3F0397-16, Application for Amend to License DPR-72 Re Once Through SG Tube Surveillance Program
| ML20137G812 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 03/27/1997 |
| From: | Richard Anderson FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20137G818 | List: |
| References | |
| 3F0397-16, 3F397-16, NUDOCS 9704010529 | |
| Download: ML20137G812 (20) | |
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\ @ l Florida Power CORPORATION na.:YE March 27, 1997 3F0397-16 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001
Subject:
Technical Specification Change Request Number 211, Revision 0
Reference:
A. FPC to NRC letter, 3F0494-09 dated April 19, 1994 B. FPC to NRC letter, 3F0595-07 dated May 31, 1995 C. FPC to NRC letter, 3F0396-19 dated March 21, 1996 D. NRC to FPC letter, 3N1096-11 dated October 15, 1996
Dear Sir:
Florida Power Corporation (FPC) hereby submits Technical Specification Change Request Notice (TSCRN) 211 requesting amendment to Operating License No. DPR-72. As part of this request, the TSCRN and the proposed new Technical Specification page changes are provided. Revised Bases pages are also provided. To assist in reviewing this request, a set of the changed pages with deleted information shown in strikeout font, and new or revised information shown in shadowed font are also provided. The TSCRN proposes to specify a group of tubes within the Crystal River Unit 3 (CR-3) "B" Once Through Steam Generator (OTSG) that exhibit similar eddy current indications characteristic of pit-like intergranular attack (IGA) te be subjected { to 100% inspection of their first span during each inservice inspection of the "B" 0TSG. No credit will be taken for this inspection in meeting minimum sample
,\
size requirements for random inspection, and degraded or defective tubes found , during this inspection will not be used in determining expanded random / 'j (q f) inspections. FPC will perform Motorized Rotating Pancake Coil (MRPC) inspection and bobbin coil inspection on all "B" 0TSG first span pit-like IGA indications. These inspections are to ensure that the damage mechanism is not changing morphology, and to ensure that the IGA indications do not mask the presence of a new damage mechanism in this region. CRYSTAL RIVER ENERGY COMPLEX 15700 W. Power une Street
- Crystal River, Florida 34428470s . (352) 7 5 6486 A norida Progress Comparty D D K O O2 h, h,Ih,h P PDR
i l t 4 , U. S. Nuclear Regulatory Commission 3F0397-16 Page 2 of 20 l FPC will continue to monitor and participate in industry and owners group efforts
- to establish a qualified sizing technique applicable to IGA indications. FPC l anticipates the adoption of a qualified sizing technique for IGA prior to our l next inspection of the "B" OTSG to complement this proposed Technical Specification change. In the event that a qualified technique is not available, the subject tubes will be repaired or plugged.
l In addition, FPC proposes to retain the lower primary-to-secondary Reactor
- Coolant System (RCS) leakage limit of 150 gallons per day that was approved with License Amendment No. 154. The previous rate of 1 gallon per minute was reduced by Amendment 154 to be in effect until the Refuel Outage 11 only. This reduced j leakage limit will provide additional assurance that if primary-to-secondary l leakage were to occur, the plant would be shutdown in a timely manner.
! Replacement pages are also provided to revert other changes made by Amendment 154 back to the requirements in effect prior to that amendment. FPC considers the proposed Technical Specification changes to be consistent with discussions between FPC and NRC personnel at our management meeting of August 21, 1996, and during informal discussions held on January 29, 1997. We intend to use these new inspection requirements during 0TSG inspections performed during the current forced outage. As such, we request approval of this TSCRN by June 30, i 1997. If questions arise during your review please contact Brian Gutherman, ! Manager, Nuclear Licensing, at 352-563-4566. i Sincerely I/ il A. Anderson nior Vice President uclear Operations 1 RAA/SCP Attachment xt: Regional Administrator, Region II Senior Resident inspector NRR Project Manager l l r l
1 4 U. S. Nuclear Regulatory Commission ; 3F0397-16 i Page 3 of 20 l 1 STATE OF FLORIDA j COUNTY OF PINELLAS Roy A. Anderson states that he is the Senior 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 _ l made and matters set forth therein are true and correct to the best of his knowledge, information, , and belief. l l 7 y A. Anderson enior Vice President uclear Operations 1 Sworn to and subscribed before me this 27 day of March,1997, by Roy A. Anderson, who is personally known to me. wE (LC Lou-Ann E. Macko Signature of Notary Public State of Florida Stamp Commissioned Name of Notary Pub k mNorMYK IDU-ANN E MACKO NOTARY PUBUCSrAlliOF RDRIDA COMMBSloN No.CC582044 MY COMMISSION EXP. SEPT 32000
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4-i U. S. Nuclear Regulatory Commission 3F0397-16 Page 4 of 20 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION i , IN THE MATTER )
) DOCKET NO. 50-302 FLORIDA POWER CORPORATION )
4 CERTIFICATE OF SERVICE Roy A. Anderson deposes and says that .the following has been served on the l Designated State Representative and Chief Executive of Citrus County, Florida, ; by deposit in the United States mail, addressed as follows:
- Chairman, Administrator, Board of County Commissioners Radiological Health Services ;
of Citrus County Department of Health and ; i Citrus County Courthouse Rehabilitative Services ! 1 Inverness, FL 34450 1323 Winewood Blvd. t Tallahassee, FL 32301 l A copy of Technical Specification Change Request No. 211, Revision 0.
