ML20072L560
| ML20072L560 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom, Limerick  |
| Issue date: | 08/24/1994 |
| From: | Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20072L561 | List: |
| References | |
| GL-94-03, GL-94-3, NUDOCS 9409010017 | |
| Download: ML20072L560 (17) | |
Text
- _ _ _ _ - - _ _ _ _ _. _ _ -
Station tuppstt Dz;Ntrtmsnt GL 9443 i
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L 965 Chesterbrook Boulevard Wayne. PA 19087-5691 August 24,1994 Docket Nos. 50-277 50-278 50-352 50-353 License Nos. DPR-44 DPR-56 NPF-39 NPF-85 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Subject:
Peach Bottom Atornic Power Station, Units 2 and 3 Limerick Generating Station, Units 1 and 2 Response to Generic Letter 94-03, "Intergranular Stress Corrosion Cracking of Core Shroud in Bolling Water Reactors"
Dear Sir:
Attached is our response to the 30 day Reporting Requirements of the subject Generic Letter (GL) 94-03, dated July 25,1994. Future responses will be provided in accordance with Reporting Requirements 2 and 3 of the GL GL 94 03 concerns the susceptibility of Boiling Water Reactors to core shroud cracking.
PECO Energy Company is an active participant in the ongoing efforts of the Boiling Water Reactor Vessel and Internals Project (BWRVIP). Future core shroud inspection plans, examination criteria, and evaluations will be revised, as necessary, based on the future recommendations of the BWRVIP.
In addition to the reporting requirements addressed in the attachments to this letter, the NRC requested that licensees volunteer cost estimates associated with implementing the actions of the GL For Peach Bottom Atomic Power Station, Units 2 and 3, we currently estimate the costs associated with inspection and contingency repair plans to be $8.3 million. For Limerick Generating Station, Units 1 and 2, we currently estimate the costs associated with inspection to be $2.5 million.
If you have any questions, please contact us.
Very truly yours, L/R,.6 Mg jV G. A. Hunger, Jr.,
Director - Licensing l
(-
Attachments I
cc:
T. T. Martin, Administrator, Region I, USNRC (w/ attachment)
W. L Schmidt, USNRC Senior Resident inspector, PBAPS (w/ attachment)
N. S. Perry, USNRC Senior Resident inspector, LGS (w/ attachment) ql r-9409010017 940824 b
4 FDR ADOCK 05000277 P
PDR J
COMMONWEALTH OF PENNSYLVANIA ss.
COUNTY OF CHESTER W. H.. Smith, til, being first duty sworn, deposes and says:
That he is Vice President of PECO Energy Company; that he has read the enclosed response to Generic Letter 94-03 dated July 25,1994, for Peach Bottom Facility Operating Ucenses DPR44 and DPR-58 and Umerick Operating Ucenses NPF-39 and NPF-85, and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowlede; information and belief.
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Vice President
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Docket Nos. 50-277 50-278 50-352 50-353 License Nos. DPR-44 DPR-56 NPF-39 NPF-85 RESPONSE TO GENERIC LETTER 94-03 PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 LIMERICK GENERATING STATION, UNITS 1 AND 2 l
REPORTING REQUIREMENT:
1.
Within 30 days from the date of this generic letter:
(a) A schedule for inspection of the core shroud.
BESPONSE:
An augmented core shroud Inspection was performed at Peach Bottom Atomic Power Station (PBAPS),
Unit 3 during the Fall of 1993 Refueling Outage 9. A final report documenting this inspection, and an evaluation of its results, was forwarded to the U. S. Nuclear Regulatory Commission (USNRC) In our letter from G. A. Hunger, Jr. (PECO Energy) to USNRC dated March 14,1994. Additionally, these results were previously reviewed with the USNRC in a meeting on November 3,1993. This evaluation demonstrates that the structural integrity of the PBAPS, Unit 3 shroud is assured for the subsequent 2 year operating cycle. During the upcoming Refueling Outage 10 at PBAPS, Unit 3, currently scheduled to begin in the Fall of 1995, further augmented inspections, or equivalent actions, will be performed in consideration of the recommendations of the Boiling Water Reactor Vessel and Internals Project (BWRVIP).
An augmented core shroud inspection or equivalent actions for PBAPS, Unit 2 will be performed during the upcoming PBAPS, Unit 2 Refueling Outage 10, currently scheduled to begin in the Fall of 1994. An augmented inspection program for the PBAPS, Unit 2 core shroud is being developed in consideration of the BWRVIP recommendations. The submittal of the inspection plan and plans for evaluation and/or repair is currently being discussed between PECO Energy Company and the USNRC Project Manager, in accordance with the Reporting Requirements 2(a) and 2(b) of the Generic Letter.
Although Limerick Generating Station (LGS), Units 1 and 2 have a low stress corrosion cracking (SCC) susceptibility ranking based on plant water chemistry, material carbon content, fabrication history, neutron fluence, and hot operating time, augmented inspections, commensurate with this ranking, will be performed during each urWs next scheduled refueling outage. The next LGS, Unit 1 Refueling Outage 6 is currently scheduled to begin January 27,1996. The next LGS, Unit 2 Refueling Outage 3, is currently scheduled to begin January 28,1995. The augmented inspection program for the LGS, Units 1 and 2 core shroud is being developed in consideration of the BWRVIP recommendations.
REPORTING REQUIREMENT:
(b) A safety analysis, including a plant-specific safety assessment, as appropriate, supporting continued operation of the facility until inspections are conducted.
RESPONSE
As stated previously, an augmented core shroud inspection was performed on Peach Bottom Atomic Power Station (PBAPS), Unit 3 during the previous Refueling Outage 9. A final report documenting this inspection, and an evaluation of its results, was forwarded to the U. S. Nuclear Regulatory Commission (USNRC) in our letter from G. A. Hunger, Jr. (PECO Energy) to USNRC, dated March 14,1994.
Additionally, these results were previously reviewed with the USNRC in a meeting on November 3,1993.
1
Docket Nos. 50-277 50-278 50-352 50-353 License Nos. DPR-44 DPR-56 NPF-39 NPF-85 inis evaluation demonstrates that the structural integrity of the PBAPS, Unit 3 shroud is assured for the subsequent 2 year operating cycle. An additional safety analysis supporting the report is contained in.
