Lof ithe welder; certification l program. _ Section.IX, 3 Article III, of > the' ASME Code' states that the^ certification of the
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Q y 9;;statesithat-the the' full [ supervision ~ welders faddicontrol o'r . welding ofl the< license ~ operators: The NR shall inspector- be tested under' Wy ,
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JdeterminedithatYthe' monitoring activities and the documentatio .
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r maintained:by.the plant quality, group.in'the' area of welder g;7j K,f ,# i _
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certification is; adequate and meets,the intent of the Code. The NRCI (fp / - , tinspecto'r had no further questions concerning this ite .. . - .
. . r
-
_j v>' V The allegers (stated ' that multiple chain-falls ;and "come-alongs" were K # $g ' utilized in" the ' fit-up of an '8" diameter stainless steel piping ? spool fyf , -^,p V ~ in startup?systeinf52-A2. The allegers ' stated;that a QC inspector-P 4- ,a; _ refused:to7 accept the' fit-up-inspection with'the'se devices in place, M&4:
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E so ~all but ohe of; the .'? corse-alongs" whre released. . The allegers
*
w MM^ >,' . stated,that'the' tack l welds l broke,findicating'that?the pipe had been .' cold-sprung" into position. The pipelw~as' allegedly manipulated backi y[' d .7 .
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G"into ~ place with the/'come-alongs'.' a'nd retacke.dk The; QC inspector-
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. allegedly accepted the;second' fit-up'in*pectionjinithe " cold-sprung"
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.im J; Teonditionb~.'The p _ jvg reluctantly " 1 allegersl ~- L7 stated # thatothis acceptance ~'
!was made "very-4-
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. in; an- attempt, to . ascertain the texact piping spool and. weld in
'
M 3 ,3 , cp , }questionia'sethesallegers could,not provide these details. .Through
=these~faterviews"and a revie'w of the record p'ackage for startu '
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: . ? system l52-A2,'itwardeterminedthatthspipespoolandweldin .
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>V ; - ^ question was in'line 6RC8121; This is'the pressure safety valve s nW " , . -
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- outlet to.the'quenchL tank as shown on drawing'150-8469-724 St : :R2/IC-899-E, ' Sheet (2,- Revi'sion 13: ;.yAlthough this line was -not S S
- $ fi ~ , qconsidered to.be. safety-related and Nequired no inspection, EBASC0;QC fjyc
.
,y ; t did; perform a -:100. percent visual inspection of all final weld surface ~ . ~ .
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SThe'NRCJinspector!determinedthat'during1theperformanceofDeshn;'
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[h M 6 # Change: Notice (DCN) MP-688 R/1 which replaced an existing [( W G !2";x 8"isafety relief valve on the pressurizer with a 6" x 8" valve,.
j@ 9 Y ,J w Dit was7secessary to cut-line 6RC8-21'. This work was accomplished ym '
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'under Condition Identification Work Authorization }(CIWA)(82A236
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] fengineering personnel who were present-during.the repa'i ,
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Thel. recordipackage ' indicates that several welds were reworked, ara f '; 7 f ~
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jresult.of'this discovery,'tofrealign the pipe without forcingdt into-
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position./ Records;indicatethatfseve'ralQCinspectors.wererinvolved
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'during th,e' original welding and the rework of this pipe but none of e sthem'was th'e.-inspector mentioned by the alleger ~
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:The'NRC 3nspector d'etermined that an. 8" line;in the startup . system -i
.s* >H 152-A2;apparently;had been " cold-sprung". prior to welding. Thi , condition; however,1 had been. discovered by EBASCO.and corrected prior-
,'. J . to the. turnover of thefsystem to;the startup and test' group. The NRC ':
^ ~ '
s a- ' inspector hsd noifurtherLquestions1concerning this item.' g
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Mk , ie. ~Theallegersstated'that,~ inane 5fo5titoacceptsubsi[ndardNeld's
~ < %,y * A, and multiple weld repairs, thel acceptance criteria 'for; the reactor *
7't . < , .coolan't; system piping "C" stops w re changed lfrom radiography,to- - _ magnetic. particle inspectionk $The-allegers feel that this;'
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4 7 ; , _ downgrading of requirements"could-affectithe functioning of.the P 1 ' s,,; , ; f"C." stop ' 4 ,
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ofsthe RCS "C"/ stops and conducted interviews with the'same: personnel'
?' - 2 , smentioned.in the evaluation of Allegation "a" above. 'The reasons for g ; '
DchangingLthe acceptance criteria and'the justification;for the k v I;i.y 7 7 ( Jchanges'were' listed on individual FCRs for each("C" s_ top.* Each was'
.
_ i,' '~
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( evaluated and;dispositione'd on a case-by-case basis by EBASCO ,
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,x g engineerin It was noted by the NRC~in'pector,that s the reasons for ~f
- 1 changing the'"C" stop' design and examination.was' due to alignment and 3 ' ,
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i fit-up difficulties and not the. weld cracking problem noted in p .
% Alle~gatioi: '?a"La bove. Following his interviews ' and record review,
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;No' violations or deviations'were identifie y . .
-
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f* 11. - QA Personnel Qualifications ' _ , y g
: , . .
