View source for 3F0199-04, LAR 243,Rev 0 to License DPR-72,requesting one-time Rev of ITS to Extend Insp Interval for OTSG Tubes to Coincide with Planned Operating Cycle.Rev 0 to Calculation M-99-0017 & Rev 0 to Calculation AES-C-3543-1 Encl

{{Adams
| number = ML20202E645
| issue date = 01/27/1999
| title = LAR 243,Rev 0 to License DPR-72,requesting one-time Rev of ITS to Extend Insp Interval for OTSG Tubes to Coincide with Planned Operating Cycle.Rev 0 to Calculation M-99-0017 & Rev 0 to Calculation AES-C-3543-1 Encl
| author name = Holden J
| author affiliation = FLORIDA POWER CORP.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000302
| license number = 
| contact person = 
| document report number = 3F0199-04, 3F199-4, NUDOCS 9902030111
| package number = ML20202E648
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 14
}}

=Text=
{{#Wiki_filter:--                                                 -
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Flrrida Power a
m taw W. NEE. mown January 27,1999 3F0199-04 i
U.S. Nuclear Regulatory Commission Attn: . Document Control Desk                                                                                        !
Washington, DC. 20555-0001 l

==Subject:==
License Amendment Request #243, Revision 0 Once Through Steam Generator Tube Surveillance Program, Inspection Interval Extension

==Dear Sir:==
l l Florida Power Corporation (FPC) hereby submits License Amendment Request (LAR)
         #243, to revise the Crystal River Unit 3 (CR-3) Improved Technical Specifications (ITS) in l
accordance with 10 CFR 50.90. This LAR requests a one-time revision of the ITS to                                    l extend the inspection interval for the Once Through Steam Generator (OTSG) tubes to coincide with the planned operating cfcle. The total interval between inspections being planned by FPC corresponds to less than 21.6 months of plant operation at a temperature of 500 F or above (measured at the hot ieg side).
CR-3 ITS 5.6.2.10 requires the OTSG inspection interval to be 24 calendar months for Category C-2 inspection results. However, the next OTSG inspection is planned to coincide with the October 1999 refueling and maintenance outage, Refueling Outage 11 (approximately 26 calendar months since the last inspection). The CR-3 operating time at a temperature of 500 F or above (measured at the hot leg side) from the last inspection, to.
the next planned inspection, will be less than 21.6 months. The conclusions reached in the operational assessments show leakage and structural integrity are maintained by substantial                       j margins until the end of the planned cycle.                                                                       I i
FPC is requesting NRC approval of this LAR by June 1,1999, with immediate implementation,                                                                                                  o\
CRYSTAL RIVER ENERGY COMPLEX: 15760 W. Power une Street
* Crystal River, Florida 34428-6708 + (352)795-6486
                                    ,         A Florida Progress Company 9902030111 990127 a PDR ADOCK 050003023 P                   PDR j

__                 -               . .         --.    -   ..,    ~.. .. _ ..    ..-
U.S. Nucley Regulatory Commission
        ' 3F0199-04
. . . Page 2 of 3 s.
If you have any qacstions regarding this submittal, please contact Mr. Sid Powe!8, Manager, .
  .j. Nuclear Licen.dng at (352) 563-4883.
Sincerely',
1 16 -
            . . Holden Director '                                                                                       .
Site Nuclear Operations                                                                          I l
JJH/lve xc:      Regional Administrator, Region II NRR Project Manager -                                                                   '
Senior Resident Inspector                                                               l l
J-                               I I
         -Attachments:                                                                                     !
1 A'.      License Amendment Request, No Significant Hazards Consideration Evaluation,    )
and Environmental Impact Evaluation                                            i B.       Proposed Technical Specification Change Page, Strikeout / Double Underline C.     : Proposed Technical Specification Change Page, Revision Bars
                  - D.      FPC Calculation M-99-0017, Deterministic Operational Assessment for Freespan ODSCC Degradation at Crystal River Unit 3 l

