3F0398-06, Forwards Response to GL 97-06, Degradation of SG Internal. Encl Contains B&Wog Integrated Response to Subject GL
| ML20217J044 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 03/30/1998 |
| From: | Cowan J FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 3F0398-06, 3F398-6, GL-97-06, GL-97-6, NUDOCS 9804060156 | |
| Download: ML20217J044 (25) | |
Text
[ Florida Power CORPORATION Yhuon.n 3 i
March 30,1998' -
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U.S. Nuclear Regulatory Commission Attn: Document Control Desk '
Washington, DC 20555-0001 i
Subject:
Generic Letter 97-06, " Degradation of Steam Generator Internals" I J
Reference:
NRC to FPC letter, 3N1297-32, dated December 30,1997 )
Dear Sir:
The purpose of this letter is to submit Florida Power Corporation's (FPC) response to Generic ,
Letter (GL) 97-06, " Degradation of Steam Generator Internals." !
Prix to issuance of GL 97-06, U.S. nuclear utilities, the Electric Power Research Institute and !
) the Nuclear Energy Institute (NEI) formed the Steam Generator-Internals Task Force in January 1997. The purpose of the task force was to develop a coordinated industry-wide.
response to the secondary side degradation issues identified in the then draft GL. The Babcock -
& Wilcox Owners Group (B&WOG) participated in the task force and the development of an action plan to assess the susceptibility of steam generator internals to degradation.
Attachment A is the Crystal River Unit 3 (CRA) response to GL 97-06 which was prepared by the B&WOG for the member utilities. The B&WOG integrated response to GL 97-06 provided in Attachment A includes the CR-3 plan, specific information which in part addresses the GL request.
Based upon the Steam Generator Internals Task For,- action plan, the B&WOG member utilities have directed Framatome Technologies Inc. t, evaluate the B&W Once Through Steam. Generator (OTSG) design versus the Electricite de France (EdF) owned, modified Westinghouse designed steam generators. The objective ' f this evaluation is to derme, or
' identify, any mode of degradation to which the secc.idary side of the OTSGs may be susceptible and to recommend appropriate inspection '.echniques to monitor that degradation.
This evaluation is scheduled to be completed by December 1998. Within 90 days of completion of the Framatome Technologies Inc. evaluation, FPC will submit an update to the response provided in Attachment A.
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l Based on results to date at B&WOG plants, the integrated response concludes that no near ,
term inspection of the internals is required, and that compliance with and conformance to the l current licensing bases for the B&WOG member utilities continues to be maintained. The future susceptibility of the OTSGs continues to be evaluated as part of the B&WOG process to develop a formal Steam Generator internals program. An overall project summary report is being prepared for the B&WOG member utilities. The report will be available for NRC inspection at CR-3.-
If you have any questions regarding this information, please contact Ms. Sherry Bernhoft, Manager, Nuclear Licensing (352) 563-4566.
l Sincerely, l
$0%
John Paul Cowan l
- Vice President I Nuclear Operations JPC/LS/LVC xc: ' Regional Administrator, Region II Senior Resident inspector NRR Project Manager Attachments:
A. B&WOG Response to GL 97-06 B. List of Regulatory Commitments
~ UrS. Nuclear Regulatory Commission 3F0398-06
. Page 3 of 3 STATE OF FLORIDA COUNTY OF CITRUS i
John Paul Cowan states that he is the Vice President, Nuclear Operations for Florida l 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.
$ SC - %
John Paul Cowan i Vice President I Nuclear Operations j i
Sworn to and subscribed before me thisM day of Geh ,1998, by !
' John Paul Cowan.
UL - Ul Signature of totary Public State of Florida' KATW R.TYLM W C0hWISS10N# CCm7M EXPSES May 11.1986 smomnutwne souwm.ec.
(Print, type, or stamp Commissioned Name of Notary Public)
Personally / Produced Known V -OR- Identification t
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 FPC LETTER NUMBER 3F0398-06 ATTACHMENT A RESPONSE TO GENERIC LETTER 97-06
" DEGRADATION OF STEAM GENERATOR INTERNALS"
U.S. Nuclear Regulatory Commission Attachment A 3F0398 06 Crystal River Unit 3 Response to Generic Letter 97-06
" Degradation of Steam Generator Internals"
Background:
In response to the issuance of a proposed Generic Letter (GL) on degradation of steam generator internals, NEI formed the Steam Generator Internals Task Force in January 1997.
