ML20086D664
| ML20086D664 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 11/21/1983 |
| From: | Rubenstein L Office of Nuclear Reactor Regulation |
| To: | Lainas G Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20085K039 | List: |
| References | |
| FOIA-83-722 TAC-52142, NUDOCS 8312020436 | |
| Download: ML20086D664 (17) | |
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UNITED STATES 8
N.UCLEAR REGULATORY COMMISSION o
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- E WASHINGTON, D. C. 20555
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'NOV 21 1983 MEMORANDUM FOR:
G. C. Lainas, Assistant Director for Operating Reactors, DL FROM:
L. S. Rubenstein, Assistant Director for Core and Plant Systems, DSI
SUBJECT:
SAFETY EVALUATION FOR TURKEY POINT UNITS 3 AND 4 TECHNICAL SPECIFICATION CHANGES RELATED TO F#
AND Fq (TACS J2142 3 D 52143)
Plant Name:
Turkey Point Units 3 and 4 Docket Numbers:
50-250 and 50-251 Responsible Branch:
Operating Reactor Branch #1 Project Manager:
D. Mcdonald Description of Review:
Technical Specification Change SER Review Status:
Complete Enclosed is the Core Performance Branch safety evaluation of a proposed amendment to the Technical Specifications contained in Appendix A of Facility Operating License Nos. DPR-31 and DPR-41 for Turkey Point Units 3 and 4.
We conclude that reasonable assurance has been provided that the proposed changes to F and F will not pose a threat to the health and safety of the public Ed that the proposed Technical Specification changes 0
are, therefore, acceptable.
( h /
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L. S. Rubenstein, Assistant Director for Core and Plant Systems, DSI
Enclosure:
As stated cc:
R. Mattson D. Eisenhut
. S. Varga D. McDonaldt R. Capra
Contact:
M. Dunenfeld, CPB:DSI Y. Hsii, CPB:DSI S. Sun, CPB:DSI X-28097 X-29473 X-29499 i
XA Copy Hai_BeX5pnt io PDR
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SAFETY EVALUATION FOR TORKEY POINT UNITS 3 AND 4 aH q (TACS 52142 AND 52143)
-TECHNICAL SPECIFICATIONS CHANGES IN F AND F 1.
INTRODUCTION By letter dated August 19,1903 (Ref.1) and as amended in a letter dated September 9,1983 (Ref. 2) F'orida Power and Light Company submitted a request for an amendment of the Technical Specifications contained in Appendix A of Facility Operating Licenses DPR 31 and 41.
These amendments propose changes to the Technical Specifications to support the integrated program for pressure vessel flux reduction and to take credit for operation with the new steam generators in an unplugged (maximum of five (5) percent tube plugging) configuration.
Changes are requested to:
(1) increase the hot channel F limit from 1.55 to 1.62; (2) increase the aH total peaking factor F limit from 2.30 to 2.32; (3) change the Overpower g
AT setpoints and thermal-hydraulic limit curves; and (4) delete restrict-ions and limits placed on the old steam generators for operation with tubes plugged in excess of five (5) perce'nt.
In connection with the review of these proposed changes, we have received Comments and a request for a Petition for Leave to Intervene in this matter from the Center for Nuclear Responsibility and Joette Lorion (Ref. 3).
We have addressed the contentions in these Comments and the Petition in the text of this Safety Evaluation Report where relevant.
In addition, we have addressed contentions concerning pressure vessel embrittlement in Appendix A to this report. Contentions concerning the use of optimized fuel assemblies (OFA), use of wet annular burnable sbsorber (WABA) rods, and increased shutdown and control rod drop times are addressed in our Safety Evaluation Report on 0FAs and WABAs perfonned for TACS 51793 and 51794.
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' 2.
NUCLEAR DESIGN EVALUATION The proposed Technical Specification changes will not impact the current nuclear design bases ~ given in the FSAR (Ref. 4) and applied to subsequent Unit 3 and 4 Reload Safety Evaluations. The standard calculational methods described in the " Westinghouse Reload Sa%ty Evaluation Methodology" (Ref. 5) continue to apply. As is current practice, each reload core design will be evaluated to assure that design and safety limits are satisfied according to this reload methodology.
As discussed in Section 4.2, a LOCA analysis has been perfonned using a total heat flux peaking factor, F, of 2.32.