        .        .                                                 .     .   -   -~         -     ..       ~.      --
e-U.S. Nuclear R:gulatory Commission 3F01994)4 -
Page 3 of 3 l               STATE OF FLORIDA COUNTY OF CITRUS John J. Holden states that he is the Director, Site 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 st. .           nade and matters set forth therein are true and correct to the best of his knowledge, information, and belief.
3 Johp4. Hdiden Director Site Nuclear' Operations Sworn to and subscribed before me this MN day of               deat a M 6-            1999,by Jolm J. Holden.
                                                                   $ M b4 e //hYlsi Signature of Notary Public State of Florida _
P
                                                                                                      ~
(Print, type, or stamp Commissioned Name of Notary Public)
Produced Personally [
Known                 -OR-     Identification

i i
FLORIDA POWER CORPORATION i,
CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 1                                                ATTACHMENT A 4
LICENSE AMENDMENT REQUEST #243, REVISION 0 ONCE THROUGH STEAM GENERATOR TUBE SURVEILLANCE PROGRAM, INSPECTION INTERVAL EXTENSION 3
1 l                                License Amendment Request, No Significant Hazards Consideration Evaluation, and Environmental Impact Evaluation i
i

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U.S. Nuclear Regulatory Commission                                                 Attachment A 3F0199-04                                                                             Page 1 of 9 ATTACHMENT A LICENSE AMENDMENT REQUEST #243, REVISION 0 ONCE THROUGH STEAM GENERATOR TUBE SURVEILLANCE PROGRAM, INSPECTION INTERVAL EXTENSION LICENSE DOCUMENT INVOLVED: Improved Technical Specifications (ITS)
PORTION:           ITS 5.6.2.10, " Steam Generator (OTSG) Tube Surveillance Program"

==SUMMARY==
OF CHANGES:
The steam generator program is being revised to propose a one-time change to the inspection interval for the Once Through Steam Generator (OTSG) tube inspections. The inspection interval defined in ITS 5.6.2.10 is based on calendar months between inspections. Florida Power Corporation (FPC) is proposing a one-time change to Crystal River Unit 3 (CR-3) ITS to coincide with the completion of the planned operating cycle.
BACKGROUND:
During Refueling Outage 10 (spring 1996), FPC performed an inservice inspection of the CR-3 OTSGs.         After restart from that outage, CR-3 came off-line on September 2,1996, to troubleshoot and repair decreasing lube oil pressure to the main turbine. The unit remained shutdown for an extended outage until February 5,1998, to complete resolution of design margin issues. On August 1,1997, while the unit was shutdown, FPC completed a baseline, 100% bobbin coil inspection of the CR-3 OTSGs. CR-3 also paformed rotating pancake coil
         . and plus-point coil inspections of the upper tubesheet roll joints. The next tube inspection is planned to coincide with the October 1999 refueling and maintenance outage (Refueling Outage 11R). The unit operating time at a temperature of 500 F or above (measured at the hot leg side), from the 1997 baseline inspection to the next planned inspection, will be less than 21.6 months. However, the time between the last inspection and Refueling Outage 11R will be approximately 26 months.
FPC submitted a combined Mode 4 and 90-day Special Report (Reference 1) outlining the tube inspections performed and the results obtained during the 1997 inspections. Included in this submittal was a detailed assessment of the indicated growth trends for the "B" OTSG first span pit-like intergranular attack (IGA) indications.
FPC has performed several operational assessments. The first operational assessment (Reference 5) was performed to determine the adequacy of the structural integrity of the steam

  '                                                                                                       1 U.S. Nuclear Regulatory Commission                                                Attachment A 3F0199-04                                                                            Page 2 of 9 l generators for the next operational cycle. The operational assessment addressed the following degradation mechanisms:
l DEGRADATION MECHANISM                       INDICATIONS IDENTIFIED IN 1997              i First span pit-like IGA                    Predominant form in "B" OTSG Free span axial cracking                       One TUBE in "A" OTSG Upper tubesheet roll transition cracking            51 indications in each OTSG           j The assessment is based on eddy current data from previous inspections as well as that           l obtained during the 1997 extended outage. The operational assessment is forward-looking and l
demonstrates, with reasonable assurance, that the tube integrity performance criteria will be maintained throughout the operating cycle until the next scheduled tube inspection. The          ,
assessment also provides a baseline for performance of condition monitoring during the next      l OTSG tube inspection.
Reference 6 provided an updated operational assessment to support License Amendment              1 Request (LAR) #228, Revision 1. This LAR requested approval for operation with tube end anomalies (TEAS) until Refueling Outage 11R.
Reference 7 granted License Amendment No.169 regarding steam generator TEAS.
Amendment No.169 allows operation until the next outage of sufficient duration with a number of indications previously identified as TEAS in the CR-3 OTSGs.
Supplemental information to the CR-3 operational assessment is provided within this submittal    l for first span pit-like IGA, TEAS and for free span axial indications (outside diameter stress corrosion cracking (ODSCC)/lGA).
DESCRIPTION OF CHANGE:
ITS 5.6.2.10, Steam Generator (OTSG) Tube Surveillance Program Add a one-time change to paragraph 5.6.2.10.3 to read as follows:
             "The above-required inservice inspections of OTSG tubes shall be performed at the following frequencies except, a one-time change for Cycle 11 is granted to modify the scheduled inspection frequency from a calendar-based interval to an interval of up to 21.6 months of operating time at a temperature of 500 F or above (measured at the hot leg side). This will allow the OTSG tube inspection to coincide with Refuel Outage llR:"
l        REASON FOR REQUEST:
l t
l        The last inspection of the OTSGs was completed on August 1,1997. Due to the recent extended shutdown, Cycle 11 will extend beyond the 24 calendar month OTSG inspection interval specified in ITS 5.6.2.10.

