ML17221A677
| ML17221A677 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 03/11/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17221A675 | List: |
| References | |
| NUDOCS 8803210483 | |
| Download: ML17221A677 (30) | |
Text
SAFFT~
EVALUATION THE OFFTCE OF NUCLEAR PEACTOR REGULATION RELATING TO TEE RFRACKING OF THE SPENT FUEL POOL AT THF. ST.
LUCRE PLANT, UNIT NO.
1 AS RFLATEP TO AMENnMENT'O.'9'1 TO UNIT 1 FACILITY OPERATING LICENSE NO.
DPR-67 FLORIDA POWER AND LIGHT COMPANY DOCKET Nl. 50-335 8803210483 8>OO0335 PDR ADOCK 0 P
P
1.
INTRODUCTION TABLE OF CONTENTS PAl'E
- 1. 1 Licensee Submitta'.
and Staff Review 1.2 Summarv Description of Reracking CRITICALITY CONSIDERATIONS 2.1 Criticality Analysis 2.P Technical Specification Changes 2.3 Conclusions 3.
MATERIAL COMPATIBILITY AND CHEMICAL STABILITY 4.
STRUCTI<PAL DESIGN 5.
SPENT FUFL POOL COOLING AND LOAD HANDLING
- 5. 1 Decay Heat Generation Rate 5.2 Spent Fuel Pool Cooling System 5.3 Heavy Load Handling 5.4 Light Load Handling 5.5 Conclusions 6.
SPENT FUEL POOL CLEANUP SYSTEM 7.
RADIATION PROTECTION AND ALARA CONSIDERATIONS 8.
ACCIDENT ANALYSES 9.
RADIOACTIVE WASTE TREATMENT 10.
SIGNIFICANT HAZARDS CONS I DERAIL'.ON COMMENTS 11.
FINAL NO SIGNIFICANT HAZARDS CONS!DFPATION 12.
FNVIRONMENTAL CONSIDERATIONS 13; CONCLUSIONS 14.
REFERENCES
?
8 q
11 11 12 13 18 21 22 22 APPENDIX A:
Technical Evaluation Report hv Brookhaven National Laboratory
p
INTRODUCTION
- 1. 1 Licensee Submittal and Staff Review This report presents the NRC staff safety evaluation for the reracking of the spent fuel pool at the St. Lucie Plant, Unit No.
1.
By letter dated June 12, 1987, the Florida Power and Light Company submitted an application to increase the storage capacity of the spent fuel pool, including the appropriate and necessary charges to the Technical Specifications.
The licensee requested the increa e in storage capacity because the pool lost full core reserve capability following a refueling outage completed in April 1987.
The June 12, 1987 request for the amendment, including the staff's proposed "No Significant Hazards Consideration,"
was noticed in the Federal Re ister on August 31, 1987 (52 FR 32852).
Further details are addresserl in ection 10 of this report.
The application is based on the licensee's "Spent Fuel Storage Facility tftodifi-cation Safetv Analysis Report" which was submitted as an enclosure to the June 12, 198? application.
During its review of the application, the staff re-quested additional information from the licensee; the additional information was provided by letters dated September 8, 1987, October 20, 1987 (three letters),
December Pl, 1987, December 22,
- 1987, and December P3, 1987 (three letters>.
In addition, the staff met with the licer see on a number of occa-sions as reported in meeting minutes dated September 11, l987, October 21,
- 1987, December 4, 1987, and December 9,
1987.
ny letter dated Januarv 29,
- 1988, the licensee submitted Revision' to the "Spent Fuel Storage Facility Modifica-tion Safety Analysis Reoort."
The purpose of the revision was to incorporate changes to certain sections resulting from the FPAL and NRC correspondence and the meetings with the staff.
The additional submittals supplemented and clarified the amendment request and did not alter the action noticed in the Federal
~ee ister or affect the staff's initial determination concerning the amenamment request (See Section 11'l.
This report was prepared by the staff of the Of<ice of Nuclear Reactor Requla-tion.
Technical assistance for the structural evaluation of the spent fuel racks and pool was provided by the Brookhaven National Laboratory,
- Upton, New York.
The principal contributnrs to this report are:
H. Ashar L.
Kopp G. DeGrassi J.
h1inns I. Spickler J. Ridgely F. Witt P.
Wu E. Tourigny Structural and Geosciences Branch Reactor Systems Branch Brookhaven National Laboratory (Consultant}
Radiation Protection Branch Radiation Protectior, Branch Plant Systems Branch Chemical Engineering Branch Chemical Enqineering Branch Pro,iect Directorate II-2
1.2 Summary Descr iptior. of Rerackina The amendment would authorize the licensee to increase the spent fuel peol storage capacity from 728 to 1706 fuel assemblies.
The proposed expansion is to be achieved hv reracking the spent
+uel pool into two discrete regions.
New, high-density storage racks (free-standing) will he used.
The existing storage racks (free-standina) will he removed, cleared of loose contamination, packaqed and shipped off-site.
Region I o< the spent fuel pool inc'rudes 4 modules (racks>
havina a total of 342 storage cells.
The nominal center-to-center spacing is 10. 17 inches.
All cr lls can be utilized for storage and each cell car accept new fuel assemblies with enriehnents up to 4.5 weight percent U-235 or spent fue'. assemblies that have not achieved adequate burnup,or Region 2.
Region 2 includes 13 modules (racks) havina a total of 1364 storage cells.
i;he nominal center-to-center spacing is 8.86 inches.
All cells can be utilized for storage and each cell can accept spent fuel assemblies with various initia'. enrichments that have accumulated minimum burnups.
Each cell in each region is designed to accom-modate a single Cembustion Engineerina or Advanced Nuc1ear Fue1s Corporation (formerly Exxon)
PMR fuel assemblv or eauivalent, from either St. Lucie Unit.
The hiah-density spent fuel storage rack cells are fabricated from 0.080 inch thick type 304L stainless steel plates.
In Region 1, strips of Bora+lex neutron absorber material are sandwiched hetween the eel~ walls and a stainless steel eoverplate.
In Region e, the Boraflex strips are sandwiched between the ad;;acent cell walls.
The cells, which form a module, are welded to a base
- plate, and a top gird'.e har is welded to the top of the module.
The new racks are not doubled-tiered ar d a>l racks will sit on the spent fuel pool fioer.
The amendment application does not involve rod consolidation.
The proposed expansion o< the spent fuel pool storage capacity to 1706 fuel assemblies should provide adequate storage ur til the year 2008, assuming full core offload capability.
In addition, the expansion should he adequate until a federal repository is available for spent fuel.
CPITICALITY CONSIDERATIONS 2.1 Critical it Ana1vs is The calculation of the effective multiplication factor, k
, makes use of the CASNO-2E two-dimensional multigroup transport theory comp5Qr code.
In addi-tion, for independent verification, criticality calculations were also performed with the KENO-IV Yonte Carlo code, as well es the EPRI-CELL and NULJF codes.
These independent verification calculations substantiate the CASNO-2E calcu>ations and resulted in a calculational bias of 0.0013 and a
95/95 probability/confidence uncertainty of 0.0018.
Ir. order to calculate the criterion for acceptable burnup for storaqe ir.
Region 2, calculations were made for fuel of several different initial enrich-ments.
At each enrichment, a limiting reactivity value, which included an additional factor for uncertainty in the burnup analysis, was established.
Burnup values that yielded the limiting reactivitv values were then determined for each enrichment
<rom which the acceptahle burnup domain for storaae in Region 2, as shown in proposed technical specification Figure 5.6-1, was obtained.
The staff finds this procedure acceptable.
For the Region I analysis, the total uncertainty is the statistical combination of the calculated bias uncertainty and manufacturing and mechanical uncer-tainties due to variations in boron loading in the Boraflex absorber
- sheets, Boraflex width tolerance, Boraflex thickness, inner stainless steel storage box dimension, flux trap water gap thickness, stainless steel thickness, fuel enrichment and density, and fuel pin pitch.
Other uncertainties due to tem-perature variations and eccentric positioninq of the fuel assemblv in the storage rack are accounted for by assuming worst-case conditions; i.e., condi-ti~ns which result in the highest calculated reactivity.
In the Region 2 analysis, the same uncertainties are considered, except there is nn water gap and,
- hence, no gap thickness uncertainty.