The purpose of the task force was to develop a coordinated industy-wide response to the secondary side degradation issues identified in the proposed GL. Participation on the task force included EPRI, licensees, and representatives of the vendors and owners groups for each domestic steam generator design. The task force developed an action plan.
Each owners group initiated a program to assist its respective owners in assessing the susceptibility of tube damage and loss of decay heat removal capability due to secondary-side degradation. An integral component in this assessment was an appreciation of the applicability of the degradation found in the French units to domestic steam generators.
EPRI responded to this need and with the assistance of Electricite de France (EdF) developed the repon, GC-109558, " Steam Generator Internals Degradation Modes of Degradation Detected in EdF Units."
The EPRI repon provides evaluations of the causal factors involved in the modes of degradation experienced in the French units. The owners groups used this report to gain insights into the applicability of the French experience to their steam' generator designs and operating history. NEI transmitted this report to the NRC via an NEI letter, dated December 19, 1997. EPRI provided copies directly to Steam Generator Management Program Technical Advisory Group Representatives.
In addition to the review of the EdF degradation . casual factors, the susceptibility assessments included consideration of design factors; fabrication and manufacturing techniques; plant operating history, including chemistry; plant inspection experience; and related degradation, such as denting. As part of the inspection experience review, the owners group compiled an assessed collective visual, video and pertinent non-destmctive examination (NDE) inspection experience information to further enhance their evaluations regarding the susceptibility to internals degradation.
The NEI' task force met with the NRC in May 1997, to gain a better understanding of the safety concerns discussed in the generic letter. As a result of these efforts, the owners group developed preliminary safety and susceptibility assessments relative to the design and operating history of their fleet. These assessments provide reasonable assurance that degradation of internds has not compromised steam generator tube integrity, or decay heat removal capability.
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U.S. Nuclear Regulatoiy Commission Attachment A -
3F0398-06 Generic letter (GL) 9' 06, Degradation of Steam Generator' Internals was issued to (1) again' alert. addressees to the previously communicated findings of damage to steam generator internals, ne nely, tube support plates and tube bundle wrappers, at foreign PWR facilities; (2) alert ad Jressees to recent findings of damage to steam generator tube support plates at a U,S. PV R facility; (3) emphasize to addressees the importance of performing comprehensive exe lninations of steam generator internals to ensure steam generator tube structural integri'/ si maintained in accordance with the requirements of Appendix B to 10 CFR Part 50: r.ad (4) require all addressees to submit information that will enable the NRC staff to verty whether addressees' steam generator internals comply with and conform to -
the currer'. licensing bases for their respective facilities.
The industry through the focused efforts of the NEI task force has provided guidance and information necessary for licensees to adequately address the potential issues'regarding steam generator internals degradation. Site specific details regarding steam ^ generator internals degradation for Crystal River Unit 3, are provided in the B&WOG Generic Response included as part of this attachment.
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.S. Nuclear Regulatory Commission Attachment A 3F0398-06 ;
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B&WOG Response to GL 97-06 j i
INTRODUCTION:
This response provides the B&W Owners Group (B&WOG) member utilities' information relative to the information requested by the Generic Letter. The B&WOG includes the j following plants: 1 e Arkansas Nuclear One Nuclear Power Plant Unit 1 (Arkansas 1, ANO-1) e Crystal River Nuclear Power Plant Unit 3 (Crystal River 3, CR-3)
- Oconee Nuclear Power Plant Unit 1,2,3 (Oconee 1,2,3 ; ONS-1,2,3) e Three Mile Island Nuclear Power Plant Unit 1 (Three Mile Island 1, TMI-1)
- Davis Besse Nuclear Power Plant Unit 1 (Davis Besse 1, DB-1)
RESPONSE TO REOUESTED INFORMATION:
The information requested by the GL is in italics; the response in normal font.
(1) Discussion of any program in place to detect degradation of steam generator l internals and a description of inspection plans, including the inspection scope, 1 frequency, methods, and equipment. l The B&WOG plants as a whole currently have no formal program to inspect / monitor steam generator internals degradation. However, as a result of other SG activities, such as sludge lancing and chemical cleaning, a significant number of secondary side inspections have been conducted at each of -ie B&WOG membo plants. Visual inspections have been performed from the lower tubesheet to the upper tubesheet and all tube support plates over a range of Effective Full Power Year (EFPY) from pre-service to 17.1 EFPY (the oldest B&W designed Once through Steam Generator, OTSG).