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U.S. Nuclear Regulatory Commission                                               Attachment A            i 3F0199-04                                                                             Page 3 of 9 JUSTIFICATION FOR REQUEST:
1 l
This request is consistent with the guidance of Section C.1.1.2 of Draft Regulatory Guide (DG)-1074 (Reference 10) regarding frequency of inservice inspections. Similar proposals for extension have been granted by the NRC (References 8 and 13). FPC has performed operational assessments of the various degradation mechanisms to justify an operational cycle of 21.6 months. The evaluation below demonstrates with reasonable assurance that OTSG tube                       I integrity will be maintained over the planned operating cycle.
                      .                                                                                              l Evaluation
           ~ Document DG-1074, " Steam Generator Tube Integrity," September 1997 (Reference 9), and                  l NRC Generic Letter (GL) 95-05, " Voltage-Based Repair Criteria for Westinghouse Steam                    l Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," were used as references during the development of the CR-3 operational assessment (Reference 5). This L
operational assessment addresses first span pit-like IGA, free span axial cracking, and upper tubesheet roll transition cracking. It also demonstrates with reasonable assurance that the OTSGs will satisfy structural and leakage integrity requirements until the end of the planned            ,
operating cycle.                                                                                         '
i Attachment B of Reference 5 provided a summary of the operational assessment performed for CR-3 based upon the 1997 eddy current inspection of the OTSGs. A period of 21.6 months at a temperature of 500 F or above was used as the length of the current cycle for the purpose of the operational assessment. The assessment indicates that the projected end of cycle degraded l'          tube conditions for a 21.6 month operating period exhibit very substantial margins relative to both structural and leakage integrity requirements.
Subsequent to submittal of the operational assessment, CR-3 submitted an exigent License Amendment Request (Reference 6) addressing the presence of tube end anomaly indications.
An evaluation of these indications is also included in the following discussion.
First Span Pit-Like IGA Indications, Deterministic Approach During several outages since 1990, FPC collected a considerable amount of site specific data for the "B" OTSG first span pit-like IGA indications based on the laboratory examination of pulled tubes from the "B" OTSG. Since 1990, field eddy current indications have been observed, monitored for growth, and compared to the laboratory eddy current indications and
_ pulled tube data (References 2,3 and 4).
A deterministic assessment was performed for those "B" OTSG first span tubes with volumetric pit-like IGA which were returned to service after the 1997 inspection outage. The assessment projected the largest indication at the end of this cycle would be less than 100%
throughwall. The volumetric pit-like IGA indications in these tubes were evaluated according L
I

