L-87-245, Applicant Exhibit A-11,consisting of Util Requesting Amend to License DPR-67 to Replace Spent Fuel Pool Racks to Ensure That Sufficient Future Capacity Exists. Proposed Tech Specs Encl
| ML20246M582 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 01/26/1989 |
| From: | Woody C FLORIDA POWER & LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| L-87-245, OLA-A-011, OLA-A-11, NUDOCS 8903270204 | |
| Download: ML20246M582 (165) | |
Text
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FLo >lD A PoWEM & LIGHT COMP ANY
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_r m, . NU. CLEAR k207t2NRY COMMISSION j D@t do.N 3#~@ Official Exh. No. O U, S. Nucleor Regulotory Commission Atin: Document Control Desk in the rnatter of Oui ek Mv * /ijSt - M dcd Washington, D. C. 20555 Staff IDENTIFIED V ;
Applicant _ RECEIVED M Centlemen: intervenor . REJECTED Re: St. Lucie Unit I cont's oftr Docket No. 50-335 Contractor - DATE / ~ M I Proposed License Amendment Other witness Spent Fuel Pool Rerock Reporter p/g h _
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In occordonce with 10 CFR 50.90, Florido Power & Light ~ Company (FPL) submits herewith a request to amend Appendix A of Focility Operating License DPR-67.
This proposed amendment permits replacement of the spent fuel pool rocks to ensure that sufficient future capocity exists for storoge of spent fuel at St.Lucie Unit 1. The new, high-density rocks increase the ovallable storage to 1706 spent fuel ossemblies and are expected to provide odequate storoge spoce until the year 2009.
0 A temporary crane will be installed and operated in the spent fuel creo during the V
rerock. Administrative procedures for the temporary construction crone will preclude the transfer of loods in excess of 2,000 pounds over the spent fuel. After l
the rerock is completed, this crone will be removed. ;
) The proposed Technical Specification changes are described below and shown on I the Attachment I marked up Technical Specificotton pages.
- 1. Technical Specificotton 3/4.9.14 Bases is revised to reflect the j assumptions used in calculations of doses based on the Decay Times. j
- 2. Technical Specification 5.6.l.a.1 is revised to correspond to the Standard Technical Speelfications for Combustion Engineering l Pressurized Water Reoctors (NUREG-0212 Rev. 2). l l 3. Technical Speelficotton 5.6.I.o.2 is revised to show the nominal center-to-center distance for the new storoge rocks.
- 4. Technical Specification 5.6.l.a.3 is edited to discuss the boron concentration only. l l
- 5. Technical Specificotlon 5.6.l.a.4 is odded to indicate the presence of Boroflex in the cells. l
- 6. Technical Specification 5.6.1.b and accompanying Figure 5.61 are ,
odded to show the increased spent fuel enrichment permitted in the j (4 pool.
8903270204 890126 PDR ADOCK 05000335 i G PDR ,
I PEOPLE St.8tviNC, PE J P.E l i
EJW4/038/l
O U. S. Nuclear Regulatory Commission L.87 245 Page two
- 7. Technical Specification 5.6.l.c is editorially changed from 't" to "c".
- 8. Technical Specificotton 5.6.3 is changed to show the copocity of the high.copocity spent fuel storage rocks, in occordance with 10 CFR 50.91(oXI), it.hos' been determined that the proposed omendment does not involve any significant hozords considerations pursuant to -
10 CFR 50.92. The No Significant Hazards Consideration determination is
- documented in Attochment 2. In addition, o Safety Analysis Report supporting the proposed omendment has been prepared and is provided as Attachment 3.
In occordance with 10 CFR 50.91(bXI), a copy of the proposed amendment is being forwarded to the State Designee for the State of Florido.
In occordance with 10 CFR 170, FPL Check No. 4797 is ottoched as remittance for the license amendment application fee.
The proposed amendment has been reviewed by the St. Lucie Focility Review -
Group and the Florida Power & Light Company Nuclear Review Board. ,
it is requested that the NRC opprove the proposed modification and proposed Techaical Specifications on on expedited schedule but no later than January 15, 1988. This schedule will suppport installation of the high. density spent fuel rocks prior to the scheduled Fall 1988 refueling outoge. Af ter that refueling outoge, rerock of the St. Lucie Unit I spent fuel pool would be extremely difficult, if not impossible, without trans shipment of spent fuel to St. Lucie Unit 2.
L Please contact us if there are questions about this submittol.
Very truly yours,
- ..h !!y !!...... ........ ......"!g.!!.!!!!!!1..!!. g !!!!!!!!.
, jj Ea
. O. W Group i President + g, INI Nuclear nergy jt:' , 5g!'m" l4,l,,, .. , u .
COW /EJW/gp Attachments (4) cc: Dr. J. Nelson Croce, Regional Administrator, Region 11, USNRC USNRC Senior Resident inspector, St. Lucie Plant Mr. Lyle Jerrett, Florido Dept. of Health and Rehabilitative Services j EJW4/038/2
- , STATE OF FLORIDA )
) ss.
l COUNTY OF PALM BEACH )
C. O. Woody being first duly sworn, deposes and says:
a Group Vice President of Florido Power & Light Company, the That he is 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.
C. O. Wood 1
Subscribed and sworn to before me this
/4 day of Ar e- _,1927 !
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A$O l NOTARY PUBLIC, in and for the County of Palm Beach, State of Florido 80f887 'llR.!C Stair Of F104mA 8' CO*n'51 ton (ap srpt gg,gggg My Commission expires: '#"I' '"> 5t*!"t 185. oc. ,
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ATTACINENT 1 MARKED UP TECHNICAL SPECIFICATION PAGES B 3/4 9-3 5-5 5-6 5-6a j
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I REFUELINGOhERATIONS BASES 3/4.9.12 FUEL POOL VENTILATION SYSTEM-FUEL STORAGE The limitations on the fuel handling building ventilation system ensures that all radioactive material released from an irradiated fuel i assembly will be filtered through the HEPA filters and charcoal adsorber I prior to discharge to the atmosphere. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assump-tions of the accident analyses.
l 3/4.9/13 SPENT FUEL CASK CRANE .
1 The maximum load which may be handled by the spent fuel cask crane is limited to a loaded single element cask which is eq:Jivalent to approxi-mately 25 tons. This restriction is provided to ensure the structural integrity of the spent fuel pool in the event of a dropped cask accident. ,
Structural damage caused by dropping a load in excess of a loaded single i element cask could cause leakage from the spent fuel pool in excess of the maximum makeup capability.
3/4.9.14 DECAY TIME - STORAGE POOL
~
The minimum requirements for decay of the irradiated fuel assemblies '
d -
in the entire spent fuel storage pool prior to movement of the spent fuel -
cient time has elapsed I cask intoradioactive to allow the fuel cask decay compartment insure of the fission that suf,f,i,The decay time of 1180 products.
hours is based upon one-third of a core placed in the'soent fuel pool each vece durino refueling......... ,.... .. .. _... ..... The decay time of 1490 hours0.0172 days <br />0.414 hours <br />0.00246 weeks <br />5.66945e-4 months <br /> is based upon one-third o,J,3,3, ore being placed in the spent fuel pool each year during refuelina ' -
- following which an entire I core is placed in the pool'to fill it.. The cask drop analysis assumes that .
all of the irradiated fuel in the filled cool - cores) is ruptured and I follows Regulatory Guide 1.25 methodology, except that a Radial Peaking Factor of 1.0 is applied to all irradiated assembi .
1
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~ UMTIL THE PC>OL.15 FILLEb j ST. LUCIE - UNIT 1 B 3/4 9-3 Amendment No. 24,40 O
DESIGN FE ATURES CONTROL ELEMENT ASSEMBLIES !
5.3.2 The reactor core shallThe contain 73 full length and no part length control element assemblies shall be designed control element assemblies.
and maintained in accordance with the original design provisions contained in Section 4.2.3.2 of the FSAR with allowance for nomal degradation pursuant to '
the applicable Surveillance Requirements.
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE !
5.4.1 The reactor coolant system is designed and shall be maintained: ,
- a. In accordance with the code requirements specified in Section 5.2 of the FSAR with allowance for nomal degradation pursuant i to the applicable Surveillance Requirements, ,
- b. For a pressure of 2485 psig, and
- c. For a temperature of 650'F, except for the pressurizer which is 700'F.
VOLUME 5.4.2 The total water and steam volume of the reacter coolant system is 11,100 : 180 cubic feet at a nominal T,yg of 567'F.
l5.5 EMERSE C CORE C00LINr. SYSTEMS I
,l5.5.1 The emergency core :ooling systems are designed and shall be main- !
i j!tainec in accorcance with the original design provisions contained in l'Section 6.3 o' the FSAR with allowance for nomal degradation pursuant to ,
the applicable Surveillance Requirements. l 5.6 FUEL STORAcE i CRITICALITY 5.6.1.a The spent fuel storage racks are designed and shall be maintained with: (yg,n p,,,g,,7)
,A 1. Ak bit $th:
t:III: equivalent to less than or equal to 0.95p :1 11;d with unborate
-4be- conservative ::;r;ti:n; :: d:::rit:1 in % ;ti;n 0.1 -
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ST. LUCIE - UNIT 1 5-5 Amendment No. 22.27,75
DESIGN FEATURES bi#
CRITICALITY (Continued) ._
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A boron concentration greater than or equal to 1720. ppm. .
-In ;dditien, f.el in tr.; ;ter;;; ;;;l ehell M ; 000 cr-k"rt :f hn th:r :- :ql t: 4.0j:ight?::rt.
The new fuel storage racks are designed for dry storage of 4
i un 'C rradiated
.Af fuel assemblies having a U-235 enrichment less than or equal to 4.0 weight percent, while maintaining a k,ff of less than or equal to p'[6 0.98 under the most reactive condition.
DRAINAGE 5.6.2 The fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 56 feet.
CAPACITY _
5.6.3 The spent fuel pool is designedfuel and shall be maintained with a storage jg, capacity limited to no more than assemblies.
1406 5.7 SEISMIC CLASSIFICATION 5.7.1 Those structures, systems and components identified as seismic Class I in Section 3.2.1 of the FSAR shall be designed and maintained to the original design provisions contained in Section 3.7 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.
5.8 METEOROLOGICAL TOWER LOCATION
( 5.8.1 The meteorological tower location shall be as shown on Figure 5.1-1.
5.9 COMPONENT CYCLE OR TRANSIENT LIMITS 5.9.1 The components identified in Table 5.9-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.9-1w r
e 5-6 Menenent No. J7,22.24.M. 75 ET l'l2SiE - unii l
m INSERTS FOR pAGE 5-6 n
U NSER1 I A
- 2. A nominal 10.12 inches center to center distance between fuel assemblies in Region 1 of the storage racks and a nominal 8.86 inches center to center distance between fuel assemblies in Region 2 of the storage racks.
INSERT B
- 4. Neutron absorber (boraflex) installed between spent fuel assemblies I in the storage racks in Region 1 and Region 2.
- b. Region 1 of the spent fuel storage racks can be used to store fuel l which has a U-235 enrichment less than or equal to 4.5 weight percent. Region {'
2 can be used to store fuel which has achieved sufficient burnup such that storage in Region 1 is not required. The initial enrichment vs. burnup requirements of Figure 5.6-1 shall be met prior to storage of fuel assemblies in Region 2. Freshly discharged fuel assamblies may be moved temporarily into Region 2 for purposes of fuel assembly inspection and/or repair, provided that the configuration is maintained in a checkerboard pattern (i.e., fuel j l
assemblies and empty locations aligned diagonally). Following such 4 inspection / repair activities, all such fuel assemblies shall be removed from Region 2 and the requirements of Figure 5.6-1 shall be met for fuel storage.
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1.5 2.0 ' ' ' ' .5 2 3.0 3.5 4.0 4.5 INITIAL ENRICHMENT, WT % U.235 FIGURE 5.6-1 INiilAL ENRICHMENT V5 '
BURNUP REQUIREMENTS FOR STORAGE OF
(], FUEL ASSEMBLIES IN REGION 2.
ST. LUCIE PLANT UNIT 1 5-68
ATTACHMENT 2 P:ge 3 of 6
/TTACHMENT 2 No Significant Hazards Consideration Florida Power & Light Company (FPL) has determined that the proposed amendment involves no significant hazards considerations, focusing on the three standards set forth in 10 CFR 50.92(c) as quoted below:
The Commission may make a final determination, pursuant to the procedures in 50.91, that a proposed amendment to an operating license under 50.2A(b) or 50.22 or for a testing facility involves no significant hazards considerations, 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. Involvr a significant reduction in a margin of safety.
FPL has determined that the activities associated with this amendment request do not meet any of the significant hazards considerations standards of 10 CFR l 50.92(c) and, accordingly, a no significant hazards considerations finding is justified. In support of this determination, the following background e information is provided, followed by a discussion of each of the above three I significant hazards considerations standards.
Background
St. Lucie Unit I has a single spent fuel pool (SFP) which at the present time contains free-standing spent fuel storage racks with 728 total storage cells.
Each cell has a center-to-center spacing of 12.53 inches. The present racks provided adequate capacity for storage of spent fuel while maintaining reserve full core off-load capacity for Cycles 1 through 7. However, beginning with startup of Cycle 8 in the Spring of 1987, St. Lucie Unit 1 lost the full-core reserve storage capebility. Therefore, to correct this situation and to ensure that sufficient spent fuel storage capacity exists at St. Lucie Unit 1 TPL has contracted with Joseph Oat Corporation for high-density spent fuel storage racks whose design incorporates a neutron absorber, Boraflex, in the cell walls and allows for more dense storage of spent fuel. The racks have an ultimate storage capacity of 1706 fuel assemblies, which is expected to extend the full-core reserve storage capability until the year 2009.
j The free standing high-density spent fuel storage racks will store fuel in two discrete regions of the STP. Region 1 includes four modules having a total of 342 storage cells. Each cell is desi.gned for storage of unburned fuel assemblies with Uranium-235 enrichments up to 4.5%, while maintaining the required suberiticality (k,ff 5 0.95).
O
QMGEp1 &
Pega 2 of 6 .
Region 2 includes thirteen modules having a total of 1364 storage cells, which ,
are available for storage of fuel assemblies. This region is designed to j
/9 store fuel which has experienced sufficient burnup such that storage in Region l D 1 is not required, id l
The high-density spent fuel storage rack cells are fabricated from 0.080 inch thick type 304L stainless steel plate. In Region 1, strips of boraflex neutron absorber material are sandwiched between the cell walls and a {
stainless steel coverplate, and the cells are separated by a 1.120 inch water '
gap. In Region 2, the boraflex strips The are sandwiched between the adjacent cells are welded together in a specified cell walls without a water gap.
manner to become a free-standing structure which is seismically qualified Each without depending on neighboring modules or fuel pool walls for support.
rack is provided with a girdle bar to absorb the energy if a rack should collide with another rack or the SFP walls. The nominal center-to-center spacings of the cells within Region 1 and 2 are 10.12 and 8.86 inches, respectively.
Since there is spent fuel presently in the St. Lucie Unit 1 STP, special l ad=inistrative controls and/or procedures will be developed to minimize The )
radiation exposure during the installation of the new spent fuel racks.
evaluation of postulated accidents with respect to nuclear criticality and/or radioactivity release has shown acceptable results, in that keff does not exceed 0.95 including uncertainties and postulated releases do not exceed 10 CFR 100 acceptance criteria.
Evaluation The f ollowing evaluation demonstrates that the proposed amendment does not The exceed any of the three significan; hazards considerations standards.
analysis of this proposed modification has been accomplished using currently l
' accepted codes and standards and the results of the analysis meet the specified acceptance criteria in these standards.
(1) Involve a significant increase in the probability or consequences of an accident previously evaluated.
In the course of the analysis, TPL has considered the following potential accident scenarios:
- 1. A spent fuel assembly drop in the spent fuel pool.
- 2. Loss of spent fuel pool cooling system flow.
- 3. A seismic event.
4 A spent fuel cask drop.
- 5. A construction accident.
J
Page 3 of 6 The probability of any of the first four accidents is not effected by tne racks themselves; thus the modification cannot increase the O probability of these accidents. 'As for the construction accident, FPL !
b does not intend to carry any rack directly over the stored spent fuel s assemblies.
All work in the spent fuel pool area will be controlled The and performed in strict accordance with specific written procedures.
crane which will be used to bring the racks into the Fuel Handling Building has been evaluated and meets the requirements of Section In 5.1.1 of NGEG-0612, Control of Heavy Loads at Nuclear Power Plants".
addition, the temporary construction crane which will be used to move racks within the spent fuel pool area will meet the design, inspection, testing, and operation requirements of Section 5.1.1 of NUREG-0612.
This program provides for the safe handling of heavy loads in the vicinity of the spent fuel pool.
Accordingly, the proposed modification does not involve a significant increase in the probability of an accident previously evaluated. !
l FPL evaluated the consequences of a spent fuel assembly drop in the spent fuel pool (scenario 1) and found that the criticality acceptance criterion, keff 5 0.95, is not violated. In addition FPL found that the radiological consequences of a fuel assembly drop are not changed from the previous analysis. The NRC also conducted an evaluation of the Both FPL and NRC potential consequences of a fuel handling accident.
analyses found that .the calculated doses are less than 10 CFR Part 100 The results of an analysis show that a dropped sp(nt fuel guidelines. assembly on the racks will not distort the racks such that they would not perform their safety function. Thus, the consequences of this type accident are not changed from the previously evaluated spent fuel aEsemMy drops which have been found acceptable by the NRC.
The consequenr# 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 (i.e., theThus, fire hose stations) in the event of a failure in the cooling system.
the consequences of this type accident are not significantly increased j from previously evaluated loss of cooling system flow accidents, The consequences of a seismic event (scenario 3), have been evaluated 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.eaisting 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, do not exceed l the structural capability of the spent fuci pool. The Fuel HandlinF Building and spent fuel pool structure have been evaluated for the increased loading from the spent fuel racks in accordance with the Thus, criteria previously evaluated by the NRC and found acceptable.
the consequences of a seismic event are not significantly increased from previously evaluated events.
v
ATTAC T2 Page 4 of 6 The consequences of a spent fuel cask drop (scenario 4) have been evaluated. The radiological consequences of the cask drop are well O within the guidelines of 10 CFR 100 and the doses are not increased as compared to the doses analysed for the presently installed racks. The cask drop analysis is based on administrative and Technical Specification controls which ensure that minimum requirements for decay of irradiated fuel assemblies in the entire spent fuel pool are met prior to movement of the cask into the cask area of the spent fuel l pool. Analyses also demonstrate that keff will always be less than the NRC acceptance criterion. In addition leakage from a cask drop will not exceed the makeup capabilities of the spent fuel pool. Thus, the consequences of a cask drop accident will not increase from previously evaluated accident analysis.