Basically, this reflects g
elimination of the need to operate the Turkey Point reactors at a reduced F because of a substantial percentage of plugged steam generator g
tubes. This results from the installation of new steam generators. There assumed as an is an approved methodology for justification that the Fq initial condition in the LOCA analysis will not be exceeded in normal operation of the power plant. This methodology is described in Ref. 5.
As a result of our questions, the licensee provided (Ref. 6) the specific results of application of this technology to Turkey Point Unit 3, Cycle 9.
These results employ radial peaking factors in conformance with the.FAH change proposed in this amendment. We have reviewed the results for Unit 3 Cycle 9 and find them acceptable. That is, based upon these results, and the power distribution monitoring Technical Specifications in place for the limit of 2.32 assumed as input to reactor, we are confident that the Fg the 1.0CA analysis will not be exceeded in nonnal operation of the power f
plant. Continued application of this methodology for future cycles of both units will permit the same conclusion to be drawn.
l limit of 2.30, the Although the previous cycle of operation had an Fg proposed change to 2.32 does not represent an increase in the potential coolant temperature change available to produce a pressurized thennal l
l shock to the reactor vessel. This is because there is no change to the authorized power level of the facility. That is, the reactor coolant
i inlet, average, and outlet temperatures do not change with a change in peaking factor. They would c.hange with a change in power level.
At full power the average linear heat generation rate is 5.58 kW/ft in the core. The product of this and the peaking factor yields 12.9 kW/ft.
This is the peak linear heat rate, which is increased from 12.8 kW/ft in the previous cycle. However, the accident analyses, particularly the LOCA, show satisfactory results with initial conditions including the slightly increased peak linear heat rate.
The F limit for Turkey Point Units 3 and 4 was 2.32 for cycles 1 through g
3 inclusive. The peak allowable operating design linear heat rate at full power in the initial cycles was 18 kW/ft. The authorized core power level is 2200 Mwt. This is the lowest for a Westinghouse 3 loop reactor.
Others operate at power levels up to 2785 Mwt.
The Turkey Point average full power linear heat generation rate of 5.58 kW/ft is the lowest of any Westinghouse designed reactor using 15x15 fuel.
Of the others, 6 run at 5.7 kW/ft, S run at 6.2 kW/ft and 3 at 6.7 kW/ft.
3.
Thermal-Hydraulic Design Evaluation Since the proposed Technical Specification amendment will increase the hot channel factor, FAH, from 1.55 to 1.62 and increase the total peaking factor, Fg, from 2.30 to 2.32, and since the future cycles will be reloaded with the 15x15 optimized fuel assemblies (OFA), the impact of operating at these higher peaking factors on thermal. margin is evaluated.
The licensee has detennined that the increase of the F from 1.55 to 1.62, AH an increase of 4.5%, will result in a' DNBR penalty of 9%.
This is i
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derived from using an estimated sensitivity factor of -2.0 for the rate of This estimated sensitivity change of DNBR with respect to the Fg.
factor is a conservative value since a study perfonned by the Battelle Pacific Northwest Laboratories (Ref. 7) has shown a sensitivity factor of -0.715 for the DNBR calculated by the W-3 CHF correlation, which would result in the DNBR penalty of 3.2% rather than 9%.
In the previous Technical Specification change the fuel rod bow effect on DNBR was calculated using the approved interim method (Refs. 8, 9,10) which resulted in a maximum rod bow penalty of 14.9%. With the approval of topical report WCAP-8691, Revision 1 (Ref.10), the licensee has recalculated the rod bow penalty using the approved method. This method applies statistical convolution of the CHF test data and inter-fuel rod gap closure data to derive the rod bow penalty on DNBR. Since rod bow and gap closure increase with fuel burnup, the rod bow penalty on DNBR increases with burnup.
However, for the purpose of calculating rod bow penalty, the maximum fuel burnup used for the calculation is 33000 MWD /MTU. This is because the physical burndown effect at higher burnup is greater than the rod bowing effects which would be calculated based on the amount of bow predicted at these burnups. The rod bow penalties at 33000 MWD /MTU are 4.7% and 5.5%, respectively, for the 15x15 LOPAR using the W-3 L-Grid CHF correlation and 15x15 0FA using the WRB-1 correlation. The difference in rod bow penalties using the old interim i
method and the new approved method are 10.2% and 9.4%, respectively, for the LOPAR and 0FA. These differences represent gains in DNBR margins which can be used to compensate for the estimated DNBR penalty of 9%
u from 1.55 to 1.62.