An evaluation was also performed for PBAPS, Unit 2. This report (" Evaluation and Screening Criteria for the Peach Bottom Unit-2 Shroud," dated December 13,1993) documents the evaluation of the PBAPS, Unit 2 shroud based on comparison with the PBAPS, Unit 3 shroud examinations results. Based on the conclus!cns of this report, continued operation was justified for the current operating cycle. PBAPS, Unit 2 is scheduled to begin Refueling Outage 10 in the Fall of 1994. An additional safety analysis supporting this report, and a copy of the report, are contained in Attachment 2.
For LGS, Units 1 and 2, safety analyses are contained in Attachments 3 and 4, respectively.
REPORTING REQUIREMENT:
(c) A drawing or drawings of the core shroud configuration showing details of the core shroud geometry (e.g., support configurations for the lower core support plate and the top guide, weld locations and configurations).
RESPONSE
Drawings for PBAPS, Units 2 and 3, and LGS, Units 1 and 2 are contained in Attachment 5.
REPORTING REQUIREMENT:
(d) A history of shroud inspections for the plant should be provided addressing date, scope, methods and results, if applicable.
RESPONSE
In the case of PBAPS, Unit 3, visual examinations (VT-3) have been performed on the outer diameter 1
shroud surfaces once per period (approximately 31/3 years) since commercial operation (1974) through the 1991 Refueling Outage 8. These examinations were performed in accordance with ASME Section XI, Category B-N 1. No indication of cracking was identified during this time period. During the PBAPS, Unit 3 Refueling Outage 9 (1993), an enhanced visual (VT-1) examination was performed in response to the issuance of SIL 572. The scope of examinations and their results can be found in the Table 4-1 and i
4-2 of the " Evaluation and Screening Criteria for the Peach Bottom Unit-3 Shroud Indications," dated December 3,1993. This re;x>rt was forwarded to you in our letter from G. A. Hunger, Jr. to USNRC dated March 14,1994.
For PBAPS, Unit 2, visual examinations (VT-3) have been performed on the outer diameter shroud surfaces once per period (approximately 31/3 years) since commercial operation (1974) through the 1992 Refueling Outage 9. These examinations were performed in accordance with ASME Section XI, Category B-N-1. No indication of cracking was identified during this time period.
For LGS, Unit 1, visual examinations (VT-3) have been performed on the outer diameter shroud surfaces once per period (approximately 31/3 years) since commercial operation (1986) through the 1994 Refueling Outage 5. These examinations were performed in accordance with ASME Section XI, Category B-N-1. No indication of cracking was identified during this period.
2
Docket Nos. 50-277 50-278 50-352 50-353 License Nos. DPR44 DPR-56 NPF-39 NPF-85 For LGS, Unit 2, visual examinations (VT-3) have been performed on the outer diameter shroud surfaces once per period (approximately 31/3 years) since commercial operation (1990) through the 1993 Refueling Outage 2. The shroud support cylinder and support leg, at azimuth 300 degrees, were also Inspected in 1991. These examinations were performed in accordance with ASME Section XI, Category B-N-1. No indication of cracking was identified during this period.
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ATTACHMENT 1 Safety Analysis Peach Bottom Atomic Power Station, Unit 3 l
1
GENERIC LETTER 94-03 SAFETY ANALYSIS PEACH BOTTOM ATOMIC POWER STATION, UNIT 3 Docket No. 50-278 Revision 0 INTRODUCTION:
This safety analysis has been developed to reaffirm the ability to safely operate the Peach Bottom Atomic Power Station, Unit 3 (PBAPS, Unit 3) facility until additional core shroud inspections are conducted. This analysis has been developed in accordance with " Requested Ucensee Actions" No. 2 of NRC Generic Letter (GL) 94-03. Additional shroud inspections (or equivalent actions) are scheduled to be conducted during the next refueling outage of PBAPS, Unit 3 (3R10), which is currently scheduled to begin in the Fall of 1995. Accordingly, this ana!ysis applies to the period of operation between the date of this submittal, and the start of the tenth refueling outage of PBAPS, Unit 3.
Core shroud examinations have already been conducted at PBAPS, Unit 3, and a resultant Safety Evaluation has been developed to address the results of these examinations. The Evaluation and Screening Criteria document, " Evaluation and Screening Criteria for the Peach Bottom Unit-3 Shroud Indications," Rev.1, GENE-523-141-1093, dated December 3,1993, (Reference 1) which formed the basis for the safety evaluation, was submitted to the NRC. Nevertheless this analysis has been developed to satisfy the requested Reporting Requirement 1(b) of the GL, and to reaffirm the conclusions of the original Safety Evaluation, in light of the most recent information on the shroud cracking phenomenon.
]
.QBJECTIVE:
The primary objective of this safety analysis is to document the basis for the expectation that severe cracking (i.e. 360* cracking,90% or greater thru-wall) is not present in the core shroud welds of PBAPS, Unit 3 at this time. Reaffirming this expectation, the facility can continue to operato under the current licensing basis as documented in the PBAPS UFSAR, Section 3.3.5 and Appendix K. Additionally, the likelihood of the simultaneous occurrence of licensing basis events and complete (360', thru-wall) shroud cracking, prior to the next refueling outage, is discussed. Further, the ability of the shroud and plant safety features to respond to the effects of these unlikely events is addressed. Finally, actions which PECO Energy Company has taken to address the impact that severe shroud cracking could have on safe plant operations are presented.
