IR 05000237/1987006
| ML17199Q962 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 07/27/1987 |
| From: | Danielson D, James Gavula, Yin I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17199Q903 | List: |
| References | |
| 50-237-87-06, 50-237-87-6, 50-249-87-11, IEB-79-14, NUDOCS 8707310045 | |
| Download: ML17199Q962 (20) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Reports No. 50-237/87006(DRS); 50-249/87011(DRS)
Docket Nos. 50-237; 50.-249 Licenses No. DPR-19; DPR-25 Licensee:
Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name:
Dresden Station, Units 2 and 3 Inspection At:
Dresden Site, Morris, Illinois Sargent and Lundy Engineers, Chicago, Illinois Nutech Engineers, Chicago, Illinois Inspection Conducted:
January 14, 22, February 18-19, 23-24, 26, March 3-4, 16-17, 30-31, April 1-2, 9, 13,.May 6, and July 15-16, 1987, at Dresden March 18, April 10, and May 7, 1987, at Sargent and Lundy June 16 and July 17, 1987, at Nutech
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Inspectors: ()J. A. lfa'~~l~
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tl. T. Yin
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Approved By:
D. H. Danielson, Chief Materials Processes Section Inspection Summary n l~l
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7)~7/1'7 Date Ins ection from Januar 14 throu 17, 1987
{Re arts No. 50-237/87006{DRS);
0-2 /87 11 DRS Areas Inspected: Special safety inspection of snubber surveillance and functional testing (70370), training (41400), licensee action on previously identified items (92702), and onsite followup of operating events (93702).
Results:
Two apparent violations were identified (inadequate design control - Paragraphs 3.e.(1) and 3.e.(2); failure ta accomplish activities in accordance with documented drawings or procedures - Paragraphs 3.a, 3.b, 3.d, 3.e.(2), and '
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8707310045 870727 PDR ADOCK 05000237 G
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DETAILS Persons Contacted Co1TUTionwealth Edison Company (CECo)
C. Reed, Vice President, Nuclear Operations L. DelGeorge, Assistant Vice President, Nuclear Engineering G. Frizzell, Principal Engineer, SNE B. Javidan, Engineer, SNED B. Christel, Modification Group Leader, K. Knudtson, Engineer, T.S *
. G. L. Smith~ Assistant Technical Staff Supervisor
- J. W. Wajciga, Production Superintendent J. D. Brunner, Assistant *superintendent, Technical Services
+I. M. Johnson, Nuclear Licensing Administrator, Dresden D. Farrar, Nuclear Licensing Director R. Mirochna, Engineer, SNED J. Welch, Engineer, G. Svenson, ISI Coordinator D. Wilgus,- Engineer, SNED
- E. D. Eeningenburg, Dresden Station Manager
- J. Achterberg, Technical Staff Supervisor
- J. R. Williams, Regulatory Assurance
+R. Flessner, Services Superintendent
+D. Butterfield, Nuclear Licensing Manager
+E. Armstrong, Regulatory Compliance
U.S. Nuclear Regulatory Co1TUTiission (USNRC)
- I. T. Yin, Senior Mechanical Engineer A. B. Davis, Regional Administrator, Region III R. F. Warnick, Chief, Reactor Projects Branch 1 J. J. Harrison, Chief, Engineering Branch G. Holohan, Assistant Director D~ Muller, Project Director, P.O. III-2 D. Jeng, Section Chief, ESGB G. Bagchi, Acting Chief, ESGB R. Lipinski, Structural Engineer Wyle Laboratories (Wyle)
L. A. Phares, Operator/Test Technician R. C. Money, QA/QC Test Technician J. Mesich, Field Engineer Sargent and Lundy Engineers (S&L)
A. Walser~ Structural Project Engineer P. C. Bhatt, Supervising Design Engineer D *. J. Gullaksen, Assistant Division Head S. l..
