ML17309A916

Application for Amend to License NPF-16 by Incorporating Attached TS Rev.Amend Will Modify TS 5.6.1 & Associated Figure 5.6-1 & TS 5.6.3 to Accomodate Increase in Allowed SFP Storage Capacity

ML17309A916
Person / Time
Site:	Saint Lucie
Issue date:	12/31/1997
From:	Stall J FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML17229A570	List: ML17229A569 ML17229A571 ML17309A916 ML17309A917
References
L-97-325, NUDOCS 9801070046
Download: ML17309A916 (55)

Text

CATEGORY 1 REGULA Y INFORMATION DISTRIBUTIOA "YSTEM (RXDS) lg ACCESS1ON NBR:9801'070046

~ DOC.DATE: 97/12/31 NOTARIZED: YES

~ ~ DOCKET, 5 FACIL:50-389 St.; Lucie Plant, Unit 2, Florida Power &, Light Co.

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05000389 AUTH. NAME AUTHOR AFFILIATION TALL,J.A.~ Florida Power S Light Co.

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RECIP.NAME

~ RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Application for amend to license NPF-16 by incorporating attached TS rev.Amend will modify TS 5.6.1 & associated Figure 5.6-1 8 TS 5.6.3 to accomodate increase in allowed SFP storage capacity.

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TITLE: OR Submittal: General Distribution I g

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PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS e OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DESK (DCD) ON EXTENSION 415-2083 DOCUMENT CONTROL TOTAL NUMBER OF COPXES REQUIRED: LTTR 14 ENCL 13

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Florida Power & Light Company. 6351 S. Ocean Drive, Jensen Beach, FL 34957 L-97-325 December 31, 1997 10 CFR 50.90 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Re: St. Lucie Unit 2 Docket No. 50-389 Proposed License Amendment SELStoragaZap Pursuant to 10 CFR 50.90, Florida Power &. Light Company (FPL) requests to amend Facility Operating License NPF-16 for St. Lucie Unit 2 by 'incorporating the attached Technical Specifications (TS) revisions. The amendment will modify Specification 5.6.1 and associated Figure 5.6-1, and Specification 5.6.3 to accommodate an increase in the allowed Spent Fuel Pool (SFP) storage capacity. The analyses supporting this request, in part, assume credit for up to 1266 ppm boron concentration existing in the SFP. As discussed with the NRC Staff, it is requested that the proposed amendment, if approved, be issued by October 31, 1998.

Attachment 1 is an evaluation of the proposed TS changes. Attachment 2 is the "Determination of No Significant Hazards Consideration." Attachment 3 contains a copy of the affected TS pages marked-up to show the proposed changes. Enclosure 1 is the "St. Lucie Unit 2 Criticality Safety Analysis for the Spent Fuel Storage Rack Using Soluble Boron Credit", and Enclosure 2 is the "St. Lucie Unit 2 Spent Fuel Pool Dilution Analysis."

The proposed amendment has been reviewed by the St. Lucie Facility Review Group and the Florida Power 5 Light Company Nuclear Review Board. In accordance with 10 CFR 50.91 (b)(1), a copy of the proposed amendment is being forwarded to the State Designee for the State of Florida.

Please contact us if there are any questions about this submittal.

Very truly yours, J. A. Stall Vice President St. Lucie Plant ggOl 980i07004b 97i23i05000389,'

PDR ADQCK JAS/RLD PDR Attachments Enclosures (see next page)

IIII IIII,IIIIIIIIIIII IIIIIIIIIIIIIIIII an FPL Group company

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St. Lucie Unit 2 L-97-325 Docket No. 50-389 Page 2 Proposed License Amendment

Enclosures:

(1) St. Lucie Unit 2 Criticality Safety Analysis for the Spent Fuel Storage Rack Using Soluble Boron Credit, CENPD-387: ABB-Combustion Engineering, October, 1997.

(2) St. Lucie Unit 2 Spent Fuel Pool Dilution Analysis, PSL-ENG-SENS-97-068, Revision 0: FPL Nuclear Engineering, November, 1997 cc: Regional Administrator, Region II, USNRC.

Senior Resident Inspector, USNRC, St. Lucie Plant.

Mr. W.A. Passetti, Florida Department of Health and Rehabilitative Services.

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Page 3 Proposed License Amendment SHoMtorag~apacity~tuble Boraa&redit STATE OF FLORIDA )

) ss.

COUNTY OF ST. LUCIE )

J A. Stall being first duly sworn, deposes and says:

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That he is Vice President, St. Lucie Plant, for the Nuclear Division of Florida Power 5 Light Company, the Licensee herein; That he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information and belief, and that he is authorized to execute the document on behalf of said Licensee.

J. A. Stall STATE OF FLORIDA COUNTY Of & LLI.C.(C Sworn to and subscribed before me this 3~ day of ttdfq by J. A. Stall, who is personally known to me.

~MnrM Signature,o;fI)oo ary Pyric>-Qggiof Florida

s; MY COMMISSION 0 CC646163 EXPlAES May 12, 2001 BOIIOEO THIIV TIef FAIN NSURANCE, INO.

Name of Notary Public (Print, Type, or Stamp)

0 8"I,, t 0

St. Lucie Unit 2

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Docket No. 50-389

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Proposed License Amendment EVALUATIONOF PROPOSED TS CHANGES for ST. LUCIE UNIT 2 SPENT FUEL POOL CAPACITY INCREASE Adapted from FPL Nuclear Engineering Safety Evaluation PSL-ENG-SENS-97-083, Revision 0, 12/17/97, 9801070046 50-389

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 1 of 29

2.0 DESCRIPTION

OF CHANGES 3.0 THERMAL-HYDRAULICCONSIDERATIONS 3.1 Decay Heat Calculations for the Spent Fuel Pool 3.1 ~ 1 Purpose and Scope of Calculation 3.1.2 Acceptance Criteria 3.1.3 Description of the Fuel Pool Cooling System 3.1.4 Calculations Performed 3.1.5 Results 3.2 Maximum Fuel Cladding Temperature 3.2.1 Purpose 3.2.2 Discussion and Results 4.0 REACTIVITY CONSIDERATIONS 10 4.1 Methodology Used in Reactivity Calculations 4.2 Acceptance Criteria 4.3 Region I - Description of Storage Arrangement Analyzed 4.4 Region II - Description of Storage Arrangement Analyzed 4.5 Calculational Assumptions and Results 4.6 Reactivity Equivalencing 4.6.1 Burnup and Decay Time Reactivity Credit 4.6.2 Gadolinium Reactivity Credit 4.7 Postulated Accidents 4.8 Criticality Analysis - Conclusions 5.0 SEISMIC AND STRUCTURAL CONSIDERATIONS

St. Lucie Unit 2

~ L-97-325 Docket No. 50-389

~ Attachment 1 Proposed License Amendment Page 2of 29 BKSBWM 6.0 ASSESSMENT OF POTENTIAL FOR INADVERTENT FUEL POOL 20 DILUTION 6.1 Description of Methodology 6.2 Boron Dilution Initiating Events 6.3 Results of Initiating Events 6.4 Spent Fuel Pool Dilution Event Conclusions 7.0 NO SIGNIFICANT ENVIRONMENTALIMPACT 23 7.1 Thermal Impact 7.2 Radiological Evaluation 7.2.1 Solid Radwaste 7.2.2 Gaseous Radwaste 7.2.3 Radioactive Releases due to Accidents

8.0 CONCLUSION

S 26

9.0 REFERENCES

27 LlSI OEXLBLES Summary of St. Lucie Unit 2 Calculated Fuel Storage Rack Stress Intensities 28 St. Lucie Unit 2 Estimated Spent Fuel Pool Capacity Requirements 29

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 3 of 29 SEP~ragaZapacit EVALUATIONOF PROPOSED TS CHANGES 1.0 The existing spent fuel storage racks at St, Lucie Unit 2 contain a total of 1584 cells, of which 1076 are currently available for storage. The estimated storage capacity requirements are illustrated in Table 2. As of November 1997, the St.

Lucie Unit 2 fuel pool contains 692 permanently discharged fuel assemblies.

By the year 2001, St. Lucie Unit 2 will have filled all fuel pool storage locations not reserved for a full core off-load of fuel; by 2006, Unit 2 will have lost the ability to discharge any fuel from the reactor. To ensure that sufficient capacity to store discharged fuel assemblies continues to exist at St. Lucie Unit 2, analyses have been performed which support an increase in the number of fuel assemblies which may be, stored in the fuel pool from the current limit of 1076 to a new value of 1360. The proposed capacity increase will extend the full-core-reserve storage capability of the Unit 2 fuel pool from year 2001 to approximately 2007. Final disposal facilities for spent fuel will not be available until at least 2010. The availability of any centralized interim facility for spent fuel storage is uncertain.

