JPN-97-028, Forwards Position Re Visual Inservice Insps of CR Drive Cap Screws During Refueling Outage 12 Along W/Plant Specific Analysis for Plant

Forwards Position Re Visual Inservice Insps of CR Drive Cap Screws During Refueling Outage 12 Along W/Plant Specific Analysis for Plant

ML20210V107
Person / Time
Site:	FitzPatrick
Issue date:	09/16/1997
From:	James Knubel POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
JPN-97-028, JPN-97-28, NUDOCS 9709230039
Download: ML20210V107 (19)

Text

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o 123 Main Street Vv%te Plains, tJew York 10801

, 914 681 6840

, 914 287 3309 (FAX)

,1 .

1. D NewYorkPower sO!"%.ni-tv Authority ""***' o" ~

September 16,1997 JPN-97-028 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, D.C. 20555

Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Control Rod Drive Reactor Pressure Vessel Housing Flange Bolt inspection and Roolacement Durina Refuelina Outaae 12

Reference:

NYPA letter, J. Knubel to USNRC dated July 15,1997 (JPN-97-023) regarding inservice Inspection Program Relief Requests for the Second Ten Year Interval Closeout and Revised Summary Report Attachment I to JPN-97-023 (Reference) stated that the Authority would provide a copy of its position regarding the visualinservice inspections (VT-1) of control rod drive cap screws during refueling out"ge 12 along with the plant specific analysis for FitzPatrick. Attachment I provides this information. This letter results in no new commitments, if you have any questions, please contact Mr. Art Zaremba at (315) 349-6365.

Very truly yours, J. Knubel Sr. Vice President and Chief Nuclear Officer att: as stated ' if cc: next page 9709230039 970916 " \

PDR ADOCK 05000333 G PDR I hl

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i

'. cc: Rrgional Administrator U. S. Nuclear Regulatory Commission

.,475 Allendale Road

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. King of Prussia, PA 19406 Office of the Resident inspector U. S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093 Ms. K. Cotton, Acting Project Manager -

Project Directorate 1-1 Division of Reactor Projects-l/Il U. S. Nuclear Regulatory Commission Mail Stop 14 B2 Washington, DC 20555 -

_ _ _ _ - . - - - - - - 1

Attachment 1 to JPN-97-028 CONTROL ROD DRIVE REACTOR PRESSURE VESSEL HOUSING FLANGE BOLT INSPECTION AND REPLACEMENT REFUELING OUTAGE 12 September 16,1997 New York Power Authority James A. FitzPatrick Nuclear Power Plant

Attachment 1 to JPN 97-028

, CONTROL ROD DRIVE REACTOR PRESSURE VESSEL HOUSING FLANGE BOLTS 1997 EVALUATION OF FLAWS FOUND IN BOLTS RETIRED AND REPLACED DURING RO12 PER GE SIL 483 INTRODUCTION:

Purpose:

The purpots of this report is to document the course of action taken by the James A.

FitzPatrick (JAF) Nuclear Power Plant in the resolution of In-Service inspection (ISI) technical issues that arose during Control Rod Drive (CRD) assembly maintenance during the 1996 refueling outage. The maintenance consisted of normal planned outage CRD assembly maintenance and the concurrent installation of CRD reactor pressure vessel (RPV) housing flange modifications. Specifically, the housing flange modifications involved replacement of original design CRD housing flange bolts with a new design bolt. During normal planned outage CRD assembly maintenance, associated RPV component ISIis also routinely scheduled and completed. During the installation of the modification, the CRD bolting ISI inspection requirements for visual examination of the removed bolts was not performed in accordance with the proper procedural change process.

This report documents the JAF engineering processes which resolved the technicalissues 4 associated with this issue while complying with Nuclear Regulatory Commission (NRC) licensing commitments and internal engineering and quality assurance (QA) procedural requirements established for JAF.

Scope:

The scope of this report is confined to the JAF CRD housing flange bolt modification and ISI inspection history performed during and after the 1996 refueling outage 12 (RO12).