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- _Sargent and Lundy Engineers (S&L)
H. McCullough, Senior Quality Assurance Coordinator H. K. Vilas, Structural Engineer M. J. Klein, Supervisor, Structural Analytical Division A. Marcos, Assistant Head, QAD J. Cullen, Structural Engineer J. E. Chittenden, Senior Structural Engineer R. A. Koncel, Structural Engineer P. A. Gazda, Senior Structural Engineer A. K. Singh, Structural Design Director Nutech Engineers (Nutech)
B. Whiteway, Gen~ral Manager, NEC G. Edwards, President A. Casillo, Engineering Manager, NEC J. Brown, Engineering Manager, NEC J. Attwood, Engineer N. Edwards, Senior Vice President R. Straebel, Project Manager Impell Corporation (Impell)
T. T. Wittig, Division Manager, Engineering
. P. S. Stoller, Section Manager, Engineering
- Attended interim exit meeting* at Dresden Site on February 24, 198 +Attended exit teleconference on July 17, 198 Action on Previous Inspection Finding (Closed) Confirmatory Action Letter (CAL) (237/85004-CC; 237/85004-lC):
On April 5, 1985, Region III issued CAL No. 85-04 requiring CECo to provide continuous transient monitoring surveillance for the main steam (MS) piping systems during plant operation, and to conduct functional tests for the MS and other piping snubbers during the next refueling outag On October 23, 1985, Amendment 1 to CAL No. 85-04 was issued to reflect the installation of linear voltage differential transformers (LVDTs) in addition to the existing strain
- gages (SGs) on the MS snubber The NRC inspector reviewed the CECo letters (May 16, 1985, to May 29, 1986) to Region III documenting the 35 recorded spurious SG and LVDT trips that were caused by electrical noises and other effects, and concurred with CEC0 1s data interpretatio Previous Region III followup inspections and data reviews were conducted in May 1985 (Inspection Report No. 50-237/85018); June 1985 (Inspection Report No. 50-237/85034); and March 1986 (Inspection Report No. 50-237/86007).
During the present plant refueling outage, the snubbers on MS Lines B, C, and D were functionally tested at the sit No abnormal or rejectable conditions were observe All snubbers met the test acceptance criteri In
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conclusion, the NRC inspector determined that CECo has effectively implemented the CAL requirements, and the continuous monitoring program provided for the MS systems is no longer a requiremen The matter is considered close (Closed) Violation (237/84027~04): Inadequate piping operability analysis performed after the low pressure coolant injection (LPCI)
system transien The NRC inspector discussed Nutech design issues with CECo technical* and QA staff during followup meeting As a result, CECo and its consultants conducted a comprehensive design audit at Nutech, San Jose, Californi The audit findings were forwarded to Region III for revie CECo a 1 so performed nondestructive examinations, vent line high point verification, and vibration te~t on the LPCI piping system (Inspection Report No. 50-237/84027, Paragraphs 4.c(2), (3), and (4)).
The NRC inspector considered the CECo actions adequate, and the matter is close (Closed) Unresolved Item (237/84027-03; 249/85013-03):
The as-built evaluations for some of the Mark I torus attached piping modifications may not be adequat Based on the analytical discrepancies found during the Core Spray System walkdowns and the piping configuration verification project (see Paragraph 3.e.(2) for details), this item will be upgraded to a violation and will be tracked under items 237/87006-02B; 249/87011-02 (Closed) Unresolved Item (237/84027-02; 249/85013-02):
The adequacy and acceptability of the --fE Bulletin 79-14 walkdowns is questionabl There is also a potential lack of structural calculation Based on the discrepancies discovered during the Core Spray system walkdowns and the piping configuration verification project (see Paragraph 3.e.(2) for details), this item will be upgraded to a violation and will be tracked under items 237/87006-02B; 249/87011-02 (Open) Violation (237/84027-01; 249/85013-01):
Support installations did not have adequate Quality Control inspection CEC0 1s response to this violation is contained in their letter from D. L. Farrar to J. G. Keppler dated August 2, 198 The NRC 1s acknowledgement letter from J. J. Harrison to C. Reed stated that pending NRR 1s approval of higher damping values for Dresden, additional evaluations and field inspections may be require To date, only limited approval of this change has been approved by NRR and only for portions of the Recirculation System pipin On this basis, additional corrective
.actions may be require In addition, the as-built configurations documented during this inspection (see Paragraph 3.e.(2)) indicate that some support installation discrepancies may be more significant than previously assumed and may cause stresses to exceed the design limit Due to the above concerns, this item will remain -Ope,. * *
3.