BACKGROUND:
The PBAPS, Unit 3 shroud was extensively examined, including use of the enhanced VT-1 Visual examination technique, during the ninth refueling outage (3R09), in October 1993. The results of these examinations revealed crack-like indications in the heat affected zone of the cylinder plates, at the H-3 and H-4 welds. No indications were found in the ring side heat affected zone of any shroud weld These Indications were characterized as intergranular Stress Corrosion Cracking (IGSCC). The identified cracking was then evaluated for its impact on the structural integrity of the shroud structure, through an additional operating cycle, including power rerate conditions. This evaluation is contained in Reference 1 Page 1 of 5
GENERIC LETTER 94-03 SAFETY ANALYSIS PEACH BOTTOM ATOMIC POWER STATION, UNIT 3 Docket No. 50-278 Revision 0 BACKGROUNDI The evaluation identified that even with the as-found cracking, the PBAPS, Unit 3 1 Cont'd) shroud would maintain its structural integrity during the enveloping design basis accident conditions (i.e. Main Steam line break plus maximum credible earthquake) for an addition cycle of operation. Revision 0 of this evaluation was presented to NRC Staff on November 3,1993, and it was found satisfactory as the basis for resuming power operations following the refueling outage. Revision 1 of the document was forwarded to the staff on March 14,1994, to officially document additional information which was identified and requested by the NRC at the November presentation.
BASIS:
The basis for the expectation that severe shroud cracking does not exist is as follows:
The conditions which could influence the initiation and propagation of IGSCC in Bolling Water Reactor (BWR) shroud welds have been identified and discussed in the PBAPS, Unit 3 Evaluation and Screening Criteria Document (Reference 1)
Sections 1,2, and 3. While PBAPS, Unit 3 is considered ' M highly susceptible to the cracking phenomenon (susceptibility factor of 0 to 9, forence 3, Table 2.3), it must be emphasized that the justification for m; peration is based on actual examination results, derived from extensive ins; c. tons of the shroud. It is also important to emphasize that the most severe cracking sound-to-date (i.e. 360*
1 deep cracks), has always been located in the end grain region of the welded plate rings. Since the PBAPS, Unit 3 shroud was fabricated with seamless, roll-forged rings, susceptibility to the extensive cracking is extremely unlikely. This has been demonstrated by actual inspection data. No indications were found on any of the j
PBAPS, Unit 3 rings, nor in the forged rings at any other plant which has performed Inspections.
While the accuracy of results for shroud inspections conducted using the enhanced VT-1 visual examination technique has been questioned, the PBAPS, Unit 3 inspections address such concerns based on the following:
The complete circumference of the inside surface (i.e.10) of both the H-3 and the H-4 weld were visually inspected, (e.g. all portions of the weld and both heat affected zones were viewed and recorded).
Questionable indications were conservatively characterized as cracks.
The length of each identified indication was conservatively assigned, realizing the inaccuracles associated with the visual technique.
Page 2 of 5
GENERIC LETTER 94-03 SAFETY ANALYSIS PEACH BOTTOM ATOMIC POWER STATION, UNIT 3 Docket No. 50-278 l
Revision 0 FASIS:
IQgnt'cQ Final plotting of all identified and sized indications was done in the horizontal plane, even if the indication was oriented at an angle to the horizontal.
ASME Section XI Proximhy Rules were then applied to join nearby indications.
Finally, crack growth was applied to each plotted indication, to account for theoretical increases in the length of the indication during the subsequent operating cycle.
All indications were characterized as through-wall cracks.
Loadings associated with power rerate conditions were used to evaluate the shroud indications, even though the plant is not currently operating at rerate conditions.
Additional conservativisms which were applied to the examination results and evaluation of the PBAPS, Unit 3 shroud are listed in Table 1-1 of Reference 1.
simultaneously with severe shroud cracking is as follows:Th The PBAPS, Unit 3 Individual Plant Examination (IPE) submittal (Referenc the NRC Safety Evaluation related to core shroud cracking at Dresden and Qua Cities, dated July 20,1994 (Reference 8), have been reviewed. Based on this review, the assessment of the risk of core damage at PBAPS, Unit 3, resulting l
significantly lower than the risk of other events analyzed in Additionally, the likelihood of the occurrence of a LOCA is influenced by the condition and quality of the reactor coolant pressure boundary components. A review of the ASME Section XI ISI Program examination results for the reactor coolant system piping provides an Indication of the condition and integrity of thi pressure boundary. Of the 126 welds which make up the Class 1 Main Steam i
piping, no significant lodications have been identified during the ISI examination to4 ate. Additionally all Recirculation System piping was replaced in 1987 The piping materials and fabrication processes used for this replacement resulted in system which is highly immune to the occurrence of iGSCC. Further, all 62 welds which join the seamless, 316 nuclear grade stainless steel piping were exam for preservice inspection, at the time of installation. Review of these and subsequent ISI examination results, revealed that no significant indications hav been identified.
Page 3 of 5 m
GENERIC LETTER 94-03 SAFET( ANALYSIS PEACH BOTTOM ATOMIC POWER STATION, UNIT 3 Docket No. 50-278 Revision 0 BASIS:
While PBAPS, Unit 3 is not expected to experience severe shroud cracking (Cont'd) during the current operating cycle, the effects of such theoretical cracking on the shroud and associated safety systems ability to respond to design basis accident conditions have been considered. The BWR Shroud Cracking Generic Safety Assessment (Reference 3), which was developed by the Bolling Water Reactor Vessel Intemals Project (VIP), and submitted to the NRC on August 5,1994, has been reviewed. PECO Energy Company endorses this document. Since the generic assessment is applicable to the PBAPS, Unit 3 shroud configuration, the conclusions of the document further support this safety analysis. Since the generic assessment indicates that the reactor will be able to be shutdown and cooled down even with the effects of the design basis accidents on a completely severed shroud, the PBAPS, Unit 3 facility is enveloped by this assessment, considering the limited extent of degradation identified on the PBAPS shroud.
Finally, to address the importance of the shroud creaking issues, PBAPS has informed plant operators of the issues to alert them es to the potential affects on plant operating parameters. This information is intended to impart a keen awareness of the potential safety significance of the shroud cracking issue.
CONCLUSIONS; Because the PBAPS, Unit 3 shroud has already been inspected, the potential for the existence of severe cracking has been eliminated. The extent of limited cracking has been identified within reasonable accuracy. This identified indications have been conservatively evaluated for impact on the performance of the shroud structure during design basis accidents. The results of the evaluation indicate that the PBAPS, Unit 3 shroud can, even in its current condition, support safe and continued operation of the facility until the next refueling outage, scheduled for the Fall of 1995.
Page 4 of 5
GENERIC LETTER 94-03 SAFETY ANALYSIS PEACH BOTTOM ATOMIC POWER STATION, UNIT 3 Docket No. 50-278 Revision 0 REFERERCESI 1.