The consequences of a construction accident (scenario 5) are enveloped by the spent fuel cask drop analysis previously performed by FPL. In addition, all movements of heavy loads handled during the rerack operation will comply with the NRC guidelines presented in NUREG-0612, 4
" Control of Heavy Loads at Nuclear Power Plants." The consequences of a construction accident are not increased from previously evaluated accident analysis.
Therefore, it is concluded that the proposed amendment to replace the spent fuel racks in the spent fuel pool will not involve a significant increase in the probability or consequences of an accident previously evaluated.
Create the possibility of a new or different kind of accident from any O (2) accident previously evaluated.
FPL has evaluated the proposed modification in accordance with the guidance of the NRC position paper entitled, "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", appropriate NRC Regulatory Guides, appropriate NRC Standard Review Plans, and appropriate industry codes and standards. In addition, FPL has reviswed several previous NRC Safety Evaluation Reports for rerack applications similar to our proposal. As a result of this evaluation and these reviews, FPL finds that the proposed modification does not, in any way, create the possibility of a new or different kind of accident from any accident previously evaluated for the St. Lucie spent fuel storage facility.
(3) Involve a significant reduction in a margin of safety The NRC 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. Nuclear criticality considerations
- 2. Thermal-hydraulic considerations
- 3. Mechanical, material and structural considerations.
ATTACHMENT 2 Page 5 of 6 The established acceptsnee criterion for criticality is that the neutron multiplication factor in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all conditions. This margin of O safety has been adhered to 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 NRC guidance and industry codes, standards, and specifications. In meeting the acceptance criteria for criticality in the spent fuel pool, such that k ff is always less than 0.95, includinguncertaintiesata95%/95$probabilityconfidencelevel,the proposed amendment to rerack the spent fuel pool does not involve a significant reduction in the margin of safety for nuclear criticality.
Conservative methods are used to calculate the maximum fuel temperature and the increase in temperature of the water in the spent fuel pool.
The thermal-hydraulic evaluation uses 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 fuel pool. The evaluation shows that the existing spent fuel cooling system will maintain the bulk pool water temperature at or below 150.80F. Thus a margin 0 of safety exists such that the maximum allowable temperature of 217 F is not exceeded for the calculated increase in pool heat load. The evaluation also shows that maximum local water temperatures along the hottest fuel assembly are well below the nucleate boiling condition values. Thus, there is no significant reduction in Se margin of safety for thermal-hydraulic or
' spent fuel cooling concernu 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 nor=al or abnormal loadings, such as an earthquake, impact due to a spent fuel cask drop, drop of a spent fuel assembly, or drop of any other heavy object. 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 of Spent Fuel Storage and Handling Applications", dated April 14, 1978, as modified January 18, 1979; Standard Review Plan 3.8.4 and other applicable NRC guidance and l
industry codes. The rack materials used are compatible with the spent fuel pool and the spent fuel assemblies. The structural considerations l of the new racks address margins of safety against tilting and deflection or movement, such that. the racks are not damaged during impact. In addition the spent fuel assemblies remain intact and no criticality concerns exists. Thus, the margins of safety are not significantly reduced by the proposed rerack.
In summation, it has been shown that the proposed spent fuel storage facility modifications 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.
- - a . . . . . .. .a ~ n -
Page 6 of 6 Thus, FPL has determined and submits that the proposed amendeants cs described do not involve a significant nazard because the criteria of 10 CFR 50.92 have Additionally, the proposed amendment is most like the example (X) been met.
of " Amendments That Are Considered Not Likely To involve Significant Hazards "
Considerations" as provided in the final NRC adoption of 10 CFR 50.92 published on page 7751 of the Federal Register Volume is not likely51, No. 44, aMarch 8, to involve i 1986. This example indicates that an amendment l significant hazards consideration as follows:
i (X) An expansion of the storage capacity of a spent fuel pool when all of )
the following are satisfied: si
- 1) The storage expansion method consists of either replacing existing racks with a design which allows closer spacing between stored spent g!
fuel assemblies or placing additional racks of the original design 8l on the pool floor if space permits.
j The St Lucie Unit I spent fuel pool rerack involves the replacement of the present low capacity racks with a design which, by incorporating a neutron absorber and requiring only burned fuel be stored in Zone 2, allows closer spacing of the stored spent fuel l Zone 1 is designed for allowing safe storage of unburned cells. )
fuel.
- 2) The storage expansion method does not involve rod consolidation or double tiering.
The St Lucie Unit i racks are not double tiered and all racks will sit on the spent fuel pool floor. Additionally, the amendment application does not involve consolidation of spent fuel.
- 3) The k eff of the pool is maintained less than or equal to 0.95.
The design of the spent fuel racks contains a neutron absorber, l Boraflex, to allow closer storage of spent fuel assemblies while ensuring that the k,ff remains less than 0.95 under all conditions with pure water in the pool. Additionally, the water in the spent fuel pool contains 1720 ppm of boron as further. assurance that k,f f remains less than 0.95.
- 4) No new technology or unproven technology is utilized in either the construction process or the analytical techniques necessary to justify the arpansion.
The rack vendor has licensed at least ten (10) other racks of the same design. The construction processes and analytical techniques Thus remain substantially the same as these other ten (10) racks.
no new or unproven technology is utilized in the construction or analysis of the high-density St Lucie Unit I spent fuel racks.
Thus, this submittal meets example (X) presented in the supplementary information accompanying publication of ?.he Final Rule and is considered as not involving significant hazards considerations.
s
O f I
i FLOPJDA POWER & LIGHT COMPANY ST LUCIE PLANT - UNIT NO.1 SPEST FUEL STORAGE FACILITY MODIFICATION SAFETY ANALYSIS REPORT DOCKET NO. 50-335 O
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TABLE OF CONTENTS-PAGE SECM cN 1-1
1.0 INTRODUCTION
License Amendment Requested 1-1 1.1 1-1 1.2 Current Status Interfaces with other Organizations 1-1 1.3 1-1
- 1. l. Summary of Report 1-2 15 Conclusions 1-2 1.6 References 2-1 i 2.0 SDDRRY OF RACK DESIGN 2-1 2.1 Existing Racks New High Density Racks 2-1 2.2 3-l' 3.0 NUCLEAR AND THERMAL-HYDRAULIC CONSIDERATIONS Neutron Multiplication Factor 3-1 3.1 Normal Storage 3-1 3.1 1 3-2 3.1.2 Postulated Accidents Caleviation Methods 3-2 3.1.3 3-9 3.1.4 Rack Modification Acceptance Criteria for Criticality 3-10 3.1.5 i
Decay Heat Calculations for the Spent Fuel Pool 3-10 3.2 (Bulk)
Spent Fuel Pool Cooling System Design 3-10 3.2.1 3-11 3.2 2 Decay Heat Analyses I Spent Fuel Fool Makeup 3-14 3.2.3 Thermal-Hydraulic Analyses.for the Spent Fuel 3-15 3.3 Pool (Localized)
Basis 3-15 3.3.1 3-15 3.3.2 Model Description Cladding Temperature 3-16 3.3.3 1
J TABLE OF CONTENTS (Cont'd)
)
O PAGE O b SECTION Potential Fuel and Rack Handling Accidents 3-16 3.4 Rack Module Mishandling 3-17 3.4.1 j 3.4.2 Temporary Construction Crane Drop 3-17 Loss of Pool Cooling (Storage Rack Drop) 3-17 3.4.3 Technical Specification Changes 3-17 3.5 3-18 3.6 References
)
MECHANICAL, MATERIAL, AND STRUCTURAL CONSIDERATIONS 4-1 l 4.0 Description of Structure' 4-1 4.1 4.1.1 Description of Fuel Handling Building 4-1 Description of Spent Fuel Racks 4-1 4.1.2 4.2 Applicable Codes, Standards, and Specifications 4-5 4.2.1 NRC Documents 4-5 l 4.2.2 Industry Codes and Standards 4-7 4.3 Seismic and Impact Loads 4-8 4.4 Loads and Load Combinations 4-9 4.4.1 Spent Fuel Pool 4-9 4.4.2 Spent Fuel Racks 4-11 4.5 Design and Analysis Procedures 4-12 4.5.1 Design and Analysis Procedures for Spent 4-12 Fuel Fool 4.5.2 Design and Analysis Procedures for Spent 4-13 Fuel Storage Racks 4.6 Structural Acceptance Criteria 4-20 4.6.1 Struc'tural Acceptance Criteria for Spent 4-20 Fuel Pool Structure 4.6.2 Structural Acceptance Criteria for Spent 4-23 Fuel Storage Racks 4.6.3 Fuel Handling Crane Uplift Analysis 4-27 ,
4.6.4 impact Analysis '4-27 4.6.5 Weld Stresses 4-27 4.6.6 Summary of Mechanical Analysis 4-28 4.6.7 Definition of Terms Used In Section 4 4-29 4.6.8 Lateral Rack Movement 4-30 O
V ii
TABLE OF CONTENTS (Cont'd)
PAGE SECTION Materials, Quality control, and Special Con- 4-30 4.7 struction Techniques 4-30 4.7.1 Construction Materials 4-30 4.7.2 Neutron Absorbing Material 4-30 4.7.3 Quality Assurance 4-30 ,,
4.7.4 Construction Techniques Testing and In-Service Surveillance 4-32 ,
4.8 1 4-32 f 4.8.1 Program Intent Description of Specimens 4-32 .
4.8.2 4-32 4.8.3 Specimen Evaluation 4-33 4.9 References ,I 5-1 l 5.0 COST / BENEFIT AND EhTIRONMENTAL ASSESSMENT 5-1 5.1 Cost / Benefit and Thermal Assessment 5.1.1 Need for Increased Storage Capacity 5-1 f 5-1 j 5.1.2 Estimated Costs 5-2 )
5.1.3 Consideration of Alternatives I 5.1.4 Resources. Committed 5-2 5-2 j 5.1.5 Thermal Impact on the Environment 5-3 l 5.2 Radiological Evaluation i
Solid Radwaste 5-3 5.2.1 5.2.2 Gaseous Releases 5-3 5.2.3 Fersonnel Exposure 5-3 5.2.4 Radiation Protection During Re-Rack 5-4 Activities 5.2.5 Rack Disposal 5-5 Accident Evaluation 5-6 5.3 5.3.1 Spent Fuel Handling Accidents 5-6 Fuel Decay 5-8 5.3.2 5.3.3 Loads'Ove.r Spent Fuel 5-9 5.3.4 Temperature and Water Density Effects 5-9 5.3.5 Conclusions 5-9 5.4 References 5-10 i
iii f]
V
m TABLE OF CONTENTS LIST OF TABLES PAGE 7 A3* I TITLE 2-3 2-1 Design Data !
2-4 ,
2-2 Table of Module Data t 2-5 2-3 Module Dimensions and Veight 3-20 3-; Summary of Critical Safety Analyses 3-21 3-2 Minimum Burnup Values !
t i 3-22 3-3 Reactivity Ef f ects of Abnormal and Accident Conditions 3-4 Fuel Burnup Values for Required Reactivities (k. ) 3-23 l with Fuel of Various initial Enrichments 3-24 3-5 Comparison of Cold, Clean Reactivities Calculated at 36.5 Mwd /kgU Burnup and 4.5%
Enrichment t 3-25 3-6 Estimated Uncertainties in Reactivity Due To Fuel Depletion Effects 1 i
3-26 3-7 Long Term Changes in Reactivity in Storage Rack 3-27 3-8 Design Basis (limiting) Fuel Assembly Specifications (CE 14 x 14) 3 3-9 Thermal / Hydraulic Cases Treated 3-29 3-10 Peaking Factor Data 3-30 l Essential Heat Transfer Data for the Fuel 3-11 Pool Heat Exchanger 3-31 3-12 Power Generation Ratio Previously Discharged Batches Bulk Pool Temperature vs. Time During 3-32 3-13 Normal Discharge Pool Bulk Temperature vs. Time Subsequent to 3-33 3-14 Completion of Discharge iv
m .. . . .
TABLE OF CONTENTS l
LIST OF TABLES (Cont'd)
PAGE TA3*.E TITLE t Loss of Cooling after Co=pletion of 3-34 3-15 Normal Refueling Discharge Bulk Pool Temperature vs. Time During 3-35 3-16 Full Core Discharge 3-36' 'l 3-17 Pool Bulk Temperature vs. Time Subsequent to Completion of Full Core Discharge Loss of Cooling After Ccmpletion of 3-37 3-18 Full Core Discharge Local and cladding Temperature Data 3-38 3-19 Boraflex Experience for High Density Racks 4-34 l 4-1 Maximum Stress Summary 4-35 4-2 Stress / Strain Summary for Liners and Anchors 4-36 4-3 4-4 Soil Bearing Stresses 4-37 ]
4-38 j 4-5 Stability Safety Factors l Degrees of Freedom 4-39 J 4-6 l i
Numbering System for Gap Elements and 4-40 4-7 Friction Elements 4-41 4-8 Rack Material Data 4-42 4-9 Adjustable Height Support Material Data Bounding Values for Stress Factors 4-43 4-10 I
Nuclear Fuel Discharge Information 5-11 l 5-1 )
St Lucie Unit 1 Annual Fuel Sa'vings Attributed to 5-12 5-2 St Lucie Unit No. 1 Gaseous Releases from Fuel Handling Building 5-13 5-3 Gamma Isotopic Analysis Spent Fuel Pool Water 5-14 5-4 v-J i
m
' TABLE OT.Cr, INTS LIST OF TABLES (Cont'd)
' TITLE PAGE
) T A 3'_ E 5-5' AnticipatedLDoses During Reracking 5-15 5-6 Effect of. Temperature and Void on Calculated 5-16 Reactivity of Storage Rack 5-7 Spent Tuel Pool Purification System 5-17
' Radionuclides Analysis Report Resin Activity 5-8 Spent Fuel. Pool Airborne' Activity Radionuclides 5-18 Analysis Report l
i l
.vi
TA3LE OF CONTENTS LIST OF FIGURES O F= == 8 2-1 Fool Layout 2-2 Typical Rack Elevation - Region 1 2-3 Typical Rack Elevation - Region 2 3-1 Acceptable Burnup Domain in Region 2 of the St Lucie Plant Spent Fuel Storage Racks 3-2 Region 1 Storage Cell Geometry 3-3 Region 2 Storage Cell Geometry 3-4 Co=parison of Depletion Calculations for Fuel of 4.5* Initial Enrichment 3-5 Bulk Pool Temperature Model for Code BU1.KTEM 36 Idealization of Rack Assembly 3-7 Thermal Chimney Flow Model 4-1 Channel Element - Regions 1 and 2 4-2 Composite Box Assembly - Region 1 4-3 Gap Element - Region 1 4-4 Typical Cell Elevation - Region 1 4-5 Typical Cell Elevation - Region 2 4-6 Adjustable Support 4-7 3 x 3 Typical Array - Region 1 4-8 3 x 3 Typical, Array - Region 2 4-9 Fuel Handling Building Spectra Envelope Curves 4-10 Mat Plan and Section 4-11 Model Overall View 4-12 North South SSE vii O
l TABLE OF CONTENTS LIST OF FIGURES
(
\ E IG"?.E TITLE e 4-13 East West SSI 4-14 Vertical SSE 4-15 Schematic Model f or DYNARACK I
4-16 Rack to Rack Impact Springs 4-17 Impact Springs Arrangement at Node i f
4-18 Spring Mass Simulation for Two-Dimensional Motion 1
4-19 Test Coupon 1
I
, l l
l l
1.0 IN3 3D'.'CTI ON 1.1 LICENSE A".ENDMENT REQUESTED Florida Power & Light Cocpany (FPL) has contracted for the design and manufacture of new spent fuel storage racks to be placed into the spent fuel pool of St Lucie Unit No. 1. The purpose of the new racks is to increase the amou.t of spent fuel that can be stored in the existing spent fuel pool. The racks are designed so that they can store spent fuel assemblies in a high density array. Theref ore, FPL hereby requests that a License Amendment be issued to the St Lucie Unit No. 1 Facility Operating License DPR-67(1) to include installation and use of new storage racks that meet the criteria contained herein. Inis Safety Analysis Report (SAR) has been prepared to support this request for license amendment.
1.2 CLEENT STATUS Tne existing racks in the spent fuel pool at St Lucie Unit No.1 have 728 total storage cells. With the presently available storage cells, St Lucie Unit No. 1 lost the full-core reserve storage capability, after the seventh refueling, which was completed in the spring of 1987. To correct this situation and provide sufficient capacity at St Lucie Unit No. 1 to store discharged fuel assemblies, FPl. plans to replace the existing storage racks with new high density spent fuel storage racks. The design of the new racks will allow f or more dense storage of spent fuel, thus enabling the existing pool to store more fuel in the spent fuel pool. The new high density racks have a usable storage capacity of 1706 cells, extending the full-core-reserve storage capability until the year 2009.
If a full core offload is required in the interim, prior to the installation of the new racks, FPL intends to transfer enough of the oldest spent fuel fro:
St. Lucie Unit 1 to St. Lucie Unit 2 to allow full core offload. A proposed license a:end ent to allow spent fuel transfer was submitted in July 1986(2) and is being reviewed by the NRC.
1.3 INTERFACES WITH OTHER ORGANIZATIONS FPL has overall responsibility for this modification. Holtec International has designed the new spent fuel storage racks. Joseph Oat (J0) is responsible for the fabrication of the new spent fuel storage racks and the evaluation of those racks under accident conditions. Ebarco Services, Inc. is responsible for the building structural analysis, the evaluation of the spent fuel cooling system and the related accident evaluations. The installer, who will be chosen later, is responsible'for the installation of the new spent fuel pool racks.
1.4 SDO:ARY OT REPORT This Safety Analysis Report follows the guidance of the NRC position paper entitled, "07 Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978, as amended by the NRC letter dated January 18, 1979(3). Sections 3.0 through 5.0 of this report are consistent with the section/ subsection format and content of the NRC position paper, Sections III through V.
1 1-1 0076L/0011L
1 The nuclear and thermal-hydraulic aspects of the report (Section 3.0) address the neutron multiplication factor, considering normal storage and handling of spent fuel as well as postulated accidents with respect to criticality and the ability of the spent fuel pool cooling system to maintain sufficient cooling. l g Movement of spent fuel stored in the spent fuel pool during removal of the present racks and installation of the new racks is also addressed.
I Sectio. 4.0, which describes the mechanical, material and structural aspects 4 of the new racks, contains information concerning the capability of the fuel d assectlies, storage racks, and spent fuel pool system to withstand the effects i of natural phenomena and other service loading conditions. l 1
Tne environmental aspects of the report (Section 5.0) concern the thermal and l radiological release from the facility under normal and accident conditions. I This section also addresses the occupational radiation exposures, generation I of radioactive waste, need for expansion, commitment of material and I non-material resources, and a cost-benefit assessment.
l
1.5 CONCLUSION
S On the basis of the evaluations and information presented in this report, plus operating experience with high density fuel storage at St Lucie Unit 2 and l
Turkey Point Unit 3, FPL concludes that the proposed modification of St Lucie l Unit No. I spent fuel storage facilities provides safe spent fuel storage, f and that the modification is consistent with the facility design and operating !
criteria as provided in the FSAR(4) and operating license.
