View source for SVP-98-376, Submits to Nrc,Comed Root Cause Analysis Concerning Flaw Indications in Unit 1 Reactor Recirculation Sys Piping Requested in .Plan Will Be Submitted by 990331, as Discussed in 981204 Telcon.With Five Oversize Drawings

{{Adams
| number = ML20202E215
| issue date = 01/26/1999
| title = Submits to Nrc,Comed Root Cause Analysis Concerning Flaw Indications in Unit 1 Reactor Recirculation Sys Piping Requested in .Plan Will Be Submitted by 990331, as Discussed in 981204 Telcon.With Five Oversize Drawings
| author name = Dimmette J
| author affiliation = COMMONWEALTH EDISON CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000254
| license number = 
| contact person = 
| document report number = SVP-98-376, NUDOCS 9902020338
| title reference date = 05-10-1996
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 34
}}

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SVP-98-376                                                                                           l l
January 26,1999 l
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555                                                                                )
                                      - Quad Cities Nuclear Power Station, Unit 1                                        I Facility Operating License No. DPR-29 NRC Docket No. 50-254

==Subject:==
Results of the Root Cause Analysis of Flaws in Induction Heat Stress Improvement Treated Recirculation System Piping

==References:==
(a)     Letter from R. M. Pulsifer (USNRC) to D. L. Farrar (Comed),
dated May 10,1996, " Flaw Evaluation of Recirculation System Piping Welds - Quad Cities Nuclear Power Station Unit I (TAC No. M95127)"
(b)     Teleconference between Comed and NRR concerning Flaws in Recirculation System Welds, held December 4,1993 The purpose of this letter is to submit to the NRC the Commonwealth Edison (Comed)
Company Root Cause Aaalysis (RCA) concerning flaw indications in Unit I reactor               ,
recirculation system piping requested in Reference (a). The root cause analysis is included in the attachment, and provides an analysis of multiple causes and contributing factors that    r together explain the observed flaw behavior. As one of the corrective actions from the root cause analysis, the Generic Letter (GL) 88-01 categorization of 16 Unit 1 Intergranular Stress Corrosion Cracking (IGSCC) susceptible Induction Heat Stress Improvement (IHSI) treated 28" pipe welds was changed from category C to category D.
These 16 welds represent all the remaining IHSI treated 28" pipe welds in Unit I that have not had weld overlays applied.
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;        ,.,   ..         USNRC                                            .2                        January 26,1999 L
SVP-98-376 l
                                                                                                                                 )
                         ~ Additionally, as indicated during a teleconference held on December 4,1998,                          ,
: l.                        Reference (b), Comed will submit to the NRC a plan to remediate the Unit 1 IGSCC l                         susceptible welds that have not been repaired with weld overlays. This plan will be                    ;
submitted no later than March 31,1999.                                                               -l
                        . If there are any questions or comments concerning this letter, please refer them to                    j Mr. Charles Peterson, Regulatory Assurance Manager, at (309) 654-2241,                                 ;
extension 3609.                                                                                       1 Sincerely, loel P. Dimmette, Jr.                                                                                  ,
                         ' Site Vice President
                                                                                                                                 )
Quad Cities Nuclear Power Station                                                                      i I

==Attachment:==
Quad Cities Reactor Recirculation System IGSCC Root Cause Report                        .  .
cc:      Regional Administrator-NRC Region III NRC Senior Resident Inspector- Quad Cities Nuclear Power Station s
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   . ''     1,      ~USNRC                                             3                     January 26,1999       l
                    . SVP-98-376                                                                                   1 i
l bcci        R. Pulsifer, Project Manager, NRR                                                 ;
W. Leech, MidAmerican Energy Company                                              j D. Tubbs, MidAmerican Energy Company
                                ~ INPO Records Center Office of Nuclear Facility Safety, IDNS N. Reynolds, Winston & Strawn R. Krich, VP Regulatory Services, Corporate, Comed                               l M. Wagner, Licensing, Comed F. Spangenberg, Regulatory Assurance Manager, Dresden C. Peterson, Regulatory Assurance Manager, Quad Cities T. Fuhs, NRC Coordinator, Quad Cities K. Bethard, Regulatory Assurance Commitments, Quad Cities R. Baumer, SRB Site Coordinator, Quad Cities G. Larson, Engineering Programs Supervisor, Quad Cities                          :
DCD License (both electronic and hard copies)
SVP Letter File                                                                   i i
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Attachment Quad Cities Reactor Recirculation System IGSCC Root Cause
                   ,        ,                                    SVP-98-376 Page 1 of 20                                          1 l
Executive Summary                                                                              i Examinations in accordance with NRC Generic Letter (GL) 88-01, "NRC Pos: tion on Intergranular Stress Corrosion Cracking (IGSCC)in BWR Austenitic Stainless Steel Piping," for purposes of detecting intergranular stress corrosion cracking (IGSCC) in the      I recirculation system welds were performed during Quad Cities Nuclear Power Station (QCNPS), Unit 1,1998 refueling outage, (i.e.,01R15). Flaws were identified in seven            j welds on 28" diameter pipe. Three of the seven welds were identified during the                j previous Unit I refueling outage (i.e., QlR14), and four of the seven were identified          ;
during Q1R15. Based on ultrasonic inspection, the indication in Weld 02BS-F14 exhibited flaw behavior that appear-d to be inconsistent with Boiling Water Reactor            l (BWR) industry experience.
1 A cause and effect analysis was performed to identify potential root causes. The major cause categories investigated were the following-I e     IGSCC Inspection accuracy
                        . Water Chemistry                                                                          '
* Original Recirculation Piping Fabrication
* Stress Improvement Process Induction Heat Stress Improvement (IHSI) and Mechanical Stress Improvement (MSIP)
                        . Abnormal Loads The investigation determined that the flaws did not exhibit behavior inconsistent to what
                      . is experienced in the BWR industry. Rather, the bounding flaw, (i.e., in weld 02BS-F14,
                      . QCl), was a long shallow, or series of shallow flaws, that grew over time and now have been detected. This investigation concluded that there were multiple causes that together explain the observed flaw behavior. The contributing factors were the following.
* The IHSI treatment of the 28" diameter piping was not fully effective.
* Undetected pre-existing flaws were likely present.
* Weld geometry masked existing flaws from Ultrasonic Testing (UT) examination.
* Normal IGSCC growth occurred in small pre-existing flaws.
* Water chemistry , (i.e., Hydrogen Water Chemistry (HWC) and Depleted Zinc Oxide (DZO)) increased flaw reflectivity to UT examinations.
Four welds received permanent repair-weld overlays during QlR15. The other three i                       cracked welds were analytically shown to be acceptable for continued operation.