BACKGROUND:

CRD

Description:

Control Rod Drives are mechanisms which are used to raise and lower neutron absorber control rods into and out of the nuclear fuel core. By controlled removal and insertion of neutron absorber control rods, the reactor operators can control nuclear reactor criticality and the resulting amount of heat generated by the nuclear fuel core. (The water that covers the nuclear fuelis a heat transfer medium, but also acts as a moderator. The coritrol rods act as " variable" neutron absorbers.) Neutron absorber control rods have limited service life, because the absorber materials and structural materials in the control rod blades are expected to degrade when exposed to high energy neutron fluxes in the reactor core for a prolonged period of time. Therefore, CRD assemblies are periodically removed, shuffled and/or replaced as part of the normal maintenance activities of the plant.

The Control Rod Drive mechanisms are part of the Control Rod Drive Hydraulic System.

1

Attachment 1 to JPN-97-028 CONTROL ROD DRIVE REACTOR PRESSURE VESSEL HOUSING FLANGE BOLTS 1997 EVALUATION OF FLAWS FOUND IN BOLTS RETIRED AND REPLACED DURING

, RO12 PER GE SIL 483 The CRD assembly consists primariy of a control rod drive to blade coupling, a drive piston, a stop piston, a collet mechanism and a check valve. The assembly is bolted to the CRD reactor vessel housing flange (nozzle) on the bottom head of the reactor vessel.

There are 137 CRD nozzles on the bottom head of the reactor vessel. The bolted closure cor.sists of an integral reactor vessel flange, a non-standard (from an ANSI perspective) eight-inch flange on the CRD assembly and eight CRD assembly fasteners, which are 1" socket head cap screws. These cap screws are called main flange bolts or CRD bolts. The CRD bolts are threaded into the reactor vessel-housing flange without nuts. Figure 1 shows the location of the CRDs in the RPV (Reference 1). Figure 2 shows a typical CRD ascembly (Reference 1). Figure 3 shows a map of control rod locations in the JAF core and the history of blade replacement or shuffling for the existing control rod configuration (Reference 1).

CRD Housina ISl:

CRD Housing ISI inspections, including botting, are performed during CRD assembly removal, shuffling and/or replacement, which is conducted as part of routine maintenance activities. Part of the CRD housing ISIinspections are visual examinations (VT-1) of CRD bolts, which attach the CRD assembly to the housing flange.

The JAF ISI program is a licensing commitment documented in the Facility Safety and Analysis acport (FSAR), Chapter 16 (Reference 1). The commitment in place during RO12 was to comply with the requirements of Section XI of the ASME Boiler and Pressure Vessel Code,1980 Edition with Addenda up to and including Winter 1981 (Reference 2). (Unless otherwise stated, further references in this report to the ASME Code will be to this section and version.) 'Mie IWB-2500-1, ASME Code, contains a requirement for a visual examination W 1-1) of the exterior surface of the CRD bolts, when they are removed during the maintenance or replacement of a CRD assembly.

CRD Bolt Crackina:

During such routine inspections at another nuclear plant in 1988, indications of cracking were identified in CRD bolts in the cap screw head to shank transition region (or shoulder).

General Electric (GE), the supplier of the CRD bolts, evaluated the problem generically for all BWR reactor vessels, and issued a Rapid Information Communication Services information Letter (RICSIL) (Reference 3.a) and subsequent Service Information Letters (SIL) to all BWR reactor vessel owners (References 3.b,3.c,3.d). The crack initiation mechanism was determined to be stress corrosion cracking (SCC) caused by the high preload stress in the CRD bolt and the presence of trace amounts of rr.anganese sulfide inclusions in the CRD bolt 4140 Cr-Mo carbon steel. Moisture on the CRD bolts from CRD removal and replacemcat inay contribute to the cracking. Moisture on the CRD bolts from CRD main flange O-Ring Irakage may also contribute to the cracking.

