3F0898-01, LAR 235,rev 0 to License DPR-72,proposing New Repair Process for Plant Otsgs.Technical Basis for Repair Roll Process Proposed by LAR Is Contained in BAW-2303P,Rev 3
| ML20238F463 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 08/31/1998 |
| From: | Cowan J FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20238F464 | List: |
| References | |
| 3F0898-01, 3F898-1, NUDOCS 9809030376 | |
| Download: ML20238F463 (19) | |
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Fisrida Power 8 F3."Ji?2" o "p. 1 ".* f ,'.'. w oen42
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l I August 31,1998 3F0898-01 U.S. Nuclear Regulatory Commissicn !
l Attn: Document Control Desk Washington, DC 20555-0001 i
Subject:
License Amendment Request #235, Revision 0 j Once Through Steam Generator Tube Surveillance Program, j Tube Repair Roll Process '
i Dear Sir Pursuant to 10 CFR 50.90, Florida Power Corporation (FPC) hereby submits License Amendment Request (LAR) #235, Revision 0, for an amendment to the Crystal River Unit 3
, (CR-3) Operating License No. DPR-72. This LAR proposes a new repair process for the CR-3 Once Through Steam Generators (OTSGs). The technical basis for the repair roll (re-roll) 1 process proposed by this LAR is contained in Topical Report BAW-2303P, Revision 3.
Proprietary Topical Report BAW-2303P, Revision 3, was previously submitted to the NRC as part of similar LARs for Oconee Nuclear Stations 1, 2 and 3, Arkansas Nuclear One and l Davis-Besse Nuclear Station (References 5,6,10 and 11 of Attachment A). A non-proprietary l version of the report was provided to the NRC by Davis-Besse (Reference 7 of Attachment A). l As part of this request, FPC is providing the LAR, No Significant Hazards Consideration Evaluation, and Environmental Impact Evaluation (Attachment A), Proposed Technical Specification Change Pages, Strikeout / Shaded (Attachment B), Proposed Technical Specification Change Pages, Revision Bars (Attachment C), and List of Regulatory Commitments (Attachment D).
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The Improved Technical Specifications (ITS) pages provided in Attachments B and C to this !
letter have been affected by previously proposed OTSG LAR #221, Revisions 0 and 1.
Proposed changes provided in Attachments B and C were made to the current ITS pages, as approved by License Amendment Numbers 158 and 169. Once a License Amendment is issued for LAR #221, FPC will provide updated ITS pages for LAR #235. o j FPC is requesting NRC approval of this LAR by July 19, 1999. This allows sufficient time for implementation prior to the next refueling outage for CR-3.
f 9809030376 900831 it f PDR ADOCK 05000302 P pm CRYSTAL RIVER ENERGY COMPLEX: 15760 W. Power une Street
J U.S. Nuclear Regulatory Commission 3F0898-01 Page 2 of 3 If you have any questions regarding this letter, please contact Ms. Sherry Bernhoft, Manager, Nuclear Licensing at (352) 563-4566.
Sincerely, MMC '
John Paul Cowan
- Vice President Nuclear Operations JPC/lve xc: Regional Administrator, Region II Senior Resident Inspector NRR Project Manager Attachments:
A. License Amendment Request, No Significant Hazards Consideration Evaluation, and Environmental Impact Evaluation B. Proposed Technical Specification Change Pages, Strikeout / Shaded C. Proposed Technical Specification Change Pages, Revision Bars D. List of Regulatory Commitments I
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U.S. Nuclear Regulatory Commission 3F08984)1 1
' Page 3 of 3 I
STATE OF FLORIDA COUNTY OF CITRUS i
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b John Paul Cowan states that he is the Vice President, Nuclear Operations for Florida Power Corporation; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief.
&&c ~
John Paul Cowan Vice President
' Nuclear Operations )
4 Sworn to and subscribed before me this 3l day of b d6t ,1998,by l John Paul Cowan.
Oihu-, .