1.6 REFERE.NCES 1 1
1 l
- 1. St Lucie Unit No. 1 Facility Operating Licenses DPR 67, Docket I i
No. 50-335.
- 3. Nuclear Regulatory Commission, Letter to All Power Reactor Licensees, from B. K. Grimes, April 14, 1978, OT position for Review and Acceptance of Spent Fuel Storage and Handling Applications," as amended by the NRC letter dated January 18, 1979.
- 4. St Lucie Plant Unit No. 1 Updated Final Safety Analysis Report, Docket No. 50-335.
8 1-2 0076L/00111
_ - _ _ _ - _ - _ _ _ _ _ - _ _ _ _ ~
- v. . .. . . .
2.0 SO m RY OF RACE DESIGN 2.1 EXISTING RACKS x_,/ The spent fuel pool at St. Lucie Unit 1 presently contains spent fuel asse=biv storage racks which are designed to provide storage locations for up to 728 fuca assemblies. The racks are designed to maintain the stored fuel in a saf e, coolatie, and suberitical configuration during normal and sbnormal conditions.
Tne present storage racks are a rectangular array composed of 14 modules.
Each storage rack module is self supportiag and rests on stainless steel !
pads. Tne present racks are free standing in that they are neither The bolted nor interface welded to the floor, nor are they attached to the pool walls.
with the pool boundaries is designed to transfer normal and sheer loads via l
tne rack supports into the pool bottom slab. I Eaci. fuel assembly storage module is composed of rectangular storage cavities fabricated from one-quarter inch thick stainless steel plate, with each cavity capable of accepting one fuel assembly. The fuel assembly storage cavities have lead-in surfaces at the top to provide guidance for insertion of fuel assemblies. Tne cavities are open at the top and bottom to provide a flow path for convective cooling of spent fuel assemblies through natural circulation. The fuel assembly storage cavities are connected by a chevron grid structure to form modules which limit structural deformations and maintain a no=inal center-to-center spacing of 12.53 inches between adjacent storage cavities during design conditions including seismic. l For further inf ormation on the existing spent fuel storage racks see Section ,
l 9.1.2 in the St Lucie Unit No. 1 updated FSAR.
(
2.2 NEW HIGH DENSITY RACKS The new high density spent fuel storage racks consist of individual cells with 8.65 inch by 8.65 inch (nominal) square cross-section, each of which accommodates a single Combustion Engineering or Exxon PWR fuel assembly or equivalent, from either St. Lucie Unit 1 or Unit 2. A total of 1706 cells are arranged in 17 distinct modules of varying siees in two regions. Region 1 is designed for storage of new fuel assemblies with enrichments up to 4.5 weight percent U-235. Region 1 is also designed to store fuel assemblies with enrichments up to 4.5 weight percent U-235 that have not achieved adequate burnup for Region 2. The Region 2 cells are capable of accommodating fuel assemblies with various initial enrichments which have accumulated minimum burnups within an acceptable b,und as discussed in this report. For example, corresponding to 4.5 and 4.0' pc?:ent initial enrichments, the minimum required burn-ups for safe storage in F4 *en 2 are 36.5 and 30.9 MWD /KgU, respectively. Figure 2-1 shc a 6ae arrangement of the rack modules in the spent fuel pool.
Tne high density racks are engineered to achieve the dual objective of maximum protection against structural loadings (arising from ground motion, thermal stresses, etc.) and the maximization of available storage locations. In general, a greater width-to-height aspect ratio provides greater margin against rigid body tipping. Hence, the modules are made as large as possible g- within the constraints of transportation and site handling capabilities.
2-1 0076L/00111
--__._________m ___
As shown in Figure 2-1, there are 17 discrete modules arranged in the fuel pool. Each rack module is equipped (see Figures 2-2 and 2-3) with girdle The nominal gap between ad p bars, 3/4-inch thick by 3-1/2 inches high.
Themodulesmakesurfacecontactbetweenthejacent r d modules is 1-1/2 contiguous walls inches.the at girdle bar locations and thus maintain a specifiedThe gap ,
Table 2-1 gives the relevant design data on each region.
modules in the two regions are of eight different types. Tables 2-2 and 2-3 between them.
succarize the physical data for esch module type.
Ine poison in Regions 1 and 2 is Boraflex. The use of this absorber material is to-preclude inadvertent criticality. i l-1
- l O ,
!O 2-2 0076L/00111
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TABLE'2-1 DESIGN DATA f
Min.' B-10 Flus Trap Losding Gap Region Cell Pitch-(nominal inch) (areal density) (nominal inch) 10 12 .020 sm/cm2 .1.12 .-
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.007 gm/ce2 0,0 2 8.86 Q
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Min. B-10 Flux Trap Loading Cap-Region Cell Pitch (nominal inch) (areal density) (nominal inch) 10.12 .020 gm/cm2' 1.12 1
8.86 .007 gm/cm2 0.0 2
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2-3 0076L/0%;'_
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TABLE 2-2 1 TABLE OF HODULE DATA l
lI NO. OF NO. OF CELLS CELLS TOTAL NO.
No. OF IN N-S IN E-W OF CELLS ;
DIRECTION - DIRECTION PER MODULE .l HODULE I.D. MODULES 1
1 Region 1 2 9 9 81 i
Al to A2 i
i Region 1 2 9 10 90 31 to B2 i
Region 2 4 13 9 117 l C1 to C4 Region 2 3 13 8 104 D1 to D3 .
Region 2 2 11 8 88 El to E2 Region 2 1 12 8 96 F1 Region 2 2 12 9 108 C1 to C2 ;
Region 2* 1 13 8 96 H1 !
i
- Cells missing in this module due to sparger. ;
i Refer to Figure 2-1.
O 2-4 0076L/0011L
O TABLE 2-3 MODULE DIMENSIONS AND WEIGHT NOMINAL CROSS-SECTION ESTIMATED DRY DIMENSIONS WEIGHT (1bs)
MODULE I .D. N-S E-W PER MODULE Region 1 90-1/4" 90-1/4" 26,700 Al to A2 .
Region 1 90-1/4" 100-7/16" 29,800 B1 to B2 Region 2 115-11/16" 80-1/6" 24.100 C1 to C4
-- Region 2 - 115-11/16" 71-3/16" 21.500
- O.' D1 to D3 V
Region 2 97-7/8" 71-3/16" 18,200 i El to E2 Region 2 106-3/4" 71-3/16" 19,800 F1 Region 2 106-3/4" 80-1/16" 22,300 G1 to G2 Region'2 115-11/16" 71-3/16" 19,800
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3.0 NUCLEAR AND THERML-HYDRAULIC CONSIDERATIONS l 3 NEUTRON MULTIPLICATION FACTOR O .1The followleg subsections describe the conditions in the apent fu are assumed in calculating the effective neutron multiplication factor e]
l
>l (k eff), the analysis methodology, and the analysis results. l 3.1.1 Normal Storage
,l The criticality analyses of each of the two separate regions of the spent fuel l I
storage pool are aussarized in Table 3-1 for the anticipated normal storage conditions. The calculated maximus reactivity in Region 2 includes a burnup-dependent allowance for uncertainty in depletion calculations and,0.0065 ,i4 furthermore, provides an additional sargin of effective multiplication factor (keff) of 0.95. As cooling time increases in long-ters storage, decay of Pu-241 results in a significant decrease in reactivity, which will provide an increasing suberiticality margin and Spacing tends to further compensate for any uncertainty in depletion calculations.
between two different rack modules is sufficient to preclude adverse nuclear interaction, since the minisus spacing between racks is greater than the design water gap spacing.
Region 2 can accommodate fuel of various initial enrichments and discharge fuel burnups, provided the combination falls within the acceptable domain illustrated in Figure 3-1 For convenience of reference, the sinimum burnup values in Figure 3-1 have been fitted by linear tangents at various values and Linear interpolation between the O the results are tabulated in Table 3-2. tabulated values will always yield value of limiting burnups.
These data will be implemented in appropriate administrative procedures to assure verified burnup as specified in draft Regulatory Guide 1.13, Revision
- 2. Administrative procedures will also be employed to confirm and assure the presence of soluble poison in the pool water at all times, providing a further margin of safety and assuring suberiticality in the event of fuel displacement during fuel handling operations, as discussed in Section 3.1.2.
3.1.1.1 New Fuel Storage in Region 2 Criticality analyses confirm that a checkerboard for pattern (fuel assemblies the storage of fresh fuel aligned diagonally) provides an acceptable k .
assemblies of 4.5% enrichment in Region 2. These calculations indicate a nominal k., of 0.819 + 0.025 (95%/95%) when fully flooded with clean unborated water. This value is substantially less than the limiting keff of I 0.95, even with the addition of a reasonable allowance for uncertainties. !
With Boraflex absorber between assemblies, conditions do not exist for the appearance of a peak in reactivity at low moderator densities, Thus, the checkerboard pattern of new 4.5% enriched fuel in Region 2 represents a safe configuration in conformance with both Standard Review Plan (SRP) 9.1.1 and 9.1.2.
3-1 0076L/0011L
- - ~ - - - _ . - _ - _ _ _ . _ _ _ _ _ _ _ _ _ _
a --
3.1.2 Postulated Accidents O Although credit for the soluble poison normally present in the spe:.t fuel pool water accident is permitted under abnormal or accident conditions *, mos e The effects on reactivity of credible 0.95) even in the absence of soluble poison. Of these abnormal /
abnormal accident conditiora, and accident conditions are summarized in Table 3-3.only on reactivity effect.
The inadvertent misplacement of a fresh fuel assembly (either into a Region 2 a storage cell or outside and adjacent to a rack module) has the potential for exceeding the limiting reactivity should there be a concurrent and independent Administrative j
accident condition resulting in the loss of all soluble poison.
l procedures assure the presence of soluble poison at all times and will preclude th I possibility of the simultaneous occurrence of these two independent accide-t conditions. The largert reactivity increase occurs for accidentally placing a new fuel assenbly into a R.gaon 2 storage cell Under with allthis other cells fully condition, loaded with the presence of fuel of the highest permissible reactivity.
approximately 500 ppm soluble boron assures that the With infinite multiplication factor the normal would not exceed the design basis reactivity for Region 2.k., is less than 0.80 concentration of soluble poison present (1720 ppa boron),
and the storage racks would not be critical even if Region 2 were to be fully TM s concentration of soluble boron loaded with fresh fuel of 4.5% enrichment.
also precludes the possibility of exceeding tb* criticality limit in the event of a dropped cask accident.
See Section 5.3 for discussions on Accident Evaluations.
3.1.3 Calculation Methods 3.1.3.1 Criticality Analysis for Region 1 3.1.3.1.1 Nominal Design Case values calculated by Urder normal conditions, with nominal dimensions, the k.,
three different methods of analysis are as follows:
Maximum k.,
Bias-corrected k. (95%/95%)
I Analytical Method 0.9313 + 0.0018 0.9331 CASMO-2E 0.9294 AMPX-KENO (27-gp SCALE) 0.9210 + 0.0084 0.9313 0.9313 Diffusion / blackness .
the ory The AMPX-KENO calculations include a one-sided For tolerance factor (13) the nominal corresponding to 95% probability at a 95% confidence limit. !
design case, the CASMO-2E calculation yields the highest reactivity and, therefore, the independent verification calculations substantiate CASMO-2E as the primary calculational method. !
l p
- Double contingency principle of ANSI N16.1-1975, as specified in the April 4, i 1978 NRC letter (Section 1.2) and implied in the proposed revision (draft) to i
(.,
Reg. Guide 1.13 (Section 1.4, Appendix A).
3-2 0076L/0011L
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - 1
l 3.1.3.1.2 Boron Loading Variation O The Boraflex absorber sheets used in Rel, ton 1 storage cells are nominally Q 0.075 inch thick, with a B-10 areal density of 0.0238 g/cm2 Independent ,
manufacturing tolerance limits are + 0.007 inch in thickness and 10 009 g/cm3 iu B-10 content. This assures that at any point where the minimum boron concentration (0.1158 gram B-10/cm3) and sinimus Borafitex thickness (0.068 inch) may coincide, the boron-10 areal density will not be lesc than 0.020 g/cm2 Differential CASMO-2E calculations indicate that these tolerance limits result in reactivity uncertainty of i 0.0021 A k for '
boron content and 1 0.0044 A k for Boraflex thickness variations.
3.1.3.1.3 Storage cell Lattice pitch variation b'
The design storage cell lattice spacing between fuel assemblies in Region 1 is 10.12 inches. A decrease in storage cell lattice spacing may or say not increase reactivity depending upon other dimensional changes that may be associated with the decrease in lattice spacing. Increasing the water thickness between the fuel and the inner stainless steel box results in a small increase in reactivity. The reactivity effect of the flux-trap water thickness, however, is more significant, and decreasing the flux-trap water thickness increases reactivity. Both of these effects have been evaluated for independent design tolerances. l The inner stainless steel box dimension, 8.650 1 0.032 inches, defines the inner water thickness between the fuel and the inside box. For the tolerance limit, the uncertainty in reactivity is 1 0.0011 A k as determined by differential CASMO-2E calculations, with k increasing as the inner stainless steel box dimension (and derivative lattice spacing) increases.
The design flux-trap water thickness is 1.120 1 0.040 inches, which results in an uncertainty of + 0.0043 4 k due to the tolerance in flus-trap water
~~
j thickness, assuming the water thickness is simultaneously reduced on all four sides. Since the manufacturing tolerances on each of the four sides s',e I I
statistically independent, then actual reactivity uncertainties would be less than 1 0.0043, although the more conservative value has been used in the criticality evaluation. .
3.1.3.1.4 Boraflex Width Tolerance Variation The reference storage cell design for Region 1 (Figure 3-2) uses a Boraflex blade width of 7.50 1 0.0625 inches. A positive increment in reactivity occurs f or a decrease in Botaflex absorber width. For a reduction in width of tne maximum tolerace. 0.0625 inch, the calculated positive reactivity -
increment is +0.0017 A k.
3.1.3.1.5 Stainless Steel Thickness Tolerances The nominal stainless steel thickness in Region 1 is 0.0B0 + 0.005 inch for the inner stainless steel box and 0.020 + 0.003 inch for the Boraflex coverplate. The maximum positive reactivity effect of the espected stainless steel thickness tolerance variations, statistically combined, was calculated (CASMO-2E) to be + 0.0010 6 k.
O 3-3 0076L/0011L
3.1.3.1.6 Fual Enrichtent and D2nsity Variation The design maximum enrichment is 4.50 1 0.05 wt% U-235. Calculations of the sensitivity to small enrichment variations by CASMO-2E yielded a coefficient O of 0.0054 4k per 0.1 wt% U-235 at the design enrichment. For a tolerance on U-235 enrichment of 1 0.05 in vtt, the uncertainty on k. is 1 0.0027 4 k.
Calculations were also made with the UO2 fuel density increased to the maximum expected value of 10.811 g/cm3 (smeared density). For the reference design calculations, the uncertainty in reactivity is 1 0.0005 4 k over the maximum expected range of UO2 densities.
3.1.3.1.7 Fuel Pin Pitch Normally, the fuel pins in the lattice are arranged on a 0.577 inch lattice spacing. For the maximum expected tolerance of 1 0.0023 inch, the calculated uncertainty is 1 0.0024 A k.
3.1.3.1.8 Eccentric Positioning of Fuel Assembly in Storage Rack l The Fuel Assembly is assumed to be normally located in the center of the storage rack cell. Calculations were also made with the fuel assemblies assumed to be the corner of the storage rack cell (four-assembly cluster at closest approach). These calculations indicated that the reactivity increases very slightly, as determined by differential PDQ-07 calculations with diffusion coefficients
- generated by NULIF and a blackness theory routine. This uncertainty is included in the evaluation of the highest possible reactivity of the Region 1 storage cells.
3.1.3.1.9 Summary of Region 1 Criticality Results j Table 3-1 demonstrates that the CASMO-2E calculated results for Region 1 storing fresh fuel at 4.50 w/o U-235 enrichment plus calculational bias and uncertainties exhibit a maximum k . of 0.9409 which allows a margin of l 0.0091 A k below the limiting effective multiplication factor of 0.95. l 3.1.3.2 Criticality Analysis for Region 2 3.1.3.2.1 Nominal Design Case The principal method of analyns in Region 2 was the CASMO-2E code, using the restart option in CASMO to transfer fuel of a specified burnup into the storage rack configuration at a reference temperature of 40C (maximum moderator density). Calculations were made for fuel of several different initial enrichments and, at'~each enrichment, a limiting k . value was established which included an additional factor for uncertainty in the burnup analysis and for the axial burnup distribution. The restart CASMO-2E calculations (cold, clean, rack geometry) were then interpolated to define the burnup value yielding the limiting k . value for each enrichment, as indicated in Table 3-4. These converged burnup values define the boundary of the acceptable domain shown in Figure 3-1
- This calculational approach was necessary since the reactivity effects are too small to be calculated by KENO, and CASMO-2E geometry is not readily O amenable to eccentric positioning of a fuel assembly.
3-4 0076L/0011L
At a burnup of 36.5 Mwd /kgU, the sensitivity to burnup is calculated to be
-0.0074 4 k per Mwd /kgU. During long-term storage, the k , values of the Region 2 fuel rack will decrease continuously from decay of Pu-241, as indicated in Section 3.1.3.3.4. o Two independent calculational methods were used to provide additional confidence in the reference Region 2 criticality analyses. Fuel of 1.69%
initial enrichment (approximately equivalent to the reference rack design for burned fuel) was analyzed by AMPX-KENO (27-group SCALE cross-section library) and by the CASH 0-2E model used for the Region 2 rack analysis. For this case, the CASMO-2E k . (0.9304) was within the statistical uncertainty of the bias-corrected value (0.9347 + 0.0064) (95%/95%) obtained in the AMPX-KENO
~
calculations. This agreement confirns the validity of the primary CASMO-2E calculations.
The second independent method of analysis used was the NULIF code for burnup analysis, and for generating diffusion theory constants (cold.. clean) for the composition at 36.5 Mwd /kgU with futt ci 4.5% initial enrichment. These constants, together with blackness theory constants for the Boraflex absorber, were then used in a two-dimensional PDQO*/ calculation for the storage rack configuration. The result of this calculation (k . of 0.8959) was somewhat lower than the corresponding CASMO-2E calculation for the same conditions (k.
of 0.9114) and thus also tends to confirm the validity of the primary calculational method.
3.1.3.2.2 Boron Loading Variation The Boraflex absorber sheets used in the Region 2 storage cells are nominally G 0.031 inch thick with a B-10 areal density of 0.0097 g/cm2 Independent manufacturing limits are 1 0.007 inch in thickness and 1 0.009 g/cm3 in B-10 content. This assures that at any point where the minimum boron concentration (0.1158 g B-10/cm3) and the minimum Borafier thickness (0.024 inch) may coincide, the boron-10 areal density will not be less than 0.007 g/cm2, Diff erential CASMO-2E calculations indicate that these tolerance limits result in an incremental reactivity uncertainty of 1 0.0036 4 k for boron content and 1 0.0111 Ak for Boraflex thickness.