Action -0n Licensee Event Reports (LERs) (Closed) LER (237/87-003-01):
Primary containment structural steel connections did not meet design requirements due to original installation error While performing a pre-modification field walkdown, CECo discovered several discrepancies in radial beam to tangent beam drywe 11 structural steel connection These discrepancies were associated with "cheek plate" connections which did not conform to the original fabrication drawing The types of discrepancies 1ncluded missing bolts, missing welds, or welds that were apparently cut s.o they were no l~nger effectiv As a result of the extent of the problems discovered during subsequent walkdowns, every structural steel connection inside the drywell was eventually reinspecte Although all of the discrepant connections met the operability criteria, a total of nine connections exceeded Final Safety Analysis Report (FSAR) stress limits and therefore, required modification An additional 53 connections were upgraded to standard fabrication details even though they met the FSAR stress limits~ Engineering Change Notices N ~865-47, No. D-875-02, No. D-875-05 and No. D-875-06 were. issued to implement the above modification (See Paragraph 3.b, LER 249/87-005-01 for additional information.)
- This is an example of a violation of 10 CFR 50, Appendix 8, Criterion V, in that the structural steel connections were not installed in accordance with the drawing (237/87006-0lA) (Open) LER (249/87-005-0l}:
Primary containment structural steel connections did not meet design requirements due to original installation error As a result of deficiencies discovered on Unit 2 drywell structural steel connections (see Paragraph 3.a above, LER 237/87-003-01), a limited walkdown was conducted on similar Unit 3 connections during a.short maintenance outag Four tangential beam connections were identified as exceeding FSAR stress limits, but these did meet the operability criteri Due to the limited nature of the outage, no modifications were performe CECo has committed to repairing these deficiencies as.w~ll as performing a comprehensive inspection of all drywell structural steel connections during the next Unit 3 refueling outag Pending the outcome of these walkdowns and a review of the significance of any other deficiencies, this item will remain ope The above occurrence is an example of a violation of 10 CFR 50, Appendix 8, Criterion V in that the structural steel connections were not installed in accordance with the drawing (249/87011-0lA)
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(Closed) LER (237/87-005-0):
Snubbers 1301, 1302 and 1303 failed functional test criteri (See Paragraph 4.c(l) for closure details.) (Closed) LER (237/87-007-0):
Drywell structural steel radial beam exceeded design requirements due to a modification erro While performing the walkdowns associated with LER 237/87-003-01, a 1 inch by 1 inch notch was discovered on the top flange of radial beam No. R-1 An analysis of this 8 inch channel indicated that the notch would cause stresses in the beam to exceed the Final Safety Analysis Report (FSAR) allowable However, the beam would meet the operability criteri A revtew of the station maintenance fil~s revealed that a portibn of the channel had been removed and reinstalled in 1974 as part of a pipe repair procedure. It was postulated that the reinstallation process mistakenly left this notch in the bea Engineering Change Notice D-875-09 was issued to return the beam to within FSAR stress limit The above occurrence is another example of a violation of 10 CFR 50, Appendix B, Criteri-0n V in that the structural steel was not reinstalled in accordance with the drawing (237/87006-0lB) (Closed) LER (237/87-010-00):
Design and construction errors on the 11A 11 Core Spray subsystem resulted in stresses in excess of FSAR design requirement During the component support portion of the routine In-Service Inspection (ISI) program, Support 1403-M-206 was found in a degraded conditio The embedment plate associated with this vertical strut had pulled away from the concrete by approximately 1/4 inch at the plate's edg An investigation was started to determine the cause of the support 1s 11failure 11 as well as the operability of the associated Core Spray system consid~ring this support degradatio (1) Embedment Plate Design Discrepancy While investigating the degraded embedment plate, it was discovered that the original fabrication drawing for some of the embedment plates was in erro Instead of specifying a 9 inch spacing for the anchor straps on the backs of the embedment plates, the drawing erroneously specified an 18 inch spacin During subsequent ultrasonic examinations (UT) to confirm this stra~ spacing, it was discovered that some of the embedment plates had 24 inch anchor strap spacing as wel These discrepancies potentially affected over 2200 supports that are attached to embedment plates at Dresden Units 2 and *
This is an example of a violation of 10 CFR 50, Appendix B, Criterion III in that the design requirements for the embedment plates were not correctly translated into the fabrication drawing (237/87006-02A; 249/87011-02A)
As a result of above discovery, CECo implemented a comprehensive review of ~upports ~ttached to all embedment plates at Dresden Units 2 and 3 and Quad Cities Units 1 and This Embedment Plate Assessment Program was conducted by S&L and consisted of the following aspect *
Embedment p1ate design and ultimate capacities were developed using worst case attachment sizes and 1 ocation *
Large bore pipe supports attached to embedment plates were collected, verified, and plotted onto hanger load summary (HLS) drawing *
Supports that exceeded the design or ultimate load capacities of the associated embedment plate were walked down to determine actual attachment size and relative locatio *
Walkdown information was utilized to determine more.realistic embedment plate stresse *
- Embedment pl ates that exceeded the FSAR stress limits were ultrasonicly tested (UT) to determine the exact location of the anchor straps*relative to the load attachment Proximate support attachments were also identified and locate *
A detailed finite element analysis was performed using the UT informatio If the stresses exceeded the FSAR limits, the embedment plate was modified or reinforce On April 16, 1987, a meeting was held in Bethesda, Maryland for CECo to present details of their embedment plate reassessment progra Thi~ meeting is document~d in an NRC memorandum from J. J. Harrison to N. J. Chrissotimos, dated May 7, 198 Based on the material presented, the NRC concluded that CECo had an
_adequate program to identify and correct any deficient embedment plate However, there were several technical issues that required additional information to determine the adequacy of the design and ultimate load capacities of the embedment plates.