Although dry storage of spent fuel may be required at the St. Lucie site in the future even with the approval of the proposed license amendment, it is prudent to maximize the storage capability of the existing fuel pool prior to initiating'he licensing and construction of an on-site dry storage facility. Deferring the requirement for dry storage at St. Lucie will permit the use of multi-purpose canisters (MPCs) which are currently being developed and licensed. Use of MPCs for on-site fuel storage and off-site fuel transport will benefit radiation workers by reducing the total occupational exposure and will minimize the generation of low level radioactive waste due to dry storage of spent fuel.

2.0 FPL proposes to modify Section 5.6 of the Unit 2 Technical Specifications, as shown in Attachment 3, to permit an increase in the storage capacity of the existing spent fuel pool storage racks from 1076 to 1360 assemblies.

Technical Specification Figure 5.6-1 will be removed and new Figures 5.6-1a through 5.6-1e will be added to describe the assembly burnup requirements for Region I and II of the spent fuel pool. The existing requirement for a fuel pool soluble boron concentration of >1720 ppm is retained.

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 4 of 29 As part of the proposed change, the permissible storage configuration of Region I will be modified; a number of cell blocking devices will be removed and the number of usable Region I cells will increase from 224 to 276. FPL expects that cell block removal and subsequent fuel repositioning operations can be performed following NRC issuance of the proposed license amendment without prior NRC approval pursuant to 10 CFR 50.59. Region I will retain the ability, for additional fuel cycles, to accept a full core discharge of irradiated fuel. Additional permanently discharged fuel which does not qualify for storage in Region II may also be stored in Region I. The proposed amendment would increase the number of usable Region II storage cells from 852 to 1084.

Justification for these proposed changes is provided in Sections 3 through 7 of this evaluation.

Approval of this proposed license amendment by the NRC may require FPL to request a modification to its current exemption from the requirements of 10 CFR 70.24. Criterion 4 of the exemption from 10 CFR 70.24 requires a k,<<of 0.95 for unborated water in the spent fuel pool. NRC rulemaking activity is underway which will obviate the need for an exemption modification request.

3.0 The thermal-hydraulic analysis is formally documented in the "St. Lucie 2 Spent Fuel Pool Thermal Hydraulic Analysis," ABB Combustion Engineering Nuclear Operations Calculation Number: 016-AS95-C-009, Rev. 0, 6/09/95, and is available from FPL Nuclear Engineering records.

3.1 DECAY HEAT CALCULATIONS FOR THE SPENT FUEL POOL 3.1.1 Because FPL is proposing to increase the quantity of spent fuel that may be stored in the fuel pool without making any modifications to the fuel pool or the fuel pool cooling system, it is necessary to ensure that the existing equipment has sufficient heat removal capacity to handle the increased load. In the course of performing the required calculations, FPL included the effects of a potential i~crease in the number of assemblies permanently discharged at each refueling outage in the event 24 month long operating cycles are implemented.

These longer cycles may result in increased batch average discharge burnups

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 5 of 29 which were also accounted for in the revised calculations. The ABB-CE PC computer code SFPOOL (Reference 6) was used to perform most calculations discussed in Section 3 of this evaluation.

To ensure that the methodology chosen for calculation of the decay heat load produced conservative results, a series of benchmark calculations were performed based on the conditions existing in the spent fuel pool as of October 1, 1994. On that date, 544 discharged assemblies were stored in the spent fuel pool. The initial enrichments, operating histories and discharge burnups of these assemblies are well known.

3.1.2 recce The performance of the fuel pool cooling system and the fuel pool temperature values derived from this reanalysis were compared to the criteria given in updated FSAR section 9.1.3.3: For a normal refueling evolution, the maximum fuel pool temperature remains at or below 150 'F; where maximum temperature values from the reanalysis exceeded the 150 'F value given in the FSAR, an additional review of the analysis of record for the fuel storage rack structure was undertaken to ensure that acceptable stress levels for rack components were not exceeded.

The maximum fuel pool temperature following the limiting postulated full core offload evolution was determined to remain less than boiling.

Normal refueling evolutions at St. Lucie Unit 2 currently employ full core fuel offloads. Compliance with pool temperature limits is ensured through the required initial conditions specified in a separate 10 CFR 50.59 safety evaluation (Reference 12); typical constraints include limits on the fuel cooling time prior to offload initiation, the rate of defueling to the fuel pool, and the maximum temperature of the ultimate heat sink. Following approval of this PLA by the NRC, Reference 12 will be revised and used to limit spent fuel pool temperature following any planned full core offload to ~150 'F.

3.1.3 The Fuel Pool Cooling System provides continuous cooling for spent fuel assemblies stored in the fuel pool ~ This permits storage of spent fuel assemblies in the pool from the time the fuel is unloaded from the reactor vessel until it is loaded into casks for shipment offsite or on-site dry storage.

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 6 of 29 The St. Lucie Unit 2 fuel pool cooling system includes two fuel pool heat exchangers cooled by shell side component cooling water and two fuel pool pumps powered from separate motor control centers providing forced circulation. Each fuel pool pump has a design flowrate of 1500 gpm.

Considering the volume displaced by a full loading of spent fuel, the net fuel pool liquid volume is approximately 300,000 gallons includin'g the fuel cask area. The cask loading area is thermally and hydraulically coupled to the remainder of the fuel pool Suction for the fuel pool heat exchanger is drawn

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from near the top of the pool and is returned after being cooled through piping which discharges near the bottom of the pool.

Normally, one fuel pool heat exchanger and one fuel pool pump are in service; two pumps may be aligned to one heat exchanger if desired. To date, no heat exchanger tubes have been plugged as a result of the cooling system's operation. Redundant fuel pool temperature and level sensors provide local readings and alarm indications in the Unit 2 control room. Fuel pool pumps and heat exchangers are located in the Fuel Handling Building but are not located in the vicinity of the fuel pool. Additional details on the fuel pool cooling system may be found in updated FSAR section 9.1.3.

3.1 4 In evaluating the capability of the fuel pool cooling system to handle the increased cooling load, four separate calculations were performed. These were selected to demonstrate that modeling of the pool cooling system and stored assemblies produced conservative results under a variety of conditions.

Sensitivity studies were performed for two of these calculations (cases 1 & 2) to quantify sources of conservatism in the methodology'.

Case 1 involved a benchmark of the computer code calculated results to fuel pool cooling system data collected in October, 1994. Actual discharged assembly burnup and enrichment information was used in this comparison.

Sensitivity studies were performed to evaluate the effects of the 2o uncertainty on power, evaporative cooling loss from the water surface, a best estimate heat transfer coefficient across the heat exchangers, and a combination of these three factors.

In Case 2, a comparison was performed between the results of the existing analysis of record for St. Lucie Unit 2 (which uses the NRC Auxiliary Systems Branch Technical Position 9-2) and the equivalent scenario using the method

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 7 of 29 documented in ANSI/ANS-5.1-1979. A sensitivity study was performed using the ANSI/ANS methodology to quantify the effect on fuel pool temperature of assumptions concerning burnup of the fuel offloaded from the core.

Case 3 calculated the fuel pool temperature resulting from the placement of a full core of offloaded fuel into the spent fuel pool 7 days after reactor shutdown with the fuel pool already containing 1300 previously discharged

. assemblies. This case used limiting values for batch average and full core discharge exposures.

Case 4 determined the fuel pool temperature resulting from the discharge of a refueling batch of 96 assemblies 5 days after reactor shutdown concurrent with an active failure in the fuel pool cooling system. The total fuel pool loading for this case was 1492 assemblies, thus precluding a full core offload.

For each case, the fuel pool water boiloff rate was also determined assuming a total loss of. fuel pool cooling, This rate was used to quantify the time available for system repairs or other remedial action prior to a decrease in fuel pool water level to a point 9 feet above the top of the fuel seated in the storage racks.

Subsequent to these calculations, Reference 10 evaluated the impact of the use of the ABB-CE "value-added" fuel design on this analysis.

3.1.5 Besults Case 1 demonstrated that the modeling scheme chosen to represent the fuel pool produced a temperature approximately 7 'F higher than that given by actual plant data. Calculations provided an expected fuel pool temperature of 92.8 'F; plant data provided a fuel pool temperature of 86.0 F. When best estimate evaporative losses, heat transfer coefficients, and the removal of the 2o 'ecay heat uncertainty were considered, the predicted fuel pool temperature decreased to 91.1 'F, thus demonstrating the conservative nature of the modeling of the fuel pool and discharged fuel. ANSI/ANS-5.1-1979 decay heat methodology was used throughout this calculation.

Case 2 demonstrated that the simplified ANSI/ANS-5.1-1979 decay heat methodology produces a higher calculated fuel pool temperature value than does NRC Branch Technical Position 9-2. This case analyzed a situation where the Unit 2 fuel pool contains 1113 discharged fuel assemblies, including a full

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 8 of 29 core discharge which has cooled for 7 days. Results give temperature values of 154.2 'F with two fuel pool pumps operating using the ANSI/ANS methodology as compared to a previously reported value of (150 'F.