3.1.3.2.3 Boraflex Width Tolerance .
The reference storage cell design for Region 2 (Figure 3-3) uses a Boraflex absorber width of 7.25 + 0.0625 inches. For a reduction in width of the maximum tolerance, the ~alcula*.ed positive reactivity increment is 0.0011 A k.
3.1.3.2.4 Storage Cell l'attice Pitch Variations The design storage cell lattice spacing between fuel assemblies in Region 2 is 8.86 + 0.04 inches, corresponding to an uncertainty in reactivity of 0.0016 A k.
3.1.3.2.5 Stainless Steel Thickness Tolerance The nominal thickness of the stainless steel box vall is 0.080 inch with a tolerance limit of + 0.005 inch, resulting in an uncertainty in reactivity of
~
10.0002 A k.
3-5 0076L/0011L
3.1.3.2.6 Fuel Enrichment, Density and Pin Pitch Variation ,
2 G Uncertainties in reactivity due to tolerances on fuel enrichment, UOdensity, and pin determined for Region 1.
3.1.3.2.7 Eccentric Positioning of Fuel Assembly in Storage Rack The fuel assembly is assumed to be normally located in the center of the storage rack cell. Calculations were also made with the fuel assemblies assumed to be in the corner of the storage rack cell (four-assembly cluster at closest approach). These calculations indicated that the reactivity decreases very slightly, as determined by PDQ07 calculations with diffusion coefficients generated by NULIF and a blackness theory routine. The highest reactivity therefore corresponds to the reference design with the fuel assemblies positioned in the center of the storage cells.
3.1.3.3 Analytical Methodology 3.1.3.3.1 Reference Analytical Methods and Bias The CASMO-2E computer code (1, 2, 3), a two-dimensional sultigroup transport theory code for fuel assemblies, has been benchmarked and is used both as a primary method of analysis, and as a means of evaluating small reactivity increments associated with manufacturing tolerance. CASMO-2E benchmarking rest'.ted in a calculational bias of 0.0013 1 0.0018 (95%/95%).
In fuel rack analyses, for independent verification, criticality analyses of the high density apent fuel storage racks were also performed with the 4s AMPX-KENO computer package (4, 5), using the 27-group SCALE
- cross-section library (6) with the NITAk'L subroutine for U-238 resonance shielding effects (Nordheim integral treatment). Benchmark calculations resulted in a bias of 0.0106 1 0.0048 (951/95%).
In the geometric sodel used in KENO, each fuel rod and its cladding were described explicitly. In Region 1 calculations, a reflecting boundary condition (zero neutron current) was used in the axial direction and at the centerline of the water gap between storage cells. These boundary conditions have the effect of creating an infinite array of storage cells in all directions. In Region 2, the zero current boundary condition was applied at the center of the Boraflex absorber sheets between storage cells. The AMPX-KENO Monte Carlo calculations inherently include a statistical uncertainty due to the random nature of neutron tracking. To minimize the statistical uncertainty of de KENO-calculated reactivity, a total of 50,000 neutron histories is normally accumulated for each calculation, in 100 generations of 500 neutrons each.
- SCALE is an acronym for Standardized Computer Analysia for Licensing Evaluation, a standard cross-section set developed by ORNL for the USNRC.
O i l
3-6 0076L/0011L
1 l
CASHO-2E is also _used for burnup calculations, with independent verification j by EPRI-CELL and NULIF calculations. In tracking long-ters (30-year)
O reactivity effects of spent fuel stored in Region 2 of the fuel storage rack, l
1 EPRI-CELL calculations indicate a continuous reduction in reactivity with time l (after Xe decay) due primarily to Pu-241 decay and An-241 growth.
A third independent method of criticality analysis, utilizinE diffusion / blackness theory, was also used for additional confidence in results l of the primary calculational methods, although no reliance for criticality safety is placed on the reactivity value from the diffusion / blackness theory technique. This technique, however, is used for auxiliary calculations of the small incremental reactivity effect of eccentric fuel positioning that would otherwise be lost in normal KINO statistical variations, or would be inconsistent with CASMO-2E geometry limitations.
I Cross sections for the diffusion / blackness theory calculations were derived from the NULIF computer code (7), supplemented by a blackness theory routine that effectively imposes a transport theory boundary condition at the surface of the Boraflex neutron absorber. Two different' spatial diffusion theory codes, PDQ07(B) in two dimensions and SNEID* in one dimension, were used to calculate reactivities.
3 1.3.3.2 Fuel Burnup Calculation Fuel burnup calculations in the hot operating condition were performed primarily with the CASMO-2E code. However, to enhance the credibility of the burnup calculations, the CASMO-2E results were independently checked by calculations with the NULIF code (7) and with EPRI-CELL (9). Figure 3-4 o compares results of these independent methods of burnup analysic under hot reactor operating conditions. The results agree with the CASMO calculation within 0.0054 d k in the hot operating condition. An archive calculation with the CHEETAH-P code is also presented in Figure 3-4 for additional confidence.
Similar comparisons were obtained in burnup calculations for other initial enrichments, as indicated in Figure 3-4.
In addition to depletion calculations under hot operating conditions, reactivity comparisons under conditions more representative of fuel to be ,
stored in the racks (cold, zenon-free) are also significant in storage rack criticality analyses. Table 3-5 compares the cold, senon-free reactivities ;
calculated by CASMO-2E, EPRI-CELL, and diffusion / blackness theory. In the rack under cold conditions, the CASMO-2E calculations gave a slightly higher reactivity value for the Region 2 fuel storage cell, and the good agreement generally observed lends credibility to the calculations.
- SNEID is a one-dimensional diffusion theory routine developed by Bisek &
Veatch and verified by comparison with PDQ07 one-dimensional calculations.
O 3-7 0076L/0011L
No definitive cathod exists for determining the uncertainty in J burnup-dependent reactivity calculations. All of the codes discussed above l have been used to accurately follow reactivity loss rates in operating
^ reactors.
CASMO-2E has been extensively benchmarked (1, 2, 3,10) against cold, clean, critical experiments (including plutonium-bearing fuel), Monte I Carlo calculations, reactor operations, and heaq -element concentration in irradiated fuel. In particular, the analyses (10) of 11 critical experiments with plutonium-bearing fuel gave an average kef f of 1.00210.011 (951/95%),
In addition, showing adequate treatment of the plutonium nuclides.
Johansson(11) has obtained very good agreement in calculations of close-packed, high-plutonium-content, experimental configurations. \
Since critical-experiment data with spent fuel is not available, it is necessary to assign an uncertainty in reactivity based on other considerations, supported by the close agreement between different calculational methods and the general industry experience in predicting reactivity loss rates in operating plants. Over a considerable portion of the ,
burnup, the reactivity loss rate in PWRs is approximately 0.01 dk By for each i Mwd /kgU burnup, becoming somewhat smaller at the higher burnups.
l !
I conservatively assuming an uncertainty in reactivity
- of 0.0005 times the burnup in Mwd /kgU, a burnup-dependent uncertainty is defined that increases with increasing fuel burnup, as would be reasonably expected. This assumption provides an estimate of the burnup uncertainty that is more conservative and bounds estimates frequently employed in other fuel rack licensing applications (i.e., 5% of the total reactivity decrement). At the design basis burnup of 36.5 Mwd /kgU, the estimate of burnup uncertainty is 0.0183 4 k; Table 3-6 summarizes results of the burnup analyses and estimated uncertainties at other burnups. These uncertainties are appreciably larger, in general, than would p be suggested by the industry experience in predicting reactivity loss rates and boron let-down curves over many cycles in operating plants. The
( ;
increesing level of conservatism at the higher fuel burnups provides an V
adequate margin in the uncertainty estimate to accommodate the possible existence of a small positive reactivity increment from the axial distribution in burnup (see Section 3.1.3.3.3). In addition, although the burnup uncertainty may be either positive or negative, it is treated as an additive term rather than being combined statistically with other uncertainties. Thus, the allowance for uncertainty in burnup calculations is considered to be a conservative estimate, particularly in view of the substantial reactivity l decrease with aged fuel, as discuseed in Section 3.1.3.3.4.
- Only that portion of the uncertainty due to burnup. Other uncertainties are accounted for elsewhere.
O 3-8 0076L/0011L
3.1.3.3.3 Effect of Axial Burnup Distribution O Initially, fuel loaded into the reactor will burn with a slightly skewed b cosine power distribution. As burnup progresses, the burnup distribution will tend to flatten, becoming more highly burned in the central regins than in the upper and lower ends. This effect may be clearly seen in the curves compiled in Reference 12. At high burnup, the more reactive fuel near the ends of the fuel assembly (less than average burned) occurs in regions of lower reactivity worth due to neutron leakage. Consequently, it is expected that distributed-burnup fuel assemblies would exhibit a slightly lower reactivity than that calculated for the average burnup. As burnup progresses, the distribution, to some extent, tends to be self-regulating as controlled by the axial power distribution, precluding the existence of large regions of significantly reduced burnup.
A number of one-dimensional diffusion theory analyses have been made based These analyses I upon calculated and seasured axial burnup distributions. i confirm the minor, and generally negative, reactivity effect of the axially distributed burnup. The trends observed, however, suggest the possibility of
]
a small positive reactivity effect at the high burnup values (estimated to be ]
due to as much as 0.006 A k at 36.5 Mwdv /ks ); (but the uncertainty in k.Section 3.1.3.3.2), is burnup, assigned at the higher burcups conservative to encompass the potential for a small positive reactivity effect i of axial burnup distributions. Furthermore, reactivity significantly J decreases with time in storage (Section 3.1.3.3.4), and, in addition, there is a further margin in reactivity (>0.006 Ak) since the nazimum calculated value (0.9435) is below the limiting keyf value (0.95). These factors would i accommodate any reasonable. reactivity effects that might be larger than expected.
3.1.3.3.4 Long-term Decay Since the fuel racks in Region 2 are intended to contain spent fuel for long periods of time, calculations were made using EPRI-CELL (which incorporates the CINDER code) to follow the long-term changes in reactivity of spent fuel over a 30-year period. C1NDER tracks the decay and burnup dependence of some 1 l
179 fission products. Early in the decay period, zenon grows from iodine l
decay (reducing reactivity) and subsequently decays, with the reactivity reaching a maximum at 100-200 hours. The decay of Pu-241 (13 year half-life) j and growth of Am-241 substantially reduce reactivity during long term storage, I as indicated in Table 3-7.
The reference design criticality calculations do not take credit for this long-term reduction iti* reactivity, other than to indicate an increasing suberiticality margin in Region 2 of the spent fuel storage pool.
3,1.4 Rack Modification The design basis fuel assembly, illustrated in Figure 3-2,is a 14 x 14 array of fuel rods with 20 rods replaced by 5 control rod guide tubes. Table 3-8 summarizes the design specifications and the expected range of significant variations. Independent calculations, with other potential fuel assembly specifications, confirmed that the 14 x 14 CE design exhibited the highest reactivity and was therefore used as the design basis.
3-9 0076L/0011L
Region 1 Storage Cells s 3.1.4.1 ,
q Ine nominal spent fuel storage cell used for the criticality analyses of C1 Region i storage cells is shown in Figure 3-2. The rack is composed of '
Boraflex absorber material sandwiched between an 8.65-inch I.D., 0.080-inch thick inner stainless steel box, and a 0.020-inch outer stainless steel i l
coverplate. The fuel assemblies are centrally located in each storage cell on a nominal lattice spacing of 10.120 1 0.05 inches. Stainless steel gap channels connect one storage cell box to another in a rigid structure and define an outer water space between boxes. This outer water space constitutes ,
a flux-trap between the two Boraflex absorber sheets that are essentially opaque (black) to thermal neutrons. The Boraflex absorber has a thickness of 0.075 _+ 0.007 inch and a nominal B-10 areal density of 0.0238 g/cm2 i
3.1.4.2 Region 2 Storage Cells l
Region 2 storage cells were designed for fuel of 4.5 wt% U-235 initial I enrichment burned to 36.5 Hwd/kgU. In this region, the storage cells are composed of a single Boraflex absorber sandwiched between the 0.080-inch stainless steel walls of adjacent storage cells. These cells, shown in Figure 3-3, are located on a lattice spacing of 8.86 1 0.040 inches. The Borafier absorber has a thickness of 0.031 + 0.007 inch and a nominal B-10 areal density of 0.0097 g/cm2, 3.1.5 Acceptance Criteria for Criticality Criticality is precluded by spacing of the fuel assemblies, which ensures that a suberitical array of keff less than or equal to 0.95 is maintained, assuming unborated pool water. The pool, however, will always contain borie acid at the refueling concentration of 1720 ppm whenever there is irradiated fuel in the pool.
l The neutron multiplication factor in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all conditions. Calculated maximum reactivity uncertainties for fuel stored in the racks are presented in Table 3-1.
Methods of initial and long-term verification of poison material stability and mechanical integrity are discussed in Section 4.8.
3.2 DECAY HEAT CALCULATIONS FOR THE SPENT FUEL POOL (BULK) ]
3.2.1 Spent Fuel Pool Cooling System Design For normal refueling discharge conditions, one fuel pool pump and the fuel l pool heat exchanger are in service. During abnormal refueling conditions, such as full core discharge, two fuel pool pumps and the heat exchanger are in service. The system is manually controlled and the operation monitored locally, except as follows. A pressure switch on the fuel pool pump discharge header annunciated low header pressure in the control room. The fuel pool high temperature alarm and low level alarms are annunciated in the control room. In the event the fuel pool pump breakers are opened, an alarm is annunciated in the control room. The component cooling water flow to the fuel s pool heat exchanger is initially adjusted to the required flow. Further i
adjustments of the component cooling water are not required. The component l '
cooling water discharge line has a flow indicator. High and low component l
' cooling water flow alarms are annunciated in the control room.
3-10 0076L/0011L 1
The clarity and purity of the water in the fuel The pool is maintained purification by the loop consists purification portion of the fuel pool system.
O of the fuel pool purification pump, ion exchanger, filter, strainers and [
surface skimmers.
Most of the purification flow is drawn through the surface '
A basket strainer is provided in the skimmers to remove surface debris.
purification line to the pump suction to remove any relatively largeThe fuel po particulate matter. filter, which removes particulate larger than Connections 5 micron for are provided size, and through an icn exchanger to remove ionic material. Fuel purification of the refueling water tank and refueling water cavity.
pool water chemistry is given in FSAR Table 9.1-2 The fuel pool piping is arranged so that the poc1 cannot be inadvertently drained to uncover the fuel in the event of a supply or discharge pipe J rupture.
All fuel pool piping is arranged to prevent gravity draining the fuel pool. To prevent siphoning of the fuel pool, the fuel pool discharge and purification suction lines have 1/2" and 1/4" holes respectively 1 foot below the normal water level.
y The only means of draining the pool below these siphon breaker holes is (
through an open line in the cooling loop while operating the pool cooling pumps. In such an event the fuel pool water level can be reduced by onlyThe 6 feet since the pump suction connection enters near the top of the pool. j remaining water in the Spent Fuel pool vill provide adequate shielding and !
heat removal capabilities at this point. The temperature and level alarms would warn the operator of such an event.
3.2.2 Decay Heat Analyses 3.2.2.1 Basis The St. Lucie Plant Unit 1 reactor is rated at 2700 megawatts thermal (MWt).
The core contains 217 fuel assemblies. Thus, the average operating power per fuel assembly, Po, is 12.44 MW. The fuel discharge can be m.sde in one of the following two modes:
- Normal refueling discharge
- Full core discharge Tables 3-9 through 3-11 give the parameters for bulk and local pool temperature analyses.
3.2.2.2 Model Description NUREG-0800 Branch Technical Position ASB 9-2, " Residual Decay Energy For Light Water Reactors For Long Term Cooling"(15) is utilized to compute the heat dissipation requirements in the pool.
l i
I v
3-11 0076L/0011L
With the long term uncertainty fcctor, K, os spacified in SRP 9.1.3 (15),
the operating power, Po, is taken equal to the rated power, even though the reactor may be operating at less than its rated power Theduring such ofand computations the
( exposure period for the batch of fuel assemblies.
results reported here are based on the discharge taking place when the inventory of fuel in the pool will be at its maximum resulting in an upper bound on the decay heat rate.
Having determined the heat dissipation rate, the next task is to evaluate the Table 3-9 identifies the time-dependent temperature of the pool water.This is a conservative representation of actual loading cases examined. BULKTEM future expected discharges such as those presented in Table 5-1.
treats the generalized pool cooling problem shown in Figure 3-5.
A number of simplifying assumptions are made which render the analysis conservative, including:
- The heat exchanger is assumed to have maximum fouling. Thus, the temperature effectiveness, P, for the heat exchanger utilfred in the analysis is the lowest postulated value calculated from heat exchanger
> technical data sheets.
- No credit is taken for the improvement in the film coefficients of the heat exchanger as the operating temperature rises due to sonotonic reduction in the water kinematic viscosity with temperature rise.
Thus, the film coefficient used in the computations are lower bounds. l
- No credit is taken for heat loss by evaporation of the pool water.
- No credit is taken for heat loss to pool walls and pool floor slab.
The basic energy conservation relationship for the pool heat exchanger system yields:
l Ct dt . gy . Q2 d7 where:
C t
- Thermal capacity of stored water in the pool t
= Temperature of pool water at time,7 01
= Heat generation rate due to stored fuel assemblies in the pool Q2
- Heat removed in the fuel pool heat exchanger This equation is solved as an initial value problem by noting that the cooler heat removal rate must equal the heat genn;2 tion rate from previously discharged assemblies. Hence, N eool P (T in - teool) = PCONS 3-12 0076L/0011L
v l
1 whsre:
Heat generation rate from previously stored assemblies PCONS:
W eool: Coolant thermal flow rate P: Temperature effectiveness of the fuel pool cooler Tin: Coincident pool water temperature (initial value before beginning of discharge) i Coolant inlet temperature tcool:
The above equation yields PCONS Tin " + tcool Wcool P The value of Tin computed from the above formula is the initial value of the pool water temperature (at the start of fuel discharge).
BULKTF11 automates the solution of the above equation using the theory presented in Reference 16. Tabulated results are presented in the next sub-sec' tion.
l 3.2.2.3 Bulk Pool Temperature Results Table 3-12 gives the total dimensionless power generation ratio of all fuel l assembly batches previously stored in the pool consisting of a total of 18 s batches. The first column in Table 3-12 gives the batch number, and the last column gives the dimensionless power, defined autthe heat generation rate of the batch divided by the nominal operating power of one fuel assembly. It is noted from Table 3-12 that the cumulative power is 0.14 times the operating l power of one fuel assembly. Tables 3-13/3-14 and 3-16/3-17 give the bulk !
temperature vs. time data.