- Another meeting was held at S&L in Chicago, Illinois on May 7, 1987, to discuss some of the technical issues associated with the embedment plate progra Four additional items were*
identified relative to the progra CECo provided responses to the above two meetings on May 13, 1987, in a 1 etter from I. M. Johnson to T. E. Mur 1 ey and on June 13, 1987, in a letter from I. M. Johnson to A. Bert Davis respectivel The fo 11 owing information is based on the final status report issued on June 19, 198 Dresden Quad Cities Total Number of Supports 2230 2000 Requiring Evaluation Number of Supports Requiring 394 183 Walkdown Information Number of Supports Requiring 198
UT Information Number of Embedment Plates
0 Requiring Modifications Number of Supports Inaccessible
22 During Plant Operation That Require Further Evaluation Pending the review of the above two responses and the final evaluation of the inaccessible supports, this will remain ope Other than noted above, no violations or deviations were identifie (2) Piping Analysis Design Discrepancies While performing the analysis to determine if the degraded support affected the operability of the Core Spray system, several dfscrepancies wer~ discovered in the original piping analysi The comparison between the subsequent as-built walkdown and the as-analyzed configuration disclosed the f o 1 lowing discrepancie *
A new snubber was modeled in the analysis, but was never designed or installe The seismic response of the associated pipe was no longer vali *
The stress intensification factor (SIF) associated with a fabricated pipe tee was incorrectly specified as a forged tee SI The calculated stresses were under estimated by a factor of 2.7 at this poin.*...
A portion of the 14 inch schedule 80 pipe was incorrectly modeled as schedule 3 The pipe weight and correlating support loads were under estimated by a factor of *
Support 1403-M-201 had not been removed as specified on the Replacement Support Drawing M-3208-0 This support restricted displacements in a direction that the analysis had assumed was unrestricted and thereby induced unpredicted stresses into the pipe and torus penetratio *
Support M-3208-08 was installed 13 inches out of toleranc The resulting pipe stresses and support loads were inaccurat After the discovery of the above discrepancies, the operability analysis was updated using the as-found condition of the entire piping subsyste The results indicated that portions of the 11A 11 Core Spray system piping would exceed the FSAR design requirements during certain design basis event However, the pipe stresses were within the operability criteria and the safety significance was considered minima As far as the cause of the original embedment plate 11failure, 11 *
the normal operating design loads, in co~junction with the fabrication deficiency of the embedment plate, were determined to be insufficient to cause the 9bserved deformatio On this basis, a transient load was postulated to have caused the original damag A review of the core spray piping configuration revealed an intermediate high point location that was not considered in the.existing operating surveillance venting procedur Based on the potential air pocket associated with this configuration, the resulting water hammer loads were calculated to be comparable to the loads necessary to obtain the observed deformation of the embedment plat During walkdowns of the rest of the Core Spray system piping, no*
other damage was note Due to similar transient event occurrences on the LPCI system, (see Paragraph 4.c(2)) this item is considered part of Open Item (237/87006-03) pending a system review during the next refueling outag The short term corrective actions associated with this specific event included the followin *
The embedment plate for Support 1403-M-206 was repaired and provided with additional anchorag *
The fabricated tee was reinforced with saddles in order to lower the SI *
Support 1403-M-201 was removed.