" Evaluation and Screening Criteria for the Peach Bottom Unit-3 Shroud Indications", Rev.1, (GENE-523-141-1093) dated December 3,1993.
2.
"BWR Shroud Cracking Generic Safety Assessment" Rev. O, (GENE-523-A107P-0794) dated July 1994. Submitted to NRC via Letter No. BWROG.
94089, dated 7/13/94.
3.
"BWR Shroud Cracking Generic Safety Assessment
- Rev.1, (GENE-523-A107P-0794) dated August 1994. Submitted to NRC via Letter No. BWROG.
94097, dated 8/5/94.
4.
"BWR Core Shroud Evaluation", Rev. O, (GE-NE-523-148-1193) dated April 1994. Submitted to NRC via Letter No. BWROG-94041, dated 4/5/94.
5.
" Safety Assessment - BWR Shroud Crack Indications. Submitted to NRC via Letter No. BWROG-93130, dated 11/9/93.
6.
" Response to NRC Ouestions on Core Shroud and Reactor Internals", Rev.
O, (GENE-523-A114P-0894) dated August 1994. Submitted to NRC via Letter No. BWROG-94100, dated 8/5/ 4.
7.
PBAPS Individual Plant Examination (IPE), submitted August,1992.
8.
NRC Safety Evaluation Related to Core Shroud Cracking at Dresden, Unit 3, and Quad Cities, Unit 1, dated July 20,1994.
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ATTACHMENT 2 Safety Analysis and Evaluation Peach Bottom Atomic Power Station, Unit 2
i GENERIC LETTER 94-03 SAFETY ANALYSIS PEACH BOTTOM ATOMIC POWER STATION, UNIT 2 Docket No. 50-277 Revision 0 lNTRODUCTION:
This safety analysis has been developed to establish the ability to safely operate the Peach Bottom Atomic Power Station, Unit 2 (PBAPS, Unit 2) facility until core shroud inspections are conducted.
This analysis has been developed in accordance with " Requested Ucensee Action" No. 2 of NRC Generic Letter (GL) 94-03. Shroud inspections (or equivalent actions) are scheduled to be conducted during the next refueling outage of PBAPS, Unit 2 (2R10), which is currently scheduled to begin in the Fall of 1994, Accordingly, this analysis applies to the period of operation between the date of this submittal, and the start of the tenth refueling outage of PBAPS, Unit 2.
Whilo core shroud examinations have not been conducted at PBAPS, Unit 2, an Eva;uation and Screening Criteria document, " Evaluation and Screening Criterla for the Peach Bottom Unit 2 Shroud,' Rev. O, GENE-523-176-1393, dated December 13,1993, (Reference 1) had been developed for PBAPS, Unit 2, because of its similarity to the Unit 3 facility, which had discovered shroud indications in 1993. This analysis has been developed to satisfy the requested reporting requirement 1(b) of the GL and to reaffirm the conclusions of that original evaluation and screening criteria, in light of the most recent information on the shroud cracking phenomenon.
.QBJECTIVE:
The primary objective of this safety analysis is to document the basis for the expectation that severe cracking (i.e. 360* cracking,90% or greater thru-wall) is not present in the core shroud welds of PBAPS, Unit 2 at this time. Reaffirming this expectation, the facility can continue to operate under the current licensing basis as documented in the PBAPS UFSAR, Section 3.3.5 and Appendix K Additionally, the likelihood of the simultanoous occurrence of licensing basis events and complete (360*, thru-wall) shroud cracking, prior to the next refueling outage, is discussed. Further, the ability of the shroud and plant safety features to respond to the effects of these unlikely events is addressed. Finally, actions which PECO Energy Company has taken to address the impact that severe shroud cracking could have on safe plant operations are presented.
BACKGROUND:
The PBAPS, Unit 2 shroud is identical, in configuration and material, to the shroud at PBAPS, Unit 3. It is however, expected to have a lower level of susceptibility to Intergranular Stress Corrosion Cracking (IGSCC) than Unit 3.
This lower susceptibility is based on the fact that PBAPS, Unit 2 has maintained a better level of water quality than PBAPS, Unit 3. This results in a lower susceptibility factor for PBAPS, Unit 2 (7.0 vs. 7.6 for Unit 3, per Table 2-3 of Reference 4). This better water conductivity record has resulted in a substantially lower frequency of occurrence of IGSCC related defects throughout the reactor coolant pressure boundary (e.g. cracking in: shroud head bolts, access hole covers, and recirculation system piping) at PBAPS, Unit 2. Nevertheless, the Reference 1 Page 1 of 5 3
GENERIC LETTER 94-03 SAFETY ANALYSIS PEACH BOTTOM ATOMIC POWER STATION, UNIT 2 Docket No. 50-277 Revision 0 BACKGROUND:
Screening Criteria document was prepared, assuming that the extent of shroud
_(. ont'd) cracking discovered in the Unit 3 shroud, was also present in the Unit 2 shroud.
C The criteria evaluated the impact of the theoretical cracking on the structural integrity of the shroud structure, through an additional operating cycle, including power rerate conditions. The evaluation identified that even with the same extent of cracking as found in the Unit 3 shroud, the PBAPS, Unit 2 shroud would maintain its structural integrity during the enveloping design basis accident conditions (i.e. Main Steam line break plus maximum credible earthquake), for an additional cycle of operation.
BASIS:
The basis for the expectation that severe shroud cracking does not exist is as follows:
The conditions which could influence the initiation and propagation of IGSCC in Boiling Water Reactor (BWR) shroud welds have been identified and discussed in Reference 1, Sections 1, 2, and 3.
While PBAPS, Unit 2 is considered highly susceptible to IGSCC, it must be emphasized that the basis for the assumed extent of cracking is the actual examination results, derived from extensive inspections of the shroud at its twin facility; PBAPS, Unit 3. It is also important to emphasize that the most severe cracking found-to-date (i.e. 360* deep cracks), has always been located in the end grain region of the welded plate rings. Since PBAPS, Unit 2 was J
fabricated with seamless, roll-forged rings, susceptibility to this extensive cracking i
is unlikely. This has been demonstrated by actual inspection data. No indications were found on any of the PBAPS, Unit 3 rings, nor in the forged rings at any other plant which has performed inspections.