The following key output data is gleaned from these tables:
Maximum pool bulk temperature:
Normal discharge: 133.30F Table 3-14 Full core discharge: 150.80F Table 3-17 Tables 3-15 and 3-18 give t'ine-to-boil data.
Time-to-boil (if coolant flow is lost upon completion of discharge and when the bulk pool temperature is nazimum):
Normal discharge condition: 13.43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br /> Table 3-15 Full core discharge condition: 5.04 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Table 3-18 O 0076L/0011L 3-13 w_____________________________. _
3.2'.2.4 Spent Fuel Pool Cooling System Summary fuel decay heat calculations were performed in accordance with the Q
Q The spentmethod provided in NRC Branch Technical Position ASB 9-2, Residual Decay .,
Energy for Light-Water Reactors for Long-Term Cooling (15).
The The existing spent fuel pool cooling system is considered to be adequate. J spent fuel pool is designed to withstand stresses associated with a steady-state water temperature of 2170 F. As shown in Table 3-17 the pool peak transient water temperature after full core discharge is less than
.151o F.
In the event of a complete loss of cooling capability, there is sufficient time to provide an alternate means for cooling.
The total increase in heat load rejected to the environment through the cooling systems due to the increased spent fuel storage over the current heat load rejected is l.7 x 106 Btu / hour. This represents an increase of The approximately 0.03 percent of the total heat rejected to the environment.
increase in heat rejected will have negligible impact on the environment.
The.increasd in heat load does not alter in any way the existing facility design bases. Thus, the heat load increase is acceptable. This decay heat analysis is also bounding for the temporary fuel storage configuration (see-Section 4.7.4) that will be utilized during rack installation.
3.2.2.4.1 Safety Evaluation The calculations for the amount of thermal energy that-may have to be removed O by the spent fuel pool cooling system are made in accordance'with Branch Technical Position ASB 9-2 (Reference 15). The resulting bulk spent fuel pool temperatures are acceptable.
3.2.3 Spent Fuel Pool Makeup p There are several sources of fresh water on the site that are available.to the ;
fuel handling building; namely, refueling water storage tank, city water storage tank via the fire main, city water storage tanks via the portable fire pump, and primary water tank. The concurrent loss of these sources and the j fuel pool cooling system is remote. Due to the fuel pool's boil-off period, there is sufficient time to obtain makeup. It should be noted that a seismic Category I backup salt water supply is available from the intake cooling water intertie. A standpipe on the fuel handling building is provided from grade to.
the operating deck elevation and hose connections are provided at both ends of the standpipe. Thus, via fire hose, the fuel pool makeup can be readily supplied by the intake cooling water pumps. The head provided by these pumps ,
is sufficient to provide the required fuel pool make up. The structural and leaktight integrity of the fuel pool will not be compromised by continuous fuel pool temperatures of up to 217o F. The results of the bulk decay heat analyses indicate that these temperatures are not exceeded. .The intake cooling water system connection via the hose connections can provide 150 gpm of makeup. See FSAR Subsection 9.1.3.4.
O 3-14 0076L/0011L
3.3 THEPNJ.-HYDRAULIC ANALYSES FOR THE SPENT FUEL POOL (LOCALIZZD)
The purpose of the thermal-hydraulic analyses is to determine the maximum fuel clad temperatures which may occur as a result of using the new high density O spent fuel racks in the St Lucie Unit 1 spent fuel pool.
3.3.1 Bases In order to determine an upper bound on the maximum fuel cladding temperature, a series of conservative assumptions are made. The most important assumptions are listed below:
- As stated above, the fuel pool will contain spent fuel with varying time-af ter-shutdown ( 7 g). Since the heat emission falls off rapidly with increasing 7 s, it is obviously conservative to assume that all fuel assemblies are fresh and they all have had the maximum postulated years of operating time in the reactor. The heat emission rate of each fuel assembly is assumed to be equal and maximum.
- As shown in Figure 2-1, the modules occupy an irregular floor space in the pool. For the hydrothermal analysis, a circle circumscribing the actual rack floor space is drawn (Figure 3-6). It is further assumed that the cylinder with this circle as its base is packed with fuel assemblies at the nominal layout pitch.
- The actual downcomer space around the rack module group varies, as shown in Figure 2-1. The nominal downcomer gap available in the pool is assumed to be the total gap available around the idealized cylindrical rack; thus, the maximum resistance to downward flow is incorporated into the analysis (Figure 3-7).
No downcomer flow is assumed to exist between the rack modules.
3.3.2 Model Description Using the bases described above, a conservative idealieed model for the rack assemblage is obtained. The water flow is axisymmetric about the vertical axis of the circular rack assemblage and, thus, the flow is two-dimensional (axisymmetric three-dimensional). Figure 3-7 shows a typical " flow chimney" rendering of the thermal hydraulics model. The governing equation to characterize the flow field in the pool is an integral equation that can be solved for the lower plenum velocity field (in the radial direction ) and axial velocity (in-cell velocity field), by using the method of collocation.
It should be added that the , hydrodynamic loss coefficients which enter into the formulation of the integral equation are also taken from well-recognized sources (17) and wherever discrepancies in reported values exist, the conservative values are consistently used. Reference 18 gives the details of mathematical analysis used in this solution process.
l l
O 3-15 0076L/0011L
After the axial velocity field is evaluated, the fuel assembly cladding temperature can be calculated. The knowledge of the overall flow field enables pinpointing of the storage location with the minimum axisi flow (i.e.,
j maximum water outlet temperatures). This is called the most "chtsked" l location. In order to find an upper bound on the temperature in a typical '
i cell, it i_s assumed that it is located at the most choked location. Knowing the global plenum velocity field, the revised axial flow through this choked cell can be calculated by solving the Bernoulli equation for the flow circuit through this cell. Thus, an absolute upper bound on the water exit !
temperature and maximum fuel cladding temperature is obtained. In view of the aforementioned assumptions, the temperatures calculated in this manner overestimate the temperature rise that will actually occur in the pool.
THERPOOL, based on the theory of Reference 18, automates this calculation.
Finally, the maximum specific power of a fuel array qA can be given by:
9A
=
q Fry where:
q = average fuel assembly specific power F3y = radial peaking factor The data on radial and axial peaking factors may be found in Table 3-10.
The maximum temperature rise of pool water in the most disadvantageous 1y placed fuel assembly is coLputed for all loading cases. Table 3-19, third l column, gives the outputs from THERPOOL in tabular form.
3.3.3 Cladding Temperature l
Having determined the maximum local water temperature in the pool, it is now l possible to determine the maximum fuel cladding temperature. A fuel rod can l produce FTot times the average heat emission rate over a small length, where l F7og is the total peaking factor. The axial heat dissipation in a rod is known to reach a maximum in the central region, and taper off at its two extremities. For added conservatism, it is assumed that the peak heat emission occurs at the top where the local water temperature also reaches its maximum. Futhermore, no credit is taken for axial conduction of heat along the rod. The highly conservative model thus constructed leads to simple algebraic equations which directly give the maximum local cladding temperature, te.
Table 3-19, fourth column, summarizes the key output data. It is found that the maximum value of the local water temperature is well below the nucleate boiling condition value. The incremental cladding temperature is too small to produce significant thermal stresses.
3.4 POTENTIAL FUEL AND RACK HANDLING ACCIDENTS The method for moving the racks into and out of the spent fuel pool is briefly discussed in Section 4.7.4.2. The methods utilized ensure that postulated l
l accidents do not result in a loss of cooling to either the spent fuel pool or the reactor, or result in a keff in the spent fuel pool exceedin; 0.95.
3-16 0076L/0011L
3.4.1 Rack Module Mishandling q The potential for mishandling of rack modules during the rerack operation has l At no time will the cask handling crane or the temporary Q been evaluated.constructim t,rane carry a rack module directly over a rack containing spent fuel. The procedures and administrative controls governing thethe Both rarack temporary operation will ensure the safe handling of rack modules.
construction crane and the cask handling crane meet the design and operational requirements of Section 5.1.1 of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants"(19).
In the unlikely event that a rack should strike the side of another rack module containing fuel assemblies, the consequences of this postulated accident would be bounded by the cask drop evaluations described in Section I
5.3.1.2. l 3.4.2 Temporary._ Construction Crane Drop During the rerack operation, a temporary construction crane will be installed !
in the Fuel Handling Building. This installation will be performed using lift rigs which meet the design and operational requirements of NUREG-0612The consequence
" Control of Heavy Loads at Nuclear Power Plants." l postulated accident during this installation are bounded by the cask drop evaluations described in Section 5.3.1.2. '
1 3.4.3 Loss of Pool Cooling (Storage Rack Drop) f During the re-racking operation, it will be necessary to raise and maneuver the old racks out of the spent fuel pool in order to install the new spent O fuel racks (See Section 4.7.4).
The handling of these heavy loads will be l C/ accomplished by the use of a temporary construction crane and the cask handling crane.
Both of these cranes meet the design and operational requirements of Section 5.1.1 of NUREG-0612. " Control of Heavy Loads at Nuclear Power Plants." 4 The consequences of dropping a rack in the Spent Fuel Pool were determined by reviewing the analysis in FSAR Subsection 9.1.4 for dropping of the spent fuel cask. The results of this cask drop analysis demonstrated that the pool This floor would remain elastic during impact and that cracks would not develop. -
cask weighs substantially more than a single rack assembly Therefore, theand rackhas a smaller drop scenario l cross sectio'al area for load distribution. ,
is bounded by the previous analysis for a cask drop scenario, and loss of !
spent fuel cooling from loss of pool water inventory will not occur as a result of a rack drop. .
3.5 TECHNICAL SPECIFICATION CHANGES Tnis proposed amendment permits replacement of the spent fuel pool racks to ensure that sufficient capacity exists for storage of spent fuel at St. Lucie Unit 1. The new racks increase the available storage to 1706 spent fuel assemblies and is expected to provide adequate storage space until the year 2009. .
O U
' 3-17 0076L/0011L
g The proposed Technical Specification changes are described below:
V 1. Specification 3/4.9.14 Bases is revised to reflect the assumptions used in calculations of doses based on the' Decay Times. ,
- 2. Specification 5.6.1.a.1 is revised to correspond to the Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors (NUREG-0212 Rev 2).
- 3. Specification 5.6.1.a.2 is revised to show the nominal l
center-to-center distance for the high capacity spent fuel storage racks.
- 4. Specification 5.6.1.a.3 is edited to discuss the boron concentration only.
- 5. Specification 5.6.1.a.4 is created to indicate the presence'of Boraflex in the cells.
- 6. Specification 5.6.1.b and accompanying Figure 5.6-1 are created to define the fuel enrichment /burnup limits for storage in each region of the high capacity spent fuel storage racks.
- 7. Specification 5.6.1c is editorially changed from "b" to "c".
_ 8. Specification 5.6.3 is changed to show the capacity of the high-capacity spent fuel storage racks.
3.6 REFERENCES
FOR SECTION 3 l 1. A. Ahlin, M. Edenius, H. Haggblom, "CASMO - A Fuel Assembly Burnup Program," AE-RF-76-4158, Studsvik report (proprietary).
- 2. A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion Code for INR Analysis," ANS Transactions, Vol. 26, p. 604,1977.
- 3. M. Edenius et al. , "CASMO Benchmark Report," Studsvik/RF-78-6293, Aktiebolaget Atomenergi, March 1978.
- 4. Green, Lucious, Petrie, Ford, White, Wright, "PSR-63/AMPX-1 (code package), AMPX Modular Code System for Generating Coupled Multigroup Neutron - Gamma Librarids from ENDF/B," ORNL-TM-3706, Oak Ridge National Laboratory, March 1976.
- 5. L. M. Petrie and N. F. Cross, " KENO-IV, An improved Monte Carlo Criticality Program," ORNL-4938, Oak Ridge National laboratory, November 1975.
- 6. R. M. Westfall et al., " SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation," NUREG/QL-0200, 1979.
O t I
3-18 0076L/0011L
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ __ _ _ _1
_ - - - - - - - - - - ~ _ _ _ _ _____ _ _ _ _ _
- 7. W. A. Wittkopf, "NULIF - Neutron Spectrum Generator, Few-Group constant Generator and Fuel Depletion Code " BAW-426. The Babcock & Wilcor Company, August 1976.
- 8. W. R. Cadwell, PDQ07 Reference Manual WAPD-TM-678, Bettis Atomic Power Laboratory, January 1967.
- 9. W. J. Eich, " Advanced Recycle Methodology Program, CEM-3," Electric Power Research Institute, 1976.
- 10. E. E. Pilat, " Methods f or the Analysis of Boiling Water Reactors (Lattice Physics)," YAEC-1232, Yankee Atomic Electric Co., December 1980.
- 11. E. Johansson, " Reactor Physics Calculations on Close-Packed Pressurized Water React.or Lattices," Nuclear Technology, Vol. 68, pp. 263-268, February 1985.
- 12. H. Richings, Some Notes on PWR (W) Power Distribution Probabilities for LOCA Probabilistic Analyses, NRC Memorandum to P. S. Check, dated July 5, 1977.
- 13. M. G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.
- 14. J. M. Cano et al., "Supercriticality Through Optimum Moderation in Nuclear Fuel Storage," Nuclear Technology, Vol. 48, pp. 251-260, May 1980.
O 15. NUREG-0800, U.S. Nuclear Regulatory Commission, Standard Review Plan, Branch Technical Position ASB 9-2, Rev. 2, July 1981.
- 16. Singh, K. P. , Journal of Heat Transf er, Transactions of the ASME, August 1981, Vol.1-3, "Some Fundamental Relationships for Tubular Heat Exenanger Thermal Performance."
- 17. General Electric Corporation, R&D Data Books, " Heat Transfer and Fluid Flow," 1974 and updates. ,
- 18. Singh, K. P. et al., " Method for Computing the Maximum Water Temperature in a Fuel Fool Containing Spent Nuclear Tuel," Heat Transf er Engineering, Vol. 7, No.1-2, pp. 72-82 (1986).
- 19. Nuclear Regulatory Commission, " Control of Heavy Loads at Nuclear Power Plants, NUREG-0612, July 1980.
4 0
3-19 0076L/0011L
.m.
-TABLE 3-1 I
SUMMARY
OF CRITICALITY SAFETY ANALYSES l
\j .
i Region 1 Region 2 1
J
~
36.5 Mwd /kgU Minimum acceptable burnup 0 0 4.51 initial enrichment 40C 40C j Temperature assumed for analysis 0.9313 0.9114 l Reference k m (nominal) 0.0013 0.0013 Calculational bias 4 1 !
Uncertainties j
1 1
+0.0018 +0.0018 Bias 70.0036 J
B-10 concentration 70.0021
~
Boraflex thickness +0.0044 70.0111
~
+0.0017 I0.0011 Borafier width 70.0016 inner box dimension +
~0.0011 W/A
- Water gap thickness 70.0043 t +0.0002 SS thickness 70.0010 70.0027 70.0027 Fuel enrichment ~
~
Fuel density +0.0005 +0.0005 70.0024 l Fuel element pitch 70.0024 l
Statistical combination (l) +0.0080 +0.0125 Eccentric assembly position
~
+0.0003 negative Allowance for N/A +0. 0183 burnup uncertainty 1 Total 0.9329 1 0.0080 0.9310 1 0.0125 Maximum reactivity 0.9409 0.9435 (with 1720 ppm soluble boron) (0.767 ) (0.760) 1 (1) Square root of sum of squares.
N/A - Not Applicable 3-20 0076L/0011L
O TABLE 3-2 MINIMUM BURNUP VALUES Initial Minimum Enrichment, % Burnup, Mwd /kgU 1.63 0 1.75 2.30 2 00 6.00 2.25 9.70 2.50 12.90 .
2 75 16.10 3.00 19.15 3.25 22.20 3.50 25.15 3.75 28.10 4.00 30.90 4.25 33.70 f- 4.50 36.50 0
S 9
3-21 0076L/0011L
O TABLE 3-3 REACTIVITY EFFECTS OF ABNORMAL AND ACCIDENT CONDITIONS Accident / Abnormal Conditions Reactivity Effect i Temperature increase Negative in both regions Void (boiling) _ Negative in both regions Assembly dropped on top of rack Negligible i
Lateral rack module movement Negligible Misplacement of a fuel assembly positive O
l 9
l l O 3-22 0076L/0011L !
l l
O TABLE 3-4 1
TUEL BURNUP VALUES FOR REQUIRED REACTIVITIES (ke )
k'ITH TUEL OF VARIOUS INITIAL ENRICtDIENTS (Reference k . = 0.9297)
Calculated ,
Initial' Uncertainty (1) Design Burnup limit.
in Burnup, 4 k Limit k . Mwd /kgU Enrichment 0.9297 0 1.6 0 5,99 2.0 0.0030 0.9267 0.9233 12,88 4 25 0 0064 3.0 0.0096 0.9201 19 13 l 3.5 0.0126 0.9171 25.15 - )
4.0 0.0154 0.9143 '30.86 4.5 0.0183 0.9115 36.50 0 \
1 d
(1) See Subsection 3.1 3.3 2 4
l i
I i
O 3-23 0076L/0011L
m TABLE 3-5 COMPARISDN OF COLD, CLEAN REACTIVITIES CALCULATED AT 36.5 Mwd /kgU BURNUP AND 4.5% ENRICHMENT
- k. Xe-free, 40C Infinite Array of . Assemblies in Calculational Method Region 2 Cell Fuel Assemblies (1) in Reactor Spacing l
CASMO-2E 1.1212 0.9114 1.1306 0.8972 DIFFUSION / BLACKNESS THEORY EPRI-CELL 1.1281(2) -
O (1) Cold, clean condition in contrast to hot operating conditions of Figure 3-4.
(2) EPRI-CELL k . at maximum value during long-term (30-year) storage.
3-24 0076L/0011L
- g O TABLE 3-6 ESTIMATED UNCERTAINTIES IN REACTIVITY DUE TO l FUEL DEPLETION EFFECTS 4 l
i Design 0.0005 Design' Reactivity Initial Burnup Times Burnup, A k k. Loss, Ak (1) l Enrichment Mwd /kgU l
0.9297 0 0 0 1.6 0.9267 0.0579 l 5.99 0.0030 2.0 0.0064 0.9233 0.1284 2.5 12.88 0.1828.
l 0.0096 0.9201 30 19.13 0.2262 l
25.15 0.0126 0.9171-
! 3.5 0.9143 0.2620 4.0 30.86 0 0154 0.0183 0.9115 0.2924 4.5 36.50 (1) Total reactivity decrease, calculated for the cold, Xe-free condition in the fuel storage rack, from the beginning-of-life to the design burnup.
l l
O 3-25 0076L/0011L
O TABLE 3-7 l
LONG-TIRM CHANGES IN REACTIVITY IN STORAGE RACK I
Storage Ak from Shutdown (Xenon-free) at 4.5% E l Time, and 36.5 Mwd /kgU years 0.5 -0.0047 1.0 0.0088 1 10.0 0.0470 20'O -0.0673 30.0 -0.0788 C
i l
f d
3-26 0076L/0011L
1 T TABLE 3-8 I
DESIGN BASIS (LIMITING)
FUEL ASSDiBLY SPECIFICATIONS (CE 14 x 14) tuel nod Data Cladding outside diameter, in. 0.440 Cladding thickness, in. 0 028 Cladding auterial Zircaloy-4 .
l 0.377 j Pellet diameter, in.