In addition, an uncertainty due to the burnup analysis is estimated and treated as an additive term in determining the burnup versus enrichment limiting reactivity values in Figure 5.6-1, rather than being combined statistically with the other uncertairties.
The staff concludes that the appropriate uncertainties have been considered and have been calculated in an acceptable manner.
In addition, these uncertainties were determined with at least a
95K probability and 95% confidence level, thereby meetinn the NRC requirements, and are acceptable.
For Region I, the rack multiplication factor is calculated to be 0.9409, including uncertainties at the 95/95 probability/confidence leve'., where fuel having an enrichment of 4.5 weight percent U-235 is stored therein.
Fuel of either the Combustion Engineering (CE) or Advanced Nuclear Fuels (ANF) type
<rom St. Lucie Unit 1 or Unit 2 may be stored.
For Region 2, the rack multiplication factor is calculated to be 0.9435 for the most reactive irradiated fuel permitted to be stored in the racks; i.e.,
fuel with the minimum burnup permitted for each initial enrichment as shown in Figure 5.6-1.
The design will accept eel of 4.5 weight percent U-235 initial enrichment burned to 36.5 MHD/kgU of either the CE or ANF type from Units I and 2.
Therefore, the results of the criticality analyses meet the staff's accentance criterion of k f no greater than 0.95, including all uncertainties at the 95/95 probability/confidence level.
Most abnormal storage conditions will not result in an increase in the k
of the. racks.
For example, loss of a cooling system will result in an incrNe in pool temperature, but this causes a decrease in the k ff value.
eff It is possible to postulate
- events, such as an inadvertent misplacement of a fresh fuel assembly either into a Region P storage cell nr outside and adjacent to a rack module, which could lead to an increase in pool reactivity.
- However, for such events, credit may be taken for the Technical Specifications minimum requiremer t of 1720 ppm of boron in the pool water.
The reduction in the k
value caused by the boron (approximately 0.24) more than offsets the rfQtivity addition caused by credible accidents.
2.2 Technical Specifications Chan es The followina Technical Speci ications (TS) changes have been proposed as a
result of the replacemert of the existina spent fuel pool racks at Unit 3.
The staf~ finds these changes acceptable.
1.
TS 5.6.l.a. 1 is revised to correspond to the Standard Technical Specifications fnr Combustion Engineering Ph'Rs (NUREG-0212, Rev. 2).
2.
TS 5.6. 1.a.2 is revised to show the nominal center-to-center spacing for the new storage racks.
3.
TS 5.6. l.a.3 is edited to discuss the boron concentration in the pool water only.
4.
TS 5.6.1.a.4 is added to indicate the presence of Boraflex in the storage cells.
5.
TS 5.6. l.b and accompanying F<aure 5.6-1 are added to show the increased spent fuel enrichment permitted in the pool.
6.
TS 5.6.l.c is editorially charaed from "b" to "c".
7.
TS 5.6.3 is charged to show the capacitv o, the high-capacity spent fuel storage racks.
2.3 Conclusions 8ased on the review described
- above, the staff finds the criticality aspects of'he design of the St. Lucie Unit 1 spent fuel racks tn he acceptable.
The staff concludes that CE or ANF fuel from Unit 3 or Unit 2 may be safely stored in Region 1 provided that the enrichment does not exceed 4.5 weight percent U-235.
Any of these fuel types may also be stored in Region 2 provided they meet the hurnup and enrichment, limits specified in Fiaure 5.6-1 of the St. Lucie Unit 1 TS.
3; MATERIAL COMPATAPILITY AND CHEMICAL STABILITY 0
The staff reviewed the compatibility and chemical stability of the high density spent fuel storage rack materials wetted by the pool water.
The proposed racks are fabricated from ASHE SA-240-3046 austenitic stainless steel sheet and plate
- material, SA-331-CF3 casting material and SA-564-630 precipitation-hardened stainless steel (to 1100'F) for supports only.
The weld filler materia~
utilized in body welds is ASMF. SFA-5.9, classification FR 308L.
The neutron absor)er material is Boraflex with a minimum 0-10 areal deBsity of 0.0238 gm/cm for the 342 Region 1 storage cells and 0.0098 gm/cm'or the 1364 Region 2 storage cells.
8oraflex is a silicone-based polymer car taining fine particles of boron carbide in a homogeneous, stable matrix.
The annulus spaces that contain the Rara lex in the high density racks are vented to the spent fuel pool.
Venting of the annuli will allow gas generated by the chemico and radiolytic decomposition of the silicone polymer binder,'hen exposed to the tl ermal and radiation environment, to escape.
This will prevent pressure buildup and possible bulging or swelling of the stainless steel absorber sheathina.
The austenitic stainless steel (304L> used in the spent fuel storage racks is not susceptible to stress co'rrosion crackino and thus, corrosion in the spent fuel storage pool environment should be of little significance during the 1'.'
nf the plant.
The spent fuel pool water is processed by filtration and demineralization tn maintain water puritv and clarity.
Dissimilar metal contact corrosion (qalvanic attack between the stainless stee~
rack assemblies and Zircalo<< in the fuel assemblies) should not be significant because the materials are protected by highly passivating oxide films and are, there~ore, at similar galvanic potentials.
gualification tests have shown that Boraflex does not possess leachable halogens that could be released into the spent fuel pool water in -the presence nf radiation.
Similar conclusions have been made regardina the leaching o<
- rom the Boraflex.
Although Boraflex has undergone extensive oualification testing to study the effects of gamma irradiation in various environments and to verify its structural integrity and suitability as a neutron absorhinq material, recent anomalies have been identified in the squad Cities and Point Beach high density spent fuel racks due to Boraflex shrinkage caused bv irradiation.
To preclude similar problems at St. Lucie Unit No. I, the specification for the handling and installation of the Boraflex requires that it not be installed in a stretched condition.
The use of adhesives in the attachment of the Roraflex to the rack cell is not permitted.
In addition, the manufacturing process avoids techniques that could pinch the Bore~lex.
Therefore, the St. Lucie Plant Unit No.
1 rack design and fabrication process allows expected shrinkage without cracking and gap formation.
Furthermore, the spent fuel rack design reau'.res that oversized Boraflex sheets be used to provide a four-inch shrinkaae allow-ance and that allowances for the elastic rebound of the Boraflex material be made before installation should the material be stretched during shipment or hahdlinq.
To provide added assurance for detection of degradation of thI. Rora~lex, the licensee has committed to conduct a long-term and accelerated survei'.lance test program.
Each surveillance coupon (5 inches hy 15 inches) contain'ng Ror aflex of a thickness similar to that used in the racks, is encased in a
stainless steel jacket, the alloy of which is identi~ied to that used in the racks.
The coupon iacket permit wetting and ventino of the specimen to the spent fuel pool water similar to that of the rack.
The long-term coupon examination frequency occurs after irradiation times of 90 days, IPQ days, I year, 5 years, 10 years, 15 years, 25 years and 35 years.
The accelerated test coupon examination freouency is after each discharge from the second tr rinth discharge rack utilization.
Acceptance criteria for continued use are dimensional changes of no more than 2.5l from the original, hardness not less than 90% of the oriqinal, and minimal areal density of boron not less than the original.
The staf~ has reviewed the proposed surveillar ce prngram for monitorina the Rrraflex in the St. Lucie Plant Unit No. I spent fuel storage pool and concludes that the program can reveal deterioration that may lead to loss of neutron absr rhing capability during the life of the spent
<uel racks.
In the unlikely event of Boraflex deterioration, the monitoring program will detect such deterioratior and the licensee will have sufficient time to take corrective action.
In the event of unanticipated degraded
- coupons, the storage racks will he inspected and then NRC will be informed if the inspection reveals Boraflex degradation in the storage racks.
Based on the above discussion, the staf~ concludes that corrosion of the high density racks due to the spent fuel pool environment should be of little significance during the life of the facility.
The staff finds that implementation of the proposed surveillance proqram and the selection of appropriate materials of canstruction by the licensee, meet the requirements of 10 CFR 50, Appendix A, General Design Criterion (GDC) 61 (regarding the capability to permit appropriate periodic irspection and testing of components) and CiDC 6? (regarding preventinq criticality by maintaining structural integrity of components and of boron absorber material) and are, therefore, acceptable.
4.
STRUCTURAL DESIGN This evaluation addresses the adequacy of the structural aspects of the pro-posed amendment.