- FLORIDA POWER CORPORATION j
~\ !
l ~ Ro A. Anderson S nior Vice President ! uclear Operations } Sworn to and subscribed before me this 7-} day of March, 1997, by Roy A. Anderson, who is personally known to me. ,
% . /16 1
bSgn~ak"u"e of*NSary Public State of Florida ulu^##^#ee NOTARY PUBLICSTATEOF FIDRIDA uvc , ne n er ntm Stamp Commissioned Name of Notary Public
< l U. S. Nuclear Regulatory Commission 3F0397-16 ;
l 'Page 5 of 20 FLORIDA POWER CORPORATION i- CRYSTAL RIVER UNIT 3 l DOCKET NO. 50-302/ LICENSE NO. DPR-72 , REQUEST NO. 211 REVISION O : ONCE THROUGH STEAM GENERATOR TUBE SURVEILLANCE PROGRAN , i i ! LICENSE DOCUNENT INVOLVED: Technical Specifications ! !- I t PORTIONS: Specification 3.4.12 "RCS Operational Leakage" ' ! Bases B 3.4.12 "RCS Operational Leakage" l Specification 5.6.2.10 "0TSG Tube Surveillance Program" l Specification 5.7.2 "Special Reports"
SUMMARY
OF CHANGES: . Limiting Condition for Operation (LCO) 3.4.12 (d) will be changed to make the 150 ; gallons per day (gpd) limit on primary-to-secondary leakage through either Once Through Steam Generator (OTSG) a permanent part of the 3.4.12 LCO. This was . changed with License Amendment 154 to be effective only until Refuel Outage 11. , This is now requested to be a permanent change. - The OTSG Tube Surveillance Program described in Technical Specification (TS) 5.6.2.10 will be changed as follows-l Revert changes made in License Amendment 154 back to the requirements in i , effect prior to Amendment 154. This affects the NOTE in Section 1 5.6.2.10.2; vocabulary terms in TS 5.6.2.10.4.a.2 and 4; and Reporting , Requirements in TS 5.7.2.c. Add new inspection criteria associated with the first span section of tubes in the "B" OTSG with pit-like intergranular attack (IGA) indications as new TS 5.6.2.10.2.e., and Revise the vocabulary term in 5.6.2.10.4.a.7 which describes pit-like ! intergranular attack. .
- 1. CHANGE TO SPECIFICATION 3.4.12, RCS Operational Leakage DescriDtion of SDecification ChanQes FPC proposes to retain the lower primary-to-secondary RCS leakage limit of 150 gallons per day that was approved with License Amendment 154. The previous rate of I gallon per minute was reduced by Amendment 154 to be in effect until Refuel l Outage 11 only. The only change required to accomplish this is to re-issue 1 1
Technical Specification page number 3.4-22 removing the Refuel 11 expiration in the page footer. DescriDtion of Bases Chances The Bases description in the LC0 section for primary-to-secondary leakage is i revised to remove reference to statistical analyses that were used to support the interim tube plugging criteria approved in Amendment 154. Editorial changes have
4 e U. S. Nuclear Regulatory Commission 3F0397-16 Page 6 of 20 been made to define previously undefined abbreviations, and page structure changes have been made to accommodate text rolled to the next page. Reason for Reouest This Technical Specification Change Reauest proposes to retain the primary-to-secondary leakaac limit of 150 gallons per day (gpd) currently approved only for use until Refuel Outage 11. The 150 gpd limit is consistent with the operational leakage limit specified in NRC Generic Letter 95-05 for plants implementing alternate repair criteria. Although FPC has elected not to pursue approval of alternate repair criteria for pit-like IGA indications in the first span of the "B" 0TSG at this time, it is recognized that a more conservative leakage limit is appropriate to ensure prompt shutdown following detection of primary-to-secondary leakage. Justification for Reauest FPC intends to pursue acceptance of alternate OTSG tub-e repair criteria that is now under development within the industry. FPC chooses to retain the more conservative leakage rate in place now for CR-3 in anticipatior of later adoption of approved alternate repair criteria.
