ML20205Q598

Safety Evaluation Supporting Amend 209 to License DPR-50

ML20205Q598
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Site:	Crane
Issue date:	04/13/1999
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UNITED STATES V

S NUCLEAR REGULATORY COMMISSION i

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WASHINGTON. D c. 20555-0001

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 209 TO FACILITY OPERATING LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GPU NUCLEAR. INC.

THREE MILE ISLAND NUCLEAR STATION UNIT NO 1 DOCKET NO. 50-289 4

1.0 INTRODUCTION

By letter dated November 25,1998, as supplemented on February 12,1999, GPU Nuclear, Inc., (the licensee) subm;tted a request to change the technical specifications (TS) at the Three Mile Island Nuclear Station, Unit 1 (TMI-1). The TS change request proposed the continuation of an attemate repair criteria, previously approved through Cycle 12, to address intergranular attack (IGA) degradation identified on the inside diameter (ID) of the TMI-1 once-through steam generator (OTSG) tubes. The repair criteria would be used in the Cycle 13 Refueling (13R) examinations and would be in effect until the end of operating cycle 13. The TS change request also proposed ch::nges to related TS eddy current testing (ECT) voltage limits and a clarification that the TS reporting period for steam generator inservice inspection reports begins with the closure of the plant's n'ain electrical generator breaker. The November 25, 1998, letter requested extension of the steam generator inspection reporting period from 90 days to 12 months. The February 12,1999, supplementalletter withdrew that request. The supplemental letter did not affect the initial" proposed no significant hazards consideration" determination.

2.0 BACKGROUND

in November 1981, while performing reactor coolant system hydrostatic testing with the reactor shut down, primary-to-secondary system leakage was detected in both OTSGs. Subsequently, eddy current examinations revealed many defective tubes. Metallographic examination of poi?ons of removed tubes confirmed that the tube degradation initiated from the primary side

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of the tubes in the form of circumferential stress-assisted intergranular cracks. The active I

chemical impurity causing the corrosion was sulfur in reduced forms, which had been i

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'. inadvertently introduced into the reactor coolant system. The vast majority (approximately 95 percent) of the defects occurred within the top 2 to 3 inches of the 24-inch thick upper tubesheet (UTS). The corrosion attacks most rapidly at the air / water interface and during lay up. The air /waterinterface was located in the UTS during a significant portion of the post-hot-functional shutdown period. To repair the defective OTSG tubes within the UTS, the licensee i

applied a kinetic (explosive) expansion repair technique. The staff previously reviewed and approved the licensee's repair of the OTSG tubing in NUREG-1019, " Safety Evaluation Report Related to Steam Generator Tube Repair and Retum to Operation - Three Mile Island Nuclear Station, Unit No.1," dated November 1983.

The kinetic expansion repair technique applied in the early 1980's addressed the existence of defects located in tubes in the UTS. However, a limited population of tubes in the TMI-1 OTSGs contained degradation located below the UTS secondary face that could not be repaired by the kinetic expansion technique. Because of the uncertainty in sizing the depth of ID IGA degradation that was not previously repaired, the NRC and the licensee agreed during a meeting held in Rockville, Maryland, on July 15,1997, that the tube repair criteria in the TMI-1 TS should be amended to address tubes identified with this mode of degradation.

Subsequently, the licensee proposed TS repair criteria to be used through the Cycle 12 operating cycle, which addressed ID IGA below the UTS. The repair criteria supplemented the 1

existing TS depth-based tube repair limit (i.e.,40 percent through-wall) with ECT voltage-based and length-based repair limits. The length-based repair limits do not account for flaw growth over the cycle because the licensee has concluded the ID IGA flaws are inac(ive. In i

addition to the proposed repair limits, the licensee proposed modifications to the reporting l

requirements for steam generator tube inspections to include the submission of a summary of steam generatorinspections and results to the NRC within 90 days following completion of the inspections and repair. The proposed TS repair criteria were reviewed and approved in a staff safety evaluation dated October 16,1997 (Reference 1).

The licensee has proposed the continuation of the TS repair criteria, previously approved through Cycle 12 as documented in Reference 1, through the end of Cycle 13. The licensee has also proposed two other related changes. The staff's evaluation of this proposal considers leakage and structural integrity margins, growth rate analysis of ID IGA indications, and eddy current technique capability of reliably implementing the length-based repair criteria.