The Rrookhaven National Laboratory (RNL) assisted the staff in reviewing various anal.yses and responses submitted by the licensee.
Attached is the technical evaluation report (TER) developed by BNL (Appendix A).
The staff accepts the findings and conclusions o; the TER by incorporating the TER as a part of this safety evaluation.
The spent fuel storage pool is located in the fuel handling building, which is a Seismic Category I structure.
The pool is 33 feet by 37 feet in plan and is 40 feet.,
6 inches deep.
The reinforced concrete foundation mat, which is 9'-6" thick except in the spent fuel cask storage area where it is 6'-0" thick, provides floor space for the spent fuel racks.
The reinforced concrete walls enclosing the spent fuel storage area vary in thickness from ?'-0" to 5'-0".
The pnal walls are lined with 3/16 inch stainless steel plates and the pool floor is lined with k inch stainless steel plates.
The proposed high-density storaqe racks consist of individual cells with 8.65 inches by 8.65 inches square cross-section, each of which would accoranodate a single Combustion Engineering or ANF PWR fuel assembly.
A total of 1706 cells are arranged in 17 distinct rack modules of various arrays of fuel cells.
Each rack module is equipped with 3/4 inch thick hy 34 inch high girdle bars at the upper end desioned to withstand the impact loads under the postulated seismic conditions.
The rack modules are free-standing, and they make surface contact at the girdle bar locations providing a nominal l,k inch gan between adjacent module cell walls.
The primary areas of review associated with the proposed application are focussed towards assuring the structural integrity of the fuel, fuel cells, rack modules, and the spent fuel pool floor and walls under the postulated (Appendix D of SRP 3.8.4) loads and fuel handling accidents.
The major areas of concern and their resolutions are outlined in the following paragraphs.
The fuel handling building analysis and design had been reviewed and accepted during the initial licensing stages.
Since the effect of the additional fuel rack load an the pool floor is limited to the mat in the pool ar ea, the licensee reanal.vzed the lower portion of the walls, the pool floor, and the effects on the underlying soil.
The design-analysis results satisfy the acceptance criteria.
Details of the analysis, design and adequacy of the pool, pool liner and its anchorages are discussed in Section 4.5 of Appendix A.
The plant is located on potentially liquefiable soil.
During the operating license review, the licensee provided sufficient data and analyses to demon-strate that the factor of safety against liquefaction under a Safe Shutdown
Earthquake (SSF) is more than 2.
Durina this review, the staff expressed a
concerr.
about the effects of added weight on liquefaction potential under the postulated seismic condition (i.e.,
SSE).
Based on the research work publishrrr by Seed, Idriss and other researchers in the puhlication, "Liquefaction of Soils During Farthquake (National Academy Press, 1985)," where it was shown that snils subjected to static shear stresses prior to an earthquake have hiaher resistance to liquefaction, the licer see concluded that the added weight would maintain or improve the resistar.ce to liquefaction.
Tl e licensee's report alsn indicated that the maximum bearing pressure on the soil under the combined effects of dead load (including the added fuel weiaht) and an SSE is less than the allowable bearing capacity of the soil.
The staff accepts the licensee's conclusion and considers the concern as resolved.
The adequacy of considering a single rack model in the seismic analysis was questioned.
The seismic motion of a single rack is coupled to the motion of adjacent racks throuch impact forces and fluid couplina forces.
The single rack model constrains the motion of a rack within an imaginary boundarv.
Maximum displacements cannot exceed one-half the gap to the adjacent racks.
For sufficiently strona seismic motion, slidinq and tiltina motions of the racks could be larger than those predicted by a cons+rained single rack model resulting in higher impact velocities than would be predicted by a single rack model.
Under worst conditions, rows of racks could slide together in one directon and pile up against a pool wall.
The additional mass of racks involved in the impact could generate laraer loads on the racks and the pool walls.
This concern may be more critical for the pool walls, since they are not desiqned to accomodate seismic impact loads from the fuel racks.
To resolve the concern, the licensee performed a two-dimensional multiple rack analysis of a single row of fuel racks to determine the extent of displacement under an SSF..
The limited multiple rack analysis indicated that the correspond-ing displacements are small (less than or equal to 1/2 inch) compared to the minimum clearance provided (3 1/2 inches) between the edge racks and the walls.
A detailed discussion of the other concerns, the comparative results of various analyses and conclusions thereof are provided in Section 4.P of Appendix A.
Based on its evaluation of the licensee's submittal, the supplementary informa-tion provided by the licensee, discussions with the licensee at meetings, and information audited by the staff and its consultant, the staf~ concludes that.
the licensee's structural analyses of the spent fuel rack rrndules and the spent fuel pool are in compliar ce with the acceptance criteria set forth in the FSAR and consistent with the current licensing practice and, therefore, are acceptable.
5.
SPENT FUEL POOL COOLING AND LOAD HANDLING
- 5. 1 Deca Heat Generation Rate e
In the June 12, 1987 submittal, the licensee stated that the calculation of the decay heat generating rate was in accordance with the guidelines nf Standard Review Plan
{SPP) Section
- 9. 1.3 and Branch Technical Position ASB 9-2.
For the normal maximum heat load condition, the licensee assumed
+he pool was filled with one-third core refuelings everv 18 months
~rom the St. Lucie Unit 1 reactor and calculated a heat. generation rate of 16.42 NBTU/Hr.
The abnormal maximum heat load condition had the same assumptions as the normal maximum hea'.
load condition, except that the PI? emptv fue'i storage locations were filled with a full core offload.
For this condition, the licensee calculated a heat generation rate of 33.70 MBTU/Hr at 169 hours0.00196 days <br />0.0469 hours <br />2.794312e-4 weeks <br />6.43045e-5 months <br /> into the refueling outage in lieu of the 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> identified in the SRP.
The staff performed an independent calculation of the heat generation rate in accordance with the guidelines in SRP Section
- 9. 1.3 and Branch Technical Position ASB 9-2 assuming the anticipated 18-month operatinq cvcle.
The staff calculated a normal maximum heat generation rate of 16.84 MBTU/Fr and an abnormal maximum heat generation rate of 33.56 MBTU/Hr at 169 hours0.00196 days <br />0.0469 hours <br />2.794312e-4 weeks <br />6.43045e-5 months <br /> into the refueling outage and 34.96 YRTU/Hr at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />.
The licensee's calculation of the normal maximum heat broad is not significantly different from the staff's calculated
- value, and thus, the staff concludes that the licensee has properly calculated the heat generation rate in accordance with the guidelines o
th~ SRP.
5.2 S ent Fuel Pool Coolin Sy stem The spent fuel pool cooling system (SFPCS) consists of one train of equipment, including two 3560 gpm centrifugal pumps and one tube-and-shell heat exchanger with a heat transfer capability of approximately 34 MBTU/Hr, as indicated in the FSAR.
After water from the spent fuel pool is cooled by the heat exchanger, it is puri ied by the spent fuel pool cleanup system.
Neither the SFPCS nor the cleanup system are seismic Category I.
In the event of a loss of SFPCS, a seism',c Cateoory I salt water makeup supply to the spent fuel pool is availah>e from the intake cooling water intertie.
The SPFCS heat exchanger is a low pressure, low temperature component.
Maintenance of the heat exchanger, such as tube cleaning or plugging, can be scheduled to be performed when the heat being generated bv the spent'fuel is low, such as immediately prior to entering a refueling outage when the time until the spent fuel pool reaches boiling will be significantly longer than the 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> calculated for the normal maximum heat load case.
Thus, the staff concludes that having a single heat exchanger is acceptable.
5.2. 1 Heat Remnval Ca abilit Under the normal maximum heat load conditions (16.84 MBTU/Hr) using one SFPCS pump, the SFPCS heat exchanger will maintain the spent fuel pool water temperature below 134'F, which is less than the I>0'F temperature guideline specified in SRP Section
- 9. 1.3.
For the abnormal maximum heat load condition (33.56 MBTU/Hr} using one SFPCS pump, the heat exchanger will maintain the spent fuel pool water temperature below 167'F, which is well below boiling.
Thus, the staff finds that the SFPCS meets the requirements of GDC 44, "Cooling Mater" with respect to providing adequate pool coolinq under normal heat load conditions following a single failure.
5.2.2 Protection A ainst Natural Phenomena The SFP cooling capability was reviewed with respect to the requirements o,
GOC 2, "Design Bases for Protection Against Natural Phenomena,"
which includes protection against earthquakes, hurricanes, tornadoes, or other natural events.