Case 3 demonstrated that for a fuel pool containing 1517 assemblies, including 217 fuel bundles offloaded from the core 7 days following reactor shutdown, the maximum fuel pool water temperature was 170.9 'F with one fuel pump in operation and 154.9 'F with both spent fuel pumps in operation. The resultant heat load for this case was 35.22 E6 Btu/hr including uncertainties.

Actual discharge exposures were used for fuel placed in the spent fuel pool prior to June 1995; fuel discharges subsequent to this time were modeled using conservative cumulative exposure values including a value of 55,000 MWD/MTUfor each of the 217 assemblies offloaded from the core. Pursuant to the guidance in NUREG 0800, this calculation has demonstrated that no spent fuel pool bulk boiling occurs and thus, the criteria of NUREG 0800, Section 9.1.3 is satisfied:

The analyses discussed in Sections 3.1.4 and 3.1.5 of this evaluation contain a number of conservatisms when compared to the actual plant conditions that will exist following approval of the proposed license amendment (see above paragraphs). These conservatisms tend to increase the calculated maximum fuel pool temperature. The maximum spent fuel pool water temperature following any planned full core fuel offload at St. Lucie Unit 2 will be maintained s150 'F by the plant restrictions specified in Reference 12. This 150 'F value is consistent with the fuel pool temperature limit given in NUREG 0843 (St. Lucie Unit 2 SER) following a full core offload with two cooling pumps in operation.

For a total loss of fuel pool cooling, Case 3 provides the limiting fuel pool boil-off rate. For this case the boil-off rate was determined to be 73.3 gallons/minute. At this rate of boil-off, 37.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> are required for the pool water level to drop to 9 feet above the top of fuel seated in the storage racks.

Case 4 demonstrates that for a series of partial core offloads and using bounding parameters for assembly burnup, fission product generation, and fuel pool cooling system heat transfer, the resulting fuel pool water temperature remains less than 150 'F. The maximum heat load calculated for this case, including the effect of decay heat uncertainties, was 19.76 E6 Btu/hr. With allowance for active component failures (both a pump and heat exchanger assumed unavailable) the fuel pool water temperature was calculated to be

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St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 9 of 29 139.8 'F. Considering only the second heat exchanger to be unavailable (2 pumps feeding a single heat exchanger) the resulting pool water temperature is 130.8 'F. Both of these values are within the 150'F criterion specified in Section 9.1.3.3 of the updated.St. Lucie Unit 2 FSAR.

Reference 10 has determined that use of the value-added fuel design at St.

Lucie Unit 2 will have no adverse effect on the conclusions of'the fuel pool cooling analysis.

3.2 IVIAXIIVIUMFUEL CLADDING TEMPERATURE 3.2.1 Purpose.

It is important to ensure that fuel rod cladding integrity will be maintained under limiting conditions in the spent fuel pool environmerit. To do this, calculations were performed, using conservative inputs, to demonstrate that film boiling does not occur at the surface of the clad in the event of a.loss of forced flow cooling; i.e. the heat transfer coefficient remains within the range of nucleate boiling.

3.2 2 The maximum local heat flux at the fuel rod surface for an assembly discharged to the fuel pool 3 days after reactor shutdown has been calculated to be 1980.9 Btu/hr-ft'. The physics and geometry parameters used as input to this calculation (peaking factors, rod burnup and rod diameter) were selected to maximize the heat flux value and to bound both current and value-added fuel designs.

The calculation of peak cladding temperature in a fuel pool environment couples the maximum calculated surface heat flux with an empirical equation for free convection that assumes a constant cladding surface temperature. In this empirical equation fluid properties are evaluated at the saturation temperature. To ensure a conservative result when applying this equation, the axial position of the maximum decay heat flux is assumed to be located at the bottom of the spent fuel assembly. The saturation temperature at a pool depth corresponding to the bottom of the fuel assembly seated in the storage racks is 252 'F. For this condition, the empirical correlation used in the SFPOOL computer code produced a peak fuel cladding temperature of 309.2 'F.

f 1 St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment 10 of 29 'age As a check on the above calculated temperature value, the Rohsenow boiling correlation (Reference 7) was also used to calculate the peak cladding temperature. The cladding temperature calculated using this method is consistent with the 309.2 'F value reported above.

The temperature values calculated using the methods and conservative input described above provide assurance that fuel rod cladding will be maintained intact in 'the event of a loss of fuel pool cooling.

4.0 The following subsections describe the proposed new configuration of the St.

Lucie Unit 2 spent fuel pool, the methodology used to generate calculated values of reactivity and effective neutron multiplication to support this configuration, and the analysis results.

4 4.1 METHODOLOGY USED IN REACTIVITY CALCULATIONS Criticality calculations to support the proposed increase in the St. Lucie Unit 2 spent fuel pool storage capacity have been performed by ABB-CE using methodology consistent with that described in WCAP-14416 (Reference 3).

This Westinghouse Owners Group report was submitted to the NRC in July, 1995 and was supplemented in October, 1996. The analysis methodology used by ABB-CE in the evaluation of the spent fuel storage rack configuration employs: (1) SCALE-PC, a personal computer version of the SCALE-4.3 code package (which includes KENO-Va, NITAWL, CSAS-2 and BON-AMI), with the updated 44 group ENDF/B-5 cross section library, and; (2) the two-dimensional integral transport code DIT with an ENDF/B-6 neutron cross section library. A detailed discussion of the application of this criticality methodology may be found in Enclosure 1 to this evaluation. Both regions of the Unit 2 spent fuel pool will credit the presence of soluble boron; most analyses of the Unit 2 pool modeled both Region I and Region II explicitly in a single calculation.

4.2 ACCEPTANCE CRITERIA The SER issued by the NRC (Reference 2) for soluble boron credit methodology requires the application of a two part acceptance criteria to the St. Lucie Unit

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St. Lucie Unit 2 L-97,-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 11 of 29 2 spent fuel storage racks. First, the 95% probability 95% confidence (95/95) value of the effective neutron multiplication factor (k,<<) for the proposed spent fuel storage array must be less than 1.0 when analyzed with 0 ppm soluble boron, including the effect of all uncertainties and tolerances. Secondly, the acceptance criteria for fuel pool conditions with soluble boron present require that the 95/95 k,<<must be less than or equal to 0.95, including the effect of all.uncertainties and tolerances..

Reactivit'y calculations for the spent fuel racks also include the effect of two biases. Computer code biases based on the derived value of k, from KENO-Va compared against experimental benchmarks are applied. In addition, the reactivity effects of possible fuel pool temperature variations encountered during normal operation are included.

Enclosure 1 (and its references) also provides a summary of the comparisons made to experimental data that were used to derive the KENO-Va reactivity bias and uncertainty.

Both calculations performed at 0 ppm soluble boron and calculations which credit the effect of soluble boron on storage rack reactivity also include the effects of tolerances in fuel assembly enrichment, fabrication and positioning parameters, fuel rack construction tolerances, and uncertainties in the calculation of storage rack reactivity, control element assembly (CEA) worth and assembly burnup.

In addition to the boron concentration required to compensate for uncertainties and tolerances in calculations of k.<< for normal storage conditions, the amount of soluble boron required to compensate for postulated accident conditions is also quantified.

4.3 REGION I - DESCRIPTION OF STORAGE ARRANGEMENT ANALYZED The storage arrangement for Region I proposed by this license amendment preserves the capability to fully offload fuel from the Unit 2 reactor vessel by providing storage space for 276 fuel assemblies. The proposed Region I storage geometry is shown in Figure 9 of Enclosure 1; a discussion of the specific storage requirements for this region is summarized below. Region I continues to make use of flux traps to increase neutron leakage (and minimize k,<<) through placement of fuel next to regions of water. Two noteworthy differences between the proposed arrangement and the existing Region I

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St. Lucie Unit 2

~ L-97-325 Docket No. 50-389

~ Attachment 1 Proposed License'Amendment Page 12 of 29 storage arrangement are: (1) that the annular water region between the exterior of the rack array and the fuel pool wall serves as an explicitly-analyzed neutron sink, and (2) that a row of vacant, water-filled storage cells can serve to neutronically decouple regions of fresh fuel placed on either side.

As shown in Figure 9 of Enclosure 1, the U-shaped rows (or rings, beginning at the outside of the array and working inward) 1, 3 and 9 of Region I will hold 120 fuel bundles (and up to 56 full strength CEAs) in a high (89%) density array. The most reactive (or lowest burnup) fuel assemblies from the core offload will be placed in these three rings,. The U-shaped rows 5, 6, 8 and 11 are used to store the assemblies from the core offload with the greatest burnup (or lowest reactivity). Depending on their burnup, certain of these assemblies may be required to contain full strength CEAs. The additional storage locations in rows 8, 11, and the 3 by 10 array at the lower center of Figure 9 may be used to store permanently discharged fuel which does not meet the criteria for storage in Region II. The U-shaped rows 2, 4, 7 and 10 serve. as flux traps and will remain vacant.