UO2 stack density, s/cm3 10.281 1 0.031 Enrichment, wt1 U-235 4.5 1 0.05 Fuel Assembly Data Maximum number of fuel rods 176 (14 x 14 array)
Fuel rod pitch, in. 0.577 1 0.0023 Control rod guide tube Number 5
- Outside diameter, in. 1.115 Inside diameter, in. 1.035 l Zircaloy-4 Material U-235 Loading grams / axial em of assembly 51.7 1 0.7
. l
, 1 O
3-27 0076L/00111 l
O TABLE 3-9 THERMAL / HYDRAULIC CASES TREATED *
- 1. Normal Batch Discharge:
Irradiation time: 54 months (1.42 x 108 secs)
Addition of the most recent batch : 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown Baten size: 80 assemblies 2 Full Core Discharge Irradiation time: 73 assemblies 90 days 72 assemblies 21 months 72 assemblies 39 months
- Fuel transfer begins 7 days after shutdown.
?
- The pool has total storage capacity of 1706 storage cells. It is conservatively assu,ed m that 18 batches of 80 assemblies have been previously discharged at 18 month interva2.s. Each assembly in these previous discharges has had 54 months of exposure at full power (12.44
!!Wt ) .
9 3-28 0076L/0011L l
TABLE 3-10 PEAKING TACTOR DATA Haximum Radial Maximum Axial Fuel Peaking Factor Peaking Factor St. Lucie Unit 1 1.67 1.32 CE 14 x 14 and Exxon 14 x 14 St Lucie Unit 2, 1.75 1.35 CE 16 x 16 O
8 O .
3-29 0076L/0011L
TABLE 5-11 ESSENTIAL HEAT TRANSTER DATA FOR THE FUEL POOL NEAT EXCHANGER Number of heat exchangers; one coolant flow rate: 3560 spm Temperature effectiveness: 0.36 (two pumps)*
0 263 (one pump)
Heat transfer surface area: 4380 sq. ft.
Overall heat transfer coefficient (fouled)
(two pumps): 260 5tu/sq.ft.-hr-OF O
- Temperature efficiency of the heat exchanger is calculated in the f ollowing manner, using the information provided in l the FSAR:
l Cooling water outlet - inlet P =
Pool water inlet - cooling water inlet 118-100 i e
150-100
= . 36-i 3-30 0076L/0011L 1
l l
- w__ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
TABLE 3-12 P0h'ER GENERATION RATIO PREVIOUSLY DISCHARGED BATCHES Reactor Exposure Non Dimensional Batch Batch Time After Shut Power Gen. Ratio Down in Days Time in Days -
No. Size 1643 5 .00487 1 80 9719.9 1643 5 .00505 80 9179.9 2
1643.5 .00523 3 80 8639.9 .00542 8099.9 1643 5 l 4 80 .00562 80 7559.9 1643 5 5 .00582 7019.9 1643.5 6 80 .00603 80 6479.9 1643.5 7 .00624 80 5939.9 1643 5 8 .00647 80 5399.9 1643.5 9
1643.5 .00670 10 80 4859.9 4319.9 1643.5 .00694 11 80 .00720 80 3779.9 1643.5 12 1643.5 .00746 13 80 3239.9 1643.5 .00776
/ 14 80 2699.9 2159.9 1643.5 .00815 15 80 .00888 80 1619.9 1643.5 16 .01097 17 80 1079.9 1643.5 1643.5 .01893 ,
18 80 540.0 CUMULATIVE DIMENS10h1ESS P0h'ER = 1.3374E - 01 l
}
i O 1 3-31 0076L/0011L )
1 A__---__-__--__-__ _ _ _ _
p;-- -
TABLE 3-13
( ')
BUlX POOL TEMPERATURE VS. TIME DURING NORMAL REFUELING DISCHARGE Time Bulk Pool- Heat Generation (Hrs.) Temp. (oF) Rate (Stu/hr) 150.00* 106.0 .5689E + 07 151.00 108.8 .1643E + 08 i
- This table contains only two lines of output data. This is due to the fact that the discharge is assumed to take place O instantaneously, simulated by one hour in this computer run. I i
I 3-32 0075L/0011L
3 1
O k- / TABLE 3-14 POOL BULK TEMPERATURE VS. TIME SUBSEQUENT TO COMPLETION OF NORMAL REFUELING DISCHARGE Time Bulk Pool Heat Generation (Hrs.) Temp. (DF) Rate (Btu /hr) l 151. PD 108.8 .1642E + 08 161.00 130 0 .1613E + 08 171.00 133.2 .1588E + 08 181.00 133 3 .1565E + 08 191.00 133 0 .1544E + 08 201.00 132 6 .1525E + 08 211.00 132.2 .1507E + 08 221.00 131.8 .1490E + 08 231.00 131.5 .1475E + 08 1 241.00 131.1 .1461E + 08 251.00 130.8 .1447E + 08
/' ,! 261.00 130 6 .1435E + 08 l 271.00 130.3 .1423E + 08 281.00 130.1 .1411E + 08 291.00 129.8 .1401E + 08 301.00 129 6 .1390E + 08 311 00 129.4 .1380E + 08 321.00 129 2 .1371E + 08 331.00 129.0 .1362E + 08 341.00 128.8 .1353E + 08 351.00 128.6 .1344E + 08 l 361.00 128 4 .1336E + 08 l
371.00 128.3 .1328E + 08 !
l l l 381.00 128.1 .1320E + 08 391.00 127.9 .1313E + 08 l
i 1
0 3-33 0076L/00111 4
O
't
- TABLE 3-15 LOSS OF COOLING ATTER COMILETION OF' NORMAL RETUELING DISCHARGE Rate of- Rate of' Time to Boil Evaporation Level Change Case (hrs) (1bs/hr) (inch /hr)
When heat 16.79 16933.0 2 67' generation is maximum When the bulk 13.43 16294.0 2.57 pool temperature i
O is maximum i
4 O
3-34 0076L/0011L
.___m__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , _
TABLE 3-16 BULK POOL TEMPERATURE VS TIME DURING FULL CORE DISCHARGE l
Time Bulk Poc1 Heat Generation (Hrs.) Temp. (oF) Rate (Btu /hr) 168.00* 113.6 .8690E + 07 169.00 117.8 .3371E + 08
- This table contains only two lines of output data. This is due to the fact that the discharge is assumed to take place instantaneously, simulated by one hour in this conputer run.
O 3-33 0076L/0011L
O TABLE 3-17 POOL BULK TEMPERATURE VS TIME SUBSEQUENT TO COMPLETION OF FULL CORE DISCHARGE Time Bulk Pool Heat Generation (Hrs.) Temp. (oF) Rate (Btu /hr)
I 169.00 117.8 .337'0E + 08 179.00 148.8 .3307E + 08 189.00 150.8 .3249E + 08 199.00 150.2 .3197E.+ 08 209.00 149.4 .3149E + 08 219.00 148.7 .3104E + 08 229.00 148.1 .3062E + 08 239.00 147.4 .3024E + 08 249.00 146.9 .2987E + 08 259.00 146.3 .2953E + 08 269.00 145.8 .2921E + 08 279.00 145.3 .2991E + 08 O 289.00 299.00 144.8 144.4
.2862E + 08
.2834E + 08 309.00 144.0 .2807E + 08 319.00 143.6 .2782E + 08 329.00 143.2 .2758E + 08 339.00 142.8 .2734E + 08 349.00 142.5 .2712E + 08 359.00 142.1 .2690E + 08 369.00 141.8 .2668E + 08 379.00 141.5 .2648E + 08 389.00 141.1 .2628E + 08 399.00 140.8 .2608e + 08 409.00 140.5 .2589E + 08 O
3-36 0076L/0011L
(
O l
l TABLE 3-18 1455 0F COOLING AFTER COMPLETION OF FULL CORE DISCHARGE l Rate of Este of Evaporation Level Otange Time to Boil (inch /hr)
Case (hrs) (1bs/hr) 7.47 34742 2 5.47 When heat generation is maximum When the bulk 5.04 33660.0 53 pool temperature is maximum l
J O
3-37 0076L/0011L l-
i O !
TABLE 3-19 LOCAL AND CLADDING TDiPERATURE DATA i
Maximum Local Maximum Water. Cladding Case Instant Temp. OF Temp. OF f 1
Normal When the pool heat 155.9 198.8 discharge generation rate is at its peak value Normal When the pool bulk 179.2 219.4 discharge temperature is at its peak value
,. Full core When the heat 162.8 209.4 discharge generation rate O in the pool is at the peak value Full core When the pool bulk 188.0 222.8 discharge temperature is at its peak value O
3-38 0076L/0011L
l l
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l 35 l 30 _
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ACCEPTABLE _
l25 BURNUP DOMAIN s g _
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8 - UNACCEPTABLE BURNUP DOMAIN -
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4.5 1,5 2.0 2.5 3.0 3.5 4.0 INITIAL ENRICHMENT, WT % U-235 F LORIDA POWER & LIGHT COMP ANY ST. LUCIE PLANT UNIT 1 ACCEPTABLE BURNUP DOMAIN IN O REGION 2 OF THE ST. LUCtE PLANT SPENT FUEL STORAGE RACKS FIGURE 31
i 8.65" 1 0.32" m. ;
- ' BOX l.D. l i
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0.0238" i 0.0038" g B-lO/cm 2 l BO i ER I
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O m LATTICE SPACING m' 8.86" 1 0.040" I
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INITIAL ENRICHMENT FIGURE 3 4
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g FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 IDEALIZATION OF RACK ASSEMBLY FIGURE 3 6
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ST. LUCIE PLANT UNIT 1 THERMAL CHIMNEY FLOW MODE L FIGURE 3 7 i
4.0 KECHANICAL, MATERI AL, AND STRUCTURAL CONSIDERATIONS
4.1 DESCRIPTION
OF STRUCTURE 4.1.1 Description of the Fuel Handling Building '
The Fuel Handling Building (FHB) consists of cast-in-place reinforced concrete interior and exterior walls. It is completely isolated from all other structures.
The floors and roof are of beam and girder construction supported by columns. A complete description of the FHB is provided in Section 3.8.1.1.2 of the St Lucie Unit No. 1 updated FSAR. The FHB general arrangement is shown on FSAR Figures 1,2-18 and 1.2-19.
The FHB has been designed as a seismic Class I structure in accordance with the criteria outlined in Sections 3.8.1.1.2 and 3.8.1.4 through 3.8.1.7 of the updated FSAR. The building exterior walls, floors and interior partitions are designed to provide plant personnel with the necessary biological radiation shielding and protect the equipment inside from the effects of adverse environmental conditions including tornado and hurricane winds, temperature, external missiles and flooding.
The spent fuel pool is a cast-in place steel lined reinforced concrete tank structure that provides space for storage of spent fuel assemblies, control ,
element assemblies, new fuel during initial core loading and a spent fuel shipping cask. The fuel pool portion of the FHB including the walls and roof )
directly above the pool is designed to withstand, without penetration, the j impact of high velocity external missiles that might occur during the passage '
of a tornado. The design missiles are further discussed in Section 3.5 of the St Lucie Unit No. 1 updated FSAR.
A The spent fuel handling system includes interlocks, travel limits and other f protective devices to minimize the probability of either mishandling or of equipment malfunction that could result in inadvertent damage to a fuel assembly and potential fission product release. The interlocks prevent movement into the walls while limit switches prevent the spent fuel handling machine from raising the fuel above a height where less than nine feet separates the surface of the water from the top of the active fuel length.
A leak detection system is provided in the spent fuel pool to monitor 100 percent of the pool liner plate weld seams. This system consists of a network of stainless steel angles attached to the outside of the pool liner walls and the underside of the pool liner floor by means of welds and sealed with epoxy material. In the event that one of the weld seams develops a leak, the liquid enters the monitor channel system and flows to one of 19 collection points at the base of the pool, from which the leak can be traced back to a specific pool area.
4.1.2 Description of Spent Fuel Racks The function of the spent fuel storage racks is to provide fer storage of spent fuel assemblies in a flooded pool, while maintaining a coolable geometry, preventing criticality, and protecting the fuel assemblies from l excessive mechanical or thermal loadings.
J O :
i 4-1 0077L/0011L 1
A list of design criteria is given below:
- 1. The racks are designed in accordance with the NRC, "07 Position for Review and Acceptance of Spent Fuel Storage and Handling O Applications," dated April 14, 1978 (as amended'by the NRC letter dated January 18, 1979) and SRP Section 3.8.4 [1].
e
- 2. The racks are designed to meet the nuclear requirements'of ANSI N210-1976. The effective multiplication factor, keff, in the spent fuel pool is less than or equal to 0.95, including all i, uncertainties and under all credible conditions.
- 3. The racks are designed to allow coolant flow such that boiling in the water channels between the fuel assemblies in the rack does not occur. Maximum fuel cladding temperatures are calculated for various pool cooling conditions as described in Section 3.3.
- 4. The racks are designed to seismic Category I requirements, and are classified as ANS Safety Class 3 and ASME Code Class 3 Component Support Structures. -The structural evaluation and seismic analyses are performed using the specified loads and load combinations in Section 4.4.
~
- 5. The racks are designed to withstand loads without violating the criticality acceptance criteria which may result from fuel handling accidents and-from the maximus uplift force of the spent fuel handling nachine.
~
- 6. Each storage position in the racks is designed to. support and guide the. fuel assembly in a manner that will minimize the possibility of application of excessive lateral, axial and bending loads to fuel assemblies during fuel assembly handling
- and storage.
- 7. The racks are designed to preclude the insertion of a fuel assembly in other than design locations within the rack array.
- 8. The materials used in construction of the racks are compatible '
with the storage pool environment and will not contaminate the fuel assemblies.
4.1.2.1 Design of Spent Fuel Racks 4.1.2.1.1 Region 1 The rack module is fabricated from ASME SA-240-304L austenitic stainless steel '
sheet and plate material, and SA-351-CF3 casting material and SA-564-630 ,
precipitation hardened stainless steel (to 11000F) for supports only. The weld filler material utilized in body welds is ASME SFA-5.9, Classification ER 30BL. Boraflex serves as the neutron absorber material. Additional information on Boraflex say be found in Section 3.1.3. The Boraflex l experience list is given in Table 4-1.
O l 4-2 0077L/0011L
A typical module contains storage cells which have an 8.65-inch nominal square cross-sectional opening. This dimension ensures that fuel assemblies with l
maximum expected axial bow can be inserted and removed from the storage cells l without any damage to the fuel assemblies or the rack modules.
f Figure 4-7 shows a horizontal cross-section of a 3 x 3 array. The cells provide a smooth and continuous surface for lateral contact with the fuel _
assembly. The anatomy of the rack modules is best explained by describing the cocponents of the design, namely:
- Internal Square Tube Neutron Absorber material (Boraflex)
- Poison sheathing Gap element
- Baseplate
- Support assembly
- Top Lead-In 4.1.2.1.1.1 Internal Square Tube -
This element provides the lateral bearing surface to the fuel assembly. It is fabricated by joining two formed channels (Figure 4-1) using a controlled seam welding operation. This element is an 8.65-inch square (nominal)
- cross-section by 169 inches long.
R 4.1.2.1.1.2 Neutron Absorber Material (Boraflex)
Boraflex is placed on all four sides of a square tube over a length of 143
(hinimum), which provides the requisite B-10 screen for all stored assemblies including a four-inch shrinkage allowance.
4.1.2.1.1.3 Absorber Sheathing The absorber sheathing (cover plate), shown in Figure 4-2, serves to position -
and retain the absorber material in its designated space. This is- /'
- accomplished by spot welding the cover sheet to the square tube along the
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former's edges at numerous (at least 20) locations. This manner of attachment
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ensures that the absorber material will not sag or laterally displace during fabrication processes and under any subsequent loading condition.-
4.1.2.1.1.4 Cap Element Gap elements, illustrated in, Figure 4-3, position two inner boxes at a predetermined distance to maintain the minimum flux trap gap required between two boxes. The gap element is welded to the inner box by fillet welds. An '
array of composite box assemblies welded as indicated in Figure 4-7 forms the honeycomb gridwork of cells which harnesses the structural strength of all sheet and plate type members in an efficient sanner. The array of composite boxes has overall bending, torsional, and axial rigidities which are an order of magnitude greater than configurations utilizing grid bar type of construction.
O 4-3 0077L/00111
j 4.1.2.1.1.5 Baseplate !
,q Tne baseplate is a 3/4-inch thick plate type member which has 6-inch diameter ;
holes concentrically located with respect to the internal square tube, except Q at support leg locations, where the hole size is 5 inches in diameter.
Secondary flow paths These are
)
l holes provide the primary path for coolant flow. l available between adjacent cells via the lateral flow holes (1 inch in 1 diameter) near the root of the honeycomb (Figure 4-4) which preclude flow blockages. The honeycomb is welded to the baseplate with 3/32-inch fillet welds.
4.1.2.1.1.6 Support Assembly Each module has at least four support legs. All supports are adjustable in length to enable leveling of the rack. The variable height support assembly consists of a flat-footed spindle which rides into an internally-threaded cylindrical member. The cylindrical member is attached to the underside of the baseplate through fillet and partial penetration welds. The base of the flat-footed spindle sits on the pool floor. Leveling of the rack modules is accomplished by turning the square sprocket in the spindle using a long arm (approximately 46 feet long) square head wrench. Figure 4-6 shows a vertical cross-section of the adjustable support assembly.
- The supports elevate the module baseplate approximately 8-7/8 inches above the pool floor, thus creating the water plenum for coolant flow. The lateral holes in the cylindrical member provide the coolant entry path leading into the bottom of the storage locations.
4.1.2.1.1.7 Top Lead-in Lead-ins are provided on each cell to facilitate fuel assembly insertion.
Cont guous walls of adjacent cells are structurally connected at the lead-ins j
l with a suitable vent opening. These lead-in joints aid in reducing the l lateral deflection of the inner square tube due to the impact of fuel assemblies during the ground motion (postulated seismic motion specified in the FSAR). Tnis type of construction leads to natural venting locations for the inter-cell space where the neutron absorber material is located.