3.0 EVALUATION 3.1 Extension Through Cycle 13 The licensee has proposed continuation of the alternate repair cnteria (TS 4.19.4.a.3(a), TS 4.19.4.a.6, TS 4.19.5.b and TS Bases], previously approved through Cycle 12, to address ID IGA identified in TMI-1 steam generator tubes. The repair criteria would be extended to be in effect during 13R steem generator examinations and through the end of operating cycle 13. ln addition to the submittals made as part of this license amendment request, the staff reviewed several other documents containing relevant information. These are listed as References 2 through 4. The staff reviewed leakage integrity, structural integrity, growth rate analysis and inspection capability to assess the effect of extending the repair criteria through Cycle 13.

'. 3.1.1 Leakage Integrity in order to demonstrate leakage integrity margins, the licensee completed in-situ pressure testing, during 12R, of steam generator tubes with the bounding ID IGA indications to demonstrate a low leakage potential for tubes containing this mode of degradation. In addition, supplemental laboratory leak tests were performed on puiled tubes in order to

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effectively simulate both peak hoop and axial stresses which might be applied to tubes containing volumetric flaws during a postulated main steamline break (the in-situ test equipment used during 12R was not able to fully simulate these stresses on steam generator tubes). None of the indications tested in-situ or in the laboratory leaked at pressures simulating up to three times the normal steam generator tube operating delta pressure.

Therefore, the licensee concluded that no increase in accident-induced leakage would be expected due to the presence of ID IGA flaws.

During 13R, the licensee plans to select tubes for leak testing if any have indications of I

degradation which appear more limiting with respect to leakage integrity than those which were tested during 12R, thereby continuing to assess the potential for leakage from all degraded tubes. Assuming the limiting tubes retain leakage integrity throughout the test, it is reasonable to conclude that other, less degraded tubes in the population would also have sufficient j

integrity to withstand accident-induced loads without failure. The licensee indicated that

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supplemental laboratory leak testing will not be necessary during 13R because the in-situ pressure test equipment that will be used has been modified since 12R and can effectively simulate both peak hoop and axial stresses. On these bases, the staff concludes the licensee has demonstrated adequate leakage integrity margins in the presence of ID IGA flaws.

3.1.2 StructuralIntegrity The primary method of ensuring structuralintegrity of ID IGA flaws is the application of the TS length-based repair criteria which were determined based on structural analyses of flawed steam generator tubing assuming through-wall defects of the limiting length, in order to demonstrate continued structuralintegrity, the licensee completed in-situ burst testing, during 12R, of steam generator tubes with bounding ID IGA indications, in addition, supplemental laboratory burst tests were performed on pulled tubes in order to effectively simulate both peak hoop and axial stresses which might be applied to tubes containing volumetric flaws during a postulated main steam line break (the in-situ test equipment used during 12R was not able to fully simulate these stresses on steam generator tubes). None of the indications tested in-situ or in the laboratory burst at pressures simulating up to three times the normal steam generator tube operating delta pressure.' In fact, the burst pressures of the tube sections containing ID IGA that were tested in the laboratory were above 10,000 psig. This indicates that substantial structural margin exists for these flaws since the burst pressure for tubes with no defects averages 11,216 psig.

During 13R, the licensee plans to se!ect tubes for burst testing if any have indications of degradation which appear more limiting with respect to structural integrity than those which were qualified by tests during 12R, thereby continuing to assess the structural integrity of tubes with ID IGA flaws. The licensee indicated that supplemental laboratory burst testing will not be necessary during 13R because the in-situ pressure test e auipment that will be used has

1 1 been modified since 12R and can effectively simulate both peak hoop and axial stresses. On these bases, the staff concludes the licensee has demonstrated, and will continue to demonstrate, adequatn structural integrity margins in the presence of ID IGA flaws.

3.1.3 Growth Rate Analysis The licensee has previously demonstrated that ID IGA degradation is not an active degradation mechanism at TMI-1. This was demonstrated through comparisons of nondestructive examination data from several outages. This is a significant aspect of the technical basis supporting this amendment request because the length-based repair criteria were developed based on the assumption of zero flaw growth. Therefore, the staff has a heightened sensitivity to the results of the licensee's growth rate analysis. The licensee's 12R growth rate analysis was based on a review of eddy current data and metallurgical and chemical analysis performed on the 12R pulled tubes and is discussed in the following paragraphs.