(c) Piping Configuration Verification Due to the extent and nature of the discrepancies found in Nutech's Core Spray analysis, a comprehensive review program was implemented by CEC The scope of the program included all large bore (greater than 4 inch) safety-related piping analyzed by Nutech during the Mark I Torus Attached Piping Program in the early 1980' The basic elements contained in the program are as.follows~ Wal kdowns All applicable p1p1ng is walked down in the field to yerify the following attribute *
Supports identified on the piping isometric drawing exist in the plant and any supports that exist in the plant are identified on the drawin **
Support configurations match support detail drawing *
Support locations conform to general key plan location *
All valves are indicated on the drawing *
All branch lines are identified and located on the isometric *
All tees greater then 2 1/2 inches are noted as fabricated or forged and any reinforcements are detaile *
All flanges are indicated on the drawings..
Any obvious apparent discrepancies between the as-built and isometric are recorde.
Analyses Reviews The as-built information obtained from the walkdowns plus information from the piping and instrumentation diagrams (P&ID) are compared to the model used in the piping analysi The following items are considered during this revie *
Support locations
Missing supports
Non-Demolished supports
Pipe size and schedule
- The embedment plate for Support M-3208-09 was reinforced such that the higher loads resulting from the as-built reanalysis could be accommodate *
The Core Spray and LPCI operating surveillances were revised to clarify venting requirement *
The piping analytical model and records were updated to reflect the above change All of the above* actions were accomplished prior to the restart of Unit The previously discussed analytical discrepancies discovered during this inspection resulted from design deficiencies as well as construction deficiencie The first three discrepancies discussed above are other examples of a violation of 10 CFR 50, Appendix 8, Criterion III, Inadequate Design Control (237/87006-028; 249/87011-028).
The fourth and fifth discrepancies discussed above are other examples of a violation of 10 CFR 50, Appendix 8, Criterion V, failure to accomplish activities in accordance with drawings (237/87006-0lC;
.249/87001-018).
In response to the above deficiencies, several corrective action programs were implemented by CEC (a) Design Interface Review The missing snubber discussed above was located on a portion of the Core Spray line where significant design interfaces occurred between Nutech and Impel Nutech was responsible for the Mark I reanalysis for a portion of this line and Impell was responsible for the IE Bulletin 79-14 reanalysis for another portio Due to some 11miscommunjcation, 11 this support was never designed or issued.by Nutec A review was performed for those subsystems where significant design interface occurred between these two firm No other deficiencies were discovered during this CECo review; (b) Field Inspections A field walkdown was performed by Nutech and Impell personnel to confirm that the supports deleted during the Mark I and 79-14 reanalyses had in fact been removed in the fiel Support 1506-M-207 was the only other support out of the 160 locations inspected, that was inadvertently left in plac This support was removed prior to the restart of Unit *
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Branch connection details El bow radii Restraint modelling
- Pipe routing Component weight Pipe segment length Support orientation Valve orientation Discrepancy Resolution All Discrepancy Reports resulting from the above analyses reviews will be reconciled. Initially, all discrepancies will be :
11screened 11 to determine if it is significant enough to affect the operability of the system. If it is found that the operability is potentially impacted, a formal analysis will immediately be implemente Otherwise, discrepancies found during the analyses are reviewed and dispositioned to verify FSAR complianc As of June 22, 1987, except for one model at Quad Cities, all of the walkdowns and analyses reviews were completed for both Dresden and Quad Citie A total of 145 and 220 discrepancies were found for each plant respectively. The first level prioritization determined that 39 and 48 of the discrepancies, respectively, appear to be significan Of these only two have required a formal operability analysis to dat The overall pipe configuration verification program is still ongoing and will not be completed until the end of 198 CECo has corrmitted to provide a biweekly status report on the above work to Region II Except as noted, no violations or deviations were identifie f~
(Closed) LER (237/87-017-00):
Embedment plate for Support M-3210-03 exceeded FSAR design limits due to original construction error During the ongoing inspection and evaluation program as a result of a previous embedment plate discrepancy, (see Paragraph 3.e.(1)
LER 237/87-010-00 for background), the embedment plate for Support M-3210-03 was found to exceed the FSAR design limits. The analysis of the embedment plate did, however, indicate that it would meet the operability criteria established under the current evaluation progra The safety significance of this was, therefore, considered minima The corrective action performed for this specific situation was the reinforcement of the embedment plate itself. Based on the comprehensive review being performed under the above program, this item is considered close.