2

Attachment 1 to JPN-97-028

~ ' CONTROL ROD DRIVE REACTOR PRESSURE VESSEL HOUSING FLANGE BOLTS 1997 EVALUATION OF FLAWS FOUND IN BOLTS RETIRED AND REPLACED DURING RO12 PER GE SIL 483 Upon receipt of GE RICSIL 019, dated May 19,1988, JAF evaluated the information, and existing plant procedures were found to contain adequate steps to identify and correct CRD bolt cracking. Upon receipt of GE Sll 483, Rev.0, dated March 17,1989, JAF re-evaluated the information, and existing plant procedures were again reviewed for adequacy. This review found that GE recommendations were already implemented in plant procedures. OC personnel were interviewed and given a heightened awareness of the potential problem. Upon receipt of GE SIL 483, Rev.1, dated October 1,1991, JAF implemented GE recommendations to perform enhanced inspections of its CRD bolts for stress corrosion cracking. The recommended enhanced inspections include a magnetic particle test (MT) or liquid penetrant test (PT) of suspect CRD bolts. During the 1992 refueling outage (RO10), these inspections observed CRD bolt cracking at JAF and initiated _,

the short term corrective actions recommended by GE in GE SIL 483. CRD bolts which showed no indications of cracking were returned to service CRD bolts which showed indications of cracking were retired from service and replaced with "new" original design CRD bolts.

Lona Term Corrective Action:

The long term corrective action to address CRD bolt cracking is modification M1-92-319.

This corrective action involves replacement of the original design bolts with a new design offered by GE. A comparison of the original design bolts and the new design bolts is summarized in Table 1 (Reference 4). Starting in 1994 (RO11), all original design CRD bolts that were in CRD assemblies removed during routine maintenance were retired and replaced by new design CRD bolts. After completion of RO12,64 of the 137 CRD mechanisms had new design CRD bolts.

DISCUSSION:

ISI ResDonSibilitle5:

The JAF nuclear plant was committed to an in-service inspection program based on the requirements of Section XI of the ASME Code for the second ten-year ISIinterval, which extended form 1987 to 1997. RO12 was in the second ten-year interval.

During RO12, the original design CRD bolts which were associated with the CRD mechanisms removed for replacement and maintenance were retired and not returned to service in accordance with modification M1-92-319. Despite their retirement, the CRD bolts were required to have ISI inspections by the ASME Code. These ISI inspections were not intended to qualify the bolts for continued use. The inspections were intended to identify any new CRD bolt problems, which might affect the new design CRD bolt, and to monitor for any changes in the extent of condition of the original design CRD bolt cracking, 3

Attachment 1 to JPN 97-028

, CONTROL ROD DRIVE REACTOR PRESSURE VESSEL HOUSING FLANGE BOLTS 1997 EVALUATION OF FLAWS FOUND IN BOLTS RETIRED AND REPLACED DURING RO12 PER GE SIL 483 which might affect the modification M1-92-319 replacement schedule.

The schedule for replacement of the original design CRD bolts with the new design CRD bolts was to replace the bolts during normal maintenance removal of the CRD mechanisms.

The replacement schedule documented in modification M192 319 was contingent upon arrested or very slow crack 3rowth, GE concluded that the crack growth was arrested or very slow based on their rnetallurgical evaluation of the shape of the crack tips (Reference 3.d.) The RO12 ISI inspections were intended to confirm the assumption by identifying any exacerbation or acceleration of the cracking. Any exacerbation or acceleration in the extent of the cracking condition would then accelerate the CRD bolt replacement schedule, in addition, Article IWB-2430, ADDITIONAL EXAMINATIONS, ASME Code, requires that the sample size be doubled if any flas indication exceeds its acceptance standard, Classifying an inspection finding fo an ASME Code component as an acceptable flaw indication is important in determinir.,, extent of corrective actions required by the ASME Code. A flaw indication is considered to be one of the following:

(1) a manuf acturing imperfection which is small enough to have met the fabrication acceptance criteria for the ASME Code component; (2) an inspection anomaly which is the result of the limits of the accuracy of the inspection method, e.g., visual inspection (VT), liquid penetrant test (PT), magnetic particle test (MT), ultrasonic test (UT) or eddy current test (ET):

(3) an actual flaw in the ASME Code component, which has started to grow as the result of the operating condiGGn; ge.g., loads and resulting stresses) and/or environment (e.g., temperature, pressure, chemistry, etc.). -

An actual flaw as described above may be below a threshold flaw size to be expected to centinue to grow significantly, if at all, between reinspection intervals. In this case, the ASME component may be monitored for flaw growth or may be scheduled to be replaced.