(a) Whether inspection records at thefacility have been reviewedfor indications of tube support plate signal anomaliesfrom eddy-current testing of the steam generator tubes that may be indicative ofsupport plate damage or ligament cracking. If the addressee has performed such a review, include a discussion ofthefindings:
No. There is currently no eddy current technique (ECT) used to identify degrada4n in B&W broached hole tube support plates (TSPs). Currently available eddy current techniques are adegrate to detect the presence of tube support plates. The many recent 100% bobbin examinations at the B&WOG plants have resulted in no instances of missing TSP indications. Therefore, it is concluded that the TSPs are properly located in the bundle. The current Page 3 of19
U.S. Nuclear Regulatory Commission Attachment A 3F0398-06 ECT guidelines require the analyst to verify the correct number of tube support plates is present during bobbin exam analysis.
(b) Whether visual or video camera inspections on the secondary side of the steam generators have been performed at the facility to gain information on the condition of steam generator internals (e.g., support plates, tube bundle wrappers, or other components). If the addressee has performed such inspections, include a discussion of thefindings:
Table 1 cor.trans a summary of the secondary side internal inspections performed at each (if the B&WOG member plants. This table presents the steam generator, EFPY, date, location, purpose, and results for each internals inspection performed. This table shows that a significant number of secondary side inspections have been conducted at each of the B&WOG member plants. These inspections span from pre-service to 17 EFPY, and include all 15 tube support plates, the upper and lower tubesheets, and one recent inspection of the upper wrapper welds. These inspections are typically performed in conjunction with cleaning processes or tube repair operations.
As part of the secondary side cleaning processes, pre-cleaning and post-cleaning visual inspections were performed on the secondary side of the OTSGs. These cleaning processes include sludge lancing, chemical cleaning and water slap. The post-cleaning visuals provided the clearest view of the
' tube support plates inspected, typically the 3"' through 6*, 9*, and 10*
support plates. These inspections, which were performed to ensure that the process os effective in removing the deposits, showed.no signs of any structural damsge to the tube support plates.
In recent years, fiberscopic inspecticns of the secondary side have been conducted following tube pull operzions. These inspections utilize the open path'left by the pulled tube to viscahy inspect the condition of the' support plates. Because the tubes are typically pulled from the lower tubesheet, most of the inspections encompassed the lower tubesheet through either the seventh, eighth, or ninth tube support plates. In one case the tube hole was inspected from the upper tube end to the 11* tube support plate as well.
None of these inspections have found any tube support plate structural damage.
In 1997, the oldest OTSG (Oconee 1) was visually inspected to determine the condition of the upper wrapper welds. For this inspection, two of the main feedwater nozzles were removed to examine the upper wrapper welds from below. No damage was found as a result of this inspection.
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. U.S.__ Nuclear Regulatory Commission . Attachment A 3F0398-06 In summary, visual inspections have been performed on the internals of all
' the B&WOG member utilities' OTSGs. These inspections include numerous
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inspections of the tube support plates and one recent inspection of the upper wrapper welds. These inspections have found no structural damage to these internal components.
( (c) _Whether degradation of steam generator internals has been detected 'at the
_ facility, and how the degradation was assessed and dispositioned.
With the exception of the internal auxiliary feedwater (AFW) header, no degradation of tube support plates, internal support structures, or other internal components that may affect tube integrity has been detected at any B&W designed plant.
There are two configurations of auxiliary feedwater header assemblies which are used on the steam generators of B&W designed 177FA plants. The_ first type uses an external distribution header moun ed outside the OTSG_ with nozzles penetrating the shell and shroud. The second type uses an internal distribution header mounted inside the OTSG. Only two of the operating i B&WOG member plants used internal AFW headers, Oconee 3 and Davis Besse 1.
In 1981 and 1982, tube leaks were experienced by Davis Besse 1 and Oconee
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3, respectively. As a result of eddy current and . visual examinations,'it was.
determined that the internal headers and the brackets which attached them to the wrapper were damaged. This degradation resulted in movement of the internal header during plant operation, damaging some peripheral tubes. The AFW internal headers were subsequently stabilized and functionally replaced by external headers at these plants. No movement or new indications of tube degradation have been noted at either plant since the internal AFW supply headers were stabilized.