Signature of tary Public l State of Florida MycoMMrN[ccf2Y802 EXPIRES - ' '
d#U 8=t*Y'*%"" (Print, type, or stamp Commissioned Name of Notary Public) i Personally / Produced Known v -OR- Identification
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302/ LICENSE NO. DPR-72 ATTACHMENT A LICENSE AMENDMENT REQUEST #235, REVISION 0 ONCE TIIROUGH STEAM GENERATOR TUBE SURVEILLANCE PROGRAM, TUBE REPAIR ROLL PROCESS License Amendment Request, No Significant Hazards Consideration Evaluation, and Environmental Impact Evaluation I
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U.S. Nuclear Regulatory Commission Attachment A 3F0898-01 Page1of15 FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302/ LICENSE NO. DPR-72 LICENSE AMENDMENT REQUEST #235, REVISION 0 ONCE THROUGH STEAM GENERATOR TUBE SURVEILLANCE PROGRAM, TUBE REPAIR ROLL PROCESS LICENSE DOCUMENT INVOLVED: Improved Technical Specifications (ITS)
PORTIONS: ITS 5.6.2.10, " Steam Generator (OTSG) Tube Surveillance Program" ITS 5.7.2, "Special Reports"
SUMMARY
OF CHANGES:
The program is being revised to require inspection of both Once Through Steam Generators (OTSGs) during each inservice inspection.
References to pre-service inspections and inservice inspections (first, second and third) are being deleted since the plant is now in its twenty-first year of commercial operation. Crystal River Unit 3 (CR-3) has performed its thirteenth OTSG inservice inspection and has entered the third ten-year inservice inspection interval.
Florida Power Corporation (FPC) is proposing a new repair process for the OTSGs called l
" repair roll" or "re-roll." This process is used to repair steam generator tubes with defects I within the upper tubesheet.
Inservice inspection and reporting requirements are proposed for tubes which have a repair roll process performed in the upper tubesheet.
Several format and editorial changes are being proposed for clarification.
BACKGROUND:
CR-3 is a Babcox and Wilcox (B&W) pressurized water reactor with Model 177 FA Once Through Steam Generators. In this design, the primary coolant enters the steam generators at the top of the tubes (the hot leg) and exits the steam generator at the bottom of the tubes (the cold leg), where the primary coolant is directed back to the reactor coolant pumps and the reactor vessel. The tubes are mill annealed Alloy 600 (Inconel) which have been sensitized as a result of the full vessel post fabrication heat treatment. The original tube-to-tubesheet joint consists of a roll expansion of one to two inches in length, with a seal weld (fillet) between the tube and the primary side tubesheet cladding. The tubesheets are nominally 24 inches thick carbon steel with l
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U.S. Nuclear Regulatory Commission Attachment A 3F0898-01 Page 2 of 15 l a minimum primary side Inconel clad of 5/16 inch. Each OTSG was fabricated with 15,531 tubes. The tubes have a nominal outer diameter of 0.625 inches and a nominal wall thickness of 0.034 inches.
Industry experience has demonstrated that the roll transition and expansion region of Alloy 600 tubing can be susceptible to stress corrosion cracking (SCC) and intergranular attack (IGA),
particularly in the hot legs of the steam generators, which are operated about 50 F warmer than the cold leg.
During the 1997 CR-3 OTSG inservice inspection, a number of tubes were plugged due to indications in the hot leg (upper tubesheet) roll expansion and roll transition area. Reference 1 provides details on the results of this inspection. Both axial and volumetric indications were identified and dispositioned using rotating coil eddy current technology. These indications appeared to be originating from the tube inner diameter, indicating that the most likely modes of degradation are primary water SCC and/or IGA.
The 1997 inservice inspection was the first extensive assessment of the upper tubesheet roll transition area for the CR-3 OTSGs. Fifty-one (51) tubes in each OTSG were plugged as a !
result of indications in the upper roll region. An operational assessment performed in the spring of 1998 (Reference 2) suggests that about 100 tubes in each OTSG are expected to have indications of degradation in the upper roll region during the next inservice inspection.
Additionally, LAR #228 (Reference 3) and License Amendment No.169 (Reference 12) addressed the potential for indications previously identified as Tube End Anomalies (TEAS) and Multiple Tube End Anomalies (MEAS) in the upper roll expansions to be within the pressure boundary of the tubes. The implementation of a repair roll process will allow CR-3 to keep most of these tubes in service by establishing a new pressure boundary for the upper tubesheet end of the tube.