The basis for determination of the threshold flaw sizes for ASME Code components is documented in EPRI NP-1406-SR (Reference 5).

Alternatively, an actual flaw may be an imperfection in an ASME Code component which has started to grow as the result of the operating conditions (e.g., loads and resulting stresses) and/or environment (e.g., temperature, pressure, chemistry, etc.), and which is of a size that could be expected to grow significantly between reinspection intervals. This kind of flaw indication requires a detailed fatigue, fracture mechanics and environmental qualification evaluation prior to returning the ASME Code component to service to ensure that the expected flaw size at the next reinspection meets the requirements of the ASME Code. IWB-3600, ANALYTICAL EVALUA flON OF INDICATION, ASME Code, contains the acceptance criteria for this kind of evaluation, in this case, the ASME Code component is 4

I

B Attachment 1 to JPN 9%028

, CONTROL ROD DRIVE REACTOR PRESSURE VESSEL HOUSING FLANGE BOLTS 1997 EVALUATION OF FLAWS FOUND IN BOLTS RETIRED AND REPLACED DURING -

RO12 PER GE SIL 483 retired, scheduled for replacement or modified to prevent continued flaw growth.

Activities Durina RO12:

JAF scheduled 22 CRD assembly replacements for RO12. The work scope included replacement of old design CRD bolts with the new design CRD bolts for the sffected CRD assemblies. Due to mistakes in detorquing incorrect CRD mechanisms, additional CRD assembly bolts were retired and replaced.

During the performance of CRD assembly replacements, vendor personnel grouped all s removed CRD bolts into a single population for disposal. No visualinspections of the CRD bolts, which were removed and retired, were performed. Visualinspections.of new design CRD bolts which were installed were completed.- This oversight was the result of an error in a procedure revision. Prior to the outage, wording in maintenance procedure MP-004.03, CRD Removal and Replacement, (Reference 7) was changed from " perform a VT-1 ISI examination of the removed flange bolts" to " perform a VT-1 ISI examination of removed or replacement flange bolts." This change was approved without appropriate reviews. The wording of the procedure change should have been " perform a VT-1 examination of removed and replacement flange bolts."

Activities After RO12:

After the outage was completed, a review of the RO12 ISIinspection records was conducted. The purpose of this review was to prepare the outage ISI summary report. No records of inspections of retired CRD bolts, which had been in service and had been replaced with new design bolts, could be found. The retired CRD bolts had been classified as radioactive waste and had been disposed of off site. The retired CRD bolts were located, and the CRD bolts were then given VT-1 and PT inspections. The problem with the maintenance procedure was identified and the procedure was revised to correct the error.

Results of the inspections indicated that 6 of the 306 inspected CRD bolts had indications which exceeded the linear indication acceptance criteria of 1/4" for non-axia!

(circumferential) indications documented in the JAF non-destructive visual examination procedure for VT-1 (Reference 8). (304 of the 306 CRD bolts had been removed during the outage. The additional two CRD bolts were previously retired and were inadvertently mixed in with the 304 CRD bolts retired during RO12.) Two bolts had linear indications 9/16" long; three bolts had linear indications 7/16" long and one bolt had aligned rounded indications caused by corrosion and pitting, whose proximity to each other required linkage into a postulated linear indication greater than 1/4" long per the ASME Code. Allindications were circumferential (Reference 9) =The consequence of the ISI findings, if accepted without further engineering evaluation, was to mandate a doubling of the ISI sample size of

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l l

Attachment 1 to JPN 97-028

, ,_ CONTROL ROD DRIVE REACTOR PRESSURE VESSEL HOUSING FLANGE BOLTS 1C97 EVALUATION OF FLAWS FOUND IN BOLTS RETIRED AND REPLACED DURING RO12 PER GE SIL 483 CRD bolts in accordance with Article IWB-2430, ADDITIONAL EXAMINATIONS, ASME Code, in light of industry experience with this issue, this was not considered prudent due to the complexity and high dose nature of the work. Also, it was uncertain whether further examinations would improve plant safety. The CRD bolt cracking mechanism had been previously identified and a corrective action, modification M192-319, was in the process I

of implementation.