Eddy current examinations of peripheral tubes in the DB-1 steam generators are performed each outage to ensure tube integrity. The internal header is visually inspected per Technical Specifications every ten years at DB-1.
Oconee 3 inspected the internal AFW headers as part of a commitment to the NRC in 1982. In addition, visual examinations were performed during the two subsequent refueling outages and then during the second 10-year ISI-outage. Oconee 3 has conducted 100% eddy current bobbin coil inspections since 1995, and has not detected any new degradation of the tubes in the periphery of the steam generators in the areas that are susceptible to damage Page 5 of 19 l
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U.S. Nuclear Regulatory Commission Attachment A 3F0398-06 from movement of the AFW supply header. Another visual inspection is planned for the third 10-year ISI outage.
(2) If the addressee currently has no program in place to detect degradation of steam generator internals, include a discussion andjustification of the plans and schedule for establishing such a program, or why no program is needed.
Prior to issuance of GL 97-06, U.S. nuclear utilities, the Electric Power Research Institute (EPRI) and the Nuclear Energy Institute (NEI) developed an action plan to assess the susceptibility of SG internals to secondary side degradation. Based upon this action plan, the B&WOG member utilities have developed a process by which a formal ~ program to detect the degradation of steam generator internals will _ be
' developed. This process was started in 1997 and is scheduled for completion by
' Decemher 1998.
The major tasks that comprise the B&WOG process are ' discussed below.
Preliminary results of the work completed are also presented. All available data fram NDE inspections, tube pull evaluations, and secondary-side visual inspections supports compliance with and conformance to the current licensing basis for the B&WOG member utilities.
l t 1.0 Owners Grous Deeradation Experience The purpose of this task is to identify any internals degradation detected at operating OTSG plants. This task focuses on internal components that may have an effect on tube integrity, and includes the review of relevant visual and ECT data.
Preliminary review of secondary side visual inspection data and available ECT data has shown no. generic internals degradation in the B&W plants (Table 1). In_ fact, the only internals degradation found is related 'to the internal AFW headers. As noted in the response to question 1, Part c, only Davis Besse Unit 1 and Oconee. Unit 3 had OTSGs of this design, and these internal AFW headers were stabilized and functionally replaced with external feedwater headers in the'early 1980s. No further tube damage associated with the internal AFW header has been detected during subsequent tube inspections at these two plants. ,
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'U.S. Nuclear Regulatory Commission Attachment A 3F0398-06' L . .
2.0 Owners Group Derradation Assessment This pan of the process is broken down into subtasks and described below:
. Evaluate Susceptibility of OTSG Relative to Electricite de France G4F)
Experience l The B&WOG member utilities are in the process of assembling and summarizing design documentation relative to steam generator secondary side components for all operating B&W plants. Existing analyses that relate to possible degradation mechanisms are also being collected and summarized.
This information is being used to determine the susceptibility of the OTSG to secondary side degradation relative to the experience of EdF plants.
EPRI released GC-109558, " Steam Generator Internals Degradation: Modes of Degradation Detected in EdF Units," in December 1997. According to this document, several types of degradation were found in the EdF SG internals. These degradation modes, and the preliminary results of the evaluations to determine possible applicability to the OTSG are presented below.
- 1. Flow assisted corrosion (FAC) of the top tube support plate (TSP) at Fessenheim 2 was noted during routine eddy current testing (ECT) in 1995. This corrosion was determined to be the result of improper placement of hoses during a chemical cleaning performed in 1992. The corrosion is not progressing, and inspections indicate that no other French units have experienced similar attack.
Four of the seven B&WOG plants have chemically cleaned their steam generators. The EPRI Steam Generator Owners Group (SGOG) chemical cleaning solvent used in these cleanings is different than the solvent utilized by EdF. Extensive qualification testing of the EPRI SGOG solvent was conducted prior to its first application, and continues to be tested for most applications. FAC was studied as part of the qualification testing. The flow rates during steam generator cleanings are designed to remain below the critical FAC rates. Post-chemical cleaning visual inspections have shown no TSP degradation at any of the four plants.