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Attachment Quad Cities Reactor Recirculation System IGSCC Root Cause
              ,        ,                                 SVP-98-376 Page 2 of 20 1.0     Background While performing the Unit 1 GL 88-01 Intergranular Stress Corrosion Cracking (IGSCC) examinations, during the 1998 refuel outage, QlR15, flaws were identified in seven welds on the 28" diameter recirculation piping. Three of the flawed welds,02AD-F12,02AS-S4, and 02BS-F14, had been identified and were determined to be acceptable by evaluation during the last refuel outage, QlR14. The flaw sizes in welds 02AD-F12 and 02AS-S4 were found to be the same in QlR15 as they were in QlR14. The flaw size in the remaining weld, 02BS-F14, was measured by UT methods to have extended 23.5"in circumferentiallength from one end of the original flaw indication. The flaw depth was reported to be the same in the area of the original indication and 0.83" in depth at two peaks in the new region of the indication. The four remaining new flaws were found in welds,02AS-F9,02AD-F8,02BS-F4 and 02BS-F7.
All of these welds had been stress improved using the Induction Heating Stress Improvement (IHSI) process in 1984. Accordingly, these welds should have an              ;
increased resistance to IGSCC. However, eight welds were found to have flaws during examinations conducted in refueling outage QlR14 in March 1996. Five were repaired        ;
by weld overlay, and as previously stated, three were found acceptable by evaluation.
This investigation was initiated because flaws were found in stress improved welds, and   1 in one case the measured flaw depth for weld 02BS-F14 appeared inconsistent with
                - industry experience.
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2.0       Problem Statement Between QlR14 and QlR15, UT examinations of the 28" diameter recirculation piping indicated that weld 02 BS-F14 exhibited flaw behavior inconsistent with BWR industry experience. Additionally, four unanticipated flaws were found in Q1R15.
3.0       Root Cause Investigators A project team was rapidly assembled with representatives from Quad Cities, Downers Grove, General Electric, LMT, Welding Services and consultants from EPRI and the Industry. The root cause analysis team was formed to identify causal basis for the identified flaw indications and growth. Industry information, data and expertise were at the disposal of the team.
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Attachment Quad Cities Reactor Recirculation System IGSCC Root Cause                         ,
SVP-98-376                                             l Page 3 of 20                                           l i
i 4.0       Root Cause Analysis 4.1       Process                                                                               I A cause and effect analysis was performed to identify potential root causes. The evaluation team decided on using a dispersion analysis type cause and effect process.           ;
Based on industry information, field inspections, station data and historical reports, major cause categories were identified. Each cause category was graphically displayed           j (See Figure 4.4) on a cause and effect diagram. The major cause categories are listed as        j follows.
                  . IGSCC Inspection / accuracy .                                                             l
                  . Water Chemistry
                  . Original Recirculation Piping Fabrication                                                 ;
* Stress Improvement Process e Abnormal Loads For each major cause category all of the possible causes related to the newly observed          l flaws in the reactor recirculation system piping welds were identified. Each possible           l cause was thoroughly evaluated to validate its contribution to the problem. Based on the collective evaluations, those causes that were determined to be significant, contributing       j factors were identified as root causes and validated accordingly.
4.2       Major Cause Categories /Possible Causes -                                             )
4.2.1     IGSCC Inspections l
4.2.1.1 Inspection Scope QlR15 is not the first time IHSI welds were found to have flaws. A comprehensive review of 25 IHSI treated welds determined to be flawed at both Quad Cities Unit I and Unit 2 were previously evaluated, (Review of UT Results for Flawed IHS1 Welds for Commonwealth Edison Quad Cities - Units 1 and 2, EPRI Letter Report, March 1997.)
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1 Attachment Quad Cities Reactor Recirculation System IGSCC Root Cause
             ,       ,                                     SVP-98-376 Page 4 of 20 During QlR15, scheduled UT examinations of austenitic stainless steel piping welds were performed in accordance with the NRC Generic Letter (GL) 88-01 Program. In addition to the routine GL 88-01 examinations, twenty nine (29) austenitic stainless steel welds that had been stress improved by the IHSI process were also selected for examination to address the concerns with the effectiveness of the IHSI process identified by the 1996 examination results. As a result of flaws detected in the original "IHSI Scope" examination sample, the inspection scope was expanded to all of the remaining IHSI welds that are in Category C, (i.e., non-resistant with stress improvement after two'
               . years of operation). The expanded examination sample consisted of twenty-nine (29)
IHSI welds. Figures 4.2.1-2a and 4.2.i-2b present the original and expanded inspection scope welds for recirculation Loops A and B. Welds identified as "not inspected yet" were completed during QlR15 with no failures, after the root cause analysis was completed. The following table provides the number of austenitic stainless steel welds examined during QlRIS. Category C' denotes Category C IHSI welds that are examined at the GL 88-01's Category D frequency. The number of welds examined under the expanded examination sample is shown in parenthesis in Table 4.2.1.1 below.
Table 4.2.1.1 Q1R15 Inspection Scope QlRIS Inspection Scope SYSTEM          SIZE     Cat. A      Cat. C     Cat. C'    Cat. D      Cat. E     Cat. F Recirculation       4"         0          0           0         0           0          0 12"        0          0        9 (11)       0           0          0 22"         0          0        4 (3)        0           0          0 28"         0          0        10 (10)      0           3'         3' RHR (SDC)         20"         0          0           3         0           0          0 RHR (LPCI)          16"        0          0        3 (5)        0           0          0 NOTE 1: Flaws identified in 1996, during QlR14.
4.2.1.2           Inspection Results The UT examinations performed on the original and expanded QlR15 inspection scope of 28 inch diameter recirculation piping welds identified IGSCC indications in four welds previously considered unflawed. These new indications are in welds 02BS-F4,02AD-F8,02BS-F7, and 02AS-F9. The UT examination of 02AD-F8 determined the weld contains an axi-1 IGSCC flaw indication. A review of the report covering the axial indications confirms many of the conclusions covered later in this report. It indicates that axial cracks are diflicult to detect and it is therefore not surprising if additional axial flaws are identified. Therefore, this discussion does not revisit axial flaw detection.
Further, since the weld 02BS-F14 flaw size bounds all of the circumferential flaws identified for both Quad Cities Nuclear Power Station units, this report focuses on circumferential flaws.