As a member of the B&W Owners Group (BWOG), CR-3 participated in and funded the development of Framatome Technologies Report BAW-2303P, Revision 3 "OTSG Repair Roll Qualification Report." This proprietary report was previously submitted and/or acknowledged by the NRC in References 5,6,10 and 11. Additionally, a non-proprietary version of the report -
was previously submitted in Reference 7. CR-3 proposes to implement the methodology and process established within this report as a repair method in the upper tubesheet. The design analysis and conclusions within BAW-2303P, Revision 3, bound the normal and accident conditions at CR-3. I BAW-2303P, Revision 3, provides a description of the repair process, design requirements, j design verification, and an evaluation of the repair roll life. The installation of a repair roll in a j l
tube will establish a new pressure boundary for that specific tube. The full length of the new roll l will constitute the new pressure boundary. The original tube-to-tubesheet joint will no longer be considered part of the pressure boundary, and can be removed from the inservice inspection program. To date, this repair process has been implemented at the Oconee Station, Arkansas Nuclear One, Unit 1, and Davis-Besse Nuclear Station.
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U.S. Nuclear Regulatory Commission Attachment A 3F0898-01 Page 3 of 15 The following considerations and limitations on BAW-2303P, Revision 3, applicability will be administered at CR-3:
- 1. The repair roll process will only be used in the upper tubesheet. !
- 2. Only the " single repair roll" methodology described in the qualification report will be implemented.
- 3. The repair roll length will be one inch, and this roll length will be controlled by the physical length of the roll expander used.
- 4. The repair roll will be situated entirely below the existing roll (there will be no overlap with the existing roll).
- 5. Only one repair roll will be performed per tube.
- 6. The repair roll area will be examined using eddy current methods following installation.
The repair roll must be free of imperfections and degradation for the repair to be considered acceptable.
- 7. All tubes which receive a repair roll will have the repair roll inspected during each subsequent inservice inspection.
- 8. If a primary-to-secondary leak is caused by degradation in a repair roll, and results in ,
, shutdown of the plant,100% of the repair rolls in that OTSG will be examined. If this l inspection falls in Category C-3, additional repair roll examinations will be performed I
in the unaffected OTSG.
Implementation of these limitations, along with the enhanced inservice inspection requirements, will ensure that leakage and stmetural integrity is maintained for these new tube-to-tubesheet joints while allowing these tubes to remain in service.
The CR-3 ITS has previously been changed to implement the 150 gallons per day (per OTSG) operating leak rate limit which is typically used in conjunction with alternate repair techniques.
License Amendment No.158 (Reference 8) was issued on October 28, 1997, to permanently establish this limit.
CHANGES TO SPECIFICATION 5.6.2.10, " Steam Generator (OTSG) Tube Surveillance Program" Description of Specification Change l Revise paragraph 1 of ITS 5.6.2.10, Page 5.0-13 to read as follows: 3
U.S. Nuclear Regulatory Commission Attachment A 3F0898-01 Page 4 of 15
Reason for Request
This change is being proposed to simplify the CR-3 ITS and to enhance inspection requirements.
Justification for Request Due to the age of the CR-3 OTSGs, and recent industry experience with OTSG tube degradation, CR-3 will inspect both OTSGs as part of the OTSG inservice inspection. This change is an enhancement to the program without impact to safety.
Description of Specification Change Revise paragraph 5.6.2.10.2, Page 5.0-13 to read as follows:
- 2. The OTSG tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 5.6.2-1. The inservice inspection of OTSG tubes shall be performed at the frequencies specified in Specification 5.6.2.10.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 5.6.2.10.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in each OTSG. The tubes selected for these inspections shall be selected on a random basis except:
Revise paragraph 5.6.2.10.2.d, Page 5.0-14, to change the reference from Table 5.6.2-3 to Table 5.6.2-2.
Reason for Request
Original Table 5.6.2-1 has been deleted and the remaining tables have been renumbered. Thus, former Table 5.6.2-2 is now Table 5.6.2-1, and Table 5.6.2-3 is now Table 5.6.2-2. The word "all" was changed to "each" since both OTSGs will be inspected during each inservice inspection.
Justification for Request Due to the age of the CR-3 OTSGs, and recent industry experience with OTSG tube degradation, CR-3 will inspect both OTSGs as part of the OTSG inservice inspection. Additionally, CR-3 has
- only two OTSGs; thus the use of "each" OTSG is more appropriate than "all" OTSGs.