While the accuracy of results for shroud inspections conducted using the enhanced VT-1 visual examination technique has been questioned, the PBAPS, Unit 3 inspections, as applied to the Unit 2 shroud, address such concerns based on the following:
The complete circumference of the inside surface (i.e. ID) of both the H-2 and the H-3 weld were visually inspected. (e.g. all portions of the weld and both heat affected zones were viewed and recorded).
Questionable indications were conservatively characterized as cracks.
The length of each identified indication was conservatively assigned, realizing the inaccuracles associated with the visual technique.
Page 2 of 5
GENERIC LETTER 94-03 SAFETY ANALYSIS PEACH BOTTOM ATOMIC POWER STATION, UNIT 2 Docket No. 50-277 Revision 0 BASIS:
Final plotting of all identified and sized indications was done in the (Cont'd) tarizontal plane, even if the indication was oriented at an angle to the horizontal.
ASME Section XI Proximity Rules were then applied to join nearby indications.
Finally, crack growth was applied to each plotted indication, to account for theoretical increases in the length of the Indication during the subsequent operating cycle.
All indications were characterized as through-wall cracks.
Loadings associated with power rerate conditions were used to evaluate the shroud indications, even though the plant is not currently operating at rerate conditions.
Additional conservativisms which were applied to the examination results and the evaluation of the PBAPS, Unit 2 shroud are listed in Table 1-1 of Reference 1.
The basis for the limited probability of the design basis accidents occurring simultaneously with severe shroud cracking is as follows The PBAPS, Unit 2 individual Plant Examination (IPE) submittal (Reference 8), and j
the NRC Safety Evaluation related to core shroud cracking at Dresden and Quad Cities, dated July 20,1994 (Reference 9), have been reviewed. Based on this review, the assessment of the risk of core damage at PBAPS, Unit 2, resulting from core shroud weld cracking (e.g. H-5 weld) and coincident system failures is significantly lower than the risk of other events analyzed in the IPEs.
Additionally, the likelihood of the occurrence of a LOCA is influenced by the condition and quality of the reactor coolant pressure boundary components. A review of the ASME Section XI Inservice inspection (ISI) Program examination results for the reactor coolant system piping provides an indication of the condition and integrity of this pressure boundary. Of the 127 welds which make up the Class 1 Main Steam piping, no significant indications have been identified during these examinations to-date. Additionally, all Recirculation System piping was replaced in 1984. The piping materials and fabrication processes used for this replacement resulted in a system which is highly immune to IGSCC. Further, all 56 welds which join the 316K Grade stainless steel piping were examined, for preservice inspection, at the time of installation. Review of these and subsequent ISI examination results, revealed that no significant indications have been identified.
Page 3 of 5 4
GENERIC LETTER 94-03 SAFETY ANALYSIS.
PEACH BOTTOM ATOMIC POWER STATION, UNIT 2 Docket No. 50-277 Revision 0 BASIS:
While PBAPS, Unit 2 is not expected to experience severe shroud cracking during
,{Qgnt'd) the current operating cycle, the effects of such theoretical cracking on the shroud and associated safety systems abHity to respond to design basis accident conditions have been considered. The BWR Shroud Cracking Generic Safety.
Assessment (Reference 4), which was developed by the BoHing Water Reactor Vessel Internals Project (VIP), and submitted to the NRC on August 5,1994, has been reviewed. PECO Energy Company endorses this document. Since the generic assessment is applicable to the PBAPS, Unit 2 shroud configuration, the conclusions of the document further support this safety analysis. Since the Generic assessment indicates that the reactor wRl be able to be shutdown and cooled down even with the effects of the design basis accidents on a completely severed shroud, the PBAPS, Unit 2 facility is enveloped by this assessment, considering the limited extent of degradation expected in the PBAPS, Unit 2 shroud.
Finally, to address the importance of the shroud cracking issues, PBAPS has informed plant operators of the issues to alert them as to the potentle' affects on plant operating parameters. This information is intended to. Impart a keen awareness of the potential safety significance of the shroud cracking issue.
CONCLUSIONS:
Because the conditions of the PBAPS, Unit 2 shroud are expected to be' enveloped by those identified in the PBAPS,- Unit 3 facility, the conclusions reached for the Unit 3 apply to the Unit 2 as well. Since the Unit 3 shroud has already been Inspected, the extent of cracking has been identified within reasonable accuracy.
This identified indications have been conservatively evaluated for impact on the performance of the shroud structure during design basis accidents. The results of -
the evaluation indicate that the PBAPS, Unit 2 shroud can, even in its current condition, support safe and continued operation of the facuity until the new refueling octage. Therefore,~ PBAPS, Unit 2 likewise can safely operate untu the next refuciing outage which is scheduled for the Fall of 1994.
Page 4 of 5
GENERIC LETTER 94-03 t
SAFETY ANALYSIS PEACH BOTTOM ATOMIC POWER STATION, UNIT 2 Docket No. 50 277 Revision 0
REFERENCES:
1.
" Evaluation and Screening Criteria for the Peach Bottom Unit-2 Shroud,"
Rev. O, (GENE-523-176-1093) dated December 13,1993.
2.
" Evaluation and Screening Criteria for the Peach Bottom Unit-3 Shroud indications," Rev.1, (GENE-523-141-1093) dated December 3,1993.
3.
"BWR Shroud Cracking Generic Safety Assessment," Rev. O, (GENE-523-A107P-0794) dated July 1994. Submitted to NRC via Letter No. BWROG-94089, dated 7/13/94.
4.
"BWR Shroud Cracking Generic Safety Assessment," Rev.1, (GENE-523-A107P-0794) dated August 1994. Submitted to NRC via Letter No. BWROG-94097, dated 8/5/94.
5.
"BWR Core Shroud Evaluation", Rev. O, (GE-NE-523-148-1193) dated April 1994. Submitted to NRC via Letter No. BWROG-94041, dated 4/5/94.
6.
" Safety Assessment - BWR Shroud Crack Indications." Submitted to NRC via Letter No. BWROG-93130, dated 11/9/93.