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l Attachment Quad Cities Reactor Recirculation System IGSCC Root Cause
                         ,       ,                                     SVP-98-376 Page 5 of 20 IGSCC flaws detected in the other three welds during Q1R15 were determined to be circumferentially oriented. Of the four welds, three welds, (i.e.,02AD-F8 axial flaw, 02BS-F7, and 02AS-F9) were repaired by the weld overlay process in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel
                           - (B&PV) Code, Code Case N-504. Weld 02BS-F4 was evaluated as acceptable for continued operation without repair and will be re-classified as a Category F weld in accordance with GL 88-01. Category F welds are examined each outage. IGSCC flaw indications in welds 02AD-F12 and 02AS-S4, which were detected in Q1R14 in 1996, showed no change in flaw size and remain acceptable for service.
l The IGSCC flaw size and behavior in the weld 02BS-F14 Heat Affected Zone (HAZ) bounds all other flaws detected in Unit 128" diameter pipe during QlR15. The flaw originally detected during QlR14 was found in Q1R15 to have increased in length by 23.5" on one end only. The depth of the flaw was reported to be approximately the same in the first 22" as measured during Q1R14; however, the new area of the flaw had two points that were 0.83" in depth.
Table 4.2.1.2 Flaw Indication Characterization                                                                 ,
Flaw Indication Characteritation 1996                  1998                      1998 WELD #           CONFIG                                                          ACTION TAKEN Length        Depth       Length        Depth 02AD-F12         Pipe to         2.05"        <10%        2.05"        <10%         Analyze
* Pump           .2.5"         0.18"        2.5"        0.18" 02AS-S4          Elbow to        1.25"       0.206"       1.25"       0.206"        Analyze
* Pipe                                                                                        o
                            .02BS-F4          Pipe to     360* Root     Geometry       26.5"        0.25"        Analyze Pipe l
02AD-F8          Elbow to    360 Root      Geometry       Axial        0.20"       ,lepair Valve 02'BS-F7          Valve to   360* Root     Geometry        8.0"        0.32"        Repair Pipe 02AS-F9           Valve to   360* Root     Geometry       360*         0.25"        Repair .
l Elbow                                Intermittent                                             l 02BS-F14         Pipe to         22.0"       0.401"       45.5"        0.83"        Repair
* l Elbow Cast Elbow
* Previously Found Acceptable by Analysis in 3/96.