4.1.2.1.2 Region 2 Design l The rack modules in Region 2 are fabricated from the same material as that used for Region 1 modules, i.e., ASME SA-240-304L austenitic stainless steel.
l As shown in Figure 4-5 a typical Region 2 module storage cell also has an 8.65-inch nominal square cross-sectional opening . Figure 4-8 shows a horizontal cross-section of a 3 x 3 array. The rack construction varies fron that for Region 1 inasmuch as the stainless steel cover plates, gap elements and top lead-ins are eliminated. Hence, the basic components of this design are as follows:
Inner tube Neutron absorber material Side strips
- Baseplate
- Support assembly 4-4 0077L/0011L
In this construction, two channel elements form the es11 of en 8.65-inch ;
nominal square cross-sectional opening. The poison material is placed between [
two boxes as shown in Figure 4-8. Stainless steel side strips are fuserted on both sides of the poison asterial to firmly locate it in the latersi direction. The bottom strip positions the poison material in the vertical )
direction to envelope the entire active fuel length of a fuel assembly (Figure
/
4-5). Two adjacent boxes and the side strip between boxes are welded together as shown in Figure 4-8, to form the honeycomb rack module.
The baseplate and support assemblies are incorporated in exactly the same manner as described for Region 1 in the preceding section. i l
I 4.1.2.2 Fuel Handling The design of the spent fuel racks will not affect the conclusions of the fuel handling accidents presented in the FSAR (Section 15.4.3) and summarized by the NRC in the Safety Evaluation Report. That is, the radiological doses for tne postulated fuel cask and fuel assembly drop accidents are well within the 10 CFR 100 criteria.
4.2 APPLICABLE CODES, STANDARDS, AND SPECIFICATIONS The design and fabrication of the spent fuel racks and the analysis of the
- spent fuel pool have been performed in accordance with the applicable portions of the following NRC Regulatory Guides, Standard Review Plan Sections, and published standards:
4.2.1 h7C Documents O April 14, 1978 NRC OT Position for Review and Acceptance of Q a.
Spent Fuel Storage and Handling Applications, as amended by the NRC letter dated January 18, 1979.
- b. St Lucie lant Unit 1 Updated Final Safety Analysis Report, Docket No. 50-335.
1
- c. NRC Regulatory Guides 1.13. Rev 2 Spent Fuel Storage Facility Design Basis Dec. 1981 (Draft) 1.25 Assumptions Used for Evaluating the Potential March 1972 Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors O
4-5 0077L/00111
1.26, Rev 3 Quality Group Classifications and Standards Feb. 1976 for Water, Steam and Radioactive Waste Containing Components of Nuclear Power Plants O) 1.29, Rev 3 Seismic Design Classification r
Sept. 1978 1.31 Rev 3 Proposed Control of Ferrite Component in Stainless Steel Weld Material 1.71, Rev 0 Welder Qualification for Areas of Limited Accessibility 1.85, Rev 22 Material Code Case Acceptability ASME Section III Division 1 l 1.92, Rev 1 Combining Modal Responses and Spatial Components in Seismic Response Analysis 1.124, Rev 1 Service Limits and Load, Combinations for Class Jan. 1978 1 Linear-Type component Supports 3.41, Rev 1 Validation of Calculational Methods for Nuclear Criticality Safety.
- d. NRC Standard Review Plan - NUREG-0800 e Rev 1. July 1981 -Section 3.7, Seismic Design i
Rev 1, July 1981 Section 3.8.4, other Seismic Category I Structures, Appendix D Rev 3. July 1981 Section 9.1.2, Spent Fuel Storage Rev 1. July 1981 Section 9.1.3, Spent Fuel Fool Cooling System Rev 2, July 1981 NRC Branch Technical Position ASB 9-2, Residual Decay Energy for Light Water Reactors for Long Term Cooling
- e. General Design Criteria for Nuclear Power Plants, Code of Federal Regulations, Title 10, Part 50, Appendix A (GDC Nos.1, 2, 61, 62 and 63)
- \
- f. NUREG-0612 Control of Heavy loads at Nuclear Power Plants.
I i
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4.2.2 Industry Codes and Standards A
U ANSI N14.6-1978 American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More for Nuclear Materials ANSI N16.1-75 Nuclear Criticality Safety in Operations with Fissionable Materials outside Reactors ANSI N16.9-75 Validation of Calculation Methods for Nuclear Criticality Safety ANSI N18.2-1973 Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants ANSI N45.2.2 Packaging, Shipping, Receiving, Storage and Handling l
of Items for Nuclear Power Plants ANSI N45.2.1 Cleaning of Fluid Systems and Aasociated Components during Construction Phase of Nuclear Power Plants ANSI N45.2.11 1974 Quality Assurance Requirements for the Design
- of Nuclear Power Plants ANSI ANS-57.2-1983 Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants ANSI N210-76 Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations ASME Section III Nuclear Power Plant Components, Subsection NT (1983 Edition i up to and in-l cluding Summer 1984 Addenda ACI-ASME Code for Concrete Reactor Vessels and Section III, Containments Division 2 (1977 Edition)
ACI 318-63 Building Code Requirements for Reinforced-Concrete
- l AISC 1980 Specification for the Design, Fabrication and i Erection of Structural Steel for Buildings, Eighth Edition AWS D1.1 Structural Welding Code ASNT-TC-1A American Society for Nondestructive Testing June 1980 (Recommended Practice for Personnel Qualification) o ,
4-7 0077L/0011L
ASME 11 Part A Material Specifications Part A Ferrous, Part C
& C- Welding Rods, Electrodes and Filler Metals (1983 Edition
()'
up to and including Summer 1984 Addenda)
ASME IX Welding 6 Brazing Qualifications (1983 Edition up to and in-cluding Summer 1984 Addenda)
ASHE Boiler and Non-destructive Examination Pressure Vessel,Section V, (1983 Edition up to and including Summer 1984 Addenda) 4.3 SEISMIC AND IMPACT LOADS
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The objective of the seismic analysis of the spent fuel racks is to determine the structural responses resulting from the simultaneous application of three orthogonal seismic excitations. The method of analysis employed is the time
, history method.
Seismic floor response. spectra for the spent fuel pool floor have been developed using the methods described in Subsections 3.7.1 and 3.7.2 of the St l Lucie Unit No 1 Updated FSAR. The parameters of the original lumped mass model of the Fuel Handling Building were adjusted to reflect the increased mass corresponding to the new high density spent fuel storage racks. The resulting floor response spectra are shown in Figure 4-9. These spectra were then used to generate statistically independent time history excitations, one for each of the three orthogonal directions. Since the spent fuel racks have I no connection with the pool walls or with each other, the pool floor time histories are used as input to the dynamic analysis of the racks, as described in Subsection 4.5.2.2.1. Fluid coupling is also considered as described therein.
Deflection or movements of racks under earthquake loading is limited by design such that the nuclear parameters outlined in Section 3.1 are not exceeded.
Impact loads have been considered as discussed in Subsection 4.6.4.
The interaction between the fuel assemblies and the rack has been considered, Particularly gap effects. The resulting impact loads are of small magnitudes so there is no structural damage to the fuel assemblies.
The spent fuel pool structure has been reanalyzed for the increased dead, thermal and seismic loading resulting from the storage of additional fuel assemblies in the pool, as described in Subsection 4.5.1.
O 4-8 0077L/0011L
m-4.4 LOADS AND LOAD COMBINATIONS 4.4.1 Spent Fuel Pool
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4.4.1.1 Loads The following design loads were considered in the spent fuel pool analysis:
a) Structural Dead Load (D)
Dead load consists of the dead weight of the spent fuel racks, the pool water and the concrete structure, superstructure, walls and miscellaneous building items within the Fuel Handling Building.
b) Live Load (L)
Live loads are random temporary load conditions for maintenance which include the spent fuel cask dead weight.
l c) Seismic Loads (SSE and OBE)
Seismic loads include the loads induced by Safe Shutdown Earthquake (SSE) and Operating Basis Earthquake (OBE). The hydrodynamic load during the earthquake events was also considered.
I d) Normal Operating Thermal Loads (T)
The load induced by normal thermal gradients existing between the building interior and the ambient external environment was
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(,, considered. Tne conditions are:
Su=mer
- Interior water temperature 1500F
- Exterior air temperature 930F
- Soil temperature 700F Winter
- Interior water temperature 1500F
- Exterior Air temperature 320F
- Soil temperature 700F For all cases, the "as constructed" concrete temperature was assu ed to be 700F. A linear gradient through the wall and sat was assumed.
i O
4-9 0077L/0011L l
1 e) Accident (Loss of Tuel Pool Cooling) Thermal Load (TA ) I J
The thermal accident temperature for the spent fuel pool water j is 2170F throughout the pool.,,At this temperature, the exterior air..
l temperature at 40oF was assumed for the critical thermal gradient I l
through the wall. '700F soil temperature was used. The thermal gradient.was assumed to be linear.
f) Fuel-. Cask Drop Load (M)
A 25 ton cask drop from the, maximum height of 58 feet above'the pool floor (Elevation 79.50') was considered. (The cask bottoa sust attain Elevation 77.00' for entry into the building.) i 4.4.1.2 Load Combinations In the spent fuel pool analysis, the following load combinations, from the St. 1
/
Lucie No. 1 Updated FSAR, Section 3.8.1.5, were considered:'
a) Normal Operation 1.5 (D + T) + 1.8 L ,
3
~
b) OBE Condition 1.25 (D + T + OBE + 0.2 L) ;
c) SSE Condition 1.05 (D + 7 + 0.2 L) + 1.0 SSE d) Accident and Cask Drop 1.05 (D + TA + 0.2 L) 1.05 (D + T + 0.2 L) + 1.0 M For the evaluation of the liner and liner anchors, the above load combinations are applicable except that load factors for all cases may.be taken equal to 1.0 (in accordance with Table CC-3230-1 of ACl-ASME Section III, Division 2) in conjunction with the structural acceptance criteria of this SAR subsection 4.6.1.1.b. l l
Linear analyses without iterations were performed initially to determine the l critical load combinations. As a result, the following loading cases were selected for the non-linear concrete cracking analysis:
- 1) 1.5 D + 1.8 L ;
ii) 1.05 (D + T winter + 0.2 L) + 1.0 SSE iii) 1.05 (D + T summer + O.2 L) + 1.0 SSE I iv) 1.05 (D + 0.2 L) + 1.0 SSE I v) 1.05 (D + TA + 0.2 L) i vi) 1.05 (D + T winter + 0.2 L) + 1.0 M i vii) 1.05 (D + 0.2 L) + 1.0 M l
4-10 0077L/0011L
4.4.2 Spent Fuel Racks 4.4.2.2 Loads The following loads were considered in the rack design:
Dead Load (D) - Dead weight-induced stresses (including fuel assembly weight).
(D') - Dead weight of empty rack.
Live Load (L) = 0 for the structure, since there are no moving objects in the rack load path.
Fuel Drop (Fd ) = Force caused by the accidental drop of the Accident heaviest load from the maximum possible height.
Load (See Section 4.6.6. )
Crane (Pf ) = Upward force on the racks caused by postulated Uplift stuck fuel assembly (4000 lbs).
Load Seismic (E) = Operating Basis Earthquake.
- Loads (E')
Thermal (To ) = Differential temperature induced loads (normal condition).
O. Loads (Ta ) = Differential temperature induced loads (abnormal design condition). For upset and emergency conditions. Ta is the differential temperature for the full core offload condition. For faulted conditions T ,is the differential temperature for the loss of cooling condition.
The conditions Ta and T o cause local thermal stresses to be produced. The worst situation will be obtained when an isolated storage location has a fuel assembly which is generating heat at the maximum postuisted rate. The surrounding storage locations are assumed to contain no fuel. The heated water makes unobstructed contact with the inside of the storage walls, thereby producing the maximum possible temperature difference between the adjacent cells. The secondary stresses thus produced are limited to the body of the rack; that is, the support legs do not experience the secondary (thermal) stresses.
4.4.2.2 Load Combinations Each component operating condition has been evaluated for the applicable loading combinations listed below:
O 4-11 0077L/00111
m ;
a) Normal Condition D+L D+L+To D+L+To+E D' + T o b) Upset Condition D + L + Te + E D+L+Ta + Pf i
D+Ta + Fp c) Energency Condition D + Ta + Pg.+ E l D+Ta+FD+E d) Faulted Condition D+L+Ta + E' l D+L+FD l D+L+Pf 4.5 DESIGN AND ANALYSIS PROCEDURES 4.5.1 Design and Analysis Procedures for the Spent Fuel Pool l 4.5.1.1 Spent Fuel Pool Structure Finite Element Analysis In this analysis, the EBS/NASTRAN progras, developed by Ebasco and linked to the consercially available NASTRAN progras, was used. Various layers'of ;
concrete and reinforcing bars were used to determine the effects of concrete '
cracking. The nonlinear analysis scheme based on the combination of stiffness '
iteration and load iteration methods, which were available in EBS/NASTRAN progras, was used to automatically determine the stresses in the concrete and reinforcing bars after the concrete cracks. The finite element model used in this analysis can be aussarized as follows:
a) Since the effect of the additional fuel rack load on the pool floor is limited to the sat in the pool area, the upper portion of the pool walls is not required for the re-evaluation. Therefore, the finite element model included the lower portion of walls, the pool floor (sat) and the underlying soil. The structural cosponents included in the model are shown on Figure 4-10. The cut-off ~
boundary of the walls is at EL. 45.25 ft.
b) The following boundary conditions were used at the model cut-off boundaries:
- 1) South end of the sat - Rotational springs representing the bending resistance of the cut-off sat were provided.
4-12 0077L/00111
i
- 11) Top of the walls - The rotation about the axis parallel to the j edge of the wall was restrained to consider the effect of the !
cut-off wall. This assumed boundary condition has little effect Of on the response of the peol mat, since the boundary is far above the mat. This waa demonstrated in the linear analysis results.
iii) South end of east and west valls - Since the rigidity of the cut-off walls is very small, a free boundary conditior. was assumed.
A computer plot of the finite element model is presented in Figure 4-11 which j shows the uverall view of the mode) indicating the composite of the four exterior and one interior walls.
4.5.1.2 Liner and Anchorage Analysis The liner and its anchors were evaluated for the temperature load, the strain '
induced load due to the deformation of the floor, and the horizontal seismic load. The program POSBUKF developed by Ebasco was used for the liner buckling analysis due to the temperature and strain induced loads. This program is I capable of dett:sining the post-buckling stress / strain if the liner plate buckles. The effect of the hydrostatic pressure was considered in this analysis. In calculating the in-plane shear due to the horizontal seismic i
- loads transmitted from the fuel rack to the liner, the maximum assumed i friction coefficient of 0.8 was used.
The liner anchors were evaluated for the unbalanced liner in-plane force due
. to the temperature and strain induced loads, as well as the horizontal seismic gJY in-plane shear force.
4.5.1.3 Foundation Stability and Soil Bearing A detailed soil bearing evaluation was performed for the increased fuel rack loading. Tne soil stresses were obtained at each sat corner and compared to the allowable value. Stability calculations were performed for overturning and sliding.
1 4.5.2 Design and Analysis Pre.Q res for Spent Fuel Storage Racks 1
Tne purpose of this subsection is to demonstrate the structural adequacy of the spent fuel rack design under normal and accident loading conditions. The.
method of analysis presented herein uses a time-history integration method similar to that previously used in the Licensing Reports on High Density Fuel Racks for Fermi 2 (Docket NU 50-341), Quad Cities 1 and 2 (Docket Nos 50-254 and 50-265), Rancho Seco (Docket No 50-312), Grand Gulf Unit 1 (Docket No 50-416), Oyster Creek (Docket No 50-219), V C Summer (Docket No 50-395),
Diablo Canyon 1 and 2 (Docket Nos 50-275 and 50-323) and Byron Units 1 and 2 (Docket Nos 50-454 and 50-455). The results show that the high density spent fuel racks are structurally adequate to resist the postulated stress combinations associated with level A, B, C and D conditions as defined in References 1 and 2.
4-13 0077L/00111
4.5.2.1 Analysis Outline Thus, they are The spent fuel storage racks are ceismic Category I equipment.
required to remain functional during and after a Safe Shutdown I Earthquake (3). As noted previously, these racks are neither anchored to the The individual rack pool floor nor are they attached to the side walls.Furthermore, a particular rack may be modules are not interconnected.
completely loaded with fuel assemblies (which corresponds to greatest rack inertia), or it may be completely empty. The coef ficient of hiction, Accordingp to between the su Rabinowicz(4) pports and pool floor is determined as follows.the results of 199 e steel plates submerged in water show a mean value of p to be 0.503 with a standard deviation of 0.125. The upper and lower bounds (based on twice the standard deviation) are thus 0.753 and 0.253, respectively. Two separate l analyses are performed for the rack assemblies with values of the coefficient of friction equal to 0.2 (lower limit) and 0.8 (upper limit), respectively.
Analyses performed for the geometrically limiting rack modules focus on limiting values of the coefficient of friction, and the number of fuel assemblies stored. Typical cases studied are:
- Fully loaded rack (all storage locations occupied),
- = 0.6, 0.2 ( y = coefficient of friction)
~
- Nearly empty rack # = 0.8, 0.2
- Rack half full # = 0.2, 0.8 Pool floor slab acceleration data developed for the Safe Shutdown Earthquake d (SSE) are shown in Figures 4-12 through 4-14. The method of analysis employed is the time-history method. The pool slab acceleration data were developed from the building response spectra.
The objective of the seismic analysis is to determine the structural response (stresses, deformation, rigid body motion, etc) due to simultaneous application of the three independent, orthogonal excitations.
The seismic analysis is performed in three steps, namely:
- 1. Development of a nonlinear dynamic model consisting of inertial mass )
elements and gap and friction elements.
- 2. Generation of the equations of motion and inertial coupling and 4 solution of the equations using the " component element time integration scheme-(6, 7) to determine nodal forces and displacements.
- 3. Computation of the detailed stress field in the rack (at the critical location) and in the support legs using the nodal forces calculated in the previous step. These stresses are checked against j
the design limits given in Section 4.6.2.2.
A brief description of the dynamic model follows.
4-14 0077L/0011L
4.5.2.2 Tuel Rack - Tuel Assembly Model Since the racks are not anchored to the pool slab or attached to the pool walls or to each other, they can execute a wide variety of rigid body motions. For example, the rack may slide on the pool floor (so-called *
" sliding condition"); one or more legs may momentarily lose contact with the liner (" tipping condition"); or theThe rack may experience a combination of structural model should permit simulation sliding and tipping conditions. Since these of these kinematic events with inherent built-in conservatists.
racks are equipped with girdle bars to dissipate energy due to inter-rack impact (if it occurs), it is also necessary to model the inter-rack impact Similarly, lift off of the support legs phenomena and subsequent in a conservative manner. impacts must be modelled using appropriate impact eleme Coulomb friction between the rack and the pool liner must be sisuisted by These special attributes of the rack appropriate piecewise linear springs.
dynamies require a strong emphasis on the modeling of the linear and nonlinear springs, dampers, and stop elements. De model outline in the remainder of this section, and the model description in the following section describe the detailed modeling technique to simulate these effects, with emphasis placed on the nonlinearity of the rack seismic response.