'
7~. . m ~During the' week of-February 25, 1985,Ethe Waterford Task Force-QA Team l,* 9/ ' ,
-completed'the' review of qualifications of QA personnel'at the Waterford- ,
3 : site. The NRC staff committed to complete thisfreview by March 1, 1985,
' ,e 'in SSER, . e
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_ . . ~ , W M f 7 .QA personne1rincludedLeanagers, supervisors," auditors, record reviewers,
- K ! clerks, and secretaries. These individuals were 'not assigned 1 I 1 responsibilities- or-performed' any functions that required .them to be
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%4 ^ qualified under the requirements of' ANSI N45.2.6;.i.e., their
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$ training.and'some, relative experience commensurate,with.their job
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description and responsibilities._ The requirements for- auditor
%n ; qualifications L are delineated .in' ANSI'N45' 2.12, Draft '3,cRevision 4 - 1974 m (LP&L's commitment). This'. standard requires training, relative
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. experience,fand independence from the area being audited.~ The NRC staff b' ', ,
7 reviewed,the site: standard practices,(procedures)' pertaining'to all QA
~J- g- - personnel and, found them
, :
to be, acceptable and to.neet' LP&L's commitment , ,
* ~ .
,
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, 6 .
24, .The LP&L1 evaluation included background verification, qualification, and
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iT ~ determinations. -~The NRC' staff also reviewed the corrective action 4
'
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& . (accomplished to resolve the problems associated.with'personnelcwho were x d [ determined.to.be not qualified by th' eLP&L evaluation.' LP&L '
'
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V D Procedure QASP 19.12 contains_the guidelines used in this evaluation of QA
~
s V J . * ' - personnel; qualifications. The' guidelines'were based on training and . 6 0 '- TJ ~ : indoctrination ~as well as' ANSI N45.1.12-1974" requirement's for auditor fE y,'1 M , ; Qualification requirements?for other QA~ personnel are not specifically
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- - defined by ANSI /ASME standards or NRC regulation Ow
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1The!NRC. staff: reviewed the'LP&L' evaluation of personnel qualifications > g ;N 2 " ~ ~ usingva statistically based sampling' plan. The following is a summary of~ ~ N
'
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&,s 47 s c -L t he'NR.C sta'ffirevie #
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LP&L(sedthe~ criteria'of'theGreenBook, WASH 1283,-Revision- .
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.(5/24/74)/ ANSI.N45.2.12 (2/22/74), Draft 3,:Revisien 4.to evaluate . *
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p [auditoriqualifications. At 'least one or.nore of the 'following. had ,to;
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(TheNRC. staff,revieNdalampls;ofauditorqualification-records,
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,. averification,,and~ disposition'ofLauditorsidetermined.not qualifie W y. %m , ,c
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WX K > Jauditor qualification , gg - % > - >
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i b.- DocumentIReviewers-
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b 's v ., ' iThe[joblresponsibilitiesofdocumentreviewers:conbstedofverifying,
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j for a.' document reviewer to have a significant technical knowledge,. ' * - lg, S,but 'did ' require that he knew what' QA information had to be contained.
Mi , a] Qd M yS ' M,g in, specific documents Consequently, the LP&L evaluation of document
-
j y "( F ereviewer's qualification consisted _of determining if those personnel
-
r % ,J^ employed'as document" reviewers received appropriate training ~t ; OcIM?..M acquire'this(knowledge. 1The results of this evaluation concluded pp E^ that allypersonnel' employed as' document reviewers did receive the
'[ltrainingneededto-qualifythemtoperformQAreviews.ofoneormore .
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? types of qualitv documents and concluded,' therefore, that thes ,
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personnel.were; adequately qualified to_ perform these tasks. The NRC-staffereview of this LP&L evaluation concurs with this conclusio' , iThis;conclus' ion? was .in ' concert with -resolution:of : Allegations' A-06, .
;C ' A-09,(A-289b,' A- 196, and A-306t as stated in.SSER '
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~,. OtherQIPe'.rsonnel-
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specialists, supervisors;and managers. Since QASP 19.12 does not- -
*
_
:contain guidelines for evaluation offthe qualifications _of the'se-
? personnel,7 the evaluation of these qualifications were in acca iarme -
'
'
. '-
with the' licensee, contractors and subcontractors, QA' programs. The ,c' , NRCfstaff reviewedithe qualifications of selected personnel listed in ~ R - - this category;and~ agreed with the LP&L-assessment that:they>had . ,
,
'
.
,
'
sufficient qualifications to perform QA tasks normally~ associated
, ;with these job' classifications. .It was,ttherefore, concluded'that- , g; ,these. personnel possessed the necessary qualifications to. perform the '
q4 . 1assigne( < - QA; task.' - ,. . <
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' R d .- 1Puso'nne17 Identified ~ as L Not ' Qualified '
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s 0f'the 657 personnel evaluated:by;LP&L, 115 were determined not, .- cqualified (questionable or indeterminate). , Corrective actio ~
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reque'sts were prepared and' issued,to document the unqualified w^ W ,
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; personnel,' an'd the following was: determined: j a m <
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'(1)--Forty individuals wer'e" determined qual'ified; based on additional"
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'information obtained;?i.eJ,nbackgroundichecks and. training
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- 32)OFortyindividualshadbeen.previously'addressedas.aresultof
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'faudits. performed ~by EBASC0_at Waterford 3. These' audits were , ,
ireviewedsby'LP&L and the staff and were'found_ acceptabl ;,
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