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resulting from the increase of Fg l
The licensee has perfonned the therr.al-hydraulic analysis with the proposed Fg of 1.62 using the same methods described in the FSAR, except for the use of the new approved fuel densification model (Ref.11).
l For the 15x15 LOPAR fuel, DNBR is calculated with the W-3 L-Grid CHF-correlation with the safety analysis minimum DNBR limit of 1.30. This value is 4.8% higher than the allowable DNBR limit of 1.24 derived from
l The themal analysis also uses the 15x15 L-Grid CHF test data.
conservative values of themal diffusion coefficient and pitch reduction There-which provide inherent DNBR margins of 3% and 3.3% respectively.
fore, a total of 11.1% DNBR margin is available to compensate for the For the OFA fuel, the safety analysis ninimum rod bow penalty of 4.7%.
This DNBR limit is DNBR limit is 1.34 using the WRB-1 CHF covelations.
This 12.7% higher than the allowable DNBR limit of 1.17 for WRB-1.
f margin is sufficient to compensate for the 5.5% rod bow penalty, the transitional mixed core penalty of 3% DNBr. imposed on the 15x15 0 to account for the mismatch in the hydraulic resistances between the ar.d 0FA fuel, and possibly small uncertainty associated with the appli-Therefore, with the cation of the WRB-1 correlation to the ISx15 0FA.
reduction of rod bow penalty calculated with the approved method of of 1.62 WCAP-8691, Revision 1, the plant operation with the proposed F AH will not result in the DNBR violating the specified acceptable fuel design limits required for the W-3 and WRB-1 CHF correlations during nomal operation and the anticipated operational occurrences.
4.
Accident Evaluation _
The licensee also provided an evaluation on the effects of the increased limits on non-LOCA and LOCA accidents.
F and F AH q
4.1 Non-LOCA Evaluation _
The Reactor Core Themal and Hydraulic Safety limits are recalculated limit of 1.62. Based on these new protection limits, l
using the new FAH f
the licensee has perfomed calculations for the Overtemperature aT.
(OT 5) and dverpower 5 (OPaT) setpoint equation constants using The results indicate that the standard Westinghouse method (Ref.12).
Overterrperature aT setpoint equation in the current Technical. Sp Therefore no change in the OT s equation and no is conservative.
reanalysis for the OT K trip events'are required. - A change in the f
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- r OPAT setpoint is required.
It was caleviated with the methods described in Ref. 12. These methods have been used to calculate the safety limit curves -
j and OTaT and OPAT setpoints for all Westinghouse initial and reload ores approved to date. We reviewed these methods as applied to the t.
ety limit curves and OPAT setpoint changes submitted for this application and find the requested changes acceptable. These changes are in the conservative direction.
4.2 Large Break LOCA Evaluation The large break LOCA analysis is performed with 102% of the rated thermal power of 2200 Mwt, a hot channel factor, F6H, of 1.62, a total peaking l
factor, F, of 2.32 and an assumed steam generator tube plugging le el of 5%.
g A sensitivity study is performed with break sizes ranging from 1 ft area to a full double ended break of the cold leg, and various Moody discharge coefficients. The results show that the double ended cold leg l
guillotine break with a discharge coefficient of 0.4 is the worst large break LOCA case.
It has the highest peak cladding temperature.
The analysis is perfonned with a modified version of the 1981-Westinghouse ECCS evaluation model (Ref.13). This evaluation model uses the revised PAD Fuel Thermal Safety Model for the calculation of the initial-fuel conditions; the SATAN-VI code for the transient thermal. hydraulic calculation during blowdown period; the WFFLOOD code for the calculation of the refill and reflood transient periods; the LOCTA-IV code for the calculation of peak cladding temperature; and the C0C0 code for the j
I calculation of the dry containment pressure history. The modified version of ECCS evaluation model uses the BART computer code (Ref.14) to calculate the reflood heat transfer coefficient normally perfonned by the WREFLOOD code. The purpose of BART is to provide a time and location dependeni.