The SFPCS is not seismic Category I. Under such circumstances, SRP Section 9.1.3 identifies an alternative method for cooling of spent fuel fo1lt winq an earthquake.
Specifically, the SRP discusses use of a seismic Category I spent fuel pooi makeup water capability and a seismic Category I ventilation system to process potential radiological releases to the pool building result'ng from pool boiling.
.*.2..
~k The St. Lucie Unit 1 FSAR identi ies several makeup water snurces.
The refueling water storage tank and the primary water tank are seismic Category I sources of water.
In addition. salt water can be provided to the spent fuel ponl rom the intake structure via the seismic Category I intake cooling water system at the rate of l50 gpm.
5.2.2.2 Buildin Ventilation
'The licensee has not taken credit for any ventilation system to mitigate the offsite releases due to boiling of the spent fuel pool water.
The licensee has provided the results nf the offsite dose consequence analysis in their submi+tal de+ed December 23, 1987, which indicates that the maximum calcula)ed adult absorbed
+hyroid dose is 0. 1235rem, the whole body dose is 1.82 x 10
Since the thyroid dose is less than ll of 10
- CFP, 100 limits (300 rem) and the whole body and skin doses are insign',iCant,,
the staff concludes that not using any ventilation system to mitiaate the release of radioactivity when the water in the spent fuel pool is boiling meets the requirements of GDC 60, "Cortrol of Releases of Radioactive Materials to the E'nvironment."
In the event that all SFP cooling is lost, the spent fuel pool temperature will increase until boiling is achieved.
The licensee has estimated the time from the loss of pool cooling until the pool water boi~s for the normal maximum heat load condition tn be approximately 16.79 hours9.143519e-4 days <br />0.0219 hours <br />1.306217e-4 weeks <br />3.00595e-5 months <br /> and for the abnormal heat load condition to be approximately 7.47 hours5.439815e-4 days <br />0.0131 hours <br />7.771164e-5 weeks <br />1.78835e-5 months <br />.
The calculated boil-off rates are estimated to be 33.9 gpm and 60.5 gpm, respectively.
The staf< finds that the intake cooling water sys+em caoability is in excess of those estimated boil-off rates and there is reasonable time tn take action to provide SFP makeup.
The staff further concludes that the makeup water system, without any ventilation svstem to mitiqate the release of radioactive materials, meets th~
requirements of GDC 2, "Design Bases for Protection Against Natural Phenomena,"
for ensuring adequate spent fue> pool cooling and prevention of unacceptable radiological releases following an earthquake.
5.3 Heav Load Handlin The new spent
~uel storage racks weigh more than a fuel assemh>y and its handling tool.
Thus, the spent
~uel storage racks are considered to be heavy loads.
The cask handling crane will be used to move the new storage racks into the fuel handling building and into the cask area within the spent fuel
- pool, and to remove the existino storage racks from the cask area to the cask decontamination area outside of the fuel handling building.
The movement o~
the cask handling crar e is phvsically limited hv the opening in the side wall and the ronf of the fuel handlina building.
This openinq is normally closed by a L-shaped door.
The cask handling crane, rIue to this limitation, cannot carry heavy loads over spent fuel.
In the previous review of compliance with the guidelines of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plarts," the staff concluded in a Safety Evaluation Report dated March 4, l985, that the cask handl<no crane met the guidelines of NURfG-0612.
I
10 Due to the physical limitations of the lifting capability of the cask hand1ina
- crane, a new, temporary crane will be installed in the fuel handling building.
The temporary crane will be used
.o move the new storage racks to their appropriate positions in the pool fromm the cask area and the existing racks from their present locations to the cask area.
The temporary crane will also be used as a platform for re-rigging of the new and existing storage racks on the cask handling crane.
In the December 22, 1987 submittal, the licensee provided installation details for the tempnrary crane.
It will be brought into the fuel hand~ing building as five separate pieces (two truck units, two girder pieces.
and the hoist unit).
The information provided'in the December 23, 1987 submittal demonstrated that no piece of the temporary crane will be carried over spent fuel or over racks containing spent fuel.
To provide assurance that nn part of the temporary crane will he carried over spent <uel, the licensee committed to park the tuel handlir g machine over the nearest spent fuel assembly as a
physical barrier tn movement.
In the December 22, 1987 submitta'.,
the licensee provided the results of evaluations n< three potential load drat accidents:
(1) the temporary crane dropping a spent fuel storage rack in the spent fuel pool; (2> the cask handling crane dropping a spent fuel storage rack into the cask area of the spent. fuel pool; and (3) the cask handling crane dropping a spent fuel storage rack onto the temporary crane.
In all cases, the radiolngical consequences of the load drop are less than that for the cask drop accident identified in the FSAR.
The rack drop accidents involving the cask handling crane would require the L-shaped door in the fue> handling building to be open.
- Thus, no credit was taken for retention of the radioactivity by the building.
The licensee committed to remove the temporary crane and to pe~form a load test on it if a heavy load is dropped onto it.
Mhen the cask handling crane is moving a rack into or out of the <uel building, the temporary crane will be located next to the north wall of the spent fuel pool.
Three new racks will be placed along the north wa>1 beside the cask area.
The three new racks will, as part of the fuel shuffling program, contain spent fuel; however, nn spent fuel will be placed under the temporary crane parking location.
Thus, if the temporary crane were to fail as the result of a load drop from the cask handling crane, no spent fuel would be impacted.
The temporarv crane is not single-failure proo<.
The licensee stated in the December 22, 198?, submittal that the safety factors for all load-hearing components of the temporary crane meet nr exceed the safety factors identifiec'n NUREG-0612.
All welds will be inspected using either liquid penetrant or maqnetic partic'.e methods.
The crane hoist will be load-tested to 150% of the rated load.
h~ith these
- measures, the staff finds that the temporary crane meets the guidelines of NUREG-0612.
The licensee provided drawings of the special liftina devices for the new and existing spent fuel storage racks.
These drawings demonstrate that the lifting devices are single-failure proof and thus meet the guidelines of NUREG-0612.
In the December 23, 1987 submittal, the licensee provided drawings that show the order in which the existing racks will be removed and the new racks wil'. be
~
1 installed.
These drawings also identi+y those storage locations that will contain spent fuel and veri.y that the racks will not be transported over spent fuel or over racks containing spent fuel.
From the above review, the staff finds that handling heavy loads during the reracking procedures is in accnrdance with the auidelines of NUREG-0617 and, therefore, the requirements of GDC 61, "Fuel Storage and Hand>ing and Radio-activity Control," are met as the>> relate to proper load handling to ensure against an unacceptable release of radioactivity nr a criticality accident as a
result of a postulated load drop.
5.4 Li ht Lead Handlina A light load is defined as any load that. weighs less than a fuel assembly and its handling tool.
In a submittal dated December 2',?,
1987, the licensee provided the results of an evaluation of light load drops for St. Lucie Unit 1.
The ',icensee reviewed the liaht load analysis that was performed for St. Lucie Unit 2 at the time. of licensing; which was approved bv the staff in NUREG-0843, Supplement 3, dated April 1983.
The licensee verified that tho e light loads evaluated for Unit 2 are applicable for Unit 1.
From that review, the licensee concluded that the consequences to spent fuel from a light load drop would be less than that for a design basis fuel handling accident, namely the failure o
all fuel pins in one fuel assembly.
5.5 Conclusions 0
Based on the above, the staff concludes that the proposed expansion of the St.
Lucie Unit 1 spent fuel pool complies with the requirements of General Desian Criteria P., 44, 60, and 61 and the guidelines of NUREG-061", and Regulatory Guide 8.8 with respect to the capability to provide adequate spent fuel pool cooling, safe loading handling, and to maintain offsite and onsite radioloaical releases within acceptable limits.
The staff, therefore, finds the proposed expansion to be acceptable.
6.
SPFNT FUEL POOL CLEANUP SYSTEM The spent fuel pool (SFP> cleanup or purification system maintains pool water clarity and purity. It consists of a 150 gpm purification Dump, a cartridge filter, a mixed bed demineralizer, and the required pipina, valves, and instrumentation.
The pump draws water from the SFP and discharqes through the cartridge filter and the demineralizer.
The water is then returned to the pool.
It is possible to operate the system with either the filter or demineralizer bypassed.