Snubber Visual Inspection and Functional Testing Background Dresden's Unit 2 snubber testing program was based on 88 safety-related snubbers for the current test perio The Technical Specification (Tech Spec) survei 11 ance requirements ca 11 for testing 10% of these snubbers each refueling cycl However, as a result of the snubber failures during the previous outage, additional testing requirements were imposed by Confi.rmatory Action Letter (CAL) 85-04 Item No. A total of 20 snubbers were specified as requiring functional testing during this outag Five of these snubbers had failed the functional tests last outage and seven others were associated with the main stream transient monitoring program (see Inspection Report No. 50-237/84027; 50-249/85013). Procedure and Documentation Review The NRC inspector reviewed the following CECo and Wyle Procedure *
DTS 020-2, "Snubber Functional Performance Criteria,"
Revision l, dated March 198 *
DTS 020.,;l, "Snubber Visual Inspection Criteria,"
Revision 6, dated October 198 *
MSS-002, "Hydraulic Snubber Testing Using the Wyle Model 100 Test Machine," Revision A, dated June 4, 198 *
MSS-004, "Procedure for Functional iesting.of Pacific Scientific Mechanical Snubbers Using the Wyle Model 100 Test Machine, 11 Revision 6, dated August 5, 198 *
MSS-008, 11Procedure for Functional Testing of Anchor Darling DynA/Damp Mechanical Snubbers Using the Wyle Model 100 Test Machine;" Revision 2, dated November 20, 198 With the exception of DTS 020-1, no adverse comments were made regarding the above procedure The list of snubbers contained in DTS 020-1, Table 2, should be updated to reflect changes made du.ring this outage as.well as information contained in the various checklist Pending review of the revised procedure, this is considered an Open Item (237/87006-04).
The calibration documentation was also reviewed for the snubber test equipmen The Wyle Snubber Test Machine, Serial No. 107, is due to be calibrated January 29, 198 All calibration documentation appeared to be adequat No violations or deviations were identifie... Functional Test Results A total of 11 snubbers failed the functional testing criteria for Unit These failures can be described as follow (1) Hydraulic Snubber Failures All three of the Unit 21 s safety-related hydraul iC snubbers (No. 1301, No. 1302, and No. 1303) exceeded the specified activation velocity of 10.inches per minute (IPM).
Since the entire population was tested, no additional test sample was required, These three snubbers are associated with the two Isolation:
Condenser steam supply and condensate return line Based on the results of the functional tests, both piping systems were reanalyzed with all three snubbers removed from the computer model The analysis performed by Impell (reference Impell Letter N ~191-002) indicated that both piping systems met IE Bulletin 79-14 operability criteria even though the FSAR criteria was exceede The safety significance was therefore judged to be minima As part of the corrective.actions, Modification Ml2-2-83-47 was initiated to replace the three hydraulic snubbers with two mechanical snubbers and one rigid strut. This work was completed prior to startup and n~ further action is considere This closes LER 237/87-005-0 (2)* LPCI Snubber Failure Snubber Nb. 1527, located on the LPCI System, failed to meet the functional acceptance criteria. This snubber also failed during the previous outag (See Inspection Report No. 50-237/84027, Paragraph 4 for details.) The cause of th current failure was attributed to the same cause as last time; that is, some type of a water hammer even After performing an autopsy of the failed PSA-10 snubber, it was determined that the current extent of internal damage was less severe than for the previous outag On this basis, the loading assumed in the previous operability analysis remains valid and no further evaluations of the piping system are require Since the snubber has failed two consecutive functional tests, it is evident that the previous corrective actions did not completely eliminate the root cause of the water hamme Although the overall extent of snubber damage indicated the
water hammer loads were reduced, the magnitude of the water hammer loads still exceeded the ultimate capacity of the snubbe To determine the actual cause of the event and to ensure the safe operation of the system in the future, CECo initiated the fo~lowing actions:
Conduct nondestructive examinations (MT) on the high stressed pipe weld joints in the syste Reevaluate the LPCf system to determine-the cause of the transient event. -
Replace the damaged PSA-10 mechanical snubber with a new unit of the same design and capacity. The new snubber will have strain gages and linear voltage differential transformers attached *
Provide continuous monitoring and surveillance of the snubber to confirm and. quantify any system transien The CECo actions will be verified and evaluated by the NRC inspector. Pending the review, this matter is considered part of the previously identified Open Item (237/87006-03) in Paragraph 3.e.(2).