- 2. CHANGE TO SPECIFICATION 5.6.2.10, OTSG Tube Surveillance Program 6
Description of Specification Chances ' A. Add new paragraph "e" to TS 5.6.2.10.2, page 5.0-14, after item "d. " This paragraph defines the inspection requirements and the disposition criteria for applicable tubes in the "B" OTSG first span. Paragraph "e" reads as follows:
- e. Inservice tubes with pit-like IGA indications in the first span of the "B" OTSG, identified in the OTSG Inservice Inspection Surveillance Procedure, must be inspected with bobbin and Motorized Rotating Pancake Coil (MRPC) eddy current techniques from the lower tube sheet secondary face to the bottom of the first tube support plate during each inservice inspection of the "B" OTSG. No credit is to be taken for this inspection in meeting minimum sample size requirements for the random inspection. 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, unless the degradation mechanism identified is a mechanism other than pit-like IGA. 1 i
The subject tubes are listed in Table 1 of this submittal. This table will be added to the CR-3 OTSI Inservice Inspection Surveillance Procedure. Changes to the table will be made in accordance with plant J procedures and the 10 CFR 50.59 change control process. J B. Revise a vocabulary term in TS 5.6.2.10.4.a.7, page 5.0-16. This change ) replaces the existing definition added by Amendment 154 with a new l 1 i l
U. S. Nuclear Regulatory Commission 3F0397-16 Page 7 of 20 vocabulary term for pit-like IGA indications as 5.6.2.10.4.a.7. The TS vocabulary term would read as follows: Pit-like Intergranular Attack (IGA) indication means a bobbin coil indication confirmed by Motorized Rotating Pancake Coil (MRPC) or other i qualified inspection techniques to have a volumetric, pit-like morphology , characteristic of IGA. C. Return several changes made in Amendment 154 to sections of 5.6.2.10 and 5.7.2 to their previous wording. These changes remove statements that ) were specific .to the interim tube plugging criteria applicable until Refuel 11, and remove the Refuel 11 expiration footers. The changes are made on pages 5.0-14 (footer), -14A (NOTE in Section 5.6.2.10.2), -16 (Vocabulary Terms 5.6.2.10.4.a.2 and a.4), -17 (footer), and -2E , (Reporting Requirements 5.7.2.c). Reason for Reaues_t TS 5.6.2.10 currently requires that a random inspection of a population of at least 3% of the tubes in each steam generator or at least 6% of the tubes in one
- steam generator, be inspected at intervals of no less than 12 months and no greater than 24 months. The purpose of the periodic inspection is to ensure the r
integrity of the reactor coolant pressure boundary through the early detection and subsequent monitoring of tube wall degradation. The requirement to select a random sample for inspection is intended to ensure that all areas of the steam generator are appropriately sampled during each inspection, with inspection 1 results providing an overall assessment of the general condition of the steam The results of completed random inspections are
; generator tube bundle.