Licensina Issues /ADDlicable ASME Code Reauirements:

The Licensing Department at JAF evaluated the plant ISI licensing commitments to the

NRC to determine which ASME Code Edition was applicable and to determine whether a

, supplemental acceptance criteria was permitted for JAF in the applicable ASME Code Edition. Results of this evaluation are contained in Reference 10 and are summarized

, below.

i Application of supplemental examination methods and techniques is permissible, if this does not constitute a modification of an acceptance standard per IWB 3410, MODIFICATION OF STANDARDS, ASME Code.

^

VT-1 indication acceptance criteria for ASME Code bolts less than or equal to 2" in diameter (i.e., CRD bolts) are not provided by IWB-3000, ACCEPTANCE STANDARDS FOR FLAW INDICATIONS, ASME Code. Analogous VT-1 indication acceptance standards are provided by IWB-3000, ACCEPTANCE STANDARDS FOR FLAW INDICATIONS, ASME Code,1989 Edition, with no Addenda. Lacking explicit guidance in the Code of Record, these standards were adopted as acceptance criteria for VT 1 indications for the second 10-year interval inspections.

IWB-2430, ADDITIONAL EXAMINATIONS, ASME Code, requires doubling the sample size in the event that any indications exceed the allowable indication standards of IBW 3000.

This explicit requirement is applicable for those indications which have acceptance standards clearly defined in the Code. This additional examination requirement may be conservatively assumed to be an implicit requirement for indications which have no clearly define) ecceptance standard.

IWB-3200, SUPPLEMENTAL EXAMINATIONS, gives the latitude to use other examination methods and techniques to determine the character of a flaw. The application of supplemental examination methods and techniques does not constitute a modification of an acceptance standard which is not explicitly defined in the Code of Record, in this case, application of supplemental examination methods and techniques supplements an existing acceptance criteria to more accurately determine the nature of a flaw, and is not a modification of an acceptance standard per IWB-3410, MODIFICATION OF STANDARDS.

6

Attachment 1 to JPN 97-028

, CONTROL ROD DRIVE REACTOR PRESSURE VESSEL HOUSING FLANGE BOLTS 1997 EVALUATION OF FLAWS FOUND IN BOLTS RETIRED AND REPLACED DURING RO12 PER GE SIL 483 Therefore, the conclusion of the Licensing Department evaluation was that it is acceptable to supplement the acceptance criteria given in the non-destructive visual examination procedure with supplemental acceptance criteria which can be applied using supplemental examination methods and techniques.

Previous Enaineerina Evaluations of CRD Bolt Crackina Performed at Other BWR Nuclear P)wer Plants:

A search of licensing submittals prepared for other BWR CRD Bolts on this issue was conducted. Docketed correspondence relevant to CRD bolt cracking evaluations performed by other license holders were reviewed. In those evaluations, the CRD bolt crack flaw indications were conservatively characterized as ASME Code component flaws, and a fatigue, fracture mechanics and environmental qualification evaluation was performed either to justify the continued use of the bolts or to limit the extent of additional inspections.

i-Results of these two evaluations showed that the flaws were stable and would not grow to a size which would exceed the acceptance criteria of the ASME Code, The flaw acceptance criteria for this type of evaluation is based on the as-found size ("a")

of the flaw, assuming continued growth over an appropriate period of time ("daldt") and subject to an appropriate number of cycles ("da/dn"), The acceptable ASME Code flaw size contains suitable factors of safety over either a brittle fracture failure or an ultimate strength failure.