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clad surface heat transfer coefficient for the reflood rates ranging frein The BART computer 0.6 to 1.5 inch /sec during the reflood stage of LOCA.
code and design application are describrd in the Westinghouse topical report WCAP-9561 (Ref. 14). The BART computer code without grid spacer model and its application in the Westinghouse evaluation model have been reviewed 6nd approved by the staff in a separate SER regarding WC Since the spacer grid model to be used in BART is still under staff review, the licensee in its letter of September 20, 1983 (Ref. 15) submitted additional analysis of the large break LOCA using the ungridded BART mod We find that the approved methods The staff has reviewed this analysis.
and computer codes are used and the results show that the peak cladding temperature, metal-water reaction and clad oxidation are within the accep ance criteria imposed in 10 CFR 50.46 for LOCA analysis.
This analysis is applicable to both a full core 15x15 LOPAR and a full For its application to the transition mixed core, core 15x15 0FA fuel.
the licensee has perfomed an evaluation to determine the effect' of the flow redistribution due to the hydraulic resistance mismatch in the mixed core configuration.
Since the 15x15 0FA increases the flow resistence by about 4.5%, the reflood flow rate for the 15x15 0FA fuel during the transitional mixed core period will be reduced by approxi-This will result in an approximately 10*r increase in mately 2.2%.
peak cladding temperature for the transition core which is still within the acceptance criteria.
4.3 Small Break LOCA Evaluation _
The small break LOCA analysis is performed with the approved computer c i.e., (1) the revised PAD thermal safety model for the calculation of the fuel initial conditions; (2) the WFLASH code for the calculation of the transient depressurization of the reactor coolant system, fuel power, mir.ture height and steam flow past the uncovered part of the core; and The (3) the LOCTA-IV code for the peak cladding temperature analysis.
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evaluation is done with.the 102% of the rated thermal power, the hot channel factor of 1.62 and the total peaking factor of 2.32 at the core midplane. Various break sizes are performed and the results show that the worst break size to be a 3 inch diameter break which results in the highest peak cladding temperature of 1605'F, well below the acceptance criterion of 2200*F. This analysis is applicable to both 15x15 LOPAR and 15x15 0FA fuels.
For a transition mixed LOPAR-OFA core, the flow redistribution due to mismatch in the fuel assembly hydraulic resistance may have a small effect on the PCT.
However, since the PCT margin is so large, this small effect will not cause the PCT to approach the acceptance criterion.
5.
Technical Specifications The specific Technical Specification changes and the reasons for their acceptability are:
Page vi Figure 3.1-1 has been added to the List of Figures. This change is editorial and has no safety significance.-
Figure 2.2-1 This figure has been modified to remove the " note", which is no longer applicable with the new steam generators. The limits were recalculated to reflect the increase in the allowable F limit, and is acceptable as aH discussed in Section 5.1.
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Figure 2.1-la & 2.1-lb These figures are no longer required with the new steam generators.
Page 2.3-2 The note is deleted as it is no longer applicable with the new steam generators.
Page 2.3-3 The multiplier is modified in the Overpowera T equation. The notes are deleted as they are no longer needed with the new steam generators.
The modified multiplier in the OverpowerA T is acceptable as discussed in Section 5.1.
a Page 3.1-7 The notes are deleted as they are no longer applicable with the new steam generators.
Page 3.2-3_
Notes and references limit and part power multiplier are increased.
The F to plugging levels are deleted as they are no longer applicable with the 6g steam generators.
is increased to 2.32 on the basis of LOCA analysis using BART methodolo Fq limit change is acceptable for che reasons discussed in Section 4.
l The Fg This change The part power multiplier on the limit is changed from.2 to.3.
linearly increasing from zero at full power to allows an increase in Ff of 20% at The.2 multiplier allowed an increase in Fg 30% at zero power.
The purpose of this multiplier is to allow an increase in FaH zero power.
with decreasing power level to account for the effect of insertion of control rods and reduction in negative feedback with decreasing power level.
It has been found generically that the.2 multiplier was too restrictive, and limits at very low power levels, when there caused violations of the F3g is clearly no safety problem. Accordingingly, licensees with Westinghouse designed reactors have been requesting, and we have been accepting, the change to the.3 multiplier. We have reviewed the licensee's evaluation of the effect of this change (Ref.1, Section 3.1) and agree with his conclusion that its effect is negligible, f
The F increase is acceptable as discussed in Sections 2 and 4.2.
q Figure 3.2-3 & 3.2-3a This figure has been combined into Figure 3.2-3 and revised to present new limits from the LOCA analyses. We calculated this figure independently and agree that it is correct, and therefore this change is < acceptable.