I Description of Specification Change f Delete paragraphs 5.6.2.10.2.b and 5.6.2.10.2.c Page 5.0-13. Renumber current paragraphs j 5.6.2.10.2.d and 5.6.2.10.2.e on Pages 5.0-13 and 5.0-14 as paragraphs 5.6.2.10.2.b and 5.6.2.10.2.c, respectively.
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U.S. Nuclear Regulatory Commission Attachment A 3F0898-01 Page 5 of15
Reason for Request
These paragraphs are not applicable to CR-3 since the plant is now in its twenty-first year of commercial operation. This change will clarify the CR-3 ITS.
Justification for Request These paragraphs currently discuss the first, second and third OTSG inservice inspections. CR-3 performed the thirteenth inservice inspection during the 1997 extended outage. Therefore, these paragraphs are outdated and no longer needed.
Description of Specification Change Add new paragraph 5.6.2.10.2.d, Page 5.0-14 as follows:
- d. All tubes which have been repaired by the repair roll process shall have the new roll area examined during each inservice inspection. No credit is to be taken for this inspection in meeting minimum sample size requirements for the random inspection.
Degraded or defective tubes found during this inspection are to be plugged or sleeved. Degraded or defective tubes found during this inspection are not to be considered in determining the inspection results category for the random inspection.
Reason for Request
This inspection is added to ensure that an appropriate level of monitoring is performed on the new roll expanded areas. Defining the repair rolls as a specific limited area is consistent with the existing ITS requirements.
Justification for Request The additional required repair roll inspections are intended to' ensure that the condition and integrity of the new roll expansion is adequately monitored and maintained.
Removing the repair roll inspections from the minimum sample size requirements and random tube inspection results category is consistent with the current ITS 5.6.2.10.2.d requirements for a specific limited area. The upper tubesheet repair rolls are distinguished from the remainder of the tube bundle by unique physical construction. Not including these inspections as part of the minimum sample size ensures that a larger population of tubes is examined. Removing the
- repair roll inspection results from the overall inspection results classification will prevent bias of l the random inspection results due to a specific limited area in the OTSGs.
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U.S. Nuclear Regulatory Commission Attachment A 3F0898-01 Page 6 of15 i
Description of Specification Change Revise ITS 5.6.2.10.3, Page 5.0-15, paragraphs a, b and c as follows: !
Paragraph a, delete ", not including the preservice inspection,"
Paragraph b, change Table references to read "...with Table 5.6.2-1 or Table 5.6.2-2 requires..."
i Paragraph c, change Table references to read "...in Table 5.6.2-1 or Table 5.6.2-2 during..."
Reason for Request
The change to paragraph 5.6.2.10.3.a clarifies the CR-3 ITS since the preservice inspection no longer applies. The referenced tables have been renumbered due to the deletion of current Table 5.6.2-1.
Justification for Request CR-3 is now in its twenty-first year of commercial operation. The next inservice inspection of the OTSGs will constitute the fourteenth inservice inspection. Thus, the reference to the preservice inspection is outdated and no longer needed, i
I Description of Specification Change Add the following new paragraph to ITS 5.6.2.10.3, Page 5.0-15:
- d. If the source of primary-to-secondary leakage in excess of ITS 3.4.12 is determined to be due to degradation of a repair roll,100% of the tubes with a repair roll in the ,
affected OTSG shall have the repair roll inspected. If the results of this inspection j fall into the C-3 category, then additional repair roll inspections shall be performed in the unaffected OTSG.
Reason for Request .
l This paragraph provides provisions for assessing the extent of condition in the event that primary-to-secondary leakage is caused by degradation in the repair roll region.
Justification for Request l This proposed paragraph is consistent with the requirements for a specific limited area. These
- inspection requirements provide an adequate and proactive methodology for addressing primary-to-secondary leakage if attributed to degradation of the repair roll.
U.S. Nuclear Regulatory Commission Attachment A 3F0898-01 Page 7 of 15 Description of Specification Change
' Revise ITS 5.6.2.10.4.a.4, ITS 5.6.2.10.4.a.5 and ITS 5.6.2.10.4.a.6, Page 5.0-16 to read as follows:
4.- Degraded Tube means a tube containing degradation 2 20% throughwall but < 40%
throughwall in the pressure boundary.
- 5. % Degradation /% Throughwall means the percentage of the tube (pressure boundary) wall thickness affected or removed by degradation.