Radioactivity and impurity levels in the water of a spent fuel pool increase primarilv during the refuelina operations as a result n<<ission product leakage from defective fuel elements being discharged into the pool and to a
lesser degree durir g other spent fuel handling operations.
The reracking of the spent fuel pool at the St. Lucie Plant, Unit No.
1 will not increase the refueling frequency and fraction of the core replaced after each fuel cycle.
There,ore, the frequency of operating the spent fuel pool cleanup system is not expected to increase.
Similarly, the chemical and radionuclide composi-tion of the spent fuel pool water will not change as a result. of the proposed reracking.
Following the discharge of spent fuel from the reactor into the pool, the, fission product inventory in the spent cruel and in the pool water will decrease by radioactive decay.
Furthermore, experience also shows that there is no significant leakage of fission products. from spent
~uel stored in pools after the fuel has cooled for several months.
Thus. the increased quantity of spent fuel to be stored in the St. Lucie Plant, Unit Nn.
1 fuel pool will not increase significantly the total ~ission product activity in the spent fuel pool water durina the operation of the ponl.
The staff has evaluated the information provided by the 1i~ensee.
Based on this evaluation and its experience with other high-density spent fuel storage fac',lities, including evaluation of operating data, the staff has determ ned that the proposed reracking of the spent ue>
pool at St. Lucie Plant, Unit No.
1 will not adversely affect the performance capability nr capacity o. the spent fuel pool cleanup system.
The radioactivity and impurities in the pool water are not expected to increase as a result o~
he rerackino.
Replacement of filters or demineralizers would offset any unanticipated increase of the radioactivity and impurity level of the water in the event of a reduction o~
the decontamination effectiveness.
On the basis of the above discussion, the spent fuel pool rerack is acceptable.
7.
RADIAT'.ON PROTECTION AND ALARA CONSIDERATIONS The additional cccupa+ional radiation exposure associated with the actual rerackir q o the pool is estimated by the licensee to be less than 15 person-rem.
In a letter dated October 20,
Some of the.ALARA activities directed tn the reduction of occupational radiation include:
(a) vacuum cleanirg of SFP floor s will be performed remotely from the surface; (b) maximum water shielding to reduce dose rates to divers, if they are used
- (c) underwater radiation surveys; (d) calibrated alarmino dosimeters and personnel monitoring dosimeters for divers, if they are used; te) hydrolasinq and cleaning of old spent fuel racks; (f> +he use of remote operations for rack removal and replacement operations; and (g)
SFP purification system augmented bv ur derwater vacuum system to maintain radioactive contamination ALARA and maintain SFP clarity.
The licensee has alsn provided a description of contained and airborne radioact'v.'ty sources related tn the SFP water, which may become airborne as a result of fa'led fuel and evaporation.
The staff has reviewed these source terms and finds them acceptable.
nased on our review of the St. Lucie's submitta~s, we conclude that the proiected activities and estimated person-rem dose for this pro,iect appear
. reasonable.
FPL intends to take ALARA considerations into account, and to implement reasonable dose-reducing activities.
Ve conclude that FPL will be able to maintain individual occupational radiation exposures within the applicable limits of 10 CFR Part 20, and maintain doses ALARA, consistent with the quidelines of Reoulatory Guide 8.8.
Therefore, the proposed radiation protection aspect of the SFP rerack is acceptable.
8.
ACCIDENT ANALYSES The staff has reviewed the accidenta~
fission product releases that could occur at the St. Lucie Unit 1 facility in coniunction with the proposed reracking of the SFP.
The only potential releases that have not been previouslv analyzed by the staff as part of the original SER are the potential offsite consequences o< the dropping of a cask into the reracked full SFP and release of fission products from the spent fuel resulting from the boiling of the pool water.
The consequences of these accidents have been reviewed by the licensee and the staff.
13 With regard to cask drop accident, the most conservative case occurs with the cask being dropped into the SFP 1490 hours0.0172 days <br />0.414 hours <br />0.00246 weeks <br />5.66945e-4 months <br /> afte~ the following fuel cycle history:
One-third of a core is placed in the SFP each year durina refueling for the next 20 years.
Following the 21st year of operation, the entire core is removed from the reactor and placed into the pool, which fills the pool.
The number of'ssemblies damaged is equal to a full-core offload plus the remainder of the pool filled with discharged assemblies from previous refuelings.
The 1490 hour0.0172 days <br />0.414 hours <br />0.00246 weeks <br />5.66945e-4 months <br /> figure is the earliest that a cask could be moved intn the SFP area with a full pool based on the TS.
It is assumed that a>l the spent fuel in the pool (8 full cores) is damaged with the release of 10K of the noble gases (except Kr-85) and the iodines and 30% o< the Kr-85 to the pool water, and with 99" of the released iodine remaining in the pool water.
The remainder of the released fission products is released to the environment.
The resulting dose to an individual at the exclusion zone boundary would be 21 thyroid-rem and less than 0. 1 rem to the whole body.
In their December 23, 1987 submittal, the licensee presented a conservative ar alysis of the radioloaical consequences of boiling of the SFP water.
The staff has reviewed the licensee's analysis.
The staff analysis differs only slightly (staff utilization of slightly rrore conservative dilution factors).
The potential doses at the exclusion boundary (0.97 miles
) and low<populatinn zone boundary (1.0 miles) are approximately 0. 1 thyroid-rem and 10 rem to the whole body.
The potential doses resulting from the cask drop ard spent fuel pool water boiling accidents are well below the allowable 10 CFR lCO guidelines doses of 300 rem to thyroid and 25 rem to the whole body.
Therefore, the accident analysis aspect of the SFP rerack is acceptable.
9.
RADIOACTIYE WASTE TREATMENT The plant contains radioactive waste treatment systems designed to collect and process the gaseous, liquid, and solid waste that miqht contain radioactive material.
The radioactive waste treatment systems are evaluated in the Final Environmental Statement (FES) dated June 1973 (US NRC 1973).
There will be no chanqe in the waste treatment systems described in the FES because of the proposed SFP r crack.
10.
SIGNIFICANT HAZARDS CONSIDERATION COMMENTS The licensee's request for amendment was noticed on August 31, 1987 (59 FR 32852), followed by a biweekly notice on September 23, l987 (52 FR 35813>.
By letter dated September 30, 1987, Mr. Campbell Rich requested a public hearing.
An Atomic Safety and Licensing Board was established on October 22, 1987 to consider the reauest.
In pleadings filed November 4 and 9, 1987, both the licensee and the NRC staff pointed out that the letter failed to meet the requirements of 10 CFR 2.714 and that, therefore, the request should he denied.
By Memorandum and Order of November 13, 1987, the Board directed the licensee and Hr. Pich to seek informal resolution of Mr. Rich's concerns and set January 15, 1988 as the deadline for filina an amended petition.
Mr. Rich met with the licensee and subsequently filed an amended petition which proffered 16 contentions.
The licensee and the staff responded to the contentions by
pleadings dated February
~,
1988 and February 4, 1988, respectively.
The Licensing Board has not yet ruled on the contentions but has scheduled oral arguments on intervention and the contentions for March 29, I988 (53 FR 5661).
The proposed contentions and the staff comments are contained below.
Contention 1:
"That the expansion o
the spent fuel pool at St. Lucie, Unit No.
1 is a significant hazards consideration and requires that a
public hearing be held before issuance of the license amendments I.sic".."
The staff may issue and make immediately effective an amendment to an operating license pursuant to the Coomission's regu>ations.
A public hearinq need not be held before issuance o< the amendment.
The staf+
has followed the Covmissinn's regulations in the licensing action.
A Final No Significant Hazards Consideration Determination is inc>uded in this safety evaluation.
Contention 2:
"Expansion of the spent fuel pool at the St. Lucie facility, Unit No.
1 constitutes a maior Federal action and requires that the Commission prepare an environmental impact statement in accordance with the National Environmental Policy Act of l969
{NEPA) and 10 CFR Par+ 5I."
The sta.
prepared an Environmental Assessment related to this licensing action.
Based r n the Assessment, the staff made a finding of no signif-icant impact pursuant to 10 CFR 5!..32. (53 FR 7065).
Therefore, no envi-ronmental impact statement need be prepared.
Content'on 3:
"That the calculation of radiological consequences resulting from a cask drop accident are I sicl not conservative, and the radiati.on releases in such an accident will no fsicj be ALARA, and will not meet with the 10 CFR Part 100 criteria."