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Attachment Quad Cities Reactor Recirculation System IGSCC Root Cause                     ,
SVP-98-376                                         '
Page 6 of 20 l
4.2.1.3 IGSCC Examination                                                                i l
The root cause investigation included efforts to identify potential changes between i'
QlR14 and QlR15. This process included a review of the UT examinations performed during both outages. The primary objective of the review was to identify conditions and process variations that could have influenced the UT examination results. In addition to reviewing the UT examination personnel, procedures, and equipment, the applicable UT data reports were reviewed for each weld including historical data. Information was       i reviewed regarding the evaluation of previously recorded indications and any scan limitations or access problems.
4.2.1.4 Manual Examination Technique IGSCC examinations performed during both outages employed manual UT examination techniques. Several automated exams were performed during QlR15. The manual examination technique has been successfully qualified for detecting and sizing IGSCC. A review of the manual examination data could not conclusively determine whether the new IGSCC flaws were missed or mis-sized in previous examinations. However, the EPRI/ Industry information suggests that discrimination ofIGSCC has improved as supplementary evaluation tools (i.e., high-angle shear and Refracted Longitudinal (RL) probes) have gained greater acceptance in the industry and UT examiners have become more familiar with their application. Thus, it is possible that improved flaw discrimination contributed to the apparent anomalous flaw behavior in that the flaw was a pre-existing condition that was undetected.
4.2.1.5 Personnel and Training Qua:ification of personnel was noted to be different between QlR14 and QlR15. This occurred due to an industry transition from the Three Party Agreement IGSCC Examiner Qualification Program to the Performance Demonstration Initiative (PDI) program for personnel qualification. The Industry and NRC has determined that use of the PDI Program is the preferred method of qualifying BWR IGSCC examination personnel for the future (Reference 2). Personnel performing UT examinations during QlR14 were qualified in accordance with the Three Party Agreement program. For QlR15 the personnel were qualified under PDI. It is probable that the changes in examination results were influenced by the additional experience gained by PDI examiners in discriminating flaw indications from geometrical reflectors such as severe root and counterbore geometries.
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Attachment Quad Cities Reactor Recirculation System IGSCC Root Cause                         ;
                  ,                                            SVP-98-376 Page 7 of 20 4.2.1.6 Inspection Procedures The IGSCC detection procedure used during QlR14 was procedure NDT-C-2, Rev. 22.                ,
During QlR15, Rev. 24 of the procedure was used along with NDT-C-55, Rev. 0 which was required for single side access examinations. During both outages, procedure NDT-C-41, Rev. 2 was used for determining the through-wall size of recorded flaw indications. Initial qualification of previous revisions of NDT-C-2 and NDT-C-41 was accomplished at the EPRI Non-Destructive Examination (NDE) Center under the Three Party Agreement. Since the procedure used for detection was revised between the two (2) outages, a comparison of the procedures was perfonned to identify changes that may have effected examination results. This comparison identified two procedural enhancements that may have contributed to differences in examination results between QlR14 and QlR15.
During both outages, the UT examination detection method used a 2.25 MHz,45-degree shear wave probe for detection. The procedure used in QlRIS required that a supplemental examination be performed when the weld configuration is limited to single side access. The supplemental examination used a different procedure, NDT-C-55, Ultrasonic Inspection of Welds Using Refracted Longitudinal Wave Techniques, designed especially for RL waves. Since the performance of shear waves is known to be adversely affected by austenitic weld metal, the purpose of the supplemental examination is to improve ultrasonic interrogation of the far side weld volume. The supplemental examinations were performed using a 2.25 MHz,60 degree RL wave probe.
Based on review of the available data, the supplemental examinations did not identify indications that were not also recorded using the 45-degree shear wave technique.
Additionally, no indications were identified on the far side of any weld examined. It is
                    -well established that RL waves are highly sensitive to deep IGSCC and can be used to assist in discriminating between confusing geometrical signals and IGSCC. Because the RL examination provides another look at the entire weld volume, it is likely that the supplementary examinations assisted the examiner in identifying additional IGSCC.
However, based on the review of QlR15 examination data reports and examiner comments, the recorded IGSCC flaws were clearly visfole using the 45-degree shear wave technique.
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Quad Cities Reactor Recirculation System IGSCC Root Cause                               i
                        ,      ,                                      SVP-98-376                                                 l Page 8 of 20                                               1 The second enhancement employed during Q1R15 was the revised approach for ultrasonically evaluating flaw length. During QlR14, the procedure required that the                   j end points of the flaw be determined by moving the probe along the flaw length until the               l corresponding signal response decreased to 20% ofits maximum amplitude. During                         i QlR15, this evaluatien was performed in a similar manner except end point was defined                   I when the signal response was reduced untilit could not be discerned from the noise level. This technique is intended to ensure that an IGSCC flaw having a correspondingly                i low amplitude response is adequately length sized. Because of this, the later technique                !
could potentially explain the increase in recorded length ofIGSCC flaws.
4.2.1.7 Previous Inspection Data                                                                       I Recirculation weld 02BS-F14 is a 28" diameter Pipe-to-Cast Elbow weld that had IHSI in 1984. Due to the cast elbow configuration, ultrasonic examination can only be performed from the pipe side of the weld (i.e. single-sided examination). :Immediately following the application ofIHSI in 1984, a " baseline" examination was performed.
Since 1984, weld 02BS-F14 was examined four more times. The examinations were performed in accordance with requirements applicable at the time. The summary of UT examination results and procedure used on weld 02BS-F14 is provided in the following table.- All indications are located in the HAZ on the pipe side of the weld. A summary of indications for all Unit I flaws on 28" diameter recirculation piping is presented in Table 4.2.1.2.
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Attachment Quad Cities Reactor Recirculation System IGSCC Root Cause
             ,     ,                                      SVP-98-376 Page 9 of 20 Table 4,2,1.7 Examination History of 02BS-F14 Examination History of 02BS-F14 UT        UT                  UT RESULTS                       PROCEDURE                 '
DATE      VENDOR 5/84       LMT       3 Linear Indications.            Comed C-2, Rev 13.                    1 100% DAC -Intermittent 360 ID    45* S-wave,2.25 MHz.
{
Geometry                                                              j UTL       Root Geometry                    UTL Procedure using a WSY-70, 2 MHz Probe Comed      ID Geometry                      Evaluation by L-III 10/87        GE       80% DAC - 360 Root Geometry      Comed C-2, Rev 15.                   1 45" S-wave,2.25 MHz.                 I 9/89       LMT       120% DAC - 360* Root Geometry    Comed C-2, Rev 17, 45* S-wave,1.5 MHz.
3/96       LMT       CircumferentialIGSCC             Comed C-2, Rev 22.
22" Long x 0.401" Deep           45* S-wave,2.25 MHz.
Comed Verified w/ WSY-70,2 MHz.
I1/98       LMT       CircumferentialIGSCC             Comed C 2, Rev 24.
45.5" Long x 0.83" Deep           4F* S-wave,2.25.Mliz.
60" RL Single Sided Supplement.
Comed Verified w/ Multiple Transducers Including 1.5 MHz S-l                                                         Wave.
A review of examination history for weld 02BS-F14 reveals that examinations have identified geometrical conditions caused by root geometry for every examination. The 02BS-F14 weld had intennittent root geometry UT signals reported in 1984. As additional examinations were performed, the root geometry signals were reported to encompass the entire ID. In 1996, a flaw was identified. Finally, in the most recent inspection, the flaw was identified again and was measured to have approximately doubled in length and had two locations with significant depth changes. The manual UT examination data did not include a detailed mapping of the flaw profile. However, the rapid change in length may be attributed to the root geometry masking the flaw. As the flaw increased in depth the existing length was then identified. The construction radiographs for this weld indicate the presence ofI.D. mismatch and areas where grinding and backwelding operations were performed. In addition, initial efforts to overlay this weld determined that the cast elbow is ovalized. Each of these conditions adds to the complexity of performing the UT examinations and discriminating small flaw indications from geometrical conditions. However, UT examiners qualified under the PDI program do receive additional experience in discriminating between small flaw indications and various geometrical conditions.
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Quad Cities Reactor Recirculation System IGSCC Root Cause SVP-98-376 -                                                i Page 10 of 20                                               j
_4.2.2 - _ Water Chemistry:                                                                               j
                                                               ~
As part the root cause investigation Comed review'ed the following water chemistry                       )
subjects for potential impacts on IGSCC growth rate.                                                    j i
                                          -e Chemical Decontamination (Decon)                                                                    !
e   ' Water Chemistry Excursions
                                                                                                                                                    -l e    Depleted Zinc Oxide (DZO) Injections                                                                j f                     .. Hydrogen ~ Water Chemistry Reference 1 provides a technical basis for the crack growth rate used in a structural                    j assessment for flaws to be left in service without repairs. Each topic is addressed below.-
                                          - 4.2.2.1 Chemical Decon:
Based on extensive testing by EPRI, Chemical Decon is not suspected to be a contributor to the IGSCC crack growth; however, Chemical Decon was considered to be a potential contributor to increased crack tip reflectivity. This would allow the UT examination to' detect flaws that were previously masked. This review determined when -
the Decon was performed in relation to the UT inspections of 02BS-F14. The UT inspection of the 02BS-F14 weld during QlR14 occurred after Decon. In QlRIS Refuel Outage the UT examination occurred prior to the Decon, thereby implying that Chemical Decon was not a contributing factor to the change in UT results between the two outages.
1 4.2.2.2 Water Chemistry Excursions.
Quad Cities Unit I had the best water chemistry quality in the plant's history during the                3
                                           -last operating cycle. This placed Quad Cities Unit 1 at the top of the industry's second                 i quartile, (i.e. WANO ' cpi of 1.03, and first quartile range from 1.0 to 1.02). In the last
                          ,               ' cycle Quad Cities experienced seven minor excursions that did not exceed EPRIs Water                     ;
                                           . Chemistry Guidelines (BWR Water Chemistry Guidelines - 1996 Revision, BWRVIP-29                         l EPRI Report TR-103515.) . Action Level 1 Criteria. The total time at Action Level 1 was less than 131 hours out of a total of 13,900 hours in the cycle, or less than 1%. The
                                          ; GE PLEDGE Model (Reference 1) predicts that the normal water chemistry maintained
: at Quad Cities Unit I for the past cycle would produce a crack growth rate three times lower than the crack growth rate in the NRC NUREG-0313 Rev 2. The crack growth rate in NUREG-0313_ was used in the previous acceptable flaw evaluation for Weld 02BS-F14 in 1996. Bdsed on this information it was concluded that water chemistry                        l t excursions could not play a significant part in creating an unusually high crack growth                  j rate.                                                                                                     1 l