7.
" Response to NRC Ouestions on Core Shroud and Reactor Internals," Rev.
O, (GENE-523-A114P-0894) dated August 1994. Submitted to NRC via Letter No. BWROG-94100, dated 8/5/94.
8.
PBAPS Individual Plant Examination (IPE), submitted August,1992.
9.
NRC Safety Evaluation Related to Core Shroud Cracking at Dresden Unit 3, and Quad Cities, Unit 1 dated July 20,1994.
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l ATTACHMENT 3 Safety Analysis Limerick Generating Station, Unit 1 j
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GENERIC LETTER 94-03 SAFETY ANALYSIS LIMERICK GENERATING STATION, UNIT 1 Docket No. 50-352 Revision 0 INTRODUCTION:
This safety analysis has been developed to document the ability to safely operate the Limerick Generating Station, Unit 1 (LGS, Unit 1) facility until core shroud inspections are conducted. This analysis has been developed in accordance with
" Licensee Requested Action" No. 2 of NRC Generic Letter (GL) 94-03. Shroud inspections (or equivalent actions) are scheduled to be conducted during the next refueling outage of LGS, Unit 1(1R06), which is currently scheduled to begin in January 1996. Accordingly, this analysis applies to the period of operation between the date of this submittal, and the start of the sixth refueling outage of LGS, Unit 1.
OBJECTIVE:
The primary objective of this safety analysis is to document the basis for the expectation that severe cracking (i.e. 360* cracking, 90% or greater thru-wall) is not present in the core shroud welds of LGS, Unit 1 at this time. Reaffirming this expectation, the facility can continue to operate under the current licensing basis as documented in the LGS UFSAR, Section 3.9.5. Additionally, the likelihood of the simultaneous occurrence of licensing basis events and complete (360*, thru-wall) shroud cracking, prior to the next refueling outage, is discussed. Further, the ability of the shroud and plant safety features to respond to the effects of these unlikely events is addressed. Finally, actions which PECO Energy Company has taken to address the impact that severe shroud cracking could have on safe plant operations are presented.
BACKGROUND:
Augmented core shroud examinations have not yet been conducted at LGS, Unit
- 1. This is based on the recommendations put forth in General Electric Company i
Service Information Letter (SIL) 572, Rev.1, which recommends inspection for facilities like LGS, Unit 1 after eight (8) years of hot operation. LGS, Unit 1 has only experienced 6.4 years of hot operation. Additionally, since the LGS, Unit 1 shroud is fabricated from stainless steel materials which contain a very low carbon content, their susceptibi'ity to the intergranular Stress Corrosion Cracking (IGSCC) is reduced. Further, good water quality control history at the LGS, Unit 1 facility, since initial operation, further retards the initiation of such cracking.
. BASIS:
The basis for the expectation that severe shroud cracking does not exist is as follows.
Since LGS, Unit 1 has only experienced 6.4 on line years of operation, initiation of IGSCC cracking is not expected at this time. This is based on parameters such as the materials of construction and environment (water chemistry, and temperature).
At LGS, Unit 1, these parameters include: 0.15 for a 5-cycle conductivity value and 6.4 on-line years; which yleids a susceptibility factor of 1.0 (per Table 2-3 of Reference 2). This value places the LGS, Unit 1 facility in the middle of the lowest susceptibility ranking group which is made up of 14 plants. Of these 14 plants,6 Page 1 of 3 i
i GENERIC LETTER 94-03 SAFETY ANALYSIS LIMERICK GENERATING STATION, UNIT 1 Docket No. 50 352 Revision 0 1
BASIS:
have already performed shroud examinations, and none have found any indications,.
(Cont'd) let alone extensive cracking. Of additional importance, is the fact that many of these plants utilize the welded plate rings. Since these types of rings have been found to have an increased level of susceptibility to the cracking phenomenon in older plants, it is important to emphasize that none of the newer plants have found cracking in this area.
The basis for the limited probability of the design basis accidents occurring simultaneously with severe shroud cracking is as follows:
The LGS, Unit 1 Individual Plant Examination (IPE) submittal (Reference 7), and the NRC Safety Evaluation related to core shroud cracking at Dresden and Quad Cities, dated July 20,1994 (Reference 8), have been reviewed. Based on th!s review, the i
assessment of the risk of core damage at LGS, Unit 1, resulting from core shroud weld cracking (e.g. H-5 weld) and coincident system failures is significantly lower than the risk of other events analyzed in the IPEs.
Additionally, the likelihood of the occurrence of a LOCA is influenced by the condition and quality of the reactor coolant pressure boundary components. A review of the ASME Section XI ISI Program examination results for the reactor coolant system piping provides an indication of the condition and integrity of this j
pressure boundary. Of the 88 welds contained in the Class 1 Main Steam piping, no significant Indications have been identified during the ISI examinations to-date.
Additionally, of the 95 welds contained in the Recirculation System piping, no significant Indications have been identified. Additionally, the piping materials and fabrication processes used for the Recirculation System are highly immune to the occurrence of IGSCC.
While LGS, Unit 1 is not expected to experience severe shroud cracking during the current operating cycle, the effects of such theoretical cracking on the shroud and associated safety systems ability to respond to design basis accident conditions have been considered. The BWR Shroud Cracking Generic Safety Assessment (Reference 2), which was developed by the Boiling Water Reactor Vessel Internals Project (VIP), and submitted to the NRC on August 5,1994, has been reviewed.
PECO Energy Company endorses this document. Since the generic assessment is applicable to the LGS, Unit 1 shroud configuration, the conclusions of the document further support this safety analysis. Since the generic assessment Indicates that the reactor will be able to be shutdown and cooled down even with the effects of the design basis accidents on a completely severed shroud, the LGS, Unit 1 facility is enveloped by this assessment, considering that no degradation is likely on the LGS shroud.
Page 2 of 3
GENERIC LETTER 94-03 SAFETY ANALYSIS LIMERICK GENERATING STATION, UNIT 1 Docket No. 50-352 Revision 0 i
RASISI Finally, to address the importance of the shroud cracking issues, LGS has informed (Cont'd) plant operators of the issues to alert them as to the potential affects on plant operating parameters. This information is intended to impart a keen awareness of the potential safety significance of the shroud cracking issue.