i classified in accordance with the definitions for results categories C-1, C-2, . and C-3 contained in TS 5.6.2.10.2, and included here for ease of reference: Cateaory Insoection Results i C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective. C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 107, of the total tubes inspected are degraded tubes. C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective. The current requirement to expand the random sample population based on the , classification of inspection results is intended to increase the sample population proportionately with the extent and severity of degradation detected during inspection of the initial sample. During previous OTSG tube inspections extending as far back as 1980, FPC has observed pit-like IGA indications in the "B" 0TSG first span region. Continuing
l l U. S. Nuclear Regulatory Commission i 3F0397-16 ! Page 8 of 20 observation of these indications during recurring inspections has demonstrated I no apparent growth of the existing indications. Destructive examination of tube samples has determined the cause for these indications to be due to a specific ; degradation mechanism, pit-like IGA. Burst tests of CR-3 tube samples with IGA have shown that tube structural capability has not been degraded below acceptable levels due to the indications. Burst pressures substantially greater than the limit of Regulatory Guide (RG) 1.121 have been recorded for all samples. l As discussed at a meeting between FPC Senior Management and NRC Staff on August 21, 1996, see Reference D, FPC is proposing to re-inspect "B" 0TSG first span pit-like IGA indications during each inservice inspection of the "B" 0TSG. Results of the re-inspection will be dispositioned in accordance with the existing through-wall based acceptance criteria contained within TS 5.6.2.10.4, including classification of each tube as imperfect, degraded, or defective. FPC is also requesting to exclude the pit-like IGA indications in the first span from consideration in determining the overall inspection results category for the random inspection sample. The reason for this exclusion is to prevent bias of the random sample inspection and subsequent inspection results classification due to known, historical indications which were caused by an identified degradation mechanism and have been present in the "B" 0TSG as far back as 1980. Justification for Reauest Backaround The CR-3 once-through steam generators (OTSG) were subjected to a post-fabrication full vessel stress relief at 1100 - 1150 F for 10 - 20 hours which produced a sensitized microstructure in the tubing, that is, increased carbide decoration in the grain boundarles, but with an accompanying chromium depletion. The sensitized microstructure increases the resistance of the tubing to caustic stress corrosion cracking, but increases its susceptibility to intergranular attack (IGA) by sulfur oxyanions. Laboratory studies and field experience with this type of IGA have shown that the attack occurs at low temperatures (less than 170 F) typical of shutdowns, in acidic, oxidizing conditions, and in the presence of reduced sulfur ions. Under this combination of conditions, nickel is preferentially dissolved at the grain boundaries due to the lower chromium content present there, thus producing the intergranular attack. Tubes pulled from the "B" 0TSG in 1992 and 1994 were destructively examined in the laboratory. Detailed results of the pulled tube examinations have been previously reported in References A and B. Numerous, small patches of IGA were found on the tubes in the first freespan located primarily 6 - 21" above the lower tubesheet secondary face. (See Figure 1-1) A total of 168 patches were observed in eight (8) expanded tube sections. The IGA was confined to spots on ! the OD tube surface where the deposit thickness was thinner than the bulk i deposit. Surface analysis of some defects by energy dispersive spectrometry (EDS) I and x-ray photoelectron spectroscopy (XPS) detected significant levels of sulfur j present (2 - 5 wt%), predominately as sulfate, with slight nickel depletion at 1 the grain boundaries. These findings suggest that IGA by sulfur oxyanions in mildly alkaline to acidic conditions was the probable cause of the degradation. '
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( U. S. Nuclear Regulatory Commission 3F0397-16 i Page 9 of 20 ; 1
; A review of limited historical eddy current data available on the pulled tubes and other tubes in the first span of the "B" 0TSG where the-IGA had occurred, suggests that the eddy current indications in this region were present as far l i
back as 1980. The results of growth studies performed on "B" 0TSG first span ' indications, which have consistently concluded that there is no apparent growth ; of these indications occurring, have been previously reported in Reference C. I Scatter plots which provide a comparison of eddy current bobbin probe voltages i for "B" 0TSG first span indications inspected in both the 1992 and the 1994 eddy current inspections, and the 1994 and the 1996 eddy current inspections, are provided as Figures 1-2 and 1-3. Since bobbin voltage is considered to be a good indicator of eddy current indication volume, the minimal scatter of the data around the zero growth line of Figures 1-2 and 1-3 supports the previously submitted growth rate study conclusions that the pit-like IGA has exhibited no i apparent growth during multiple inspection intervals. The average change in voltage is less than 0.1 volts, well within the repeatable accuracy of bobbin coil examinations , Two tube pull campaigns have been performed at CR-3 to obtain samples of tubing with IGA. Five sections of tubing with a combined total of approximately 127 indications were subjected to burst testing to evaluate structural integrity. ) I All burst well above RG 1.121 limits. Burst data for one tube section was : particularly relevant to deQonstrating structural integrity of 0TSG tubing as it relates to the proposed amendment