As Low As Is Reasonably Achievable (ALARAl applies to normal plant operations.
CLARA is not a consideration in accident analysis consequences determination.
The licensee addressed the cask drop accident in the licensing submittal.
The staff reviewed the licensee's analysis (including input assumptions) and agrees with the licensee's conclusions.
Section 8 of this evaluation contains the details nf the staff's independent evaluation.
Contention 4:
"That the consequences of a cask drop accident or an accident similar in nature and effect are greatly increased due to the presence of a large crane to be built inside the spent fuel pool building in order to facilitate the reracking."
The large crane that will be "built" in the fuel handling building is considered a temporary construction crane.
The crane will be used to remove the existing racks and install the new racks.
The crane will be in the fuel handling building for only a few months.
Once the rerack modification is completed, the crane will be removed from the building.
The spent fuel cask and the temporary construction crane wi 11 never be in the building at the same time.
Thus, there is no possible accident as a result of the temporary construction crane and cask being in the building at the same time.
15 The contention also refers to "an accident similar in nature."
This was also evaluated by the staff as follows.
The staf~ ev'aluated the use of the temporary construction crane to be used during the rerack modif'.ca-tion.
The staff postulated various load drop accidents, such as the drop of a rack during the rerack modification, in spite of the fact that no heavy load will be carried over spent fuel or over ary rack which contains spent fuel.
The staf< concluded that in all cases, the radiological consequences of the load drop accident are less than that for the cask drop accident evaluated.
Section 5.3 of this evaluation contains the details of the staff's evaluation.
Contention 5:
"~hat FP~L has not provided a site specific radiological analysis of a spent fuel boiling event that proves that off-site dose limits and persoral I.sic" exposure limit. will not be exceeded in a>lowing the pool to boil with makeup water from only seismic Cateoory
] sources."
The licensee and the staff used the Standard Review Plan (SRP) as guidance in the spent
<uel pool cooling analysis.
The SRP specifies that the pool water temperature should not exceed 140'F (a single active failure to +he system is assumed>
under normal refueling condi-tions and not exceed boiling (a single active failure need not be considered) under full core discharge conditions.
Independent calcula-tions per ~nrmed by the staff and licensee concluded that the SRP acceptance criteria are met.
Nevertheless, the licensee and the staff, as a further precaution, postulated the site-specific pool boiling event and evaluated the radiological consequences and makeup water sources.
The staff concluded that the radio>laical consequences were well within the guidelines of 10 CFR Part 100 and seismic Category I makeup water sources were available to supply makeup water to the pool.
Section 5.2. 1 contains the staff's evaluation of heat removal capabil-ity.
Section 5.2.2 contains the staff's evaluation o< makeup water sources.
Section 8 contains the staff's evaluation of the radioloaical consequences as a result o~ pool boiling.
Contention 6:
"The Licensee and Staff have not adequatel.v considered or analyzed materials deterioration or failure in materials integrity resulting from the increased generation of heat and radioactivity as as I.sic", result of increased capacity and long-term storage in the spent fuel pool."
The staff reviewed materials integrity n all materials used in the spent fuel pool.
The corrosion of the high density racks due to the spent fuel pool environmen+
should be of little significance during the life of the facility.
The long-term durability of Boraflex is ensured by the proposed surveillance program.
Section 3 of this document contains the staff evaluat'.on to support these conclusions.
Contention 7:
"That there is no assurance that the health and safet!~
of the workers will be protected during spent fuel pool expansion, ar d that the NRC estimates of between 80-130 rem/person will not meet ALARA requirements, in particula~,
those in 1.0 CFR Part 20."
The staff evaluated the occupational radiation doses to workers involved with reracking the spent fuel pool.
The staff concludes that the occupational radiation exposure is less than 15 person-rem, within
the applicable limits to 10 CFR Part 20, and is ALARA.
Section 7
contains the staff's evaluation of doses to workers.
?r addition, Section 3.2 of the Environmental Assessment also add~esses doses to workers.
Contention 8:
"That the high density design of the fuel storage racks will cause higher heat loads and increases in water temperatur~
which could cause a loss-of-coolirg accident and/or challenge the reliability and testability o~ the systems designed for decay heat and other residual heat removal, which could, in turn, cause a major release of radioactivity into the environment."
The staff's comments are the same as those in response to Contention 5.
Contention 9:
"That the cooling system will be unable to accommodate the increased heat load in the pool resulting from the high-density storage system and a full core discharge in the event of a sir gle failure of any of the pumps or the electrical power supply tn the pumps on the shell side o
the cooling svstem and/or in the case nf a single failure of the electrical power supply to the pumps on the pool side of the spent pool cooling system.
This inability will, therefore, create a greater potential for an accidental release of radioactivity into the environment."
The staff's comments are the same as those in response tn Contention 5.
Contention 10:
"That in calculating tire to boil after loss of cooling after completion of full core discharge with the presence of the proposed 1706 assemblies, FPSL utilized a different set of assumptions than in determining the original figures +or time to boil as indicated in the Final Safety Analysis Report for the St. Lucie plant, Unit No. 1.
(9.1-49.
Table 9.1-3)."
The staff's comments are the same as those in response to Contention 5.
Contention ll:
"That the proposed use of high-density storage racks designed and fabricated by the Joseph Oats Corporation is uti~ization of an essentially new and unproven technology."
The staff does not agree that the proposed use of high-density storage racks designed and fabricated by Joseph Oats Corporation is utilization of an essentially new and unproven technology.
A large number of high-density storage
- racks, which utilize Boraflex, have been abricated by J.
Oats for other utilities.
These racks have been installed and are currently storing spent fuel.
Similar statements can be made of other fabricators of spent fuel storage racks.
Section 3 of this evaluation contains the staff's rack materials evaluation.
Contention 12:
"That the presence of degraded Boraflex specimens or absorber sheets on the floor of the pool will pose an increased hazard in promoting the propagation of cladding
<re to low power bundles and thus promote a far larger spent fuel pool accident."
Boraflex specimens or absorber sheets will not be located on the floor o, the pool.
The Boraflex will be installed as part of the racks, within the rack structure.
See Section 3.
Contention 13:
"The Licensee has rot analyzed the effect that a
hurricane or tornado could have on
..he spent fuel storage ~acility or its contents, and that the SEP nealects cer tain accidents that could be caused by such natural disasters."
The staff evaluated the fuel handlina building structure under natural phenomena conditions when the unit was originally licensed.
The staff's SER evaluation dated November 8,
- 1974, SER, Supplement 'ated May 9, 1975, and
- SER, Supplement 7 dated March 1, 1976, served as the licensinc basis to approve the St. Lucie 1 safety-related structures, including the fuel handling buildinr, and considered natural phenomena.
The rerack itself will not involve any changes to the fuel handling building/spent fuel pool; thus, natural phenomena need not he reanalyzed as part of this review.
Contention 14:
"That FPhL has not properlv considered or evaluated the radiological consequences to the environment and surrounding, human population of an accident in the spent fuel pool'."
Section 8 of this document contains the staff's evaluation of postulated acc'.entS..
The consequences nf the postulated accidents are within the guidelines of 10 CFR Part 100.
Contention 15:
"That. the increase of the spent
+uel pool capacity, which includes fuel rods which have experienced fuel failure and fuel rods that are more highly enriched, will cause the reouirements of ANSI-N16-]975 not to be met and will increase the probability that a
criticality accident will occur in the spent fuel pool and will exceed 10 CFR Part 50, A 6P. criterion."
The staff used the Standard Review Plan to evaluate the criticality aspects
- o. the spent fuel pool rerack.
The results showed that the rerack is acceptable from a criticality perspective.
The staff's criticality evaluation is contained in Sectic r 2 of this document.
Contention 16:
"That FPAL has not responded to the concerns as presented by the NRC by outlinino a loading schedule for the spent
.uel pool detailing how the most recentlv discharged spent
- uel will he isolated from other recently discharged fuel and/or a full core discharge in order to mitigate potential risks from fires in the spent fuel pools fsicI resultina in releases of radioactivity into the environment in excess of the IO CFR 100 criteria."
The staff did not express a concern in reaard to a loading and storaae configuration for discharged fuel in connection with this rerack appli-cation.
The licensee proposed limiting the spent fuel assemblies having minimum burnup per proposed Technical Specification Figure 5.6-1.