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Attachment Quad Cities Reactor Recirculation System IGSCC Root Cause
                ,      ,                                    SVP-98-376 Page 11 of 20 4.2.2.3 Hydrogen Water Chemistry:
The recently completed Unit operating cycle (i e., cycle 15) history reveals availability      )
above 90% for the hydrogen injection system. The hydrogen injection resulted in an approximate concentration of 1.5 ppm hydrogen in the feed water piping. GE PLEDGE Model predicts that this amount of hydrogen, in conjunction with the Unit I water chemistry, would lower the crack growth rate by a factor of eight over the NRC established crack growth rates in NUREG-0313 Rev. 2.                                           '
However, hydrogen injection produces a reducing environment that will thin existing            j oxide layers on the recirculation piping. It is postulated in the GE report (Reference 1)      l that the reducing environment may have caused the flaw to be more reflective which, in         i turn, allows UT examination to detect the flaw indications.
i 4.2.2.4 Desolved Zinc Oxides (DZO)
Unit I started to inject DZO September 1998. Zinc injection is used to lower dose rates        I by replacing Co-60 concentrations in the lattice locations of the spinel oxide crystals.
Zinc injection is also credited with creating a thinner more protective oxide film (per Experience With Depleted Zinc Oxide Injection in BWRs, EPRI Report TR-1071883).
That may also have caused the flaw reflectivity to increase, which allows UT examinations to more readily detect these flaws (Reference 1).
4.2.2.5 Water Chemistry Conclusions The current operating water chemistry at Unit I supports the use of crack growth rates         ,
that are significantly lower than the rates given in NUREG-0313, Rev. 2. Therefore, the        '
use of the NUIEG rate in a structural margin assessment is conservative. The understanding ofIGSCC behavior and the specific water chemistry conditions at Unit 1 are inconsistent with the occurrence of significant crack growth in IHSI-treated piping welds.
Explanations for the changes in the UT-determined depth are most likely due to changes in the visibility of an existing crack. These changes are consistent with the changes that are known to occur due to hydrogen injection coupled with zinc addition, (i.e., reduced oxide layer). During the piping decontamination activities conducted during the recent l                 Unit I refueling outage, the first pass decontamination was more efTective than previous outages, indicating a reduced oxide layer.
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Attachment Quad Cities Reactor Recirculation System IGSCC Root Cause SVP-98-376                                               l Page 12 of 20 l
4.2.3 Original Recirculation Piping Fabrication 4.2.3.1 Radiography Review The original fabrication radiographs for weld 02BS-F14 were reviewed by the investigation team and a Comed Level III radiographic interpreter. The 02BS-F14 weld             !
was identified as FW-1-PS2B/202-1B during original construction.                                 !
Examination of the weld radiographs revealed evidence ofID mis-match, ID grinding and back welding at the root.
l 4.2.3.2 Coldworking and Grinding                                                                 i The ID mis-match could be related to the ovality in the elbow observed when the automatic welding equipment track was set-up. ID mis-match could lead to lack of fusion defects. In large diameter pipes, such as these 28" diameter recirculation lines, access could be gained to grind the ID surface and perform back welding to pass radiography. Any back grinding of the ID surface will induce significant cold work and residual tensile stresses. Both of which enhance the potential to initiate IGSCC.
4.2.3.3 Back Welding The back welding as well as ID mis-match will cause problems during later UT examinations of the weld. These type of geometric discontinuities result in the UT signals identified as root geometry and can mask flaw indications.
4.2.3.4 Field Welds vs. Shop Welds IGSCC experience at Unit I does not indicate significant differences in crack initiation between field and shop welds in the 28" diameter piping.
4.2.4 Stress Improvement Process There are two types of stress improvement processes that have been employed on Unit 1. One is the Mechanical Stress Improvement Process (MSIP). The other is Induction Heating Stress Improvement (IHSI). Figures e 2.1-la and b present the recirculation welds treated with IHSI and MSIP procera for recirculation loops A and B, respectively.