In the rack criticality analysis, most fuel placed in rows 1, 3 and 9 is assumed to have an initial enrichment'of 4.5 w/o U"'; two assemblies each in rows 1 and 3 and one assembly in row 9 are assumed to have a 1.82 w/o U"'resh fuel equivalent. The analysis of fuel stored in rows 5, 6, 8, 11 and the 30 assembly center array also utilizes a conservative credit for the reactivity depletion of the offloaded fuel placed here. In all core offload scenarios applicable to St. Lucie Unit 2, some portion of the fuel will have accrued appreciable burnup and need not be analyzed as fresh fuel.

4,4 REGION II - DESCRIPTION OF STORAGE ARRANGEMENT ANALYZED The proposed fuel storage arrangement for Region II of the spent fuel pool increases the storage density of the region to 95.4% from the current value of 75%. All of Region II continues to require a minimum value of assembly burnup for storage. This required burnup value is a function of the initial fuel assembly enrichment and its decay (or cooling) time. The relationship between fuel cooling time and required burnup is primarily due to the 14.3 year half life of Pu"'. Over time, this fissile isotope decays to Am"', which is primarily a neutron. absorber. Decay of Pu'4'dds a significant amount of negative reactivity to the fuel pool.

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 13 of 29 The conservative initial neutron source distribution assumed for subcritical multiplication ensures that Region II controls the reactivity of the entire fuel pool. To ensure sufficient reactivity margin to the 1.0 k,<< limit, each Region II rack module retains 4 vacant interior cells, detailed in Figure 10 of Enclosure 1, which serve as flux traps. The proposed Region II storage arrangement also recognizes that, because of differences in neutron leakage, the burnup requirements for fuel storage in the interior of Region II are more restrictive than those for fuel storage in the outer row of Region II where an assembly faces a vacant Region I cell, or faces the water gap separating the storage racks from the fuel pool wall.

Spent fuel rack storage cells located in the interior of Region II require an assembly burnup equivalent to 1.3 w/o U"'resh fuel, prior to any credit for actinide decay, to permit fuel storage. Storage cells located on the periphery of Region II with at least one surface facing water require an assembly burnup equivalent to 1.5 w/o U"'resh fuel, prior to crediting actinide decay.

4.5 CALCULATIONALASSUMPTIONS AND RESULTS The assumptions listed below were used for St. Lucie Unit 2 in calculations of the k,<<applicable to the spent fuel pool storage racks:

Fuel assemblies contain uranium dioxide at the nominal enrichment over the entire length of each rod. The reactivity effect of a a0.05 w/o variation in UO, enrichment has been included in the reactivity

'olerances and uncertainties.

2. Fuel rods have been modeled considering the design characteristics of

. the ABB-CE "value-added". pellet, which is planned for insertion into Unit 2 for cycle 11. Modeling this design conservatively bounds the current fuel design relative to reactivity equivalencing. The reactivity effects of a 2% uncertainty in the fuel rod stack density have been considered; this uncertainty value is conservative compared to the observed history of variations in this parameter.

3. All fuel assemblies are assumed to contain 236 fuel rods in a 16x16 fuel rod lattice. Table 3 of Enclosure 1 tabulates the fuel parameters utilized in the fuel pool criticality analysis.

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 14 of 29

4. Tolerances due to uncertainties in the thickness of L-inserts (Region I) and rack storage module walls have been considered.

5. Tolerances due to uncertainties in positioning of fuel assemblies within the storage cells have been considered. For nominal calculations, fuel assemblies were assumed to be centered within each storage cell type.

Using infinite arrays of both Region I and Region II type cells, the reactivity effects of off-center assembly placement were examined.

6. The reactivity effects of variations in storage cell interior dimensions (ID) and cell pitch have been considered.

7. The moderator is water containing either 0 ppm (for comparison to 1.0 k, limit) or 350 ppm soluble boron (for comparison to 0.95 k,<< limit) at a temperature of 50 'F.

8, A 95/95 confidence level assessment of calculated CEA worth was developed for application to the Region I storage rack array. This value was based on CEA worth comparisons between predictions and measurements from operating reactors and comparisons between DIT and KENO-Va calculations of rod worth.

Using these assumptions, the KENO-Va model of the St. Lucie Unit 2 storage racks calculated a k,of 0.97001 for the 0 ppm soluble boron condition, prior to the application of any biases, tolerances or uncertainties. Including the effect of these factors, the resulting k,<<value is 0.99801. This value is less than the'k~ acceptance criteria value of 1.0 for 0 ppm conditions. Page 30 of Enclosure 1 provides a detailed tabulation of the reactivity effect for each bias or uncertainty.

Table 5 of Enclosure 1 provides a detailed accounting of the reactivity effect of each bias or uncertainty for the calculation of the storage rack k,<< in the presence of soluble boron. At 350 ppm, prior to application of any biases or uncertainties, k,<<was determined to be 0.91497. After application of all biases, tolerances and uncertainties, k,<<equals 0.94797. This value is less than the 0.95 acceptance criteria for fuel pool k,<< in the presence of soluble boron.

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 15 of 29 4.6 REACTIVITY EQUIVALENCING Reactivity equivalencing is used to define the conditions under which fresh, burned and shimmed fuel assemblies are interchangeable on an overall reactivity basis. At St. Lucie Unit 2, this strategy is used to translate the array of unshimmed fuel assemblies and their enrichments that have been demonstrated acceptable for the no soluble boron condition into an array of burned fuel assemblies with different initial enrichments, decay times, and burnable absorber concentrations.

4.6.1 Storage of fuel with higher enrichments than that identified as acceptable for the no soluble boron case relies on credit for the decrease in fuel assembly reactivity that results from reactor power operation. To derive a burnup credit curve, a series of reactivity calculations are performed to generate a set of initial enrichment/assembly burnup ordered pairs which all yield an equivalent k,when placed in the spent fuel storage racks. Any burnable poisons present in the fuel lattice may also be factored into the initial composition. Figures 11 through 15 of Enclosure 1 show the constant reactivity contours generated for Regions I and II of the St. Lucie Unit 2 spent fuel storage racks.

Uncertainties associated with burnup credit include an allowance for the uncertainty in the burned composition of a fuel assembly and a 5% allowance on the calculated fuel assembly burnup.

The effects on fuel assembly reactivity of axial burnup distributions have been considered in the development of isotopic concentrations for burned fuel assemblies. To maximize the conversion ratio and the reactivity of a depleted assembly, St. Lucie Unit 2 fuel assemblies have been burned using a conservatively hard neutron spectrum. Comparison of the reactivity of an assembly burned with this harder spectrum in the fuel pool rack lattice and an assembly depleted at actual Unit 2 operating conditions shows that the hard spectrum assembly is approximately 0.7% more reactive at end of life. Axial reactivity effects in depleted fuel assemblies are bounded by this spectral shift treatment.

Following its discharge from the reactor and the decay of short lived fission products, the reactivity of a burned fuel assembly will decrease due to the decay of actinides and long half-life fission products. The most important decay chain involves the decay of Pu"'nto Am"'. As noted previously, Pu"'

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 16 of 29, is a fissile isotope which contributes to positive reactivity at high burnup, whereas Am'4's primarily a neutron absorber. With a half-life of approximately 14 years, decay of Pu"'ver the duration of assembly storage in the fuel pool is significant and contributes to a reduction in bundle k,<< in the fuel pool environment. Credit for actinide decay is used to reduce the minimum burnup required to meet reactivity requirements. Table 6 of Enclosure 1 summarizes the decay time/required burnup ordered pairs as a function of fuel assembly initial enrichment out to a decay time of 20 years.

Section 6.0 of Enclosure 1 indicates that 170 ppm soluble boron is required to compensate for reactivity equivalencing methodologies used at St. Lucie Unit 2.

4.6.2 The criticality analysis described in this evaluation and Enclosure 1 was performed assuming that all fresh fuel contained no burnable absorbers and that the maximum fresh fuel enrichment is 4.5 w/o U"'. If the reactivity hold-down due to the presence of burnable absorbers is considered, then the fresh fuel enrichment can be increased above 4.5 w/o until the assembly reactivity matches that of an unshimmed, 4.5 w/o assembly.