- 6. Defective Tube means a tube containing degradation 2 40% throughwall in the L pressure boundary. Any tube which does not permit the passage of the eddy-current inspection probe shall be deemed a defective tube.
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Reason for Request
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This proposed change is to provide clarification to these definitions.
> ' Justification for Request The current definitions of imperfection, degraded tube, and defective tube as shown in ITS l 5.6.2.10 are confusing, and in some cases, contradictory.
The current description of a degraded tube is "a tube containing imperfections 2 20% of the nominal wall thickness caused by degradation except where all such degradation has been
- spanned by the installation of a sleeve." This definition is confusing in that the term imperfection is defined in 5.6.2.10.4.a.2 as "an exception to the dimensiens, finish or contour of
.a tube from that required by fabrication drawings or specifications. Eddy-current testing
- indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections." Thus, the use of the term " imperfection" is inappropriate (a contradiction) in defining degradation as 2 20% throughwall.
The addition of % Throughwall to the definition of % Degradation is to identify that these terms i are identical in meaning. !
Revision of 5.6.2.10.4.a.6 is provided to simplify this paragraph.
, The phrase "except where the imperfection has been spanned by the installation of a sleeve" is l redundant wording, and is thus deleted. Paragraph 5.6.2.10.4.a.1 defines tubing or tube as "that !
portion of the tube or sleeve which forms the primary system to secondary system pressure !
boundary." The phrase delineating that these limits apply in the " pressure boundary" reinains in these definitions. This inclusion is intended to ensure that both the parent tube and sleeve pressure boundary degradation is categorized as either degraded or defective.
U.S. Nuclear Regulatory Commission Attachment A 3F0898-01 Page 8 of15 These changes will clarify the contents of the technical specification and improve specificity; thus, resulting in the following simplified terminology when communicating the results of tube inspections: '
Imperfection: Indications < 20% throughwall or manufacturing artifacts 1 Degraded Tube: Tubes with service-induced degradation 2 20% throughwall but < 40%
throughwall Defective Tube: Tube with service-induced degradation 2 40% throughwall
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Description of Specification Change Revise ITS 5.6.2.10.4.a.8, ITS 5.6.2.10.4.a.10, and add paragraph 5.6.2.10.4.a.11, Page I 5.0-17 as follows:
- 8. Plugging / Repair Limit means the extent of pressure boundary degradation beyond which the tube shall either be removed from service by installation of plugs or the area of degradation shall be removed from service (a new pressure boundary established) using an Approved Repair Technique. The plugging / repair limit is 40% throughwall for all pressure boundary degradation.
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- 10. Tube Inspection means an inspection of the OTSG tube pressure boundary.
- 11. Approved Repair Technique means a technique that has been submitted to, evaluated by, and accepted by the NRC as a methodology to remove degraded or defective portions of the pressure boundary from service and to establish a new pressure boundary in place of the degraded or defective portions which have been removed. The following are Approved Repair Techniques and are acceptable for use:
a) Sleeve installation in accordance with the B&W process (or method) described in report BAW-2120P. No more than live thousand sleeves may be installed in each OTSG.
b) Installation of 1" repair rolls in the upper tubesheet in accordance with the Framatome process (or method) described in report BAW-2303P, Revision 3. This repair process may be performed once per tube. The new roll area must be free of imperfections and degradation in order for the repair to be acceptable.
U.S. Nuclear Regulatory Commission Attachment A 3F0898-01 Page 9 of 15
Reason for Request
The changes to paragraphs 5.6.2.10.4.a.8 and 5.6.2.10.4.a.10 are for clarification. The addition of paragraph 5.6.2.10.4.a.11 is intended to establish a single paragraph in which Approved Repair Techniques will be listed. This approach will limit the amount of ITS pages that will need to be changed in the future as additional repair methodologies are implemented at CR-3.
Justification for Request When a steam generator tube has pressure boundary degradation ;t 40% throughwall, the tube must either be removed from service via plugging the tube, or the degraded portion of the pressure boundary must be repaired. The options currently available for OTSG tube repair
) include sleeving and installation of a repair roll. As the OTSGs age, additional repair technologies are being developed and considered for use. Referring to these alternate repair technologies generally as " Approved Repair Techniques" and simply listing them as they are approved for use will reduce subsequent changes needed to the ITS.