The results of these evaluations indicated that there was significant margin in the 1989 ASME Code linear acceptance criteria, The results also indicated that the JAF RO12 inspection findings might be treated as flaw indications below a threshold flaw size of concern rather than flaw indications for which a detailed analytical evaluation was required, if a suitable supplemental acceptance criteria, based on crack depth / length ratio, could be developed. if the inspections findings could be characterized as flaw indications below a threshold flaw size of concern in accordance with the ASME Code, the sample size of inspected CRD bolts for RO12 would not have to be increased. Uninspected CRD bolts could then be left in service until the periodic maintenance and replacement of the applicable CRD assembly, when they would be replaced in accordance with modification g M1-92 319, SuDDiemental AcceDtance Criteria:

As discussed above, the threshold size for a non-axial (circumferential) surf ace linear indication was determined to be 1/4" in EPRI NP-1406-SR This small flaw indication provides a large factor of safety for RPV head studs, which are > 2" in diameter and are 7

Attachment 1 to JPN 97-028

, CONTROL ROD DRIVE REACTOR PRESSURE VESSEL HOUSING FLANGE BOLTS 1997 EVALUATION OF FLAWS FOUND IN BOLTS RETIRED AND REPLACED DURING RO12 PER GE SIL 483 torqued to a high stress. This small flaw indication is based on a flaw which was judged not to be relevant for the RPV head studs. In addition, for these large diameter studs, further volumetric inspections (UT) would be performed to disposition any reportable VT-1 flaw indications, using a length-versus-depth planar approach.

Larger VT-1 flaw indications were not considered permissible, because in the analysis of the postulated larger indications, the assumption was that they would be found at the root of the bolt or stud threads. Analysis of such postulated larger indications did not provide sufficient f actors of safety to justify larger acceptance criteria for circumferential surface linear indications. The postulated larger indications were unacceptable due to the thread notch stress intensity effect and indication / thread depth effect contributions in the flaw analysis, in addition, in evaluating the postulated larger indications for acceptance criteria, the bolt or stud yield stress was conservatively assumed to be the applied stress at the flaw, which is a reasonable assumption for RPV head studs but is overly conservative for CRD bolts, in contrast, the indications in the CRD bolts were not in the threads, but were in the shank of the cap screw at the cap screw shank to-cap shoulder. This is probably due to the as installed orientation of the CRD bolts. Any flange leakage will not wet the thread surf aces, but will wet the shank-to-cap shoulder. Also, the indications were found in bolting < 2" in diameter. In addition, the 350 +/-25 f t-lb torque on the original design CRD bolts resulted in a nominal tensile stress of approximately 27 ksi at the indication location, which is 31 %

of the yield stress of 86 ksi at 575 F. These differences masked factors of safety in the original design CRD bolts, which could be utilized in developing supplemental acceptance criteria applied using supplemental examination methods and techniques.

JAF staff contacted StructuralIntegrity Associates (SIA) for assistance in developing additional acceptance criteria based on supplemental examination methods and techniques.

A Structural Integrity Associates principal had served as the chairman of the ASME Working Group on Acceptance Standards, which was instrumentalin developing the acceptance standards of Section XI of the ASME Code and in reviewing and documenting these standards in EPRI NP-1406-SR SIA confirmed that the approach proposed by JAF is not only acceptable under the Code, but is consistent with the generally accepted Section e XI philosophy that, if a more thorough inspection is performed, then the acceptance criteria consistent with the more thorough inspection technique may be used.

SIA prepared supplemental acceptance criteria for the CRD bolt flaw indications for JAF (Reference 11). Fracture mechanics and strength of materials analyses consistent with those described in EPRI NP-1406-SR were used in the development of the criteria. This criteria was based on a length-ve sus-depth planar approach, rather than the surface linear acceptance standard of 1/4" non-axial (circumferential) linear indication currently in use with the JAF VT-1 procedure. This supplemental acceptance criteria requires destructive 8

Attachment 1 to JPN-97-028

, , CONTROL ROD DRIVE REACTOR PRESSURE VESSEL HOUSING FLANGE BOLTS 1997 EVALUATION OF FLAWS FOUND IN BOLTS RETIRED AND REPLACED DURING RO12 PER GE Sll 483 metallurgical sectioning of the worst case retired CRD bolts in order to determine the planar geometry of the indications (flaws). The VT-1 data combined with the metallurgical sectioning data can then be used to determine whether the flaw indication size was below a threshold flaw size of concern.

JAF staff prepared a Nuclear Safety Evaluation to verify that there were no unreviewed safety questions associated with providing supplemental VT 1 acceptance criteria to evaluate flaw indications in the CRD bolts (Reference 12).