The licensee performed assessments of volumetric ID IGA growth rates based on changes in bobbin coil voltage, motorized rotating pancake coil (MRPC)-indicated axial length and MRPC-indicated circumferentiallength. When the 12R bobbin coil voltages were compared to previous voltage responses at the same location, a +0.1 volt average increase (0.26 volt standard deviation) was observed. These voltage changes are comparable with voltage variations from seven previous outages and indicated no statistically significant growth.

MRPC-indicated axial and circumferential length measurements were made for all ID IGA I

indications detected during 12R. These values were compared to measurements from examinations performed during the previous two outages (where available). The licensee determined there was essentially no change in axial or circumfererstial dimensions for the values compared. The dimensions used for this comparison were obtained using the same type of eddy current probe for all three outages.

1 The licensee described the ECT equipment and calibration techniques used in 11R and 12R and planned for 13R. The licensee discussed differences between outages and the expected effects on the ECT data and growth rate analysis, and concluded the differences would not i

adversely affect flaw growth studies because procedures were implemented to assure consistency. Two new practices were implemented in 12R which provide enhanced l

consistency; bobbin coil probe wear is now being monitored for examinations where ID IGA is being evaluated; and a Babcock and WIcox Owners Group (BWOG) Mother ASME standard is being utilized for calibrations (versus individual calibration standards) in order to eliminate subtle voltage differences between calibration standards.

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The licensee performed chemical and metallurgical analyses on the 12R pulled tube to supplement the growth rate analysis. The following procedures were utilized specifically to investigate the micro chemistry of the volumetric ID IGA (pit-like) degradation found in the pulled tube: X-ray Photoelectron Spectroscopy, Scanning Auger Microprobe and Scanning Electron Microscopy with Energy Dispersive Spectroscopy. The strongest evidence derived from these analyses that supports the historical nondestructive examination record of inactivity l

of the ID degradation is the absence of aggressive chemical species and the absence of j

metallographic evidence of any new or different corrosion mechanisms. These analyses did not reveal unusual surface chemir,try or reduced forms of sulfur, such as that which caused the 1

. original ID damage. In addition, the analyses did not identify other contaminants in amounts that have been linked to steam generator tube pitting-type degradation. Microscopic studies of the IGA pits confirmed their volumetric pit-like geometry, which had been characterized dunng the field ECT exams. The ID IGA defect characterization of the 1997 tubes was similar to that of the ID IGA defects of tubes pulled in 1986. If this IGA were continuing to propagate, as was observed in the 1981 degradation, it would be expected to ultimately develop into stress corrosion cracks (SCC). There was no evidence of ID-initiated SCC at the ID IGA flaws examined in the laboratory by metallographic grinding or microscopy.

Based on eddy current data comparisons, measures taken to validate comparisons of ECT data between outages, and chemical and metallurgical analyses, the staff concluded the licensee's growth rate analysis supports implementation of the ID IGA repair criteria for another cycle.

3.1.4 Eddy Current inspection Capability During 12R, a steam generator tube was pulled that contained several volumetric ID IGA flaws.

The purpose of the tube pull was, in part, to demonstrate the inspection capability by confirming the ability of rotating pancake coil eddy current to conservatively estimate axial and circumferential extents of flaws in the field. The licensee compared axial and circumferential lengths based on eddy current tests in the field, to axial and circumferential lengths based on destructive examinmion results. The licensee found that in all cases the eddy current measurements were conservative relative to actuallengths. Based on this, the licensee concluded that eddy current probes (i.e., MRPC probes) are able to conservatively assess the axial and circumferential extents of TMI-1 ID IGA flaws. On this basis, the staff concludes the eddy current technique (MRPC) is capable of adequately implementing the TS length-based repair criteria.

3.1.5 Conclusions Based on the four criteria discussed above (leakage integrity, structural integrity, growth rate analysis, and inspection capability), the staff concludes that extension of the repair criteria through the end of Cycle 13 is acceptable and will not affect public health and safety.