U.S. Nuclear Regulatory Commission                                                 Attachment A 3F0199-04                                                                             Page 4 of 9 to the deterministic structural performance criteria of DG-1074 (Reference 10), and assumed a safety factor (SF) of 1.4 for accident conditions and a SF of 3.0 for normal operations. As part of the assessment, the pit-like IGA was analyzed as if each pit were a throughwall crack having a maximum crack length equal to the diameter of the largest pit (0.3 inch diameter).
1 The largest pits returned to service were determined to be capable of withstanding maximum structural tensile loads experienced during main steam line break (MSLB) accident conditions.
Leakage is projected to be zero gallons per minute (gpm) for both normal and accident conditions. In addition to this analytical determination, in-situ pressure testing performed in 1996 (Reference 2), and the burst testing performed on tubes pulled from the "B" OTSG in 1992 and 1994 (References 3 and 4) continue to bound the indications left inservice.
i Based on the deterministic assessment, tube structural and leakage integrity requirements will continue to be satisfied at the end of the planned operating cycle. Therefore, there is no risk that OTSG first span pit-like IGA indications will result in a tube rupture, inservice leakage during normal conditions, or post-accident leakage (References 1 through 5).
Upper Tubesheet Roll Transition Indications Upper tubesheet roll transition indications were first detected at CR-3 during the 1997 inspection and were plugged upon detection. The operational assessment (Reference 5) relies upon data collected from other plants. The projected leak rate at accident conditions is 0.001 gpm. This is below the allowable limit of 1 gpm at accident conditions.
Because of the restraint the tube sheet imposes upon the tube end, there is no risk of a tube rupture as a result of primary water stress corrosion cracking (PWSCC) within the upper             '
tubesheet roll transition.
Free Span Axial Indications (ODSCC/ IGA), Deterministic Approach During the 1997 CR-3 outage, two free span axial IGA indications were detected in the "A" OTSG tube 4-39 using a bobbin probe. This tube was plugged. Babcock & Wilcox (B&W) plants have performed laboratory examinations on more than 12 sections of tubing containing free span axial IGA indications. These data include leak and burst testing of the sections with no indication of leakage and margin to burst at three times normal operating differential pressure (3 AP). Also, B&W plants have recently performed over 20 in-situ pressure tests on OTSG tubes containing free span axial indications. This includes data from two tubes containing free span axial IGA indications which were left inservice for approximately 32 effective full power months (EFPM) due to a tube plugging error. None of the OTSG tubes             l leaked when tested at 3 AP.
L 1
To address this degradation mechanism, an operational assessment has been performed (Reference 5). This operational assessment followed the probabilistic performance criteria of       l DG-1074, September 1997 (Reference 9). To supplement that assessment, a deterministic

r      ,.
U.S. Nuclear Regulatory Commission                                                  Attachment A 3F0199-04                                                                              Page 5 of 9      i evaluation (Attachment D) has been performed in accordance with the deterministic performance criteria of Section C.4.1.1 of DG-1074, March 1998 (Reference 10).
The basic calculational technique for the deterministic evaluation of free span ODSCC/ IGA followed industry accepted methods for calculating tube integrity using a set of bounding conditions on material properties, flaw size (length and depth), growth, and detection. The i         differential pressure safety factors of 1.4 for accident conditions and 3.0 for normal operations l-        outlined in the DG-1074 (Reference 10) structural performance criteria guidelines will be l         satisfied throughout the CR-3 operational cycle.
l         The structural limit for axial ODSCC/ IGA was determined as the maximum allowable flaw l         depth satisfying the above safety factors, a 95%/95% minimum flow stress for the tube l         material, and a 95% upper limit on axial flaw length. These statistical criteria are consistent with DG-1074 (Reference 10).
The limiting flaw depth at end-of-cycle (EOC) was established by statistical analysis using Monte Carlo simulation models to project the progression of axial degradation at CR-3.
Projected flaw numbers, depths, and growth of axial cracks allowed computation of the limiting flaw size at EOC. The limiting flaw depth was established from extreme value distribution of the largest flaw depths observed in the simulation. Again, a 95%/95% upper tolerance limit on the largest depth distribution was used tc determine the limiting flaw depth of 54.9% at EOC.
The deterministic analysis gave a structural limit for free span ODSCC/lGA as 61.1%
throughwall. This limit was established for 3 AP conditions. Therefore, the limiting flaw size at the EOC does not exceed the structural limit for axial degradation and the structural performance criteria of DG-1074 will be satisfied for an operating cycle length of up to 21.6 months of l         operating time at a temperature of 500 F or above (measured at the hot leg side).
l         Projected leak rates at postulated steam line break (SLB) conditions due to free span ODSCC/ IGA are expected to be minimal and will not challenge radiological dose limits. This is based on the 95%/95% flaw size at EOC and upper bound flaw shape considerations indicating that throughwall penetration is not likely during the planned operational cycle. Therefore, no leakage from free span ODSCC/ IGA should occur during this cycle.
Tube End Anomalies and Multiple Tube End Anomalies, Deterministic Approach CR-3 also has single tube end anomalies (SEAS) and multiple tube end anomalies (MEAS) in the l         upper roll portion of the OTSGs. These indications are in the upper area of the tube roll joint in the cladding region of the tubesheet. Becaus,e of the restraints imposed by the tubesheet, there is no potential of burst and the indications are structurally acceptable. The cumulative EOC accident induced leakage from the SEAS and MEAS is 0.010 gpm in the "A" OTSG and 0.030 gpm in the "B" OTSG (Reference 11).