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- 2. In response to the Fessenheim 2 experience, all other cperating units in France with drilled hole carbon steel TSPs, i.e., all units with Model L 51A and 51M steam generators, were inspected for TSP damage. Both
! ECT and television (TV) visual inspections were performed and TSP Page 7 of19
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U.S.~ Nuclear Regulatory Commission Attachment A -
3F0398 .
ligament cracking was identified. Review of inspection data has shown that the ligament cracking was present at the first inspection for which interpretable data is available', i.e., at either pre-service or early in-service inspections. The cracking has not increased with time. It is believed that the cracking is the result of mechanical loads applied to the TSPs during manufacture, shipping or early operation.
At present, no eddy current technique -is used to identify TSP degradation in B&W broached hole TSPs. Currently, used eddy current techniques indicate that the tube support plates are properly located and show no indications of gross degradation. No TSP degradation _has been observed in any of the numerous visual inspections performed over the service lives of the OTSGs ( Table 1).
- 3. FAC induced thinning of the top TSP has been detected by ECT and TV j in three units (Gravelines 2, 3, and 4) that operated from startup in the early 1980s until recently using ammonia water chemistry. The FAC occurred at the periphery of the TSP at the location of the largest radius U-bends. It occurred mostly on the hot leg side, but some was also observed on the cold leg side. The French concluded that the FAC is associated with the lower pH of ammonia water chemistry.
M1 of the OTSG plants have operated on All Volatile Treatment I (AVT) feedwater chemistry. Early in plant life, the chemistry l consisted of condensate polished feedwater treated with hydrazine and I ammonia for oxygen and pH control, respectively. Later, the pH I control additive was switched to alternate amines, such as j morpholine. The objective was to reduce FAC in the balance of plant ' j system piping where such problems had been observed. Based on NDE inspections, tube pull evaluations, and the numerous secondary side inspections listed in Table 1, FAC has not been identified as a damage mechanism in the OTSGs at this time. The primary contributors to FAC - flow' velocity, quality, and solution pH - will be evaluated in more depth as part of the B&WOG process.
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- 4. Wrapper drop was detected in 1994 at two steam generators in Blayais 3. !
The drop occurred because of the failure of the wrapper supports located i at the first TSP elevation. Two scenarios for the wrapper drop had been j
proposed by the French: (1) failure of the supports mainly due to !
thermal expansion loads developed during transients associated with the j use of cold auxiliary feedwater to expedite plant cool-down, aggravated by poor quality of the welds joining the support blocks to the wrapper and possibly by fatigue induced crack propagation, and (2) failure of the i
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i U.S. Nuclear Regulatory Commission Attachment A 3F0398-06
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supports mainly due to fatigue cracking of the support to wrapper valds, aggravated by poor quality of the welds and by axial load development by thermal transients.
l The B&W designed OTSGs have a different wrapper design than the EdF plants with the noted degradation. The wrapper consists of two shells, one upper and one lower, separated by a small gap. The lower wrapper of the OTSGs is supported by the lower tubesheet. The upper wrapper is supported by an annular ring which is welded to the !
shell. I During 1997, the oldest operating B&W designed plant (Oconee,1) conducted a visual inspection of the upper wrapper assembly, welds, and internal components in the vicinity of the welds. During this inspection, no signs of movement or slippage of the wrapper were noted. There was no evidence of damage to the welds or components in the vicinity of the wrapper assembly which would indicate that the :
L wrapper had shifted or dropped. Therefore, wrapper drop is not considered a significant near-term problem for the OTSGs. However, s during the continuing evaluation of SG internals, this potential degradation mechanism will be investigated further.
- 5. Fatigue cracks emanating from support blocks have been detected in the same two Blayais 3 steam generators that experienced wrapper drop, and j in one steam generator of similar design at Blayais 2. The fatigue cracks l appear .to be the results of flow induced vibration of the wrapper. The French indicate that they.are evaluating the possible. occurrence of this j mechanism- at other units and designs, since the root cause of this j cracking has not been demonstrated to be Blayais 2 and 3 specific. l l As noted above, mspection of the upper wrapper supports at Oconee 1 j did not identify signs of degradation. This potential damage I mechanism continues to be evaluated for possible future impact on the OTSGs. .
- 6. Some cases of TSP wedge block cracking have been observed. The.
causes of the wedge block cracking are believed to be the same as those i causing TSP ligament cracking. Some tie rod lock welds have been l found to be cracked. These cracks are considered to have no safety l significance.