CONCLUSIONS:
Because the LGS, Unit 1 shroud is highly resistant to IGSCC by virtue of both its i
materials of construction, and its history of good water quality, the likelihood of the l
initiation of any cracking in the core shroud, let alone severe cracking, is extremely low. Therefore, the LGS, Unit 1 shroud can support safe and continued operation, within its current licensing basis, until the next refueling outage, scheduled for January 1996.
REFERENCES:
1.
"BWR Shroud Cracking Generic Safety Assessment" Rev. O. (GENE-523-A107P-0794) dated July 1994. Submitted to NRC via Letter No. BWROG-94089, dated 7/13/94.
I 2.
"BWR Shroud Cracking Generic Safety Assessment" Rev.1, (GENE-523-A107P-0794) dated August 1994. Submitted to NRC via Letter No. BWROG-94097, dated 8/5/94.
3.
"BWR Core Shroud Evaluation", Rev. O, (GE-NE-523-148-1193) dated April 1994. Submitted to NRC via Letter No. BWROG-94041, dated 4/5/94.
4.
" Safety Assessment - BWR Shroud Crack Indications. Submitted to NRC via Letter No. BWROG-93130, dated 11/9/93.
5.
" Response to NRC Ouestions on Core Shroud and Reactor internals", Rev.
O, (GENE 523 A114P-0894) dated August 1994. Submitted to NRC via Letter No. BWROG-94100, dated 8/5/94.
6.
General Electric Company Service Information Letter (SIL) 572, Rev.1 dated October 4,1993.
7.
LGS Individual Plant Examination (IPE), submitted July,1992.
8.
NRC Safety Evaluation Related to Core Shroud Cracking at Dresden, Unit 3, and Quad Cities, Unit 1, dated July 20,1994.
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ATTACHMENT 4 Safety Analysis Limerick Generating Station, Unit 2 l
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GENERIC LETTER 94-03 SAFETY ANALYSIS LIMERICK GENERATING STATION, UNIT 2 Docket No. 50-353 Revision 0 INTRODUCTION:
This safety analysis has been developed to document the ability to safely operate the Limerick Generating Station, Unit 2 (LGS, Unit 2) facility until core shroud inspections are conducted. This analysis has been developed in accordance with
" Requested Licensee Action" No. 2 of NRC Generic Letter (GL) 94-03. Shroud inspections (or equivalent actions) are scheduled to be conducted during the next refueling outage of LGS, Unit 2 (2R03), which is currently scheduled to begin in January 1995. Accordingly, this analysis applies to the period of operation between the date of this submittal, and the start of the third refueling outage of LGS, Unit 2.
OBJECTIVE:
The primary objective of this safety analysis Is to document the basis for the expectation that severe cracking (i.e. 360' cracking,90% or greater thru-wall) is not present in the core shroud welds of LGS, Unit 2 at this time. Reaffirming this expectation, the facility can continue to operate under the current licensing basis as documented in the LGS UFSAR, Section 3.9.5. Additionally, the likelihood of the simultaneous occurrence of licensing basis events and complete (360*, thru-wall) shroud cracking, prior to the next refueling outage, Is discussed. Further, the ability of the shroud and plant safety features to respond to the effects of these unlikely events is addressed. Finally, actions which PECO Energy Company has taken to address the impact that severe shroud cracking could have on safe plant operations are presented.
RACKGROUND:
Augmented core shroud examinations have not yet been conducted at LGS, Unit
- 2. This is based on the recommendations put forth in General Electric Company Service Informatica Letter (SIL) 572, Rev 1, which recommends inspection, for facilities like LGS, 'Init 2, after eight (8) years of hot operation. LGS, Unit 2 has only experienced 3.6 years of hot operation. Additionally, since the LGS, Unit 2 shroud is fabricated from Stainless steel materials which contain a very low carbon content, their susceptibility to the Intergranular Stress Corrosion Cracking (IGSCC) is reduced. Further, good water quality control history at LGS, Unit 2 since initial operation further retards the initiation of such cracking.
BASIS:
The basis for the expectation that severe shroud cracking does not exist is described below:
Since LGS, Unit 2 has only experienced 3.6 on line years of operation, initiation of IGSCC cracking is not expected at this time. This is based on parameters such as the materials of construction and environment (water chemistry, and temperature).
These parameters include: 0.123 for a 5-cycle conductivity value and 3.6 on-line years; which yield a susceptibility factor of 0.4 (per Table 2-3 of Reference 2). This value places the LGS, Unit 2 facility as the least susceptible plant in the entire fleet Page 1 of 3
GENERIC LETTER 94-03 SAFETY ANALYSIS LIMERICK GENERATING STATION, UNIT 2 Docket No. 50-353 Revision 0 BASIS:
of BWR plants. Of the 14 plants in the same (lowest) susceptibility grouping,6
.(C_ont'd) have already performed shroud examinations, and none have found any indications, let alone extensive cracking. Of additional importance, is the fact that many of these plants utilize the welded plate rings. Since these types of rings have been found to have an increased level of susceptibility to the cracking phenomenon in older plants, it is important to emphasize that none of the newer plants have found cracking in this area.
The basis for the limited probability of the design basis accidents occurring simultaneously with severe shroud cracking is described below:
The LGS, Unit 2 Individual Plant Examination (IPE) submittal (Reference 7), and the NRC Safety Evaluation related to core shroud cracking at Dresden and Quad Cities, dated July 20,1994 (Reference 8), have been reviewed. Based on this review, the assessment of the risk of core damage at LGS, Unit 2, resulting from core shroud weld cracking (e.g. H-5 weld) and coincident system failures is significantly lower than the risk of other events analyzed in the IPEs.
Additionally, the likelihood of the occurrence of a LOCA is influenced by the condition and quality of the reactor coolant presswe boundary components. A review of the ASME Section XI ISI Program examination results for the reactor coolant system piping provides an Indication of the condition and integrity of this pressure boundary. Of the 88 welds contained in the Class 1 Main Steam piping, no significant indications have been identified during the ISI examinations to<! ate.