(3) _ Reci rcul at ion Pump Bowl Snubber Failure Snubber No. 9 on the 2B Recirculation Pump Bowl locked up and failed to release during initial activation tests. The snubber passed both the running drag and breakaway drag tests prior to thi In order to investigate the failure, supplemental-activation tests were conducted at various rated load Although some signs of lockup were observed at 25% of rated load, the snubber never completely locked up agai The engineering evaluation of the snubber performance concluded that the snubber would have acted normally under thermal loads and would have restrained the system under dynamic load The system was considered operable at all load levels. The snubber was replaced with a new PSA-35~ The cause of the test anomaly was thought by PSA to be a misalignment of the capstan spring retainer cli No generic implications were conclude (4) Anchor Darling Snubber Failures There are 11 Anchor Darling (AD) Size 5500 mechanical snubbers attached to the Emergency Core Cooling System (ECCS) Suction Header on Unit Of the two AD snubbers tested in the original sample, one failed the running drag test and the other marginally
passed the running drag tes Sin~e this type of failure had been reported at other facilities, all of the AD snubbers wer removed and functionally tested.. stx bf the 11 snubbers exceeded the running drag acceptance criteria. All of the test data for the 11 snubbers were reviewed by AD and the problem was determined to be a previously identified ball skating proble However, after reworking a 11 of the snubbers for this prob 1 em seven of the 11 snubbers still failed the running drag tes On this basis, it was decided.to replace.all 11 AD 5500 snubbers with PSA-35 snubbers~
Due to the generic implications~ the 11 AD snubbers attached to the Unit 3 ECCS suction header were also tested at this tim Five of these snubbers also failed the running drag test. All 11 of these snubbers were subsequently replaced with PSA-35 snubber From an analytical standpoint, the operability of the suction header was demonstrated by a previous analysis that assumed all 11.snubbers were locked up during thermal expansio There was no operability concern resulting from this conservative analysi Since all the snubbers-passed the activation test, the system was declared operabl No violations or deviations were identifie Training The certification and training records were reviewed for the following Wyle Test Personne All certifications were current and appropriate for the functions performed by the individual Observation of test technicians indicated they were familiar with the applicable test procedures a~ well as the testing apparatu L. A. Phares
VT-4 Level II
Mechanical Test Specialist R. C. Money
VT-3
Mechanical Test Specialist
QA/QC Specialist
Visual Weld Examiner - Level II
!SI Specialist NOE - Level II J. Mesich
Field Engineer Specialist
- VT-4 (Snubbers) - Level II
VT-3 (Snubbers) - Level II No. viol at ions or deviations were identifie Visual Examinations As required by Technical Specification 4.6.I.l, all safety-related snubbers were visually inspected within the approp~iate schedule (18 months in this case). Although five recordable indications were reported, these were dispositioned by Impell as all being operable (reference Impell Letters 0590-194-004 and 0590-194-010).
No vi-0lations or deviations were tdentifie.
In-Service Inspection (ISI) for Pipe Supports
.As required by Technical Specification Section 4.6.F.1 and in accordance with the American Society of Mechanical Engineers (ASME),Section XI, Part IWF, component supports associated with piping being examined und~r ASME Section XI Parts IWB, !WC and IWD are also examine Procedure and Documentation Review The following CECo documents were reviewed by the NRC inspecto *
Procedure VT-3-1, 11Visual Examination - Component Supports,
Revision 1, May 29, 1985-.
IS! Notebook 11Dresden 2, !SI December 1986, 11 Book 3, Category F-No adverse comments were made relative to the above procedur During the review of the ISI Notebook, the. NRC inspector made several observation (1) Spring Hanger Settings
Inspection Sheet V-13, Support 1302-W-10 The remarks indicated the present setting was 7.5, the hot setting was 6.0 and the cold setting was Although the spring can setting was outside of the specified range, there were no procedures prescribing any appropriate acceptance criteri *
Inspection Sheets V-11 and V-12, Supports 3001A-W-101, 3001A-W-102, 3001B-W-107, 3001B-W-105, 3001B-W-103, 3001C-W-107, 3001B-W-10 Spring settings were recorded but no range was give All supports were dispositioned with no recordable indicatio Based on discussions with the ISI coordinator, the support drawings are not utilized at all during the inspection proces Because -0f this, the inspections must rely on the field settings indicated on the spring can itsel *
In many instances, the field settings are completely missing or may be incorrect due to updated piping analyse Without any backup documentation, all spring.cans should be dispositioned with recordable indications by the visual inspectors. so that an evaluation can be done by engineering _
to determine if it is within tolerance and acceptabl The procedures controlling the !SI activities did not address this issu The above two occurrences are other exainples of a violation of 10 CFR 50, Appendix B, Criterion V, inadequate procedure (237/87006-0lD; 249/87011-0lC)
(2) !SI Drawings The !SI isometrics utilized during the component support visual inspections are outdate Drawing ISI-200 Sheet 1 of 3 was last revised in 198 Since that time, significant support changes.had been implemented due to IE Bulletin 79-14 and Mark I torus attached piping reanalyse Over 20 changes were penciled in on the above drawing indicating supports were added, supports were deleted, supports were modified and support drawing numbers were change Support 1406-M-201 was erroneously noted as being changed to Drawing M-3208-0 Upon investigation by the NRC inspector it was discovered that the new drawing number should have been M-3208-08 and as specified on this drawing, Support 1406-M-201 should have been remove (See Paragraph 3.e for further discussions on this.)