Page B2.1-2 The F and part power multiplier have been increased.
g Page 83.2-4 The increased F limit is noted.
m The last two changes are consequences of the changes on page 3.2-3 and are acceptable as discussed above.
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6.0 CONCLUSION
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We have reviewed the proposed changes a the Turkey Point Units 3 and 4 Technical Specifications involving the increase of the hot channel factor i!
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F limit from 1.55 to 1.62, increase of the total peaking factor Fg aH limit from 2.30 to 2.32, recalculation of the Overpower AT setpoints and thermal-hydraulS curves, and other minor changes and find they pose no significant hazard.
In reaching our conclusions, we have utilized the General Design Criteria, Standard Review Plan, and the experience of staff members who are familiar with the requirements and criteria app?ied to Westinghouse designed reactors to ensure their safe operation. These changes will not cause any change in the types or increase in the amount of effluents or any change in the authorized power level of the facility.
The F limit increase and the corresponding changes in the Overpower AT AH setpoints and themal hydraulic limit curve do not lead to a departure from nucleate boiling in the core during namal operation and anticipated operational occurrences, and there would not be any significant increase in tis probability or consequences of accidents previously analyzed.
The F limit increase does not significantly increase the probability or q
consequences of accidents previously analyzed because the LOCA analyses show results below the acceptable limits of performed for this Fq 10 CFR 50.46.
These chhnges also do not create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a signifi-cant reduction in a margin of safety. The amendment is therefore accept-able according to the Commission's requirements for Significant Hazards considerations.
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7.0 REFERENCES
R. E. Uhrig (FP&L) letter, L-83-344 to D. G. Eisenhut (NRC), " Turkey Point 1.
Units 3 and 4, Docket Nos. 50-250 and 50-257 Proposed License Amendment, F /F ", August 19, 1983.
aH g R. E. Uhrig (FP&L) letter, L-83-477, to D. G. Eisenhut (NRC), " Turkey 2.
Point Units 3 and 4. Docket Nos. 50-250 and 50-257, Proposed License Amendment, F /F," September 9,1983.
AH q M. Hodder letter to Secretary (NRC) " Comments and Request for Hearing 3.
and Petition for Leave to Intervene" November 7,1983.
Turkey Point Units 3 and 4 Final Safety Evaluation Report, Docket Nos.
4.
50-250/50-251, updated December 1981.
F. M. Bordelon, et al, " Westinghouse Reload Safety Evaluation Methodology".
5.
WCAP-9272 (proprietary) and WCAP-9273 (non-proprietary), March 1978.
R. E. Uhrig (FP&L) letter, L-83-507 to D. G. Eisenhut (NRC), " Clarification 6.
Technical Specification Changes", October 4,1983.
to F /F mq G. M. Hesson and J. M. Cata, " Analysis of the Sensitivity of Calculated 7.
MDNBR to Eight Selected DNB Parameters". FATE-79-101, March 1979, Battelle Pacific Northwest Laboratories, Richland, Washington.
D. Ross and D. Eisenhut, Memorandum to D. Vassallo and K. Goller, 8.
" Revised Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing 16, 1977.
on Thennal Margin Calculations for Light Water Reactors", February R. Meyer, memorandum to D. Ross, " Revised Coefficients for Interim Rod I
9.
Bowing Analysis", March 2, 1978.
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- 10. J. Skaritka (Ed.) " Fuel Rod Bow Evaluation", WCAP-8691, Revision 1.
July 1979.
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- 11. J. M. Hellman, " Fuel Densification Experimental Results and Model for Reactor Application" WCAP-8218-P-A, March 1975.
- 12. S. Ellenberger, et al., " Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions", WCAP-8745, March 1977.
- 13. E. P. Rahe, " Westinghouse ECCS Evaluation Model,1981 Version",
WCAP-9220-P-A, Revision 1, 1981.
- 14. M. Y. Young, et al., "BART-A1: A Canputer Code for the Best Estimate Analysis of Reflood Transients", WCAP-9561, January 1980.