Attachment Quad Cities Reactor Recirculation System IGSCC Root Cause SVP-98-376 Page 13 of 20 4.2.4.1 Mechanical Stress Improvement Process There are fifteen welds within the recirculation system (including ten nozzle welds) that   3 have been treated using MSIP. The fact that there have been no post-MSIP welds with flaws identified in the domestic BWR fleet, leads to the conclusion that MSIP treated welds are behaving as anticipated, that IGSCC has been mitigated by MSIP, and that the welds treated with MSIP are appropriately classified as Category C, as dermed in            >
GL 88-01, 4 2.4.2 Induction Heating Stress Improvement As was noted earlier in this report, IGSCC was detected in the large bore piping (28")
on Unit I during the 1996 QlR14 refueling outage. This was unexpected in that the piping had been treated with the IHSI process. This process uses an induction heating coil to create a magnetic flux field to generate heat within the pipe. This heating changes the stress patterns within the pipe such that the inner surface of the pipe is placed in compression, thereby reducing the susceptibility to IGSCC. Given that flaws occurred in piping treated by the IHSI process, it was determined that the process effectiveness needed to be evaluated.
An evaluation was first conducted during the 1996 Unit 1 outage. The evaluation was limited to the eight 28" diameter welds where flaws had been detected. The evaluation determined that the IHSI treatment fully met the industry criteria for temperature and
.                  coil placement for only two of the eight welds. For three other welds, the temperature I
criteria were not met and additional calculations were necessary to confirm proper temperature gradients were developed. Three welds were determined to have received a treatment that did not meet the guidelines due to coil placement, thus limiting the region that was treated.
During QlR15, additional IGSCC cracking was detected in 28" diameter recirculation piping. One 28" diameter weld, (i.e.,02BS-F14) that was determined to be flawed in 1996 had measured growth that exceeded what would reasonably be expected. In light of these results and the results of the 1996 outage, it was determined that all 28"         ,
diameter piping IHSI records should be reviewed.
The key to the effectiveness of the 1HSI treatment is the achievement of the proper through thickness temperature gradient over a suflicient length of pipe. This is impacted by the temperature achieved at the outer surface of the pipe, the flow rate and temperature of water through the pipe, coil length, and placement. The procedure used measured surface temperatures and established a minimum water flow rate. If the measured temperatures were outside the procedural criteria, heat transfer calculations that considered the water flow were performed to assess the through thickness temperature gradient. Coil placement either met the specified lengths or did not. Coil 1

Attachment Quad Cities Reactor Recirculation System IGSCC Roe t Cause SVP-98-376 Page 14 of 20 length was affected by component geometry and materials. Generally, the desired coil length was not attained when piping was attached to a fitting, (i.e., valve, pump, or tee) or a dissimilar metal component. If the desired coil lengths are not obtained, it is now known that IHSI effectiveness is questionable.
In the current IHSI review, a simple grading system was used to classify the results of the IHSI assessment. If a weld treatment was determined to meet both the temperature and coil placement / length criteria, it was identified as " effective." If a weld treatment required heat transfer calculations to confirm proper temperature gradient, it was identified as " limited - temperature." In a similar manner, if coil placement did not meet the procedure criteria for a weld, that treatment was identified as " limited - coils."
Limitation of the coil placement is judged to be more significant than temperature limitations. Finally, if the coil placement and the temperature both were outside the procedure criteria, the IHSI was identified as " ineffective." The results of this review and those from the 1996 evaluation are contained in Table 4.2.4.2.
Examination of the results in Table 4.2.4.2 reveals that the majority of the welds either received a limited treatment or an ineffective treatment. Of the four welds with new or changed cracking identified this outage, two welds, (i.e.,02AS-F9 and 02AD-F8) received IHS1 considered as ineffective and one, (i.e., weld 02BS-F7) was limited based on coillength. The other weld,02BS-F4, was classified as having received an ineffective IHSI. Weld 02BS-F14 that had apparent significant flaw change during the last operating cycle was previously evaluated to have received an ineffective IHSI.               i An industry study has concluded that the IHSI process, when applied properly, provides the intended mitigation ofIGSCC in both small and large diameter austenitic stainless steel BWR pipe welds. It further concluded that the most likely cause(s) for the post-IHSI IGSCC observed in the four plants described in the study is the result of pre-existing IGSCC which was undetected following the IHSI treatment. Nothing was discovered in the course of the investigative team's evaluation that would challenge those conclusions. In fact, given that weld 02BS-F14 apparently received an inefTective IHSI treatment, it is reasonable to conclude that the observed crack indication is due to a flaw indication that was missed years earlier and has slowly grown to a size that can now be detected with improved UT techniques. Further, the review of the IHSI process employed on Unit 1 indicates that many of the 28" diameter piping welds may not have received an effective IHSI treatment and that IGSCC was not an unexpected condition at the time ofIHSI application.
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Quad Cities Reactor Recirculation System IGSCC Root Cause SVP-98-376                                             l Page 15 of 20                                           l l
l TABLE 4.2.4.2 Weld Number vs. IHSI Effectiveness                                        >
WELD NUMBER                                              IIISI EFFECTIVENESS 02AS-F2                                             Limited- Temperature 02AS-S3                                                     Effective 02AS-S4                                             Limited - Temperature 02AS-F5                                             Limited - Temperature   1 02AS-F8                                                    Ineffective 02AS-F9                                                    Ineffective 02AS-S12                                                     Effective       !
02AS-F14                                                    Ineffective 02AD-Fi                                              Limited - Temperature 02AD-S2                                              Limited -Temperature 02AD-S6                                                      Effective        l
,;                                          02AD-F8                                                     Ineffective 02AD-F9                                              Limited - Temperature 02AD-F12                                                   Limited - Coil 02BS-F2                                             Limited-Temperature
                                                                                                                       ~
02BS-S3                                                      Effective 02BS-F4                                                      Effective 02BS-F6                                                     Ineffective 02BS-F7                                                   Limited - Coil 02BS F14                                                    Ineffective 02BD-Fi                                              Limited - Temperature 02BD-S2                                              Limited - Temperature 02BD-S6                                                      Effective       !
02BD-F8                                                     Ineffective 02BD-F9                                                     Ineffective 02BD-F12                                                   Limited - Coil Notes:
: 1. Effective = Temperatures and coil length met procedure criteria.
: 2. Limited - Temperature = Heat transfer calculations were required.
: 3. Limited - Coil = Coil length criteria was not satisfied.
: 4. Ineffective = Coil length criteria not met and heat transf:r calculations required.