The staf~ finds the proposed controls for placement o< spent fuel assemblies in Region 1 and Region 9 acceptable, and concludes that no other loadina and storage controls are necessary.
See Section 2.0 of, this document.
In addition, the staff has aenerally addressed the potential for cladding fires in Section 5 of the Environmental Assessment.
11.
FINAL NP 'SIGNIFICANT HAZARDS CON"IDERATION The licensee's request for amendment to the operating license for the St. Lucie Plant, Unit No. 1, including a proposed determination by the staff of no sigr ificant hazards consideration, was individually noticed in the Federal Renister on Auoust 31,
- 1987, fol>owed by a biweekly notice on Septemmmer H, 1~81.
This is the staff's finai detemainatinn nf nn significant hazanrls consideration.
The Commission's regulations in 10 CFR 50.92 include three standards used hy the NRC staff to arrive at a determination that a request for amendment involves no significant hazards considerations.
These regulations state that the Comnission may make such a final determination if operation of a facility in accordance with the proposed amendment would not (>} involve a significart increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new nr different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin o+.safety.
The proposed spent fuel pool expansion amendment is simi,lar to more than 100 earlier requests from other utilities f'r spent fuel pool expansion's.
The ma ority of these requests have alreadv been granted by the NRC; others are under staff review.
The knowledge and experience pained by the NRC staff in reviewing and evaluating these similar reauests were utilized in this evaluation.
The licensee's request.
does not use any new or unproven technolooy in either the analytical techniques necessary to suoport the expansion or in the construction process.
The staff has determined that the licensee's request for amendment to expand the spent fuel pool storage capacity for the St. Lucie Plant, Unit No.
1 by reracking to allnw close~ spacing of spent fuel assemblies does not significantly increase the probability or consequences of accidents previously evaluated; does not create new accidents not previously evaluated; and does not result in any significant reduction in the margins of safety viith respect to criticality, cooling or structural considerations.
The following staff evaluation in relation to the three standards demonstrates that the proposed amendment for the SFP expansion does not involve a significant hazards consideration.
First Standard "Involve a significant increase in the probability or consequences of an accident previouslv evaluated."
The following postulated accidents and events involving spent fuel storage have been identified and evaluated by the licensee.
The staff likewise evaluated the same accidents and events.
l.
A spent fuel assembly drop in the spent fuel pool.
2.
Loss of spent fuel pool cooling system flow.
3.
A se',smic event.
19 4.
A spent fuel cask drop.
5.
A construction accident.
The probability of any of the first four accidents is not affected by the racks themselves; thus the modification cannot increase the probability n<
these accidents.
As for the construction accident, the licensee will not carry any racl directly over the stored spent fuel assemblies.
All work in the spent
<uel pool area will be controlled and performed in strict accordance with specific-written procedures.
The crane that will be used tn br~kg +he racks into the Fuel Handlinc Building has been evaluated and found acceptable.
In addition, the temporary construction
- crane, which will be used to move racks within the spent cruel pool area, has been evaluated and four d acceptable.
Section 5.0 of this safety evaluation contains the details of the staff's ana'ysis.
Thus, the probability of a construction accident is not signifi-cantly increased as a result of reracking.
Accordingly,, the proposed modi-fication does not involve a significant increase in the probability of an accident previously evaluated.
As noteC in Section 2.0 of this safety evaluation, the consequences of a spent fuel assembly drop in the spent fuel pool (scenario
- 1) was evaluated and it was found that the criticality acceptance criterion, k ~
less than or equal to 0.95, is not violated.
In addition, the radiologi38f conseouences of a fuel assembly drop are not changed from the previous analysis.
The staff also conducted an evaluation of the potential consequences of a fuel handling accident.
The staff analysis found that the calculated doses are less than 10 CFR Part 100 guidelires.
The results of the analysis show that a dropped spent
<uel assembly on the racks will not distort the racks such that they would not perform their safety ~unction.
Section 8.0 contains the details o~
the staff's accident analysis.
Thus, the consequences of this type accident are not changed from the previously evaluated spent fuel assembly drops which have been
<otind acceptable, The consequences of a loss of spent fuel pool cooling system flow (scenario 2) have been evaluated and it was found that sufficient time is available to provide an alternate means for cooling fi.e., the fire hose s ations) in the event of a failure in the cooling system (see Section 5.0 of this safety evaluation}.
Thus, the consequences of this type of accident are not signif-icantly increased
~rom previously evaluated loss of cooling system flow accidents.
The consequences of a seismic event (scenario
- 3) have been eva>uated and are acceptable.
The new racks will be designed and fabricated to meet the requirements of applicable portions of the NRC Regulatory Guides and published standards.
The new free-standing racks are designed, as are the existing free-standing racks, so that the floor loading from racks completely filled with spent fuel assemblies, partially filled, or empty at the time of the incident, does not exceed the structura> capability of the spent fuel pool.
The Fuel Handling Building and spent fuel pool structure have been evaluated
<nr the increased loading from the spent fuel racks in arcordance with the criteria previously evaluated by the staff and ound acceptable.
Section 5.0 contains the details of the staff's analysis.
Thus, the consequences of a seismic event are not significantly increased from previously evaluated events.
The consequences of a spent fuel cask drop (scenario
- 4) have been evaluated (see Section 8.0 of this safety evaluation).
The radinlogical consequences o~
the cask drop are well within the guidelines o< '.0 CFR 100 and the doses are not increased as compared to the doses analyzed for the presently installed racks.
The cask drop analysis is based on administrative and Technical Specifi-rations controls which ensure that minimum requirements
".or decay of irradiated fue'. assemblies in the entire spent fuel pool are met prior to movement of the cask intn the cask area of the spent fuel pnol.
Analyses also demonstrate that k
will always be less than the NRC acceptarce criterion.
In addition, luggage from a cask drop wil> not exceed the makeup capabilities'f the spent fuel pool.
- Thus, the consequences of a cask drop accident will not increase from previously evaluated accident analyses.
The conseouerces of a construction accident (scenario
- 5) are enveloped bv the spent fue> cask drop analysis.
No rack (old or new) weighs more than a single 25 tor cask.
In addition, all movements of heavv loads handled dur'.ng the rerack operation will complv with the NRC guidelines presented in NUREG-0617.,
"Control of Heavy Loads at Nuclear Power Plants."
The consequences of a construction accider t are not increased from previously evaluated accident analyses.
Therefore, it is concluded that the proposed amendment to replace the spent fuel racks in the spent fuel pool will not involve a siqnificant increase in the probability or consequences of an accident previously evaluated.
Second Standard "Create the possibi litv of a new or different kind of accident from any accident previously evaluated."
As noted in various sections of this safety evaluation and the consultant's Technical Evaluation Report description of acceptance criteria (Section P.O),
the staff evaluated the proposed modification in accordance with the guidance of the YPC position paper entitled, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," appropriate NPC Regulatory Guides, appropriate NPC Standard Review Plar s, and appropriate industry codes and standards.
In addition, the staff has reviewed several previous NRC Safety Evaluations
<or rerack applications similar to this proposal.
No unproven techniques and methodologies were utilized in the analvsis and design o
the proposed high density racks.
No unproven technology will be utilized in the fabrication and installation process of the new racks.
The basic reracking technology in this case has been developed and demonstrated in numerous applications for a fuel pool capacity increase which have already received NRC staff approval.
Third Standard "Involve a significant reduction in a margin of safety."
The staff Safety Evaluation review process has established that the issue of margin of safety, when applied to a reracking modification, should address the following areas:
1.
2.
3.
Nuclear criticality considerations Thermal-hydraulic considerations Mechanical, material and structural considerations.
21 The established acceptance criterion for criticality is that the neutron multi-plication factor in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all cor ditions.
This margin o< safety has been adhered'o in the criticality analysis methods for the new rack design.
The methods used in the criticality analysis conform with the applicable portions of the appropriate staff guidance and industry codes, standards, and specifications.
In meeting the acceptance criteria for criticality in the spent fuel pool, such that k
is always less than 0.95, including uncer-tainties at a
95%/95% probab'fffty/confidence level, the proposed amendment to rPrQck the spent fuel pool does not involve a significant reduction in a margin of safety for nuclear criticality.
Section P.O contains the detai' of the staf 's analysis.
Conservative methods were used to calculate the maximum fuel temperature and the increase
~n temperature of the water in the spent fuel pool.