request. Tube section 68-46-3 contained a ; defect in the lower tube sheet region which exhibited an axial extent of 0.228 _l inch with 75% through wall penetration. Despite the relatively large size of this indication in comparison to all other IGA indications identified, the burst pressure for the tube section was 7000 psi, which is still substantially greater than the RG 1.121 limit of 4050 psi. j Evaluation The random inservice inspection required by Technical Specification 5.6.2.10 is designed to serve as one element in an overall strategy to minimize the probability of a steam generator tube rupture, and to ensure the integrity of the reactor coolant pressure boundary. This is accomplished through early detection of steam generator tube degradation, and subsequent repair of defective tubes. The other elements of the strategy consist of chemistry, operational, and maintenance considerations. All tubes in the random sample are required to receive full length examination with results of the sample classified into Technical Specification defined categories for purposes of determining the need for sample expansion. Exclusion of tubes with pit-like IGA in the "B" 0TSG first span from the random inspection j and inspoction results categorization will not reduce the effectiveness of the random sample as it currently exists. This will not increase the probability of a steam generator tube rupture, since the proposed change to TS 5.6.2.10 will require more extensive examination of the "B" 0TSG first span IGA indications than is currently required. i Removal of the first span portion of tubes with IGA from consideration for random sample inspection will not decrease the effectiveness of the inspection since these will now be inspected during each inservice inspection interval. Eddy current indications in this region have been present since as early as 1980, and
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U. S. Nuclear Regulatory Commission 3F0397-16 Page-10 of 20 have been tracked over multiple inspection intervals by eddy current bobbin inspection and by Motorized Rotating Pancake Coil (MRPC) sample examination. The extent and prevalence of damage by IGA in the "B" first span is considered to be substantially bounded. The first span region of tubes adjacent to those identified in the OTSG Inservice Inspection Surveillance Procedure, and the balance of tube length outside the
, first span for those tubes will remain in the population of tubes from which the 5
random sample population is selected. The random sampling of tubes adjacent to those included in the "B" 0TSG first span special inspection area will continue i to provide a trigger for eddy current sample expansion if the degradation mechanism has not been appropriately bounded or later proves to be active. t Furthermore, the proposed specification requires that if a "B" 0TSG first span indication identified on a tube listed in the- 0TSG Inservice Inspection Surveillance Procedure is found by MRPC to be representative of a degradation
- mechanism other than IGA, and the tube is classified as degraded or defective, i
then the tube will be considered in classification of inspection results for expanded random inspection. This provision ensures that the special . inspection and classification of " B OTSG steam generator first span pit-like IGA indications'will not " mask" the presence of any new degradation mechanisms. The proposed 150 gpd limit is intended to ensure prompt shutdown following detection of primary-to-secondary leakage. Prompt shutdown at a lower leakage limit will reduce the likelihood of the source of the leak propagating to a potential tube rupture and will ensure that tubes initially leaking during normal operation would not contribute excessively to total leakage under post accident
- conditions. The lower. leakage limit and prompt shutdown will also reduce impact on the plant secondary side due to the presence of primary-to-secondary leakage.
Based on these considerations:
- 1. The damage mechanism to which the new inspection criteria will be j l applied is a known degradation mechanism (pit-like IGA) which is not ;
structurally significant, and har, been confirmed by two tube pull '
- campaigns, and the destructive examination of approximately 168 IGA
- indications. (Detailed information in support of this consideration
[ has been previously reported in References A and B.) i
- 2. Pit-like IGA is readily identifiable and can be distinguished from other damage mechanisms such as cracking. (Detailed information in support of this consideration has been previously reported in Reference C.) FPC commits to performance of MRPC and bobbin inspection on all identified "B" 0TSG first span pit-like IGA indications. This is to ensure that the provisions of the proposed 1 special inspection for "B" 0TSG first span pit-like IGA indications .
- does not mask the presence of a new damage mechanism in the same l region.
] 3. Re-inspection of the "B" 0TSG first span pit-like IGA indications ! , will provide continued confirmation that no apparent growth of these !
- indications is occurring in this region.
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's 1 U. S. Nuclear Regulatory Commission 3F0397 l Page 11 of 20
- 4. The random inspection of tubes adjacent to tubes listed in the OTSG Inservice Inspection Surveillance Procedure (Table 1 herein), and I the balance of tube length outside the first span for those tubes, .l will provide continued confirmation that the population of "B" OTSG i tubes which contain multiple pit-like IGA indications in the first j span has been bounded.
- 5. No changes are being proposed to the existing repair criteria in TS 5.6.2.10 for disposition of indications. Tubes with pit-like IGA i indications which equal or exceed 40% through-wall based on use of 1 a qualified sizing technique, or which must be assumed to equal or exceed 40% through-wall due to a combination of MRPC confirmation and a lack of a qualified sizing technique, will be plugged or sleeved.