When burnable poisons are required at St. Lucie Unit 2, fuel loading patterns typically utilize Gadolinium (Gd) loadings of 4 w/o or 6 w/o, with between 4 and 16 burnable absorber rods per assembly. Including an allowance for axial cutback of the Gd, the initial reactivity of a 5.0 w/o U"'ssembly with the lightest Gd loading used in Unit 2 (4 rods at 4 w/o) is equal to the reactivity of a fresh, unshimmed 4.5 w/o assembly. A fresh 5.0 w/o assembly containing any Gd shim loading above the minimum will be less reactive than a fresh, unshimmed 4.5 w/o assembly.

As the Gd depletes, the kof the shimmed, 5.0 w/o assembly will approach and eventually cross above the reactivity burndown of an unshimmed 4.5 w/o assembly. At exposures less than this crossover point the 5.0 w/o gadolinium

'assembly can replace any fresh 4.5 w/o assembly. At exposure values greater than this crossover point, burnup versus enrichment curves (Figures 11 through 15 of Enclosure 1) must be adjusted using Figures 16 and 17 of Enclosure 1 to determine the required assembly location in the spent fuel pool.

FPL is not requesting an increase in the spent fuel pool Technical Specification enrichment limit at this time.

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 17 of 29 4.7 POSTULATED ACCIDENTS The proposed storage configurations of both Region I and Region II racks at St.

Lucie Unit 2 have been examined to identify potential accidents that could result in an increase in the rack multiplication factor. Most accident conditions will not result in an increase in rack k,<<. For example, a fuel assembly drop accident that results in an assembly lying across the top of the storage modules will not result in any significant increase in k,<<of the system due to the large separation distance between the active region of fuel assemblies within the specified storage locations and the fuel assembly lying atop the modules. However, two accidents can be postulated that could increase reactivity beyond the analyzed condition: (1) a total loss of the fuel pool cooling system or, (2) the misload of an assembly into a cell for which restrictions on burnup, enrichment or location are not satisfied.

For an occurrence of either of these postulated accident conditions, the double contingency principle of ANSI/ANS 8.1-1983 can be applied. This states that it is not necessary to assume two unlikely, independent and concurrent events to ensure protection against a criticality accident. Thus, for these postulated accident conditions, the presence of additional soluble boron in the fuel pool water (above the concentration required to ensure 95/95 conditions and burnup credit) can be assumed as a realistic initial condition because not assuming its presence would represent a second unlikely event.

The total loss of fuel pool cooling has the potential of increasing the temperature of the pool coolant to boiling conditions. Calculations performed for both 0 ppm and 350 ppm conditions showed that the change in kwas less than 0.0040 for the 0 ppm case and 0.0066 for the 350 ppm case. These reactivity values reflect an increase in fuel pool temperature from 155 'F to 240 'F.

A variety of scenarios were examined to assess the consequences of a postulated fuel assembly misload event. Each scenario involved the misplacement of a fresh, unrodded and unshimmed 4.5 w/o fuel assembly.

Three types of misload positions were identified for this assembly: a misload into a position reserved for a 4.5 w/o fresh fuel assembly containing a CEA; a misload into a position designated for a highly burned (1.3 w/o fresh fuel equivalent) fuel assembly; and a misload into selected water cell locations. The largest bk observed for any of the postulated assembly misloads was 0.1016 for a type 3 misload. Type 1 assembly misloads resulted in hk values less than

St. Lucie Unit 2 L-97-325 8o. 50-389'ocket Attachment 1 Proposed License Amendment Page 18 of 29 50% as large as a type 3 misload; type 2 misloads generated h,k values approximately:75% as large as the type 3 misload.

A comparison of the reactivity values presented above demonstrates that the potential increase in k,<<due to a misloaded fuel assembly'is substantially greater than the increase in k,due to a loss of all fuel pool cooling. The boron concentration required to compensate for the >10% increase in k,<<due to the limiting assembly misplacement has been determined to be 746 ppm.

4.8 CRITICALITYANALYSIS - CONCLUSIONS Section 6 of Enclosure 1 summarizes the fuel pool soluble boron requirements for tolerances and uncertainties, reactivity equivalencing and postulated accidents.= The sum of these requirements totals 1266 ppm.

St. Lucie Unit 2 Technical Specification 5.6.1 requires that the spent fuel pool contain at least 1720 ppm soluble boron at all times. This Technical Specification requirement is greater than the total fuel pool soluble boron requirement from Enclosure 1.. Thus, current Technical Specifications will ensure that k,<<of the proposed spent fuel pool storage configuration will be maintained (0.95 in the presence of the most adverse assembly misload event.

As summarized in Section 4.5 of this evaluation, the spent fuel storage rack array was determined to remain subcritical with 0 ppm soluble boron at a 95/95 probability/confidence level, considering the effect of all applicable biases and uncertainties. In the presence of 350 ppm soluble boron the 95/95 k,of this array was determined to be (0.95, including applicable biases and uncertainties. Thus, the proposed spent fuel pool storage array conforms with acceptance criteria provided in Reference 2.

5.0 The St. Lucie Unit 2 spent fuel storage racks and fuel pool structure are designed to withstand forces generated by normal plant operation as well as those forces generated during a seismic event. Except for the removal of certain storage cell blocking devices, the proposed license amendment does not involve any change to the existing storage racks. The analyses supporting

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page.19 of 29 the use of these storage racks at Unit 2 were developed as part of a pool reracking license amendment submitted to the NRC in 1984 (Reference 4); their validity was independently confirmed by the NRC (Reference 5). These analyses have been examined to determine if the'change in storage capacity proposed here would adversely impact their conclusions or result in an unanalyzed condition. The calculational review is formally documented in the "QA Review of St. Lucie Unit 2 Spent Fuel Pool Capacity Increase," ABB Combustion Engineering Nuclear Operations Design Analysis Number: A-SL2-FE-0064, Rev. 02, 6/12/95, and is available from the plant records.

The 1984 analyses considered partial loadings of the spent fuel rack consistent with the present Region I and Region II storage configurations. These analyses also considered the complete loading of all storage locations within the fuel racks without regard to the Region I or Region II storage limitations. The 1984 analyses utilized a "consolidated" fuel weight (approximately equal to twice the weight of a single fuel assembly) in each storage rack location to conservatively estimate the resulting loads on the spent fuel rack structure and the fuel pool floor. Therefore, the assumed weight per storage cell location, and the resulting structural and seismic analyses of record are conservative c'ompared to the results that would be obtained for a single fuel assembly.

As discussed in Section 3.1.5 of this'evaluation, a conservative calculation of the maximum fuel pool temperature resulting from a complete off-loading of the reactor core produced a water temperature (T.) of approximately 155 'F with two cooling pumps in operation. Section 4.4 of Reference 4 defines T. as the highest temperature associated with the postulated abnormal design conditions. This newly calculated value of T. is greater than the 150 'F value for T. used in the 1984 seismic analysis. The impact of this 5 'F temperature increase on storage rack stresses was examined using Section III of the 'f983 ASME code. To ensure bounding results, rack stresses were evaluated assuming a fuel pool temperature of 300 'F. The results of this evaluation are given in Table 1 of this evaluation. These results show that for the plates and support bars that comprise the spent fuel pool racks, stress intensities are less than allowable values for both normal and faulted conditions at 300 'F.

The evaluation of a higher T. value was performed to ensure acceptable rack stresses under worst conditions. However, as noted previously actual fuel pool temperatures during core offload evolutions will be limited to a maximum of 150 'F; therefore the previous analysis of record for the racks and the fuel pool structure remains bounding.

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 20 of 29 The results of this review demonstrate that the St. Lucie Unit 2 spent fuel racks and fuel pool floor are qualified for the increased storage capacity proposed in this license amendment.

6.0 As discussed in Sections 4.5 and 4.6, the revised fuel pool criticality analysis credits the presence of 520 ppm soluble boron to ensure that k,<< for the new storage configuration remains ~0.95 including the effects of uncertainties, biases and reactivity equivalencing. Because credit for fuel pool soluble boron is assumed, it is necessary to identify the plant systems interfacing with the spent fuel pool that could, through a malfunction or. operator error, credibly initiate a dilution event. This dilution analysis has been performed to ensure that sufficient time remains available to detect and mitigate a dilution event before the spent fuel pool criticality analysis design basis value of k.<< ~0.95 is violated.

Enclosure 2 to this evaluation contains additional details on the evaluation of these interfacing systems, including a quantification of the time required for the loss of reactivity margin to k,<< 0.95.

6.1 DESCRIPTION

OF METHODOLOGY The boron dilution analysis performed for St. Lucie Unit 2 includes an evaluation of the following plant-specific features:

Dilution Sources Boration Sources Fuel Pool Instrumentation Fuel Pool Related Plant Procedures

St. Lucie Unit 2 L-97-325 Docket No: 50-389 Attachment 1 Proposed License Amendment Page 21 of 29 Piping Impact of a Loss of Offsite Power Boron Dilution Initiating Events Boron Dilution Times and Volumes Based on a review of the systems that interface with the spent fuel pool, each potential dilution path was identified. Next, the activities required to change each potential dilution path into an actual fuel pool dilution path were determined; this determination included identifying the plant procedure(s) that controlled each evolution. The quantity of makeup available to the fuel pool through each dilution pathway was determined and compared to the quantity of unborated water required to'dilute the fuel pool to a k,<< 0.95. An upper bound flow rate through each dilution pathway was determined.