U.S. Nuclear Regulatory Commission                                                 Attachment A 3F0199-04                                                                             Page 6 of 9 Wear Indications, Deterministic . Approach Comparisons of eddy current data for wear indications were made between the data from previous outages and that from the 1997 outage. Little or no growth was observed from the evaluation. This supported industry experience which considers wear to be a relatively slow Ldegradation process. Therefore, wear was not addressed.in the operational assessment.
Because there is little or no growth for wear indications, there is no risk of a tube rupture or inservice leakage during normal or accident conditions due to wear.
Operational Assessment Results and Conclusion Operational leakage at CR-3 is limited to 150 gallons per day (gpd), in accordance with the
,            .CR-3 ITS. This operational leakage limit was implemented on a permanent basis by License Amendment No.158, and is consistent with the recommended value in GL 95-05.
i Based upon the operational assessment (Reference 5, Attachment B), reasonable assurance is provided that the CR-3 OTSG tube integrity performance criteria will be maintained until the            j end of the planned operating cycle of up to 21.6 months of operating time at a temperature of           '
500 F or above (measured at the hot leg side). The results determined the probability of burst at accident conditions is far below the DG-1074 acceptable limits of 0.025, and the projected
,             leakage rate is less than the allowable limit of I gpm. Accident condition primary-to-secondary         )
4 leakage is limited to 1 gpm, per the CR-3 ITS Bases and accident analyses. This value is also consistent with DG-1074, Section C.2.3.
The current CR-3 ITS 5.6.2.10.3 establishes the allowable inspection interval to be based on calendar months between inspections. The last inservice inspections were completed August 1, 1997. However, the CR-3 OTSGs were not returned to operation at a temperature of 500 F or '
above until January 26, 1998. When the steam generators are not operating, the CR-3 lay-up program maintains the secondary side of the steam generators in an environment to minimize corrosion. The lay-up program follows, and in some areas exceeds, the generally accepted industry practices as specified in EPRI Report TR-102134, Revision 4, "PWR Secondary Water Chemistry Guidelines," November 1996. During the time from August 1,1997, to January 26, 1998, water chemistry was maintained per EPRI guidelines. These guidelines were established to reduce equipment corrosion and enhance steam generator reliability. On that basis, degradation during that period is expected to be insignificant..
            ~ Operations at a temperature of 500 F or above from January 26, 1998, until the start of
            - Refueling, Outage 11R will result in an operational cycle of less than 21.6 months. As
,             demonstrated in the eperational assessment, the potential for burst and leakage during an
,           ? operational cycle of this length is considerably below the limits contained in DG-1074 and
            - within the CR-3 licensing basis.

U.S. Nuclear Regulatory Commission                                                  Attachment A 3F0199                                                                            Page 7 of 9 NO SIGNIFICANT HAZARDS CONSIDERATION:
Florida Power Corporation (FPC) has reviewed the requirements of 10 CFR 50.92 as they apply to the proposed License Amendment and considers the changes not to involve a significant hazards consideration. In support of this conclusion, the following analysis is provided:
: 1. Involve a significant increase in the probability or consequences of an accident previously evaluated?
The last Crystal River Unit 3 (CR-3) Once Through Steam Generator (OTSG) tube surveillance was completed in August 1997. Both standard and enhanced eddy current techniques were used to inspect 100% of the OTSG tubes. Operational assessments performed for CR-3 provide reasonable assurance that the OTSG performance criteria meet the leakage and structural requirements in Draft Regulatory Guide-1074. These performance criteria will be maintained until the end of the planned operating cycle.
These operational assessments demonstrate that operation is acceptable for an operating cycle length of up to 21.6 months of operating time at a temperature of 500 F or above (measured at the hot leg side).
The operational assessments concluded that the projected cumulative leakage for the limiting OTSG would be less than I gallon per minute (gpm) under the limiting accident conditions at the end of the planned operating cycle. Thus, the accident analysis assumptions bound the condition of the OTSGs, and structural and leakage integrity will be maintained for the proposed operating cycle. Therefore, the proposed one-time change does not involve a significant increase in the probability or consequences of any accident previously evaluated.
: 2. Create the possibility of a new or different kind of accident from previously evaluated accidents?
No new failure modes or accident scenarios are created by changing the inspection from a frequency based on calendar months, to a one-time interval based on up to 21.6 months of operating time at a temperature of 500*F or above (measured at the hot leg side). Plant systems and components will not be operated in a different manner as a result of this change. Thus, this change does not increase the risk of a plant trip or present a challenge to any other safety system. For all known degradation mechanisms in the CP-3 OTSGs, the most recent operational assessments bound the probability of tube buru and project primary-to-secondary leakage at accident conditions for the end of Operating Cycle 11 to be less than 1 gpm. Therefore, the proposed one-time change does not create the possibility of l              a new or different kind of accident from any previously evaluated.
l