3.2 Voltage Limits The licensee has proposed to revise two voltage limits [TS 4.19,4.a.2 Note 1 and TS 4.19.4.a.3(a)) which apply to the ID IGA repair criteria. This revision is a result of a proposed change in voltage normalization for bobbin probe examinations. The licensee has used a 10-volt normalization for the steam generator tube bobbin coil eddy current examinations since the tube damage that occurred in 1981. Other nuclear plants have also used this normalization; however, a greater number of plants have adopted a 4-volt normalization. The licensee has proposed to convert to the 4-volt normalization, which is becoming standard in the industry. The licensee indicated that this change will allow a more direct comparison between TMI-1 eddy current data and that of industry. This change will affect TS sections in which an eddy current voltage is specified. In effect, the measured voltages representative of the eddy current signals under the proposed 4-volt normalization will be 40 percent of the measured voltages under the 10-volt normalization. This voltage normalization change is

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1 applicable for the bobbin probe examinations only, and does not affect signal quality or phase angle analysis of the bobbin probe eddy current signals, rotating probe eddy current, or other aspects of eddy current signal analysis. The licensee has also stated they will adjust all previous voltage measurements obtained under the 10-volt normalization to a voltage measurement that would have been obtained under a 4 volt normalization in order to make comparisons of 13R voltages to prior outage voltages equivalent. On these bases, the staff concludes the proposed TS change to voltage limits is acceptable.

3.3 Clarify Event Signifying End of Outage's Steam Generator inspection and Repair Work The existing TS fo TMl-1 state that the licensee will submit to the NRC complete results of the steam generator tube inservice inspection within 90 days following completion of the inspection and repairs. The licensee has proposed to modify this requirement (TS 4.19.5.b) by clarifying that inspection and repairs are considered complete when the main generator breaker is closed. This modification will not impact the staff's ability to adequately assess any steam generator tube integrity issues stemming from the inspection in a reasonable period of time.

Therefore, the staff concludes the proposed change to specify that a complete inservice inspection report is due 90 days from main generator breaker closure (i.e., restart) is acceptable.

4.0 FUTURE CONSIDERATIONS The licensee stated they submitted this TS change request as a request for a one-cycle change primarily because the upcoming outage (13R) will be the second of two consecutive outages in which a very large number of ID IGA flaws will have been examined using the same eddy current testing techniques.12R provided a 100 percent bobbin probe examination with follow-up MRPC examinations of all of the ID IGA flaws identified by the bobbin probe. In previous outages, less than 100 percent examination of allID IGA flaws by MRPC was performed.13R will provide the same scope as 12R, using the same eddy current acquisition equipment that was used in 12R. The results from these two outages, taken together, will provide a significant amount of eddy current flaw length data with which to analyze the growth rate of the ID IGA flaws. The licensee expects this data should further substantiate the i

insignificant growth rate of ID IGA flaw lengths already seen.

Additionally, the licensee stated that several regulatory and industry initiatives are currently being considered for implementation. The industry and NRC staff are actively working j

together to achieve agreement on now steam generator tube integrity requirements, which the licensee believes may require revision of current TS requirements related to steam generators.

It is expected that these initiatives and the required supporting programs and criteria will be in a more mature status before the next TMI-1 outage (14R), and, therefore, it is a more appropriate time for the licensee to submit a request '7r a permanent amendment.

The staff strongly encourages the licensee to pursue the development of a qualified eddy current technique which can reliably depth size ID IGA in accordance with the original 40 percent tube repair limit. If this path were pursued, further TS amendments would not be required to address this mode of degradation. As documenteo in Reference 5, the Nuclear Energy institute (NEI) indicated that each licensee would evaluate its existing steam generator program and, where necessary, revise and strengthen program attributes to meet the intent of j

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. the guidance provided in NEl 97-06, Steam Generator Program Guidelines, no later than the first refueling outage starting after January 1,1999. NEl 97-06 states that each utility shall follow the inspection guidelines contained in the latest revision of the EPRI PWR Steam Generator Examination Guidelines (Reference 6). Reference 6 contains guidelines for developing eddy current techniques which are qualified for detecting and sizing steam generator tube degradation.

Regardless of which path is pursued, NRC staff previously identified in Reference 1 areas of weakness in the licensee's ID IGA growth rate study. A number of variables were identified which were not specifically addressed in the growth rate studies and are as follows: (1) bobbin probe wear, (2) calibration practices and standards, (3) differences in data acquisition hardware, and (4) data analyst uncertainty. Based on submittals made for the Cycle 13 TS change request, it appears the licensee has addressed some of these factors for the current ID IGA repair criteria. The licensee should ensure that these are completely addressed in the future.