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4.2.5 Abnormal Loads The investigation reviewed operating experience during Unit I cycle 15 for abnormal l                 loads that'may have contributed to the flaws found in this examination. It also looked at loads that may have been generated during maintenance activities that took place between the QlR14 and QlR15 examinations. The following potentialloads were investigated.
* Operating transients e   Design Changes                                                                       :
* Pump /Line Vibrations e    Pipe Whip Restraint Interferences e    Locked Snubbers
* Rigging / Maintenance Loads e    Weld Overlay Application and Shrinkage 4.2.5.1 Operating Transients
                ' Based on discussions with the system engineer, the recirculation system operated normally during fuel cycle 15. No thermal or hydraulic transient events that would generate abnormalloads occurred.
4.2.5.2 Design Changes Based on discussions with design engineering, operating loads in the recirculation system did not change significantly between QlR14 and QlR15 as a result of system modifications. The operators for the pump discharge valves were modified during the previous refueling outage, QlR14. The design evaluation for the change in operators determined the impact on piping stresses to be insignificant.
4.2.5.3 Pump /Line Vibrations Recirculation pump vibration levels were monitored on-line during fuel cycle 15. The vibration engineer has confirmed that pump vibration levels were consistent with prior levels and normal for the system.
4.2.5.4 Pipe Whip Restraint Interferences l
l_                Design engineering reviewed piping clearances with pipe whip restraints and confirmed that the whip restraints did not restrain the piping during normal operation.
Consequently, the whip restraints did not cause any abnormal loads.
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I Attachment Quad Cities Reactor Recirculation System IGSCC Root Cause                             ,
SVP-98-376                                              l Page 17 of 20                                           )
4.2.5.5 Locked Snubbers l
During the QlRIS snubber testing, two snubbers on a lateral trapeze support on the RHR suction line that branches off the Loop B recirculation suction piping were found             ;
to be locked. The engineering evaluation determined the impact on piping stress values            i during the fuel cycle as being low and within code acceptance criteria. The thermal pipe          ;
stress at weld 02BS-F14 was reviewed by the investigation team and determined to be               l 3780 psi. This stress level is low and is not considered to have had an affect on the flaw        l growth at 02BS-F14.
4.2.5.6 Rigging / Maintenance Loads
                   . Based on discussions with design engineering and the structural engineer, the recirculation piping did not experience a significant change in pipe stress due to                i miscellaneous rigging or lead shielding during the prior or current outage.
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4.2.5.7 Weld Overlay Application and Shrinkage l
l Five weld overlay repairs were applied during QlR14 to repair welds with indications in the recirculation system. The system evaluation of the piping axial shrinkage from these overlays was evaluated and found to produce insignificant piping stresses; e.g.,200 -             !
300 psi.
Weld 02BS-S12 was repaired during QlR14 and is approximately 12" upstreer> afweld 02BS-F14, Figure 4.2.4-3. Since the end of the overlay repair for 02BS-S12 was only 7" from the 02BS-F14 weld, the thermal through-wall bending stresses generated during the weld overlay application were evaluated for their impact on 02BS-F14. EPRI Report NP-944, " Studies on AISI Type-304 Stainless Steel Piping Weldments for Use in BWR Application," December 1978, documents the results of butt welding studies, which included thermocouple measurements of a 24" diameter, schedule 80 pipe while being welded using very high heat inputs; e.g., pass 1 - 25,000 joules /in, pass 2 - 35,000 joules /in and pass 3 - 60,000 joules /m. The welding was performed without water in the           ;
pipe and the temperature measurements of the piping inner surface were recorded for several axial positions moving away from the weld. This test showed that the pipe wall peak temperature decreased rapidly, i.e. 700 F in 0.5", and will reach ambient temperature in less than 2" from weld. Based on these test results, it was concluded that the through-wall bending stresses would be insignificant at distances greater than VRt from the ambient temperature position. For this piping, the VRt distance is 4.5" 02BS-F14 weld is 7" away from the end of the overlay repair for 02BS-S12. Based on this distance, the stress associated with the application of the overlay repair should not have affected the flaw in weld 02BS-F14.