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U.S. Nuclear Regulatory Commission                                                            Attachment A 3F0199-04                                                                                           Page 8 of 9
: 3.      Involve a sigmficant reduction in a margin ofsafety?
Improved Technical Specification (ITS) Bases 3.4.12 contains relevant information pertaining to the limitations on reactor coolant system (RCS) leakage. The ITS Bases l               discuss the 1 gpm primary-to-secondary leakage assumed for a main steam line break accident, as well as for a steam generator tube rupture accident. The evaluation provided by this license amcmiment request shows that tube structural integrity is maintained, thus the required structural margins specified in NRC Regulatory Guide 1.121 are satisfied.
The operational assessments performed show the maximum accident leakage, assuming all these indications leak, is less than 1 gpm. Therefore, all known OTSG tube degradation                    l mechanisms have been assessed, and the proposed one-time change does not involve a                       q significant reduction in the margin of safety.                                                          H l
1 ENVIRONMENTAL IMPACT EVALUATION:
10 CFR 51.22(c)(9) provides criteria for identification of licensing and regulatory actions eligible l
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.
FPC has reviewed this proposed LAR and concludes it meets the eligibility criteria for categorical exclusion set ferth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(c), no environmental impact statement or environmental assessment needs to be prepared in connection with this request.
J u

U.S. Nuclear Regulatory Commission                                             Attachment A 3F0199-04                                                                         Page 9 of 9 REFERENCFS:
: 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 Current          :
Outage (90-Day Report)."                                                                 !
: 2. FPC to NRC letter, 3F0496-04, dated April 8,1996, " Technical Specification Change Request No. 203, Revision 2, Notifications Required Prior to MODE 4 and Responses to Request for AdditionalInformation."
: 3. FPC to NRC letter, 3F0494-09, dated April 19,1994, " Refuel 9 Inspection Plan for        i Once Through Steam Generators."
: 4. FPC to NRC letter, 3F0595-07, dated May 31,1995, "Once Through Steam Generator Inspection, Technical Specification Twelve-Month Report."
: 5. FPC to NRC letter, 3F0598-08, dated May 18,1998, "An Operational Assessment of            !
Steam Generator Tube Degradation at Crystal River Unit 3."
: 6. FPC to NRC letter, 3F0698-33, dated June 30,1998, " Exigent License Amendment            3 Request #228, Revision 1, Once Through Steam Generator Tube Surveillance Program."        l
: 7. 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)."
: 8. NRC to Indian Point No. 3 letter, dated June 19,1996, " Issuance of Amendment far Indian Point Nuclear Generating Unit No. 3 (TAC No. M94980)."
l        9. Draft Regulatory Guide DG-1074, Steam Generator Tube Integrity September 1997.
: 10. Draft Regulatory Guide DG-1074, Steam Generator Tube Integrity, March 1998.
t
: 11. FPC Calculation No. M-99-0016, " Evaluation of CR-3 Tube End Anomaly Indications."
: 12. EPRI Report TR-102134, Revision 4, "PWR Secondary Water Chemistry Guidelines,"
November 1996.
: 13. NRC to Millstone letter, dated September 23,1998, " Issuance of Amendment - Millstone i            Nuclear Power Station. Unit No. 3 (TAC No. MA2483)."
l l
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l FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUM'BER DPR-72 l
ATTACHMENT B i
LICENSE AMENDMENT REQUEST #243, REVISION 0                          l ONCE TIIROUGH STEAM GENERATOR TUBE SURVEILLANCE PROGRAM, INSPECTION INTERVAL EXTENSION                            I i
i Proposed Technical Specification Change Page, Strikeout / Double Underline i
i I
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l t}}