As repair techniques are implemented, new pressure boundaries are established in the affected tubes. The tube inspection is also intended to cover the new pressure boundary. Thus, the use of the phrase "the entire OTSG tube as far as possible" is nebulous, and does not adequately describe the necessary inspection.
Installation ofInconel 690 sleeves at CR-3, in accordance with Topical Report BAW-2120P, was evaluated and accepted by the NRC for CR-3 use in Reference 9. The sentences in paragraph 5.6.2.10.4.a.11.a were simply moved from other areas of the specification to this location.
Installation of repair rolls in accordance with BAW-2303P, Revision 3, is being requested by this LAR. This repair roll qualification repon has previously been submitted to the NRC in References 4 and 5.
Description of Specification Change Revise paragraph 5.6.2.10.4.b, Page 5.0-17 to read as follows: i
- b. The OTSG shall be determined OPERABLE after completing the I corresponding actions (plug or repair all tubes exceeding the plugging / repair limit) required by Table 5.6.2-1 (and Table 5.6.2-2 if the provisions of Specification 5.6.2.10.2.b are utilized).
Reason for Request
This change will simplify the CR-3 ITS.
U.S. Nuclear Regulatory Commission Attachment A 3F0898-01 Page 10 of 15
, Justification for Request Changing the references to plugged or repaired is consistent with previous changes presented in l this LAR. The referenced table numbers were changed to be consistent with other changes presented. The phrase " plug or sleeve all tubes exceeding the plugging / sleeving limit and all tubes containing through-wall cracks" was restated as shown above. The portion pertaining to throughwall cracks was redundant, since an area with a throughwall crack would exceed the defined plugging / repair limit by definition.
Description of Specification Change Delete current Table 5.6.2-1, Minimum Number of Steam Generators (OTSGs) to be Inspected During Inservice Inspection, Page 5.0-24.
Reason for Request
This change will simplify the CR-3 ITS.
Justification for Request This ITS table describes the number of steam generators that must be inspected during preservice, first, second, and subsequent inspections. This table also provides the option to inspect only one OTSG if previous inspection results indicate that the OTSGs are performing in a similar fashion.
Because of the age of the CR-3 OTSGs and recent industry experience with tube degradation, CR-3 will inspect both OTSGs to meet inservice inspection requirements in the future.
References to preservice, first, and second inservice inspections are not needed, since the plant has already passed through those periods.
Description of Specification Change l
Current Table 5.6.2-2, OTSG Tube Inspection, Page 5.0-25, is being renumbered to 5.6.2-1, due to the deletion of the current Table 5.6.2-1. References to " plug or sleeve" are being changed to " plug or repair." Inferences to more than two OTSGs are being revised to reflect that CR-3 has only two OTSGs.
Reason for Request
This change will simplify the CR-3 ITS. Also, as additional tube repair techniques are developed and submitted to the NRC for approval, fewer changed ITS pages will be required in the future.
U.S. Nuclear Regulatory Commission Attachment A
- 3F0898-01 Page 11 of 15 s
Justification for Request Specification 5.6.2.10.4.a is being changed to include a definition of Approved Repair Technique. This definition is intended to encompass all techniques used to keep tubes in service such as sleeving, re-rolling, welded tube repair, etc. This definition requires the techniques to be reviewed, and their use approved, by the NRC prior to field use at CR-3. Removing inferences to more than two OTSGs will simplify the specification.
l Description of Specification Change Current ITS Table 5.6.3-3, Specific Limited Area Inspection, Page 5.0-26, is being renumbered to 5.6.2-2, due to the deletion of the current Table 5.6.2-1. References to " plug or sleeve" are '
being changed to " plug or repair."
Reason for' Request This change will simplify the CR-3 ITS. Additionally, as tube repair techniques are developed and submitted to the NRC for approval, fewer changed ITS pages will be required in the future.
Justification for Request Specification 5.6.2.10.4.a is being changed to include a definition of Approved Repair Technique. This definition is intended to encompass all techniques used to keep tubes in service such as sleeving, repair roll, welded tube repair, etc. This definition requires the techniques to be reviewed, and their use approved, by the NRC prior to field use at CR-3.