Evaluation and Acceptance of Indications:

The six CRD bolts, which were removed during RO12 and which failed to meet the 1/4" non-axialindications, were sent to Carolina Power and Light metallurgical lab facilities for the destructive sectioning of the bolts. NYPA directed CP&L to examine the worst two of the six bolts (Reference 13). Results of the CP&L destructive sectioning (Reference 14) are as follows:

Bolt 3 A bolt with the largest linear indication, which had a composite of two linear indications (total 7/8" long), had a maximum identified crack depth of 0.011" Bolt 5- A bolt with the worst corrosive attack, which had a composite of two linear indications (total 3/8" long), had a maximum identified crack depth of 0.032". A pit defect in this same bolt was determined to have a pit depth of 0.036" None of these bolt flaw depths exceeded the deepest flaw (0.077") reported by GE SIL 483 (Refererice 3.d).

Using the supplemental acceptance criteria, SIA evaluated the planar flaw data reported by CP&L (Reference 14) Based on their evaluation, they concluded that the as-reported flaws are acceptable by the SIA supplementary criteria to the ASME Code, in addition, they asserted that the flaws as characterized above are bounded by the flaws described in GE SIL 483.

Therefore, no increase in inspection sample size was necessary to evaluate the current status of CRD bolt cracking at JAF. No exacerbation or change in the extent of condition of CRD bolt cracking had occurred, so no acceleration of the CRD bolt replacement schedule was required.

SUMMARY

AND CONCLUSION:

The CRD bolts, removed and replaced during CRD mechanirm maintenance and replacement activities in RO12, contained acceptable VT and PT indications. No increase 9

4 Attachment 1 to JPN 97 028

, , CONTROL ROD DRIVE REACTOR PRESSURE VESSEL HOUSING FLANGE BOLTS 1997 EVALUATION OF FLAWS FOUND IN BOLTS RETIRED AND REPLACED DURING RO12 PER GE SIL 483 in sampling population size was required.

There is no evidence of a significant change in the extent of condition of cracking in the original design CRD bolts still in service at JAF.

PLANNED FUTURE CORRECTIVE ACTIONS:

RO13 Course of Action:

Tentative plans for RO13 are to replace all remaining original design CRD bolts with new design CRD bolts. The number of CRD mechanisms involved is 73, so this is an extensive effort (584 CRD bolts). The rilajority of these CRD bolts will be replaced individually, one at a time, without removal of the CRD mechanism or replacement of the O-Ring seal. This represents a conservative acceleration of the schedule for implementation of modification M1-92 319.

New design CRD bolts, which were previously installed and have seen service, will continue to be inspected in accordance with ASME Code requirements. These new design CRD bolts contain design improvements over the original design CRD bolts, but they do not have sufficient operating experience to demonstrate that they are not susceptible to SCC in the operating environment of the CRD flange housing. Continued enhanced monitoring of the new design CRD bolts for SCC will remain part of the maintenance and ISI procedures until the CRD bolt cracking issue is closed. Enhanced monitoring includes requirements for PT of the CRD bolts, if there is visual evidence of linear indications.

10

Attachment 1 to JPN.97-028

' ' CONTROL ROD DRIVE REACTOR PRESSURE VESSEL !10USING FLANGE BOLTS 1997 EVALUATION OF FLAWS FOUND IN BOLTS RETIRED AND REPLACED DURING RO12 PER GE Sll 483

REFERENCES:

(1) James A. Fitzpatrick, Facility Safety and Analysis Report, May 1997.

(2) American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,1980 Edition with Addenda up to and including Winter 1981, i

(3) General Electric Communications:

a) Rapid Information Communication Services information Letter No. 019 dated May 19,1988.

b) Services information Letter No 483, Rev. O, dated March 17,1989, c) Services Information Letter No 483, Rev.1, dated October 1,1991, d) Services Information Letter No 483, Rev. 2, dated August 5,1992.

(4) JAF Modification M192 319 File.

(5) EPRI Report NP-1406-SR, Special Report, Nondestructive Examination Acceptance Standards, Mechanical Basis and Develooment of Boiler and Pressure Vessel Code, ASME Section XI Division 1. May 1980.