- 15. R. Uhrig (FP&L) letter to D. Eisenhut (NRC), " Turkey Point Units 3 and 4, Docket No. 50-250 and 50-251, Additional Analyses (LOCA with Ungridded BART Model) for FAH/BART Technical Specification Amendment", September 20, 1983.
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F APPENDIX A TO SER ON Fg g
PRESSURIZED THERMAL SHOCK FOR TURKEY P 1
Introduction _
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The staff has been actively studying the pressurized thennal sh i
than problen for the operating pressurized water re l
for PTS and the identification of those plants which were in ne two years.
The staff presented the results of its reduction to the pressure vessel.-
The PTS studies to the Commission on December 9,1982 (SECY-82-4 t ff screening criteria were the basis of a proposed rule developed b l
15, 1983 (SECY-83-288).
and sent to the Commission for approval on July l
In SECY-82-465 the staff noted that most plants could avoid re screening criteria throughout their service life by timely impleme The staff indicated that it planned to meet i
of flux reduction programs.
d tion with licensees of plants which appeared to need near-tann flux re uc factors greater than two to ensure that the screening criteria wou The infonnation obtained was summarized f
be exceeded through service life.
25, 1983.
liinally the staff for the Commission in SECY-83-79, February for updated its infonnation to the Commission on flux reduction progr 28, 1983. Twc of these eight leading plants in SECY-83-443 on October plants were Turkey Point Units 3 and'4.
'i Flux Reduction at Turkey Point Units 3 and 4
's for Turkey Point Units 3 and 4 As of January 1, 1983 tne calculated LTNDT tial) weld, were 263*F.and 264*F, respectively, for the critical (cirr:umferen I
Lowleakagecores(i.e.,acore compared to a screening criterion of 300*F.
with loading including spent or partially spent assemblies on the c l
4 periphery) were installed in Turkey Point Units 3 and 4 durin l
l through 8, i.e., before the PTS issue placed emphasis on pressu I
The current cycle for Unit 3 is Cycle 8 and for Unit 4 is
.l embrittlement.
The operating license expires in 2007 for both units. -If the that'
, core configuration of Cycle 8.were maintained, calculations in Cycle 9.
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t both units would reach the screening criterion in 1989, assum
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According to the latest presentation of the licensee to the load factor.
staff (August 2,1983) progressively higher flux reduction factors are plann for both units with future reloads.
Cycle 9 for Unit.4 is a modified low leakage core, that is estimated to reduca peak flux by a factor of 2.2 and Cycle 10 (scheduled for May 1984) will reduce the flux by a factor of 3.3 Similarly Cycle 9 for Unit 3 (scheduled for December 1983) will also reduce The 2.2 flux reduction factor will extend the date flux by a factor of 3.3.
A flux reduction factor of 3.3 for reaching the screening criterion to 1994.
Flux reduction is the most effective way to will extend vessel life to 2007.
Power reduction can extend the calendar extend the life of the pressure vessel.
time of operation but not the time in effective full power years (EFPYs).
Power derating was never considered by the staff as a means to remedy the PTS The large flux reductions to the pressure vessel at Turkey Point have two objectives (a) to reduce the flux at the peripheral weld seams of the pressure vessel in the belt region and (b) to lower the peak exposure and produce a j
more uniform integrated fluence by the end of the service life of the plant.
The licensee proposed to accomplish these objectives utilizing highly burne assemblies and assemblies heavily loaded with burnable hisons in a lim This results in a suppressed section which corresponds to the peripheral weld.
neutron source which in turn results in a reduced irradiation rate Reduction of the neutron source in the outer row of fuel pressure vessel.
rate of irradiation assemblies is the most efficient means to reduce th(
j because about 85 percent of the neutrons which reach the pressure vessel Burnable poison loaded assemblies originate in the outer row of assemblies.
have been used for many years in power reactors and what is new for the Turkey Point Units is the proposed app'lication at locations corresponding 4
The behavior the weld seam in a manner suitable for achieving flux reduction.
i of such assemblies from the nuclear.and materials point of view is known a predictable especially under the proposed conditions, i.e., in a locatio very low power.
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. The modified low leakage core will, therefore, redistribute the flux and power in the core so as to minimize power at selected peripheral assemblies in the portion corresponding to the pressure vessel weld seams of importance to PTS Concerns.
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