In addition, most of the 11corrections 11 indicated on the isometric drawing were not incorporated into the master hanger list in the
!SI document Based on the above discussions, it was.unclear to the NRC inspector how the component support portion of the ISi program could be effectively implemente Not only did th inspectors not know exactly what they were supposed to inspect, but once they located a support, they could not be sure if it was supposed to be there or no This is another example of a violation of 10 CFR 50, Appendix B, Criterion V in that inadequate drawings were used to accomplish the component support portion of the !SI progra (237/87006-01-E; 249/87011-0lD)
It was noted by the NRC inspector that the inadequacies of the IS! drawings were previously recognized by CEC A project is currently underway to update and correct these drawing (3) Deficiency Evaluations On February 2, 1987, the visual examination data form for Core Spray Support 1403-M-206 noted that the support 1 s base plate was pulled away from the building 1/4 inch at the edg Work Request 61683 was subsequently issued to correct the
...,. proble An engineering evaluation was also performed in order to determine the cause of the support 11failure.
Durin the course of this evaluation, two significant issues were
encountere The first issue was related to embedment plate anchor strap spacin (See LERs 237/87-010-00 and 237/87-017-00 in Paragraphs 3.e and 3.f for further discussions.) The second issue pertained to discrepancies between the analytical models and as-built condition of the syste (See LER 237/87-010-00 in Paragraph. 3~e.(2) for further discussions.)
Although these two issues were significant concerns in and of themselves, an additional concern was created due to the lack of coordination with the ISI visual inspection report Data Sheet V-92 completed on February 20, 1987, reported three anchor bolts missing on Core Spray Support 1406-M-20 (This report also mistakenly identified the support as M-3208-03.)
Although this is a potentially significant observation, this report was not communicated to the engineers performing the operability analysis discussed abov As discussed during a meeting on March 20, 1987, Nutech was unaware that the ISI program had identified any other support deficiencies on the Core Spray Syste Eventually, it was discovered by the NRC inspector that the support should have been remove However, the fact that the engineer was not informed about a deficient support within 20 feet of a previously identified deficient support exhibits insufficient attention to corrective action requirements. This inadequacy was attributed to the lack of procedures controlling ISI activities~ This is another example of a violation of 10 CFR 50, Appendix 8, Criterion V, inadequate procedure (237/87006-0lF; 249/87011-0lE)
Other than noted above, no violations or deviations were identifie Training The certification records were reviewed for the following CECo visual inspector J. Cain VT-3 Level II Certified January 21, 1987 VT-4 Level II Certified January 21, 1987 c. Bortmir VT-3 Level I * Certified January 21, 1987
. VT-4 Level II Certified January 21, 1987 R. Spiven VT-3 Level II Certified January 21, 1987 VT-4 Level II Certified January 21, 1987 Based on the limited review of the records, the NRC inspector had no adverse comment No violations or deviations were i denti fi.e.
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I Open Items Open items are matter~ which haVe been discussed with the licensee, which will be reviewed further by the inspector, and which involves some action on the part of the NRC or licensee or bot The two open items disclosed during this inspe~tion are discussed in Paragraphs 3.e.(2), 4.c.(2), and.
Exit Interview Exit interviews with the licensee representatives (denoted in Paragraph 1)
were conducted on February 24 and July 17, 198 The inspectors summarized the purpose and findings of the inspectio The licensee representatives acknowledged this informatio The inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed during the inspectio The licensee representatives did not identify any such documents or processes as proprietar