- 6. In the event that any of these listed tubes demonstrate a change in damage mechanism morphology, i.e. the pit-like indications evolve ;
into crack-like indications as determined by the MRPC inspections, the NRC will be explicitly notified of the occurrence. The proposed change will ensure that existing tube degradation is effectively monitored and that any new tube degradation is detected and appropriately dispositioned such that the integrity of the reactor coolant pressure boundary is maintained. DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION i In accordance with 10 CFR 50.91 (a)(1), the.following analysis is provided to : demonstrate that the proposed changes do not represent _a significant hazards ' consideration. According to 10 CFR 50.92 (c), the proposed changes discussed above are deemed to involve a significant hazards consideration if there is a l positive finding in any one of the following areas , l '. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences 'of an accident previously evaluated? FPC Response: No. The CR-3 component addressed by this proposed change is the "B" Once Through Steam Generator (OTSG), identified by plant tagging procedures as RCSG-18. OTSGs are straight tube, straight shell heat exchangers which allow for heat removal and the subsequent production of steam as a result of heat transfer from the primary side reactor coolant to the secondary side feedwater. Based on review of Chapter 14 of the CR-3 Final Safety Analysis Report , (FSAR), analyses have been performed to assess the consequences of a steam - generator tube rupture event, including the complete severance of a steam generator tube. The analyses concluded that CR-3 was sufficiently designed to ensure that in the event of a steam generator tube rupture the radiological doses would not exceed the allowable limits prescribed by 10CFR100, and wou1( not result in additional tube failures and further
i U. S. Nuclear Regulatory Commission I 3F0397-16 Page 12 of 20 degradation of the integrity of the reactor coolant pressure boundary. In addition, these change include continuing the currently accepted primary-to-secondary leakage limit that was previously approved for the current operating cycle only. This value is conservative relative to existing safety analyses, and would result in lower doses than currently calculated and found acceptable. The proposed change to the OTSG inspection criteria establishes that future inspections will include 100% inspection of the first- span of specific tubes which are known to have indications of degradation. The degradation of these tubes is attributed to a common non-random mechanism. The results of inspections of these tubes will be dispositioned using to the same criteria as all other OTSG tubes for determination of the need for plugging or sleeving. Therefore the proposed change will not increase the probability or consequence of an accident previously evaluated, as all tubes degraded beyond acceptable limits will be subject to consistent ' corrective actions.
- 2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
FPC Response: No. The purpose of 0TSG tube inspection is to identify tubes that may have a higher potential for failure due to degradation that results in a reduced ability to withstand operating conditions. Neither the type of inspection of OTSG tubes nor the process for performing inspections will be changed by this amendment. Consistent criteria will be applied to disposition inspection results, and consistent corrective actions will be taken for tubes that exceed this criteria. These changes do not alter the design or operation of the OTSGs. Therefore, no new or different kind of accident will be created as a result of this change. i
- 3. Will operation of the facility in accordance with this proposed change involve a significant reduction in margin of safety?
FPC Response: No. The analyses that have been performed on the effects of OTSG tube i failures as reported in the CR-3 FSAR have demonstrated that internal and ]; offsite consequences are within allowable limits. This change will not alter the acceptance criteria for inspection results. Since this change ! will assure that a group of tubes with existing-first span pit-like IGA indications are inspected each inspection period, the likelihood of detecting active degradation, as well as the probability of repairing degraded tubes prior to the degradation resulting in a tube rupture or through-wall opening is increased. Therefore, this change will not involve a reduction in the margin of safety. __ _ _ - __ .-- -- -____- J