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Quad Cities Reactor Recirculation System IGSCC Root Cause                        {
SVP-98-376                                          l Page 18 of 20
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4.2.5.8 Applied Stresses for Weld 02BS-F14                                                   1 The table below presents the design basis stresses acting on weld 02BS-F14:
1 Table 4.2.5.8                                        ;
1 Weld                       Pressure     ' Dead Weight          Thermal         WOL Shrinkage ;
02BS-F14                     (psi)            (psi)             (psi)              (psi)     J Design                       6439              851              3840                204      i Locked Snubber                ---              ---
3780                 ---
l 1
These pipe stresses are very low and are not considered large enough to change the           ;
residual stress field induced by the original welding or as improved by IHSI.                {
                    ' Consequently, the residual stress field will be the governing mechanism for IGSCC           j growth rate at this flaw.
l After reviewing the potential causes for abnormal loads, it was determined that there        j were no sources of unanticipated loads large enough to have caused a flaw of the size        i found in 02BS-F14. The normalloading applied also would not explain the unexpected           j flaw size in weld 02BS-F14.                                                                  J 4.2.5.9   Previous Industry Events                                                           i A search ofindustry experience revealed only one plant with a similar experience,            ,
Millstone Nuclear Power Station Unit 1. A Licensee Event Report (LER) was issued              I when they found cracked welds on previously IHSI treated piping. This information was considered in this investigation. The root cause investigation associated with the LER concluded that similar causal factors were found in their investigation (e.g. improved inspection techniques and procedures).
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Attachment                                                l Quad Cities Reactor Recirculation System IGSCC Root Cause                            I
                      ,    ,                                     SVP-98-376 Page 19 of 20 l
5.0      Conclusions 1
As noted before, the indication behavior of 02BS-F14 bounds all the other flaws                   I detected. The following summarizes the results of this investigation that support this conclusion.                                                                                       i There are multiple causes which are believed to have acted together to result in the observed flaw behavior.                                                                           I i
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                                  -The IHSI treatment of the 28-inch diameter pipmg was not fully effective.               ;
                                  -Undetected pre-existing flaws were present.                                             l
                                  -Heavy root geometry signals masked existing flaws.
                                  -Water chemistry (HWC and DZO) increased flaw reflectivity.
                                  -Normal Growth Rate added to small pre-existing flaw size.
1 This investigation concludes that there is reasonable likelihood that a small pre-existing         '
flaw was present and undetected following the IHSI that was performed in 1984. An existing flaw may have been present and was closed by IHSI, thus making it more difficult to detect. There would continue to be root geometry signals identified in subsequent UT examinations as a result of the fabrication process. These factors could limit the detection of existing flaws when shallow,                                                ;
The changes both in inspection procedures and personnel qualifications used in QlR15 provided enhanced flaw detection capability. This increase in flaw detection capability coupled with increased reflectivity from minor flaw growth and changes to the oxide layers acted together to make any existing flaws more easily detectable.
It is concluded that the flaw in weld 02BS-F14, did not grow in length from 22 to 45.5 inches. Nor did the flaw grow in depth from 0 to 0.83 inch. Rather, this was a long shallow or series of shallow flaws that grew over time and have now been detected.
Although the growth rate may have varied at times based on water chemistry, the rate is within the range that could be expected for normal BWR operation.                                  ,
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Attachment Quad Cities Reactor Recirculation System IGSCC Root Cause SVP-98-376 Page 20 of 20

==References:==
: 1. Letter from J.P. Dim nette, Jr. (Comed), SVP-98-353, to USNRC, dated November 24,1998, " Request for Approval of Pipe Weld Overlays and Pipe Flaw Evaluations,"
docketing " Assessment of Crack Growth Rates Applicable to Induction Heating Stress Improvement (IHSI) Recirculation Piping in Quad Cities Unit 1" GE Nuclear       ,
Energy Report GE-NE-B13-01980-030-2 November 1998
: 2. Letter from USNRC to Chairman of Boiling Water Reactor Owners Group, dated March 1,1996.
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FIGURE 4.2.4-3                                   %
1B Pump Suction Dimensions
                                                               -                                ~
m,          ,,                                                       ,
02BS.F6                                                                                                                                                                                        :
R EC:: RC                                                               !
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                                                                                ~-~
l   28" SUCTION JUV3 MO 1-202-48                           g SUCTION VALVE                                                                                         02BS.F7 592 elv.
_                              ___            __________________t                                                                                                                                              >_ _ _ _ _ _ _ _
i 7'- 1 1"                                                                                                                                                                                                                                                              !
Overlav Not Water Backed                                                                                                                                     .
I 02BS.S9                                                                                                          C--                                                                             ,
I                                                                                                                                                                '
U 7
7"                   ,        ___    /-02BS.S 15 wn m                                                                                                                              J                d 02BS.511                                                                                                                                                                                                                     7 _4" gi                                                                                                                                           i                                                                      ;
DECON FLANGE l
                                                                                                                                                                                                                       -iJn                                                                            :
t 02BS.S1'2                                                                                                                          02BS.F14 3 ' - 6"_                                                   _
13"                                                               _
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t

t Figure 4.4 Detection and Mitigation of IGSC                                                                                                                                                                                                                                                                                           3 7
IGSCC inspections                                                                                                                                                                                                                                                                                    *
                                                                               .        Ch stry i,
1
               ,i                                                              .
                                                    .UT Tochnology Choice Previous inspection Data                                             Operational Events __
Procedures and                                                              HW Chemistry Manual Detection Technique
                                            --           NEssentialVariables
                                                                 '- PDiQualification                                                                                    DZO Evaluations and Growth Estimate Personnel and Training ___r_                                                                                       _ - -- Decontamination                                                                                                                                                                   ,
i                   <
PDI Qualification _/                                                                                                                                                                                                                                                              l                   l i
GL 88-01                                                                                                                                                                                                                                                                                            m I ident:fication of '
inspections                                                                                                                            y                                                                                                                             Operating q New Weld Cracks !
Back Welding                                                                                                                                                                                              Transients                        l Field Welds vs Shop Welds
* IHSI                                                                                                                                                                                  -
l Water Temp. & Flow                                    >,7 - >                                                                                                                                                              Overlays Rigging and Component Geometrics
                                                                                                                                  /                                                                                                                                       Maintenance-Fabrication Techniques _                _p                                                           e'                                                                                                               Des.ign Snubbers (locked)
Delta Temp -'-
Changes Contruction RT Review-__. _.                   Cold Work / Grinding Coil Length -- >/                                                                                                                                                                                       _ _._.__ Whip Restraints f                                                  p Vibratens
____y                                      - . _ _
Original Recire.                                     .
Stress             ,                                                                                                                                                                   i Pipe Fabrication                                      !      Improvement           j                                                                                                            Abnonnal                                               ;
i                                                                                                                       Loads                                       ;
                                                                                                         .__J
_ --                               w DRIVING FORCES FOR CRACK PROPAGATION

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