(6) NYPA Internal Memorandum to Outage Distribution from R. Wiese, Jr., JPLN 016, Rev.1, dated February 4,1997.

(7) JAF Maintenance Procedure MP 004.03, "CRD Removal and Replacement."

(8) JAF Nondestructive Examination Procedure, NDEP 9.51 (J), "lSI VT-1 Visual Examination Procedure."

.(9) NYPA Liquid Penetrant Examination Report 97S034 (M.A. Benson), dated 4/23/97.

(10) NYPA Internal Memorandum to A. Smith from M. Abramski, JLIC-97-064, dated May 28,1997.

(11) Structural Integrity Associates (H.L. Gustin and P.C. Riccardella) 'etter to NYPA (J.

Goldstein), dated May 28,1997, " Flaw Acceptance Standards for CRD Botting Examinations at J.A. FitzPatrick," PCR-97-025.

(12) JAF Nuclear Safety Evaluation, JAl SE-97-018, dated May 28,1997, 11

Attac,hment 1 to JPN 97 028

, . CONTROL ROD DRIVE REACTOR PRESSURE VESSEL HOUSING FLANGE BOLTS 1997 EVALUATION OF FLAWS FOUND IN BOLTS RETIRED AND REPLACED DURING RO12 PER GE SIL 483 i

(13) NYPA (J. Goldsteln) letter to Carolina Power and Light (R. Bloch), dated May 28,1997, " Inspection Plan for CRD Bolts," PEP JSG 97170.

(14) Structural Integrity Associates (H.L. Gustin) letter to NYPA (J. Goldstein), dated May 20,1997, Review and Acceptance of CRD Flange Bolt Flaw Sizing Results,"

HLG 97 038.

12 1

Attachment 1 to JPN 97 028 CONTROL ROD DRIVE REACTOR PRESSURE VESSEL HOUSING FLANGE BOLTS 1997 EVALUATION OF FLA'NS FOUND IN BOLTS RETIRED AND REPLACED DURING RO12 PER GE SIL 483

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1. First number in cell represents the Refueling Outage when Control Blade was introduced to the cor EEQt Bladelyng Number in Core at BOC13 0 Original Equipment 61 7 Duralife 190 20 8 Duralife-230 25 9 Duratife-230 18 10A ABB CR 82M 4 10G MARATHON 4 12 MARATHON 15

2. Second number in a cell represents the location of a control blade prior to shuffling.

3. The number in parenthesis is the refueling outage when the control blade was thuffled.

Figure 3. Control Rod Blade Mapping (Reference 1) 15

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Attachment 1 to JPN 97-02B

, , . CONTROL ROD DRIVE REACTOR PRESSURE VESSEL HOUSING FLW3E BOLTS 1997 EVALUATION OF FLAWS FOUND IN BOLTS RETIRED AND REPLACED DURING RO12 PER GE SIL 483 TABLE 1 COMPARISON OF ORIGINAL DESIGN CRD BOLTS WlTH GE NEW DESIGN CRD BOLTS Design Data Original Design CRD Bolts GE New Design CRD Bolts Material ANSI 4140 Cr Mo Carbon Steel, ANSI 4340 Cr Mo V Carbon Steel, SA 193, Grade B7 SA 540, Grade B21, Class 3 Yield Strength / 105 ksi min /125 kal min 130 ksi min /145 kai min Tensile Strength Chemistry 0.045 + / 0.005 0.025 + /. 0,005 Sulfur, max %

Geometry 1* Diameter Cap Screw per 1" Diameter Cap Screw per Geometry Requirements of Geometry Requirements of ANSl/ASME B1B.3 ANSI /ASME B18.3 with an increased Blend Radius in the Cap to Bolt Shoulder Transition Region to Reduce the Local Stress Concentration Factor and a Slotted Washer to Permit Draining of Moisture from the Aree Underneath the Cap Preload Torque Original Design Preload Torque u Same as Original Design No 350 + / 25 f t.-lb. Change Operating Temperature, Pressure and Same as Original Design No Environment Chemistry Associated with the CRD Change Some improvement in Assembly Location Underneath the Moisture Drainage is Expected with Reactor Vessel (See Figure 1) the Slotted Washer 16

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