4 U. S. Nuclear Regulatory Commission 3F0397-16 Page 13 of 20 ENVIRONMENTAL IMPACT EVALUATION l Radiolooical Evaluation While 10 CFR 51 requires an environmental assessment (EA) or environmental impact statement (EIS) for any " major Federal action significantly affecting the quality of the human environment," it does allow the NRC discretion in evaluating the , extent to which EA's or EIS's are necessary. EA's or EIS's are not- required for ' any action included in the list of " categorical exclusions" set forth in 10 CFR ) 51.22(c). Specifically,10 CFR 51.22(c)(9), provides that an EA is not required for the issuance of an amendment provided that: (i) the amendment involves no significant hazards consideratici, (ii) there is no significant change in the types or significant increase i in the amounts of any effluents that may be released offsite, and (iii) there is no significant increase in individual or cumulative j occupational radiation exposure. FPC considers that the provisions of 10 CFR 51.22(c)(9) are applicable to this request for a Technical Specification change to the "B" OTSG tube inspection requirements. For the reasons described in this submittal. FPC believes that the three criteria of 10 CFR 51.22(c)(9.) are satisfied. Therefore, this Technical : Specification amendment should be considered under the " categorical exclusions" provisions of 10 CFR 51.22(c)(9). There will be no environmental impact from approval of this change to the "B" OTSG tube inspection requirements. For the reasons given in this submittal that there will be no increase in offsite consequences due to this action, its impact is bounded by the impacts assumed in the existing Final Environmental Statement (FES) for CR-3. By adopting a lower allowed primary-to-secondary leakage rate, radioactively contaminated effluents i from a possible primary-to-secondary leak will be reduced. Even if the NRC l chooses to perform an EA, information provided in the FES, together with this i submittal should assist the NRC in making a " finding of no significant impact" in accordance with 10 CFR 51.32. { Non-Radioloaical Evaluation The non-radiological environmental concerns for the Crystal River Energy Complex that impact CR-3 are discharge canal water temperature and flow. In developing the site permit for Units 1, 2, and 3, FPC negotiated with the Environmental Protection Agency (EPA) to establish a set of effluent limitations and monitoring requirements which serve to protect the environment. The agreements between FPC and the EPA for the National Pollutant Discharge Elimination System (NPDES) establish limits on discharges to the canal. These are reflected in Permit No. FL00001'59 which states: (1) the combined condenser flow from Crystal River Units 1, 2, and 3 l shall not exceed 1897.9 million gallons per day (MGD) during the i period May 1st through October 31st of each year, nor 1613.2 MGD during the remainder of the year; and
I U. S. Nuclear Regulatory Commission 3F0397-16 Page 14 of 20 (2) the discharge temperature at the site point of discharge shall not exceed 96.5'F for a period of more than three consecutive hours, or a daily maximum value of 97.0*F. This NPDES permit was issued on September 30, 1993 by the EPA to assure that the environmental consequences of Units 1, 2, and 3 remain within acceptable limits. Operation of CR-3 in accordance with the NPDES permit limits assures that the ' provisions of the Non-Radiological Technical Specifications contained in Appendix B-Part II of the CR-3 Operating License are met. FPC has considered the non-radiological effects of this Technical Specification amendment. There will be no change in the CR-3 circulating water discharge flow rate or temperature. Therefore, no changes are required in the Environmental Protection Plan because of this action.
4 U. S. Nuclear Regulatory Commission 1 3F0397-16 i
~Page 15 of 20 TABLE 1 (page 1 of 3)
B OTSG TUBES WITH FIRST SPAN PIT-LIKE INDICATIONS nou ! Tunc nou ! TusE nou ! Tust mau ! TumE Row ! TUsE
]
. , , , I 4 40 42 l 42 49 48 54 51 58 44
!. !, !, l 4 , !,
6 ! 35 42 !, 45 49 l 49 54 82 58 45
!, !, I I 25 l 4 42 l 48 49 l 50 55 ! 32 58 ! 83 l
< 25 f 10 43 f 40 50 f 33 55 f 41 [ 58 f 92 26 !, 71 43 !, 42 50 34 55 49 59 '25 27 71 43 ! 45 50 ! 39 55 81 59 26 31
' 37 43 49 50 48 55 82 59 31 l l s4 l 37 44 l 42 50 l 56 56 l 31 59 ! sc
, , ,I
, 9 .
35
' 38 44
' 44 ' ' '
51 34 56 32 59 45 l 36 39 44
!, 46 51 35 56 35 59 49 36 l 40 45 l 37 51 ! 42 56 l 42 60 38 36 44 45 45 51 l 47 56 ! 44 60 89 37 !, 40 45 l 46 51 ! 48 56 49 61 25
- 37 !, 41 46 !, 37 51
!, 49 56 !, 50 61 26 37 44 46 l 41 51 l 55 56 51 61 29 37 l 93 46 ! 44 51 l 65 56 ! 53 61 38 I 37 94 46 l 46 51 !, 79 56 !, 82 61 48 38 41 46 49 51 80 57 27 61 82 39 l 41 46 !, 75 52 l 34 57 38 62 l 27 39 1 42 47 ! 34 52 !, 36 57 l 40 62 l 33 39 44 47 47 52 39 57 ! 43 62 l 40 39 !, 45 47 !, 48 52 !, 40 57 !, 44 62 ! 50 39 46 48 l 7 52 41 57 45 62 l 99 40 42 48 !, 38 52 !, 43 57 !, 47 63 !, 17 40 44 48 41 52 81 57 51 63 !, 27 40 ' 4r 48
' 47 53
' 39 57 ' 52 63
' 29 l
40 l 47 48 ! 49 53 l 49 57 l 89 63 ! 39 41 45 49 35 53 81 57 96 63 44 41 !, 47 49 !, 38 54 !, 33 58 l 27 63 !, 45 41 ! 72 49 ! 41 54 ! 35 58 ! 33 64 ! 28 42 I 39 49 I 42 54 I 37 58 ) 37 64 39 i i i i i l 42 1 41 49 ! 47 54 ! 40 58 ! 41 64 ! 46 i 1
U S. Nuclear Regulatory Commission 3F0397-16 Page 16 of 20 TABLE 1 (page 2 of 3) B OTSG TUBES WITH FIRST SPAN PIT-LIKE INDICATIONS ROW s TUBE R0W TUBE ROW TUBE ROW
! ! TUBE ROW ! TUBE
. . . 1 64 51 82 l 95 91 l 37 96 29 100 94 l
!. . . !. l 65 27 83 l 96 91 43 96 l 30 1'01 31 l
\
65 l 28 83 l 100 91 93 96 39 101 32 I l ! !