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CHANGES TO SPECIFICATION 5.7.2, "SPECIAL REPORTS" Description of Specification Change Revise ITS 5.7.2, Special Reports, paragraphs 5.7.2.c.1 and 5.7.2.c.2; relocate and label the paragraph from Page 5.0-29A regarding reports of OTSG tube inspections resulting in Category C-3 as paragraph 5.7.2.d; label the paragraph for the 90-day report as paragraph 5.7.2.e, and revise and label subparagraph 4 of the 90-day report as follows:
5.7.2.c.1 Number of tubes plugged and repaired, L
5.7.2.c.2 Crack-like indications in the first span,
U.S. Nuclear Regulatory Commission Attachment A 3F0898-01 Page 12 of 15 5.7.2.d Results of OTSG tube inspections that fall into Category C-3 shall be reported to the NRC prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine the cause of tube degradation and corrective measures taken to prevent recurrence.
5.7.2.e Following each inservice inspection of OTSG tubes, the complete results of the OTSG tube inservice inspection shall be submitted to the NRC within 90 days. The report shall include:
5.7.2.e.4 Identification of tubes plugged and tubes repaired and specify the repair methodology implemented on each tube.
Reason for Request
This change simplifies the reporting requirements by requiring all tubes plugged and repaired to be reported. As additional tube repair techniques are submitted to, evaluated by, and accepted by the NRC, fewer changed ITS pages will be required in the future by not listing each repair technique in these paragraphs.
Identifying the OTSG inservice inspection 90-day report in a separate paragraph is intended to clarify this section of the ITS.
Justification for Request Specification 5.6.2.10, paragraph 4.a is being changed to include a definition of Approved Repair Technique. This definition will list the techniques which have been submitted to, evaluated by, and accepted by the NRC for use at CR-3. As future repair techniques are implemented at CR-3, the reporting requirements will not need to be changed. New repair techniques will need to be added only to the definition of Approved Repair Technique.
Identification of the OTSG inservice inspection 90-day report requirement as a distinct paragraph (5.7,2.d) is appropriate, since this special report is usually submitted separately from the MODE 4 report described in paragraph 5.7.2.c. Additionally, placing this paragraph and the requirements after the paragraph on inspection results that fall into Category C-3 is practical, because the C-3 paragraph describes providing a report to the NRC prior to the resumption of plant operation. Thus, these reporting requirements will now be presented in chronological order of expected occurrence (i.e., MODE 4 reporting requirements, plant operation requirements, then 90-day requirements).
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i l U.S. Nuclear Regulatory Commission Attachment A 3F0878-01 Page 13 of 15 l
NO SIGNIFICANT HAZARDS CONSIDERATION EVALUATION i
An evaluation of this proposed LAR has been performed in accordance with 10 CFR 50.91(a)(1) regarding significant hazard considerations, using the standards in 10 CFR 50.92(c). A discussion of these standards as they relate to this LAR follows:
(1) Involve a sigmficant increase in the probability or consequences of an accident previously evaluated.
l The proposed LAR addresses several editorial and format changes which do not impact accident analyses. LAR #235 also proposes to implement the repair roll (re-roll) process.
The qualification of the re-roll joint is based on establishing a mechanical roll length which will carry all stmetural loads imposed on the tubes with required margins. A series of tests I and analyses were performed to establish this length. Tests that were performed included l leak, tensile, fatigue, ultimate load and eddy current measurement uncertainty. The analyses evaluated plant operating and faulted loads in addition to tubesheet bow effects.
Any tube leakage will be bounded by the main steam line break (MSLB) evaluation {
1 presented in the Final Safety Analysis Report (FSAR). The propo:ed change also requires ]
inspections of the joints created by the repair roll process. The addition of this inspection i l does not change any accident initiators. The proposed inspections after re-roll installation, I and during future inservice inspections, assure continuous monitoring of these tubes such that inservice degradation of tubes repaired by the re-roll process will be detected. Based on the Framatome Technologies qualification, as well as the history for similar industry repair rolls, there are no new safety issues, as defined in BAW-2303P, Revision 3, associated with the repair roll. Therefore, this change does not involve a significant increase in the probability or consequences of any accident previously evaluated.