Available sources of borated makeup to the fuel pool were also identified.

For each dilution path, the time required to reach the fuel pool design value of k,<<was compared to the frequency of fuel pool boron sampling and the frequency of operator rounds in the vicinity of the spent fuel pool. Any local or control'room indications that an inadvertent dilution might be in progress were also identified.

The effect of a potential loss of offsite power on fuel pool dilution and boration pathways was identified.

6.2 BORON DILUTION INITIATINGEVENTS The initial screening of fuel pool dilution pathways identified six potential dilution scenarios requiring additional review. These are:

Primary Water System makeup through valve V-15322 Primary Water System makeup through valve V-15538 Primary Water addition through resin flush line

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 22 of 29 Primary Water addition through local fire hose station Precipitation event through open FHB L-shaped Door Dilutions resulting from Seismic events or random pipe breaks Subsequently, FPL evaluated each of the potential event initiators in greater detail to determine whether each initiator posed a credible challenge to fuel pool design reactivity margin.

6.3 RESULTS OF INITIATINGEVENTS Following a detailed review two potential dilution paths had characteristics that warranted consideration as a potential challenge to fuel pool reactivity margins.

For these pathways, the time required to achieve a dilution such that k,<< =

0.95 was quantified.

One pathway examined was the procedurally-specified makeup flow path through valve V-15538. Assuming an initial fuel pool boron concentration equal to the Technical Specification limit of 1720 ppm, more than 79 hours9.143519e-4 days <br />0.0219 hours <br />1.306217e-4 weeks <br />3.00595e-5 months <br /> would be required to dilute the pool to a k,<<of 0.95 using this flowpath.

Assuming an inadvertent dilution of the fuel pool through the resin flush line, approximately 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> is required to reduce the pool boron concentration from an initial value of 1720 ppm to a value such that fuel pool k.<< 0.95. As discussed in Enclosure 2, this quantity of makeup, through any flow path without a coincident letdown flow, would result in the overflow of the fuel pool. This overflow onto the fuel pool operating deck would be readily observed by operations personnel during their on-shift rounds; at least seven sets of operator rounds would be made during the time this dilution was in progress.

6.4 SPENT FUEL POOL DILUTION EVENT CONCLUSIONS The boron dilution analysis of the St. Lucie Unit 2 spent fuel pool discussed in Enclosure 2 of this evaluation has concluded that an unplanned or inadvertent dilution of the fuel pool boron concentration from 1720 ppm to conditions such

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 23 of 29 that k,<< = 0.95 is not a credible event. This conclusion is based on the following:

'More than 358,900 gallons of unborated water is required to dilute the Unit 2 spent fuel pool to the design k,<<value of 0.95. To actually achieve this dilution,'lant personnel would be required to take continued, manual actions to assure this quantity of water would be delivered to the spent fuel pool.

2. The normal makeup path to the spent fuel pool from the primary water system (V-15538) is maintained locked closed. The alternate primary water makeup path is capped.

3. In-place administrative controls on the primary letdown path from the spent fuel pool (the return line to the RWT) ensure that any prolonged, inadvertent fuel pool makeup would result in pool overflow.

4. The large volume of water required to achieve this dilution would be readily detected by plant personnel through installed alarms, overflow of the spent fuel pool and flooding in the fuel handling building, or by operations personnel on their normal rounds on the spent fuel pool operating deck and elsewhere in the plant.

5. Available flow rates to deliver unborated water to the spent fuel pool ensure that sufficient time is available for operations personnel to detect and respond to any dilution event.

7.0 FPL has reviewed the environmental impacts of the proposed license amendment. This review 'demonstrates that the overall radiological and nonradiological impacts of the proposal are insignificant. The review is summarized below.

7.1 THERMAL IMPACT The thermal analysis of the effect of the proposed change on the spent fuel pool cooling system is presented in Section 3.1.5 of this evaluation. That analysis included a determination of the maximum spent fuel decay heat load following a partial core offload and a full core offload. The proposed increase

St. Lucle Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 24 of 29 in storage capacity will change the maximum decay heat load for a partial core offload from 16.9 E6 Btu/hr to 19.76 E6 Btu/hr and for full core offload conditions from 31.7 E6 Btu/hr to 35.22 E6 Btu/hr. This increased heat loa'd results in an increase of approximately 3 'F in the maximum fuel pool water temperature for the partial core offload case, and an increase of approximately 5 'F in water temperature for storage of the limiting full core offload (note:

maximum fuel pool temperature will be maintained s150 'F). Because the evaporation rate from the pool is assumed to be zero, the increased decay heat load is also the increased load on the cooling system and the increased heat rejected to the environment. The total heat load rejected to the environment by St. Lucie Unit 2 is about 6.2 E9 Btu/hr. The percentage increase in the heat rejected to the environment due to the increase in spent fuel storage capacity is on the order of 0.05% for partial core discharges and 0.06% for fuel storage following a full core offload.

7.2 RADIOLOGICAL EVALUATION 0 7.2.1 The net effect of increasing the St. Lucie Unit 2 spent fuel pool storage capacity is that older fuel elements will be retained in wet storage beyond the time when they would have otherwise been loaded into casks for dry storage on-site. The concentration of radionuclides in the spent fuel pool is controlled by the actions of the fuel pool purification system and by the decay of short-lived radioactive isotopes. Most of the contamination collected by the fuel pool purification system originates either from discharged fuel freshly emplaced in the fuel pool or from the intermixing of spent fuel pool water with primary water during refueling evolutions. Retaining already-aged fuel in wet storage for an extended period will not appreciably increase the activity in the fuel pool water or the amount of solid radioactive waste which must be disposed of because the short-lived isotopes associated with these fuel bundles will have had an opportunity,to decay. Therefore, increasing the fuel pool storage capacity as proposed for St. Lucie Unit 2 will have no significant effect on the quantity of radioactive waste collected.

7.2.2 Storage of additional quantities of long decayed discharged fuel in the spent fuel pool will not significantly increase the release of gaseous fission products

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 25 of 29 such as Kr". Fuel rod integrity at St. Lucie Unit 2 has been very good, with most fuel cycles evidencing no leaking fuel rods. Additionally, the rod pressure which tends to act as driving force for fission product release, is substantially decreased after long periods of fuel cooling.

7.2.3 The proposed license amendment does not involve any changes to the method of operating or range of motion of the spent fuel cask handling crane. No movement of loads in excess of the nominal weight of a fuel assembly, CEA and associated handling tool is permitted over other fuel assemblies in the storage pool. Protection against dropping the spent fuel cask into the spent fuel storage pool is provided by the basic layout of the Fuel Handling Building.

As noted in UFSAR Section 9.1.4.3.2, additional protection is afforded by the trolley bumpers and a set of limit switches which work together with bridge and trolley brakes to prevent movement of the crane hook into the restricted area.

The proposed amendment will also not involve any changes in the mode of operating or range of motion of the spent fuel handling machine. Changes in fuel assembly weight due to the use of value-added fuel have been evaluated and determined to be acceptable (Reference 8). As noted in Reference 11, during movement of a fuel assembly, the load on the hoist cable is monitored to ensure that movement is not restricted. Installed interlocks will continue to restrict movement of the handling machine when the hoist is withdrawing or inserting an assembly.

The existing analyses of record pertaining to the radiological consequences of a fuel handling accident within the Fuel Handling Building (FHB) and the postulated drop of a spent fuel cask just outside the FHB have been examined to assess the impact of the proposed license amendment, including the use of the value-added fuel pellet design. The review is formally documented in the "QA Review of St. Lucie Unit 2 Spent Fuel Pool Capacity Increase," ABB Combustion Engineering Nuclear Operations Design Analysis Number: A-SL2-FE-0064, Rev. 02, 6/12/95, and in Reference 10, both of which are available from FPL Nuclear Engineering records.

The assumptions and parameters previously employed in evaluating the fuel mishandling accident were consistent with Regulatory Guides 1.13 and 1,25.

The previously analyzed consequences of dropping a spent fuel cask were

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 26 of 29 based on the guidelines provided in Section 15.7.5 of the Standard Review Plan.

Based on the results of two ORIGEN-II assembly depletions, FPL has concluded that the gap activities resulting from the use of value-added fuel are essentially identical to those resulting from the use of the standard pellet design. FPL's review of the existing analysis of the fuel handling accident has concluded that the gap activities provided in the analysis of record for the fuel handling accident conservatively bound those values expected to occur at assembly discharge burnups of up to 60,000 MWD/MTU (Reference 1). As defined by Section 15.7.4 of the Standard Review Plan, calculated dose values are well within the guidelines if the calculated whole body dose is s6 rem and the

,calculated thyroid dose is s75 rem. As indicated in Table 5-9 of Reference 4, these guideline dose values were easily achieved.