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1 U.S. Nucl:ar Regulatory Commission Attachment A 3F0398-06 No degradation has been noted as a result of visual inspections to date. However, during the evaluation of SG internals degradation, this degradation mechanism will be investigated further.
Assess Susceptibility to Other PotentialDamane Mechanisms i
Preliminary information indicates that recent internals damage found at the San Onofre Nuclear Generating Station'(SONGS) has been attributed to FAC associated with fouling.
The steam generators at SONGS are Combustion Engineering recirculating steam generators which have few design and performance similarities to an OTSG. The concentrating mechanisms, bulk water chemical concentrations, and pH control additives behavior are variables that are typically different in the two types of steam generators. The B&WOG member utilities are aware of the FAC damage at SONGS, and will evaluate it further as part of the B&WOG process. 1 The objective of this subtask is also to address damage mechanisms which have not yet been observed in operating plants, but which may be possible based on operating experience not considered during original plant design.
Structural, chemical, and thermal hydraulic performance experience will be reviewed to identify these potential forms of degradation. Any mechanisms which are identified as requiring further investigation will be identified and additional work will be conducted as appropriate.
3.0 Inspection Reautrements and Methodology Based on'the results of task 2, internal components identified as being susceptible to internals degradation which could affect tube integrity will be-identified. For each of these components, a recommended inspection scope and methodology to monitor for potential future secondary side degradation will be determined.
4.0- Industry Response The final deliverable for this project will be a document that will serve as the overall project summary report for the B&WOG member utilities. All areas of the OTSG identified as being potentially susceptible to degradation (if any) ;
will be discussed in the' report, along with recommended inspection i procedures and frequencies, and disposition criteria.
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TJ.S. Nuclear Regulatory Commission Attachment A
- 3F0398-06 SilMMARY:
= The B&W Owners Group has completed a preliminary review of EPRI GC-109558, " Steam
-Generator Internals Degradation: Modes of Degradation Detected in EdF Units," and the SONGS egg crate support degradation relative to the design and operation of Once Through Steam Generators. For each category of degradation, the B&WOG has concluded that the OTSGs are not significantly at risk for the same degradation in the near term. The future susceptibility of the OTSG to these or other forms of degradation continues to be evaluated as part of the B&WOG process to develop a formal SG internals program.
Table l provides a summary of secondary side visual inspections co'nducted at each B&W' designed plant. The B&WOG concludes that the number of plants that have been inspected and the visual inspection results demonstrate with a high degree of confidence that there is currently no significant degradation of SG internals in OTSGs. Currently, two of the B&WOG plants monitor-periphery tubes for damage from internal AFW headers. The other B&WOG plants do not have this design feature and thus are not susceptible to this
- type of SG internals degradation.
Based on results to date, it is concluded that' no.near term inspections of the internals are required, and that compliance with and conformance to the current licensing bases for the
'B&WOG member utilities has been -maintained. - When ' the program is complete, recommendations may be made for future periodic inspections if needed. These recommendations will be included in a report which will be distributed to all the B&WOG member utilities upon completion of this process.
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FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 FPC LETTER NUMBER 3F03984)6 ATTACHMENT B LIST OF REGULATORY COMMITMENTS j
4
, U.S., Nuclear Regulatory Commission Attachment B
- 3F0398-M - Page1of1 l l
I Attachment B l
List of Regulatory Commitments i
The following table identifies those actions committed to by Florida Power Corporation in this document. Any other actions discussed in the submittal represents intended or planned actions by Florida Power Corporation. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Manager, Nuclear Licensing of any l questions regarding this document or any associated regulatory commitments.
ID NUMBER - COMMITMENT COMMITMENT DATE OR OUTAGE 3F0398-06-1 Framatome Technologies Inc. has been directed 90 days after to evaluate OTSG designs versus EdF steam completion of j generators. This evaluation is scheduled to be evaluation by j complete by December 1998. Within 90 days of Framatome completion of the Framatome Technologies Inc. Technologies Inc.
evaluation, FPC will submit an update of the i response in Attachment A. I 1
3F0398-06-2 An overall project summary report is being 30 days after prepared for the B&WOG member utilities. The completion by.
report will be available for NRC inspection at Framatome CR-3. Technologies Inc. ,
i