I i
(2) Create the possibility of a new or dgerent kind of accidentfrom any accident previously l evaluated. )
l No new failure modes or accident scenarios are created by the re-roll process. The new l pressure boundary joint created by the repair roll process has been shown by testing and l analysis to provide structural and leakage integrity equivalent to the original design and i construction for all normal operating and accident conditions. Furthermore, the testing and analysis demonstrate the repair roll process creates no new adverse effects for the repaired
- tube and does not change the design or operating characteristics of the OTSGs. In the i unlikely event that a tube with a repair roll should fail and sever completely at the transition of the re-roll region, the tube would remain engaged in the tubesheet bore, preventing interaction with other surrounding tubes. In this case, leakage is bounded by the steam generator tube rupture (SGTR) accident analysis. Therefore, this change does not create a possibility of a new or different kind of accident from any previously evaluated.
f
U.S. Nuclear Regulatory Commission Attachment A 3F0898-01 Page 14 of 15 (3) Im'olve a sigmficant reduction in a margin ofsafety.
The repair roll process effectively removes the defective / degraded area of the tube from service. The' new roll expanded interface created with the tubesheet shtisfies all the necessary stmetural, leakage and heat transfer requirements. The joint is constrained within the tubesheet bore; thus, there is no additional risk associated with tube rupture.
The accident leakage is shown to be well within the initial assumption of the MSLB analysis of one gallon per minute primary-to-secondary leakage. Therefore, the FSAR analyzed accident scenarios remain bounding, and the use of the repair roll process does not reduce the margin of safety.
ENVIRONMENTAL IMPACT EVALUATION 10 CFR 51.22(c)(9) provides criteria for identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not:
(i) involve a significant hazards consideration, (ii) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and (iii) result in a significant increase in individual or cumulative occupational radiation exposure.
l FPC has reviewed this proposed LAR and concludes it meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(c), no j environmental impact statement or environmental assessment needs to be prepared in connection with this request.
REFERENCES:
- 1. FPC to NRC letter, 3F1297-22, dated December 5,1997, "Special Report 97-05 Once Through Steam Generator (OTSG) Notifications Required Prior to MODE 4, and Complete Results of OTSG Tube Inservice Inspection Performed During the Cunent Outage (90-Day Report)"
- 2. FPC to NRC letter, 3F0598-08, dated May 18,1998, "An Operational Assessment of Steam Generator Tube Degradation at Crystal River Unit 3"
1' U.S. Nuclear Regulatory Commission Attachment A 3F0898-01 Page 15 of 15
- 3. FPC to NRC letter, 3F0698-28, dated June 18,1998, " Exigent License Amendment Request #228, Revision 0, Once Through Steam Generator Tube Surveillance Program"
- 4. Duke Power Company to NRC letter, dated October 20,1997, " Request for Technical l
Specification Amendment, Steam Generator Tubing Surveillance" [ Repair Roll License Amendment Request, initial submittal based on BAW-2303P, Revision 2]
- 5. Duke Power Company to NRC letter dated, November 3,1997, " Request for Technical Specification Amendment, Steam Generator Tubing Surveillance" [ Repair Roll License Amendment Request, revisions and answers to NRC request for additional information, BAW-2303P, Revision 3]
- 6. First Energy to NRC letter dated, February 26,1998, " Proposed Modification to the l
Davis-Besse Nuclear Power Station Operating License NPF-3, Appendix A Technical-Specifications to Incorporate a New Repair Roll Process for Steam Generator Tubes with Defects Within the Upper Tube Sheet" l
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I 7. First Energy to NRC letter dated, March 13, 1998, Non-Proprietary Version of BAW-2303, OTSG " Repair Roll Qualification Report" l
Evaluation and Issuance of Amendment Re: Steam Generator Tube Intergranular Attack 4
- Degradation (TAC No. M98262)" [ License Amendment No.158 to Facility Operating License DPR-72] '
Issuance of Amendment Re: (TAC No. 76640)" [ License Amendment 136 to Facility Operating License DPR-72]
- 10. . Entergy to NRC letter dated February 9,1998, " Technical Specification Change Request j Allowing the Use of OTSG Repair Roll Technology"
- 11. NRC to Entergy letter dated April 10,1998, " Issuance of Amendment No.190 to t Facility Operating License No. DPR-51 Arkansas Nuclear One, Unit No.1 (TAC No.
MA0827)"
- 12. NRC to FPC letter, 3N0798-28, dated July 30,1998, " Crystal River Unit 3 - Issuance of Amendment Regarding Steam Generator Tube End Anomalies - (TAC NO. MA2123)"
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