The thermal-hydraulic evaluation used the methods used for evaluations of the present spent fuel racks in demonstrating the temperature margins of safety are maintained.
The proposed modification will increase the heat load in the spent
<uel pool.
The evaluation shows that the spent fue> will be adequately cooled.
Section 5.0 contains the details of the staff's analysis.
Thus, there is no significant reduction in the margin of safety for thermal-hydraulic or spent fuel cooling concerns.
The main safety function of the spent fuel pool and the racks is to maintain the spent fuel assemblies in a safe configuration through all normal nr abnormal
- loadings, such as an earthquake, impact due tn a spent fuel cask drop, drop of a spent fuel assembly, or drop of any other heavy obiect.
The mechanical,
- material, and structural design of the new spent fuel racks is in accordance with applicable portions of the "NRC Position for Review and Acceptance o.
Spent Fuel Storage and Handling Applications," dated April 14,
- 1978, as modified January l8, 1979; Standard Review Plan 3.8.4:
and other appl'.cable NRC guidance and industry codes.
The rack materials used are compatible with the spent fuel pool and thF spent fuel assemblies (see Section 3.0 of this safety evaluation}.
The structural considerations of the new racks address margins of safety against tilting and deflection or movement, such that the racks are not damaged during impact (see Section 4e0 of this safety evalua-tion).
In addition, the spent
.uel assemblies remain intact and no criticality concerns exist.
Thus, the margins of safety are not significantly reduced by the proposed rerack.
Sunsnaru Based on the foregoing and the fact that the reracking technolooy in this instance has been well-developed and demonstrated, the Comnission has concluded that the standards of 10 CFP 50.92 are satisfied.
Therefore, the Commiss4on has made a final determination that the proposed amendment for spent fuel pool expansion does not involve a signi icant hazards consideration.
12.
ENVIRONHENTAL CONSIDERATIONS A separate Environmental Assessment has been prepared pursuant to 10 CFR Part 51.
The Notice of Issuance'of Environmental Assessment and Finding of No Significant Impact was published in the Federal
~Re ister on March 4, 1988 I53 FR 78551.
0 e
pp
~
f 13.
CONCLUSIONS The staff has reviewed and evaluated the licensee's request for amendment
~or the St. Lucie Plant, Unit 1 regarding the expansion of the spent fuel pool.
Based on the considerations discussed in this safety evaluation, the staf<
concludes that:
this amendment will not (a) significantly increase the probability or consequences of accidents previouslv evaluated, (b> create the possibilitv of a new or different accident from any accident previously evaluated,
{c) significantlv reduce a margin of safety; and therefore, the amendment does not involve significant hazards considerations, there is reasonable assurance that the health and safety nf the public will not be endangered by operation in the proposed
- manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
14.
REFERENCES -
FP&L 4,
5.
6.
7.
8.
9.
10.
11.
FP&L letter No. L-87-2~5, June 12, 1987 from C. 0.
Mnodv (FP&L) to US NRC, Subiect:
Proposed License Amendment, Spent Fuel Pool Perack.
FP&L letter No. L-87-374, September 8,
1987 from C. 0.
Woodv (FP&L} to US NRC,
Subject:
Spent Fuel Rerack.
FP&L letter No. L-87-42P, October 20, 1987 from C.
O.
Woody (FP&L) to US NPC,
Subject:
Spent Fuel Pool Rerack-Design and Analysis.
FP&L letter No. L-87-424, October 20, 1987, from C. 0.
Woodv {FP&L} to US NRC, Subiect:
Spent Fuel Pool Rerack - Boraflex and Pool Cleanup.
FP&L letter No. L-87-425, October 20, 1987, from C. 0.
Moodv
(.FP&L> to US NRC,
Subject:
Spent Fuel Pool Perack - Radioactive
- Snurces, Dose
- Rate, and Dose Assessment.
FP&L letter No. L-87-519, December 21, 1987, from C. 0.
Moody (FP&L} to US NRC, Subiect:
Environmental Effects of Transportation of Fuel and Waste.
FP&L letter No. L-87-538, December 22, 1987, from C. 0.
Woody (FP&L) to US NRC,
Subject:
Spent Fuel Pool Rerack.
FP&L letter No. L-87-535, December 23,
- 1987, from C. 0. Woodv, (FP&L) to US NRC,
Subject:
Spent Fuel Pool Rerack - Desion and Analysis.
FP&L letter No. L-87-536, December 23, 1987, from C. 0. Woody,
<FPAL) to US NRC,
Subject:
Spent Fuel Pool Rerack - Design and Analysis.
FP&L letter No. L-8>-537, December 23, 1987, from C. 0. Woody, (FP&L) to US NRC, Subiect:
Spent Fuel Rerack.
FP&L letter No. L-88-38, January 29, 1988, from C. 0. Woodv,, (FP&L) to US NRC,
Subject:
Spent Fuel Rerack.
REFEPENCES -
NRC P3 12.
13.
14.
15.
16.
17.
18.
19.
20.
21.
22.
23.
U.S. Nuclear Regulatory Commission, letter dated.!uly 16, 1987 from E.
G. Tourigny (NRC) to C. 0.
- Woody, (FP&L),
Subject:
Request for Additional Information.
U.S. Nuclear Regulatory Cormission, letter dated August 20, 1987 from E.
G. Tour~gny (NRC) to C. 0.
- Woody, (FP&L),
Subject:
Reouest for Additional Information.
U.S. Nuclear Regulatory Commission, letter dated August 25, 1987, from N. Ber kow (NRC) to C.
O.
Woody fFP&L),
Subject:
Spent Fuel Pool Expansion.
Also:
Federal
~Re ister Notice, 52 FR 32852, August 31, !o87.
U.S. Nuclear Regulatory Commission, letter dated September 1, 1987, from E.
G. Tourigny (NRC) to C. 0.
Woody (FP&L),
Subject:
Request for Additional Information.
U.S. Nuclear Regulatory Commission, meeting minutes dated September ll, 1987 from E.
G. Tourigny (NRCl,
Subject:
Summary o+ Septemher 2,
1987 Meeting with FP&L and NRC Staff Regarding the Reracking of the Spent Fuel Pool.
U.S. Nuclear Regulatory Commission, letter dated September 21, 1987, from E.
G. Tourigny (NRC) to C. 0.
Woody (FP&L),
Subject:
Request
+or Additional Information.
U.S. Nuclear Regulatory Commission, meeting minutes dated October 21, 198>
from E.
G. Tourigny (NRC), Sub;;.ect:
Summary of October 2, 1987 Meeting with FP&L and NRC Staff Regarding the Reracking of the Spent Fuel Pool.
U.S. nuclear Regulatory Commission, letter dated October 23, 1987 from E.
G. Tourigny
{NRC) to C. 0.
Woody (FP&L),
Subject:
Request for Additional Information.
U.S. Nuclear Regulatory Commission, letter dated Novembe~ "5, 1987 from E.
G. Tourigny (NRC) to C. 0.
Woody (FP&L),
Subject:
Reauest for Additional Information.
U.S. Nuclear Regulatory Commission, meeting minutes dated December 4, 1987 from E.
G. Tourigny (NRC),
Subject:
Suranary of October 29 and 30, 1987 Audit of J.
Oats and HOLTEC in Support of Reracking of the Unit 1 Spent Fuel Pool.
U.S. Nuclear Regulatory Commission, meeting minutes dated December 9,
l,987 from E.
G. Tourigny (NRC),
Subject:
Summary of November 24, 1987 Meeting between FP&L and NRC Staf+ Regarding the Reracking of the Spent Fuel Pool.
U.S. Nuclear Regulatory Cnmnission, letter dated February 29, 1988 from E.
G. Tourigny (NRC) to C. 0.
Woodv !FP&L),
Subject:
Environmental Assessment and Finding of No Significant Impact - Spent Fuel Pool Expansion, St Louie.Plant, Unit No.
1.
Also:
Federal
~Re ister Notice, 53 FR 7065, March 4, 1988.
REFERENCES -
OTHER
~
~
~
~
~
~
~
R4.
Campbell Rich to U.S. Nuclear Regulatory Commission, Secretary to the Commission, letter dated September 30, 1987.
25.
Campbell Rich to U.S. Nuclear Regulatory Commission, undated letter, enveloped postmarked january 15, 1988.
Dated:
l4arch 11, 1988
Attachment:
Appendix A