FPL has also examined the existing analysis of an accident involving the drop of a spent fuel cask containing 10 irradiated fuel assemblies. This review has determined that conservative input assumptions were used and that the results of the existing analysis as shown in Table 5-6 of Reference 4 are well within the acceptance criteria for a Limiting Fault-2 event.

Increasing the storage capacity of the St. Lucie Unit 2 spent fuel pool as described in this proposed license amendment will have no effect on the radiological consequences of an assumed fuel mishandling event or on the consequences of the drop of a loaded spent fuel cask. For each of these events, the calculated doses are small relative to the guideline values.

8.0 The impact of the proposed increase in St. Lucie Unit 2 spent fuel storage capacity and the implications of the use of reactivity credit for fuel pool soluble boron have been examined in the above discussion. Each of the impacts of the proposed change has been quantified and determined to be within acceptable limits by comparison to established acceptance criteria.

Based on this examination, FPL has determined that the proposed changes to St. Lucie Unit 2 Technical Specifications do not constitute a significant hazards determination (see Attachment 2).

~ i St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 27 of 29 9.0 Safety Evaluation by the Office of Nuclear Reactor Regulation, Facility Operating License No. NPF-16; Amendment 21, May 29, 1987.

2. Safety Evaluation by the Office of Nuclear Reactor Regulation Relating to Topical C - 8-0

3. W CAP-14416-NP-A; EastinghnusW Nlethndnlngy, Revision 1, Westinghouse Electric Corporation, November 1996.

4. FPL letter L-84-47(and attachments), J. W. Williams, Jr. to Darrell G. Eisenhut, St. Lucie Unit No. 2 Docket No. 50-389, Pra 30.~

88

5. C . 3- -

Analysis, Florida Power and Ught Company, St. Lucia Generating Station Unit 2, prepared by, Franklin Research Center, September 19, 1984.

~

6. ABB CENO Computer Program, SFPOOL Version 1, Verification and Validation Report No. 00000-AS95-CC-010, Rev. 00, June 1, 1995. (ABB-Combustion Engineering Nuclear Operations) 7.

369 - 373.

8. i i 0.

'gal:hanges. (Available from FPL Nuclear Engineering Records)

'f

9. St. Lucie Unit 2 Updated FSAR, through Amendment 10.

10. ABB-CE letter F2-97-149, R.J.'Land to R.J. Rodriguez (FPL), Dis October 15, 1997. (Available from FPL Nuclear Engineering records) 0

12. Safety Evaluation' PSL-ENG-SENS-97-006, Revision 1, JEJnads, 4-4-97. (Available from FPL Nuclear Engineering records)

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 28 of 29 SEP ~IMag~pacit~oluble J3oraxxCxedif.

Summary of St. Lucie Unit 2 Calculated Fuel Storage Rack Stress Intensities A. Cam Normal Operation Faulted Operation Condition Stress Intensity Allowable Stress' Stress Intensity Allowable Stress'

( si) si) ( si) si)

Pm 19,713 20,000 28.056 30,000 Pm+ Pb 29.670 30,000 33,262 45,000 Pm+ Pb+ Pe 49,414 60,000 N/A N/A Allowable Stress based on a temperature of 300 'F. Note that allowable stress intensity Sm is 20,000 psi at both 200 'F and 300 'F.

B. Com sembly.

'onJIL1983)

At 300'F Sy or yield strength = 22,500 psi. The allowable stress for normal operation is 13,500 psi (0.6 Sy); this is less than the allowable stress for the faulted condition (1.2 "Sy). The faulted condition stress has been calculated to be 4965 psi. Therefore, the spent fuel racks will meet allowable stresses with SFP water temperatures of 300 'F.

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment t Proposed License Amendment Page 29 of 29 SF~~~apacity SatubleJ3ararxCredit Xab~

St. I ucie Unit 2 Estimated Spent Fuel Pool Capacity Requirements Cycle ff Approximate Total Number Spaces Requirod Total Number of Excess Storago Cycle Startup of Assemblies for Full Core Spaces Needed Available Date in Pool from all Reserve During this Cycle Previous Cycles Existing'ew Ca eci't Caoaci 1/5/96 628 217 845 231 N/A 10 5/26/97 692 217 909 167 NIA 1 2/19/98 764 217 981 95 379 12 5/27/00 836 217 1053 23 307 13 12/19/01 908 217 1125 0 235 14 5/27/03 980 217 1197 0 163 15 1 2/19/04 1052 217 1269 0 91 16 5/27/06 1124 217 1341 19 17 12/19/07 1196 217 1413

'icensed Capacity = 1076 assemblies

'roposed Licensed Capacity = 1360 assemblies

St. Lucie Unit 2 Docket No. 50-389 Proposed License Amendment SEEZMM DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION Adapted from FPL Nuclear Engineering Safety Evaluation PSL-ENG-SENS-97-083, Revision 0, 12/17/97.

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 1 of 7 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION Description of amendment request: The amendment will amend Technical Specification 5.6.1 and associated Figure 5.6-1, and Specification 5.6.3, to permit an increase in the allowed Spent Fuel Pool (SFP) storage capacity. The analyses supporting this request, in part, assume credit for up to 1266 ppm boron concentration existing in the SFP.

Pursuant to 10 CFR 50.92, a determjnation may be made that a proposed license amendment involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

FPL has determined that the activities associated with this proposed license amendment do not meet any of the significant hazards consideration standards of 10 CFR 50.92(c) and, therefore a no significant hazards consideration finding is justified.

In support of this determination, the following background information is provided, followed by a discussion of each of the significant hazards consideration standards presented above.

St. Lucie Unit 2 has a single spent fuel pool with a total of 1584 storage cell locations in 2 distinct fuel pool storage rack regions. Region I of the Unit 2 spent fuel pool storage racks contains 448 storage cells on an 8.965 inch pitch. Presently, 50% (or 224) of these storage cells are available to store fuel with an initial enrichment of 4.5% U"'r less; the remaining vacant storage cells are used as flux traps to control reactivity. Region II of the spent fuel pool storage racks contains 1136 storage cells on an 8.965 inch pitch of which 75% (or 852) are currently usable. As of May, 1997, five permanently discharged assemblies are stored in Region I; 687 discharged fuel bundles are stored in Region II.

With the present limitations on storage capacity in the Unit 2 spent fuel pool and the existing inventory of discharged assemblies awaiting shipment offsite to a Department of Energy (DOE) facility, St. Lucie Unit 2 will lose the ability to fully offload the reactor core to the fuel pool in year 2001; it will lose the ability to discharge any spent fuel at all in approximately 2007. Therefore, to ensure that sufficient storage capacity continues to exist for discharged fuel, FPL has performed analyses to

St. Lucie.Unit 2 L-97-325 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 2 of 7 support an increase in the utilization of the existing spent fuel storage racks at St.

Lucie Unit 2. These new analyses support the storage of up to 1360 fuel assemblies in the spent fuel pool, including the presence of 217 assemblies resulting from a complete offload of the St. Lucie Unit 2 reactor core. The additional storage cells made available by this proposed license amendment will permit a 6 year deferral in the need for on-site dry storage of discharged fuel at St. Lucie.

Deferring the requirement for on-site dry storage at St. Lucie Unit 2 allows additional time for the full commercialization of multi-purpose canister (MPC) technology prior to the selection of a specific cask system.

With this license amendment request, FPL proposes to modify the requirements of Sections 5.6.1 and 5.6.3 of the St. Lucie Unit 2 Technical Specification Design Eeatures to describe the revised fuel storage configuration in the spent fuel pool and to reflect the maximum storage capacity of the revised configuration.

The following evaluation demonstrates that the proposed license amendment involves no significant hazards considerations. Reference is made to the discussion contained in the attached safety evaluation.

Analyses to support the proposed fuel pool*capacity increase have been developed using conservative methodology.'he analysis of the potential accidents summarized below has shown that there is no significant increase in the consequences of any accident previously analyzed. A review of relevant plant operations has also demonstrated that there is no significant increase in the probability of occurrence of any accident previously analyzed. This conclusion is also discussed below.

Previously evaluated accidents that were 'examined for this proposed license amendment include: Fuel Handling Accident, Spent Fuel Cask Drop Accident, and Loss of all Fuel Pool Cooling.

There will be no change in the mode of plant operation or in the availability of plant systems as a result of this proposed change; the systems interfacing with the spent fuel pool have previously encountered borated pool water and,are designed to interact

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 3 of 7 with irradiated spent fuel and remove the residual heat load generated by isotopic decay. The proposed amendment does not require a change in the maintenance interval or maintenance scope for the fuel pool cooling system or for the spent fuel cask crane. The frequency of cask handling operations and the maximum weight carried by the crane is not increased as a result of the proposed license amendment.