5.0 DESCRIPTION

OF PROPOSED TECHNICAL SPECIFICATION CHANGES In order to incorporate the proposed changes to permit continuation of the altemate repair criteria for ID IGA below the UTS, the licensee has proposed the following changes to the technical specifications.

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Proposed change to TS 4.19.2 Note (1) i The threshold voltage increase at which a tube whose ID IGA indication is counted in the degraded tube population is revised from 0.6 volts to 0.24 volts. This revision is a i

result of the proposed change in voltage normalization for bobbin probe examinations.

b)

Proposed change to TS 4.19.4.a.3(a)

The threshold voltage at which tubes whose ID IGA bobbin coil indication is considered degraded is revised from 0.5 volts to 0.2 volts. This revision is a result of the proposed change in voltage normalization for bobbin probe examinations.

The definition of a degraded tube is also revised so that the criteria pertinent to ID IGA indications are extended from the "12R Outage" and " Cycle 12 operation" to the "13R Outage" and " Cycle 13 operation" respectively.

c)

Proposed change to TS 4.19.4.a.6 The definition of " repair limit" is revised to ' hat the criteria pertinent to ID IGA indications are extended beyond Cycle 12 operation. the end of Cycle 13 operation.

d)

Proposed change to TS 4.19.5.b The parenthetical phrase "(main generator breaker closure)"is added to define this event as the start date of the reporting period for submittal of the complete results of the 13R steam generator tube inspections.

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Proposed change to TS Bases section Several references to " Outage 12R" are revised to indicate " Outage 13R" and references to " Cycle 12" operation are revised to indicate " Cycle 13" operation.

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Discussion on in-situ pressure testing is modified to reflect that in-situ pressure testing may be performed during 13R.

A Note is added to the basis to reflect that TS voltage values related to ID IGA indications are based on a 4-volt normalization procedure.

Based on its review of the licensee's proposal, the staff has determined the proposed changes to the TMI-1 TS will continue to provide adequate assurances of steam generator tube integrity. The extension of the ID IGA repair criteria through the end of cycle 13 is acceptable based on the licensee's leakage and structuralintegrity analyses, growth rate analysis and eddy current inspection capability analysis. In addition, the changes to TS voltage values and the clarification regarding when steam generator inspection result reports are to be submitted to the NRC do not impact steam generator tube integrity or the staff's ability to assess steam generator tube integrity on a timely basis. Therefore, the staff has determined that the proposed modifications will not impact the public health and safety.

7.0 STATE CONSULTATION

in accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendment. The State official had no comments.

8.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The amendment also changes reporting or recordkeeping requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a propo::ed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (63 FR 69342). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (c)(10). Pursuant to 10 CFR 51.22(b) no environmentalimpact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

9.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

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10.0 REFERENCES

1)

Safety Evaluation by the Office of Nuclear Reactor Regulation Related to License Amendment No. 206 to Facility Operating License DPR-50, GPU Nuclear Corporation,

- Three Mile Island Nuclear Station, Unit No.1 Docket No. 50-289, dated October 16, 1997.

2)

Letter from J.W. Langenbach (TMI) to NRC, "TMI-1, NRC Notification on Completion of 12R Outage Steam Generator Examinations," dated December 17,1997.

3)

Letter from J.W. Langenbach (TMI) to NRC, "TMI-1, Cycle 12 Refueling (12R) Outage Once Through Steam Generator (OTSG) Tube Inspection Repon with ASME NIS Data g

Reports for incervice Inspections (ISI)," dated January 12,1998.

4)

Letter from J.W. Lanoenbach (TMI) to NRC, "TMI-1, Results from Cycle 12 Refueling (12R) Outage Pulled Tube Examinations," dated May 19,1998.

5)

Letter from Ralph E. Beedle (Ndl) to L. Joseph Callan (NRC), dated December 16, 1997.

6)

PWR Steam Generator Examination Guidelines, EPRI Report TR-107569 (Rev. 5, September 1997).

l Principal Contributor: C. Beardslee t

Date: April 13,1999 l

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