. , , , , \
65 38 84 ! 95 91 97 96 l 40 101 37 l 65 l 50 84 l 98 92 l 25 96 ! 41 101 41 l l 66 l 28 84 l 99 92 l 44 96 l 42 101 42 l l 66 36 84 100 92 45 96 43 101 43 I 66 l 52 85 l 43 92 l 93 96 44 101 45 67 l 35 85 ! 93 92 l 96 96 l 45 101 l 48 i
, I , . .
67 36 85 l 95 93 27 96 47 101 l 90 67 l 43 85 l 97 93 l 32 97 l 27 101 l 91 68 ! 35 85 98 93 l 41 97 l 45 101 l 93 68 ' 38 85 99 ' ' 93 79 97 95 101 98 l l 68 ! 99 86 l 24 94 l 39 98 36 102 l 43 69 l 42 86 l 30 94 l 41 98 l 38 102 l 91 69 l 99 86 l 32 94 l 4? 98 l 43 103 l 34 70 ! 42 86 !, 35 94 43 98 ! 47 103 35 l 73 l 39 86 l 94 94 l 44 98 ! 92 103 l 37 i 74 l 45 86 l 99 94 l 45 98 l 93 103 l 44 74 46 87 94 95 28 98 95 103 90 74 47 87 l 98 95 l 32 99 34 103 l 93 74 l 51 89 l 43 95 l 33 99 ! 41 104 ! 31 78 25 89 95 95 36 99 42 104 33 78 41 89 96 95 42 99 43 104 36 78 l 45 90 l 40 95 l 44 99 l 94 104 37 l 79 l 97 90 l 43 95 l 45 99 l 95 104 l 51 80
' 22 90
' 44 46 100 95 27 104 77 81 l 94 90 l 94 95 l 47 100 l 32 104 89 c1
' 96 90
' 96 95
' 92 100
' 33 105 32 81 l 98 90 98 l 96 100 l 37 105 l 34 81 l 104 90 l 99 95 l 97 ) 100 l 41 105 l 36 e i i i i i 82 94 91 23 96 Ro ' 100 91 105 43
! ] l t 1
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l U. S. Nuclear Regulatory Commission 3F0397-16
- Page 17 of 20~ )
2
-TABLE 1 (page 3 of 3) i B OTSG TUBES WITH FIRST SPAN PIT-LIKE INDICATIONS 1 i i i. .. i a0W ! TusE . ROW ! TURE ROW ! TUBE ROW ! TURE ROW ! TURE 105 90 107 ' 50 110 ' - 29 '
!, 112 ! 43 117 43 106 ! 33 10s 31 110 ! 41 113 ! 39 117 -44 106 l 35 108 l 33' 110 l 4s 113 ! 41 118 36 106 ! 42 10s ! '42 110 ! 47 113 l 4s tis' 39 l-106 47 108 ' 44 111 33 114 44 tis 40 106 ! so' 10s ! 47 111 ! 41 114 ! 47 129 41 )
E I 5 5 !. 106 l 74 109 l 29 111 l 43 116 l 41 l 106 ! 87 109 l 31 111 ! 45 116 42 - 107 ! 32 109 ! 32 111 ! 47 116 ! 43 '! I e i i i 6 i 107 ! 48 109 l 45 112 l 40 116 ! 44 ! I i l I j i
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