Thus, there will be no increase in the probability of a loss of fuel pool cooling or in the probability of a failure of the cask crane as a result of the proposed amendment.

There will not be a significant increase in the frequency of handling discharged assemblies in the fuel pool as a result of this change; any handling of fuel in the spent fuel pool will continue to be performed in borated water..lf the license amendment is approved, there will be a one-time. repositioning of certain discharged assemblies stored in the fuel pool to comply with the revised positioning requirements, but the increased pool storage capacity will permit the deferral of 'spent fuel handling associated with cask loading operations. Fuel manipulation during the repositioning activity will be performed in the same manner as for fuel placed in the spent fuel pool during refueling outages. There will be no changes in the manner of handling fuel discharged from the core as a result of refueling; administrative controls will continue.

to be used to specify fuel assembly placement requirements. The relative positions of Region I and Region Il storage locations will remain the same within the fuel pool.

Therefore, the probability of a fuel handling accident has not been significantly increased.

The consequences of a fuel handling accident have been evaluated. The radioactive release consequences of a dropped fuel assembly are not affected by the proposed increase in fuel pool storage capacity. They remain bounded by the results of calculations performed to justify the existing St. Lucie Unit 2 fuel storage racks and burnup limits. At the limiting fuel assembly burnup, radioactive releases from a.

dropped assembly would be only a small fraction of NRC guidelines. The input parameters employed in analyzing this event are consistent with the current values of fuel enrichment, discharge burnup and uranium content used at St. Lucie Unit 2 and with future use of the "value-added" fuel pellet design. Thus, the consequences of the fuel assembly drop accident would not be significantly increased from those previously evaluated.

The capability of the fuel pool cooling system to handle the increased number of discharged assemblies has been examined. The impact of a total loss of spent fuel pool cooling flow on available equipment recovery time and on fuel cladding integrity has also been evaluated. For the limiting full core discharge, sufficient time remains available to restore cooling flow or to provide an alternate makeup source before

St. Lucie Unit 2 L-97-325 Docket No. 50-389. Attachment 2 Proposed License Amendment Page 4 of 7 boiloff results in a fuel pool water level less than that needed to maintain acceptable radiation dose levels.. Analysis has shown that in the event of a total loss of fuel pool cooling fuel cladding integrity is maintained. Therefore, the consequences of a loss of fuel pool cooling event, including the effect of the proposed increase in fuel pool storage capacity, have not been significantly increased from previously analyzed results for this type of accident.

The analysis of record pertaining to the radiological consequences of the hypothetical drop of a loaded spent fuel cask just outside the Fuel Handling Building was examined to determine the impact of the increased fuel storage capacity on this accident's results. The results of the previously performed analysis were determined to bound the conditions described by the proposed license amendment, thus the consequences of the cask drop accident would not be significantly increased as a result. of this change.

It is concluded that the proposed amendment to increase the storage capacity of the St. Lucie Unit 2 spent fuel pool will not involve a significant increase in the probability or consequences of any accident previously evaluated.

2.

In this license amendment FPL proposes to credit the negative reactivity associated with a portion of the soluble boron present in the spent fuel pool. Soluble boron has always been present in the St. Lucie Unit 2 spent fuel pool; as such the possibility of an inadvertent fuel pool dilution has always existed. However, the spent fuel pool dilution analysis demonstrates that a dilution of the Unit 2 spent fuel pool which could increase the pool k,<< to greater than 0.95 is not a credible event. Neither implementation of credit for the reactivity of fuel pool soluble boron nor the proposed increase in the fuel pool storage capacity will create the possibility of a new or different type of accident at St. Lucie Unit 2, If An examination of the limiting fuel assembly misload has determined that this would not represent a new or different type of accident. None of the other accidents examined as a part of this license submittal represent a new or different type of accident; each of these situations has been previously analyzed and determined to produce acceptable results.

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 5 of.7 The proposed license amendment will not result in any other changes in the mode of spent fuel pool operation at St. Lucie Unit 2 or in the method of handling irradiated nuclear fuel. The spatial relationship between the fuel storage racks and the cask crane range of motion is not affected by the proposed change.

As a result of the evaluation and supporting analyses, FPL has determined that the proposed fuel pool capacity increase does not create the possibility of a new or different type of accident from any accident previously evaluated.

3. Eh~ra

~ately.

FPL has determined, based on the nature of the proposed license amendment that the issue of margin of safety, when applied to this fuel pool capacity increase, should

~

address the following areas:

Fuel Pool reactivity considerations

~

1.

~

2. Fuel Pool boron dilution considerations

3. Thermal-Hydraulic considerations

4. Structural loading and seismic considerations The Technical Specification changes proposed by this license amendment, the proposed spent fuel pool storage configuration and the existing Technical Specification limits on fuel pool soluble boron concentration provide sufficient safety margin to ensure that the array of fuel assemblies stored in the spent fuel pool will always'remain subcritical. The revised spent fuel storage configuration is based on a Unit 2 specific criticality analysis performed using methodology consistent with that approved by the NRC. Additionally, the soluble boron concentration required by current Technical Specifications ensures that the fuel pool k,<<will be always be maintained substantially less than 0.95.

The Unit 2 criticality analysis established that the k,<<of the spent fuel pool storage racks will be (1.0 with no soluble boron in the fuel pool water, including the effect of all uncertainties and tolerances.. Credit for the soluble boron actually present is used to.offset uncertainties, tolerances, off-normal conditions and to provide margin such that the spent fuel pool k,<< is maintained s 0.95. FPL has also demonstrated

St. Lucie Unit 2 L-97.-325 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 6 of 7 that a decrease in the fuel pool boron concentration such that k,<<exceeds 0.95 is not a credible event.

Current Technical Specifications require that the fuel pool boron concentration be maintained >1720 ppm. This boron value is substantially in excess of the 520 ppm required by the uncertainty and reactivity equivalencing analyses discussed in this evaluation and the 1266 ppm value required to maintain k,<<<0.95 in the presence of the most adverse mispositioned fuel assembly.

The St. Lucie Unit 2 fuel pool boron concentration will continue to be maintained significantly in excess of 1266 ppm; the proposed license amendment will not result in changes in the mode of operation of the refueling water tank (RWT) or in its use for makeup to the fuel pool. Thus, operation of the spent fuel pool following the proposed change, combined with the existing fuel pool boron concentration Technical Specification limit of 1720 ppm, will continue to ensure that k,<<of. the fuel pool will be substantially less than 0.95.

Even if this not-credible dilution event was to occur, no radiation would be released; the only consequence would be a reduction of shutdown margin in the fuel pool. The volume'of unborated water required to dilute the fuel pool to a k,<<of 0.95 is so large (in excess of 358,9GG gallons to dilute the fuel pool to 520 ppm boron) that only a limited number of water sources could be considered potential dilution sources. The likelihood that this level of water use could remain undetected by plant personnel is extremely remote.

In meeting the acceptance criteria for fuel pool reactivity, the proposed amendment to increase the storage capacity of the existing fuel pool racks does not involve a significant reduction in the margin of safety for nuclear criticality.

Calculations of the spent fuel pool heat load with an increased fuel pool inventory were performed using ANSI/ANS-5.1-1979 methodology. This method was demonstrated to produce conservative results through benchmarking to actual St.

Lucie Unit 2 fuel pool conditions and by comparison of its results to those generated by a calculation using Auxiliary Systems Branch Technical Position 9-2 methodology.

Conservative methods were also used to demonstrate fuel cladding integrity is maintained in the absence of cooling system forced flow. The results of these calculations demonstrate that, for the limiting case, the existing fuel pool cooling system ca'n maintain fuel pool conditions within acceptable limits with the increased inventory of discharged assemblies. Therefore, the proposed change does not result

St. Lucie Unit 2 L-97-325 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 7 of 7 in a significant reduction in the margin of safety with respect to thermal-hydraulic or spent fuel cooling considerations.

The primary safety function of the spent fuel pool and the fuel storage racks is to maintain discharged fuel assemblies in a safe configuration for all environments and abnormal loadings, such as an earthquake, a loss of pool cooling or a drop of a spent fuel assembly during routine spent fuel handling. The proposed increase in spent fuel inventory on the fuel pool and the existing storage racks have been evaluated and show that relevant criteria for fuel rack stresses and floor loadings have been met and that there has been no significant reduction in the margin of safety for these criteria.

To'summarize, it has been shown that the proposed increase in capacity of the existing St. Lucie Unit 2 spent fuel pool storage racks and the proposed Technical Specification changes do not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. Involve a significant reduction in a margin of safety.

Therefore, FPL has determined that the proposed license amendment involves no significant hazards considerations.