Additionally of the 95 welds contained in the Recirculation System piping, no significant indications have been identified. Additionally, the piping materiais and fabrication processes used for the Recirculation System are highly immune to the occurrence of IGSCC.
While LGS, Unit 2 is not expected to experience severe shroud cracking during the current operating cycle, the effects of such theoretical cracking on the shroud and associated safety systems ability to respond to design basis accident conditions have been considered. The BWR Shroud Cracking Generic Safety Assessment (Reference 2), which was developed by the Bolling Water Reactor Vessel Intemals Project (VIP), and submitted to the NRC on August 5,1994, has been reviewed.
PECO Energy Company endorses this document. The generic assessment is i
applicable to the LGS, Unit 2 shroud configuration, the conclusions of the document further support this safety analysis. Since the generic assessment indicates that the reactor will be able to be shutdown and cooled down even with the effects of the design basis accidents on a completely severed shroud. The LGS, Unit 2 facility is enveloped by this assessment, considering that no degradation is likely on the LGS shroud.
Page 2 of 3
GENERIC LETTER 94-03 SAFETY ANALYSIS LIMERICK GENERATING STATION, UNIT 2 Docket No. 50-353 Revision 0 AA_S$1 Finally, to address the importance of the shroud cracking issues, LGS has (Cont'd)
Informed plant operators of the issues to alert them as to th9 potential affects on plant operating parameters. This information is intended to impart a keen awareness of the potential safety significance of the shroud cracking issue.
CONCLUSlQNS:
Because the LGS, Unit 2 shroud is highly resistant to IGSCC by virtue of both its materials of construction, and its history of good water quality, the likclihood of the initiation of any cracking in the core shroud, let alone severe cracking, is extremely low. Therefore, the LGS, Unit 2 shroud can support safe and continued operation, l
within its current licensing basis, until the next refueling outage, scheduled for January 1995.
l
REFERENCES:
1.
"BWR Shroud Cracking Generic Safety Assessment" Rev. O, (GENE-523 A107P-0794) dated July 1994. Submitted to NRC via Letter No. BWROG-94089, dated 7/13/94.
2.
"BWR Shroud Cracking Generic Safety Assessment" Rev.1, (GENE-523-A107P-0794) dated August 1994. Submitted to NRC via Letter No. BWROG-94097, dated 8/5/94.
3.
"BWR Core Shroud Evaluation", Rev. O, (GE-NE-523-148-1193) dated April 1994. Submitted to NRC via Letter No. BWROG-94041, dated 4/5/94.
4.
" Safety Assessment - BWR Shroud Crack Indications. Submitted to NRC via Letter No. BWROG-93130, dated 11/9/93.
5.
" Response to NRC Questions on Core Shroud and Reactor Internals", Rev.
O, (GENE-523-A114P-0894) dated August 1994. Submitted to NRC via Letter No. BWROG-94100, dated 8/5/94.
6.
General Electric Company Service Information Letter (SIL) 572, Rev.1 dated October 4,1993.
7.
LGS Individual Plant Examination (IPE), submitted July,1992.
8.
NRC Safety Evaluation Related to Core Shroud Cracking at Dresden, Unit 3, and Quad Cities, Unit 1, dated July 20,1994.
Page 3 of 3
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ATTACHMENT 5 2
Drawings of Core Shroud Configuration for Peach Bottom Atomic Power Station, Units 2 and 3 Limerick Generating Station, Units 1 and 2
REACTOR PRESSURE VESSEL - SHROUD PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 pDRYER/ SEPARATOR SUPPORT RING WELO NO.
\\
M ITEM 7 A182-F 304 0.035 % C Hijf V1 V2 g
ITEM 5 A240 TP. 304 0.052 %, C (MAX)
H2%$
7,0P GUIDE SUPPORT RING, T
k%%1 ITEM B A182-F 304 0.028 % C (UNIT 2) l 0.030 % C (UNIT 3)
ITEM 4 A240 TP.304 0.060 % C (MAX) f V3 V4 H4 V5 ys ITEM 14 A240 TP.304 0.060 % C (MAX)
H5 gCORE PLATE SUPPORT RING 5////A ITEM 13 A182-F 304 0.030 % C (UNI" 2) 7/
l 0.035 % C (UNIT 3)
ITEM 12 A240 TP.304 0.059 % C (MAX)
H7 V7 V8 j
SHROUO UP R
YLINDER pswswxj 9
W-REF. DWG. M-1-B-26
REACTOR PRESSURE VESSEL - SHROUD LIMERICK GENERATING STATION UNIT 1 DRYER / SEPARATOR SUPPORT RING T
WELD NO.
k13 PC.12 A240 TP.304L 0.026 % MAX.C Hij V1 V2 i
PC.1 A240 TP. 304L 0.020 % C H2 m eTdP GUIDE SUPPORT RING m
h\\\\1 PC.13 A240 TP.304L 0.024 % C H3 PC.2 A240 TP.304L 0.023.% C V3 V4 H4 V5 V6 PC.3 A240 TP.304L 0.018 % C t
gCORE PLATE SUPPORT RING H5 V///gA PC.14 A240 TP.304L 0.024 % C H6 PC.4 A240 TP.304L 0.024 % C H7 V7j V8
["
(kLL 600 Am
/
REF.DWGS:
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M-1-B-11-0001-C-24-2 8031-M-108
REACTOR PRESSURE VESSEL - SHROUD LIMERICK GENERATING STATION UNIT 2 DRYER / SEPARATOR SUPPORT RING WELD NO.
/
M PC.12 A240 TP.304L 0.026 % MAX.C Hij V1 V2 PC.1, A240 TP. 304L 0.020 % C H2 m
,eTOP OUIDE SUPPORT RING h%1 PC.13 A240 TP.304L 0.024 % C H3 PC.2 A240 TP.304L 0.023 % C V3 V4 H4 V5 V6 i
PC.3 A240 TP.304L 0.018 % C l
gCORE PLATE SUPPORT RING H5-V//////A PC.14 A240 TP.304L 0.024 % C H6 PC.4 A240 TP.304L 0.024 % C H7 V7 V8
[
(L0 00)
Assss
/
REF.DWGS:
/
M-1-B-11-D001-C-24-2 d-8031-M-108
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