4.5.2.2.1 Outline of Model
- a. The fuel rack structure is a folded metal plate assemblage welded to a baseplate and supported on four legs. The rack structure itself is a very rigid structure. Dynamic analysis of typical multicell racks has shown that the action of the structure is captured almost completely by the behavior of a six degrees-of-freedom structure; therefore, the movement of the rack cross-section at any height is O described in terms of the six degrees-of-freedom of the rack base.
- b. The seismic motion of a fuel rack is characterized by random rattling of fuel assemblies in their individual storage locations.
Assuming that all assemblies vibrate in phase obviously exaggerates the computed dynamic loading on the rack structure. This assumption, however, greatly reduces the required degrees-of-freedom needed to model the fuel assemblies which are represented by five lumped sasses located at different levels of the rack. We centroid of each fuel assembly mass can be located, relative to the rack structure centroid at that level, so as to simulate a partially loaded rack.
l
- c. The local flexibility of the rack-support interface is modeled conservatively in.the analysis.
- d. The rack base ;upport may slide or lift off the pool floor.
- e. The pool floor and walls have a specified time-history of seismic accelerations along the three orthogonal directions.
- f. Fluid coupling between rack and assemblies, and between rack and adjacent rac.ks, is simulated by introducing appropriate inertial roupling into the system kinetic energy. Inclusion of these effects uses the methods of References 4 and 6 for rack / assembly coupling and for rack / rack coupling (see Section 4.5.2.2.3 of this report).
4-15 0077L/0011L
- g. potential impacts between rack and assemblies are accounted for by appropriate " compression only" gap elements between masses involved.
b h. Fluid damping between rack and assemblies, and between rack and adjacent rack, is conservatively neglected.
l
- i. The supports are modeled as " compression only" elements for the The l vertical direction and as " rigid links" for dynamic analysis.
bottom of a support leg is attached to a frictional element as described in Section 4.5.2.2.2. The cross-section inertial properties of the support legs are computed and used in the final computations to determine support leg stresses,
- j. The ef fect of sloshing has been shown to be negligible at the bottoc of a pool and hence is neglected.
- k. Inter-rack impact, if it occurs, is simulated by a series of gap elements at the top and bottom of of the rack in the two horizontal directions. The most conservative case of adjacent rack movement is assumed; each adjacent rack is assumed to move completely out of phase with the rack being analyzed.
- 1. The form drag opposing the motion of the fuel assemblies in the storage locatione is conservatively neglected in the results reported herein.
- m. The form drag opposing the motiva of the fuel rack in the water is also conservatively neglected in the results reported herein.
- n. The rattling of the fuel assemblies inside the storage locations causes the " gap" between the fuel assemblies and the cell wall to change from a maximum of twice the nominal gap (to a theoretical zero gap.
basedHowever,thefluidcouplinf9).
on linear vibration theory coefficients 8) utilized Studies in the are literature show that inclusion of the nonlinear effect (viz., vibration <
amplitude of the same order of sagnitude as the gap) drastically lowers the equipment response (10),
Figure 4-15 shows a schematic of the model. Six degrees-of-freedos are used j to track the motion of the rack structure. Figures 4-16 and 4-17 respectively, show the inter-rack impact springs and fuel assembly / storage cell impact springs.
The production run model for simulating fuel assembly motion incorporates five lumped masses. The lower mass is assumed to be attached to the baseplate and to move with the baseplate. The four rattling masses are located at quarter Two height, half height, three quarter height and top of the rack. j degrees-of-freedom are t. sed to track the motion of each rattling mass.
The solution procedure described in the following is implemented in computer code DYNARACK, which is a validated computer code under Holtec's Q A program.
1 O !
4-16 0077L/0011L l
1
I 4.5.2.2.2 Model Description The absolute degrees-of-freedom associated with each of the saas locations are O shown in Figure 4-15. As shown, the discrete mass fractions are located at heights z=0, 0.25H, 0.5H. respectively. Table 4-6 gives the degrees-of-freedom and the associated generalized coordinates. !
Ug (t) is the pool floor slab displacement seismic time-history. Thus, as tabulated in Table 4-6 and shown in Figure 4-15, there are fourteen degrees-of-freedom in the system. Not shown in Figure 4-15 are the gap elements used to model the support legs and the impacts with adjacent racks.
4.5.2.2.3 Fluid Coupling An effect of some significance requiring careful modeling is the so-called
" fluid coupling effect". If one body of mass (a) vibrates adjacent to another body (sass m), and both bodies are submerged in a frictionless fluid sedium, then Newton's equations of motion for the two bodies have the form:
(21+M11) 1-H12 2 = applied forces on mass si
-M21 kl +(s2+M22) h2= applied forces on mass s2 X,X2 1
denote absolute accelerations of mass si and s2, respectively.
H11, M12. M 21 and M 2 2 are fluid coupling coefficients which depend on )
Fritz(9 the shape of the two bodies, their relative disposition, etc.
It is to be gives data for M ij for various body shapes and arrangements.
noted that the above equation indicates that the effect of the fluid is to add O a certain amount of mass to the body (M11 to body 1), and an external force which is proportional to the acceleration of the adjacent body (sass s2)-
Thus, the acceleration of the one body affects the force field on another.
This force is a strong function of the interbody gap, reaching large values for very small gaps. This inertial coupling is called fluid couplink. It has an important effect in rack dynamics. The lateral motion of a fuel assembly inside the storage location will encounter this effect. So will the motion of a rack adjacent to another rack. These effects are included in the equations of motion. The fluid coupling is between fuel array node i and cell wall in I Figure 4-17. Furthermore, the rack equations contain coupling terms which model the effect of fluid in the gaps between adjacent racks. The coupling f terms modeling the effects of fluid flowing between adjacent racks are computed assuming that all adjacent racks are vibrating 180 degrees out of phase from the rack being analyzed. Therefore, only one rack is considered surrounded by a hydrodynamic
- mass computed as if there were a plane of symmetry located in the middle of the gap region.
Finally, fluid virtual mass is included in the vertical direction vibration equations of the rack; virtual inertia is also added to the governing equation corresponding to the rotational degree-of-freedom, q(t).
I O
l 4-17 0077L/0011L l
4.5.2.2.4 Damping In reality, damping of the rack motion arises from material hysteresis (material damping), relative intercomponent motion in structures (structural damping), and fluid drag effects (fluid damping). In the analysis, a maximum of 2% structural damping is imposed on elements of the rack structure during SSE seis: : simulations. This is in accordance with the St Lucit Unit 1 FSAR(13). Material and fluid damping are conservatively neglected. The dynamic model has the provision to incorporate fluid damping effects; however, no fluid damping has been used for this analysis.
4.5.2.2.5 Impact Ref erring to Figure 4-18, any fuel assembly code may impact the corresponding structural mass node. To simulate this impact, four compression-only gap elements around each rattling fuel assembly node are provided (see Figure 4 -17 ) . As noted previously, fluid dampers may also be provided in parallel with the springs. The compressive loads developed in these springs provide the necessary data to evaluate the integrity of the cell wall structure and stored array during the seismic event. Figure 4-16 shows the location of the impact springs used to simulate any potential for inter-rack impacts. Section 4.5.2.4.2 gives more details on these additional ispact springs.
. 4.5.2.3 Assembly of the Dynamic Model s
The cartesian coordinate system associated with the rack has the following nomenclature:
o x = Horizontal coordinate along the short direction of rack rectangular platform o y
- Horizontal coordinate along the long direction of the rack rectangular platferm o z = Vertically upward As described in the preceding section, the rack, along with the base, supports, and stored fuel assemblies, is modeled for the general three-dimensional (3-D) motion simulation by a fourteen degree-of- freedom model. To simulate the impact and sliding phenomena expected, 60 nonlinear gap elements and 16 con 11near friction elements are used. Gap and friction elements, with their connectivity and purpose, are presented in Table 4-7.
If the simulation model is restricted to two dimensions (one horizontal motion plus vertical motion, for example) for the purposes of model clarification only, then a descriptive model of the simulated structure which includes gap and friction elements is shown in Figure 4-18. (Note that only the top rattling mass is shown for clarity.)
The impacts between fuel assemblies and rack show up in the gap element, having local stiffness K ,I in Figure 4-18. In Table 4-7, gap elements 5 through 8 are for the vibrating mass at the top of the rack. The support leg spring rates Kg are modeled by elements 1 through 4 in Table 4-7. Note that 4-1B 0077L/0011L i I
j To siculate the local compliance of the concreto floor is included in E g. l sliding potential, friction elements 1 through 8 in Table 4-z are empicyad. '
Friction elements 2 and 8, and 4 and 6 (Table 4-7) are represented as Ky in O Figure 4-18. The friction of the support /licer interface isf up modeled by a to the j
\' piecewise linear spring with a suitably large stiffness K , ;
limiting lateral load, N, where N is the current compression load at the interface between support and liner. At every time step during the transient analysis, the current value of N (either zero for Finally, theliftoff condition, support or a rotational I
compressive finite value) is computed.
friction springs K R reflect any rotational restraint that may be offered by the foundation. This spring rate is calculated using a modified Boussinesq equation (4) and is included to simulate the resistive soment This' of therotation support '
to counteract rotation of the rack leg in a vertical plane.
spring is also nonlinear, with a eero spring constant value assigned after a ;
certain limiting condition of slab soment loading is 7eached. I :
The colinearity of these springs (friction elements 9,11,13 and 15 in Table 4-7) reflects the edging limitation imposed on the base of the rack support legs. In this analysis, this effect is neglected; any support leg bending, induced by liner / baseplate friction forces, is resisted by the leg acting as a beam cantilevers from the rack baseplate. ,
For the 3-D simulation, all support elements (listed in Table 4-7) are included in the model. Coupling between the two horizontal seismic motions is ,
provided both by the offset of tne fuel assembly group centroid which causes ]
the rotation of the entire rack and by the possibility of liftoff of one or i more support legs. The potential exists for the rack to be supported on one or more support legs or to liftoff completely during any instant of a complex 3-D seismic event.
All of these. potential events may be simulated during a q
3-D motion and have been observed in the results.
4.5.2.4 Time Integration of the Equations of Motion J
4.5.2.4.1 Time-History Analysis Using 14 DOF Rack Model Having assembled the structural model, the dynamic equations of motion corresponding to each degree-of-freedom can be written by using Newton's second law of motion; or by using Lagrange's equation. The system of equations can be represented in matrix notation as:
[M) (q) = (Q) + (G) l where the vector (Q) is a function of nodal displacements and velocities, and (G) depends on the coupling inertia and the ground acceleration.
Premultiplying the above egiistions by [M]-1 renders the resulting equation uncoupled in mass.
We have: (q) = [M]-1 (Q) + [M)-1 (G)
As noted earlier, in the numerical simulations run to verify structural integrity during a seismic event, all elements of the fuel assemblies areThis will p assumed to move in phase.
induce additional conservatism in the time-history analysis.
4-19 0077L/0011L
i Tnis equation set is mass uncoupled, displacement coupled, and isThe ideally computer suited for numerical solution using a central difference scheme.
program "DYNARACK"a is utilized for this purpose.
Stresses in various portions of the structure are computed from known element forces at each instant of time.
Dynamic analysis of typical sulticell racks has shown that the action of the structure is captured almost completely by the behavior of a six degree-of-freedom structure; therefore, in this analysis model, the movement of the rack cross-section at any height is described in terms of the rack base degrees-of-freedom (qi(t), ...q6(t)). The remaining degrees-of-freedomIn this are associated with horizontal movements of the fuel assembly masses. j dynamic model, four rattling masses are used to represent fuel assembly !
novement. Therefore, the final dynamic model consists of air. j degrees-of-freedom for the rack plus eight additional mass degrees-of-freedom '
for the four rattling masses.
The remaining portion of the fuel assembly is assumed to move with the rack base. Thus, the totality of fuel mass is j included in the simulation. 1 4.5.2.4.2 Evaluation of Potential for Inter-Rack Impact Since the racks are closely spaced, the simulation includes impact springs to
- model the potential for inter-rack impact, especially for low values of,the friction coefficient between the support and the pool liner. To account for this potential, five inter-rack gap elements were located at each side of the rack at the top and at the baseplate. Figure 4-16 shows the location of these gap elements. Loads in these elements, computed during the dynamic analysis, g are used to assess rack integrity if inter-rack impact occurs.
4.6 STRUCTURAL EVALUATION CRITERIA 4.6.1 Structural Acceptance Criteria for Spent Fuel Pool Structure 4.6.1.1 Criteria The stresses / strains resulting from the loading combinations described in Section 4.4.1 satisfy the following acceptance criteria:
a) Spent Fuel Pool Concrete Structure The design stress limits described in Section 3.8.1.6 of St i Lucie Unit No. 1 Updated FSAR were used for the evaluation of the spent fuel pool reinforced concrete structural components.
The capacity of all sections was computed in accordance with ACI 318-63 Part IV-B, Ultimate Strength Design.
- The numerical procedure underlying DYNARACK has been previously utilized in licensing of similar racks for Fermi 2 (Docket No 50-341), Quad cities 1 and 2 O (((Docket Nos 50-254 and 265), Rancho Seco (Doc Docket No 50-219), V C Summer (Docket No 50-395), and Diablo Canyon 1 and 2 Docket Nos 50-275 and 50-323).
4-20 0077L/0011L
b) Liner and Liner Anchors
/
The acceptance criteria for the liner and liner anchors is in accordance with the requirements specified in Paragraph CC-3720 and CC-3730 of ACI-ASME Section III, Division 2, Subsection CC and can be summarized as follows:
- 1) Liner The strain in the liner induced by thermal loads and the deformation of the pool structures is limited ~to the allowables presented in Table CC-3720-1 of ACI-ASME Section III Code. Load Combinations (a) and (b) presented in Section 4.4.1.2 of this report are considered as Service Load category and (c) and (d) as Factored Load category as these terms are used in that table.
- 11) Liner Anchors i
The displacement of the liner anchors induced by thermal {
J loads and deformation of the pool structures is limited to the allowable presented in Table CC-3730-1 of ACI-ASME Section III Code. Load Combinations (a) and (b) presented in Section 4.4.1.2 of this report are considered as Normal Load category, (c) as Extreme Environmental Load category and (d) as Abnormal Load category as these terms are used in that table.
4.6.1.2 Material Properties The following saterial properties were used in the analysis of the ;
spent fuel pool structure:
a) Concrete - (f'c = 5,200 psi) )
Young's modulus Ec = 3.85 x 106 pst Poisson's ratio F e = 0.17 Thermal Expansion coeff a e = 5.5 x 10-6 1/or b) Rebar Steel -
Young's modulus Es = 29 x 106 psi Poisson's ratio Fs = 0.30 Thermal Expans, ion coeff a s = 6.5 x 10-6 1/or Yield Strength = 40,000 psi !
c) Liner Plate -
Young's modulus Ep = 28.8 x 10 6 psi Poisson's ratio Fp = 0.3 Thermal Expansion coeff a p = 6.5 x 10-61/oF Yield Strength = 27,500 psi 4.6.1.3 Results a) Spent Fuel Pool Floor
~
For the nonlinear analysis of the selected loading cases (Section 4.4.1.2), the maximum stress results in the concrete and rebars are summarized in Table 4-2.
4-21 0077L/0011L
It is observed that the stresses in wall reinforcement are significantly affected by those loading combinations which j include temperature effects. It is further observed that
\ loading case y, which has the largest temperature gradients, has i I
the worst effect on concrete compressive stresses for both sat and wall locations. Also, the average reinforcement stress for )
locations is greatest for this loading case while the sat rebar stress at the cask storage area is greatest for loading case vi.
The safety factor (SF) is deffned as ultimate stress divided by saximum actual stress including load factors. The safety factors for the maximum stress are also presented in Table 4-2.
The smallest safety factors for the reinforcement tension and concrete compression are 1.10 and 3.65, respectively, which resulted frca loading case v, while the smallest safety factor for concrete shear is 1.05 resulting from loading case vi. This clearly it.dicates that the shear stress in the concrete is the governing component. The critical location of this shear stress is at the cask storage area of the mat, since the thickness of j I
the sat is the smallest (5 feet) here.
b) Liner and Anchorage
~
The critical loading case for the liner evaluation was loading case v which produced marious compressive stress in the liner plate. This compressive stress was due to temperature and the deformation of mat. The buckling analysis result indicated that the liner plate would not buckle, due to the stability effect of r the hydrostatic pressure.
Two loading conditions were considered necessary in the liner anchor evaluation; one was the strain-induced load which produced the unbalanced in-plane force at the edge of the pool area, and the other was the horizontal seismic load transmitted through the friction between the rack support and the liner.
This horizontal seismic. load was assumed to the uniformly distributed at the liner anchors. A maximum friction coefficient of 0.8 was used in calculating this horizontal force. In the liner anchor analysis, the load-deflection l
relationship for the liner anchor subjected to the liner in-plane force, which is usually obtained from actual test data, is required. Since there were no test data available for the actual anchor,, size of W8 x 24, the load-deflection test data for a lesser strength anchor angle 3 x 2 x 1/4 were used in this analysis. This is considered to be a conservative approach.
The results of the liner and liner anchor evaluation are summarized in Table 4-3. The minimum safety factors for liner and liner anchor are 5.20 and 1.33 respectively. It should be noted that the actual safety factor for the liner anchor would be greater than 1.33 if the load-deflection data for the actual anchor size of W8 x 24 were used.
O 4-22 0077L/0011L
c) Foundation Stability and Soil Bearing O A detailed soil bearing evaluation was performed. The soil stresses were obtained at each sat corner and coepared to the allowable value.
The results for the critical loading case are suasarised in Table 4-4. The minimum safety factor for soil bearing for this loading condition is 1.0.
Stability calculations were performed for overturning and sliding. The results for the critical loading case are summarized in Table 4-5.
The minimum safety factors for overturning and sliding for this loading condition are 4.59 and 3.10 respectively.
4.6.2 Structural Acceptance Criteria for Spent Fuel Storage Racks 4.6.2.1 Criteria l e
There are two sets of criteria to be satisfied by the rack modules:
- a. Kinematic Criterion This criterion seeks to ensure that the rack is a physically j I
stable structure. St Lucie racks are designed to sustain certain inter-rack impact at designated locations in the rack modules. Therefore, physical stability of the rack is
' considered along with the localized inter-rack impacts.
Localized permanent deformation of the module is permissible, so l j
long as the suberiticality of the stored fuel array is not violated.
l I b. Stress Limits The stress limits of the ASME Code,Section III, Subsection NT, 1983 Edition up to and including Susser 1984 Addenda are used since this code provides the most appropriate and consistent set of limits for various stress types and various loading conditions.
4.6.2.2 Stress Limits for Specified Conditions The following stress limits"are derived from the guidelines of the ASME Code,Section III, Subsection NF, in conjunction with the material properties data of Subsection 4.6.2.3.
O V
4-23 0077L/00111
4.6.2.2 1 Normal and ljpset Conditions (Level A or Level B Service Limits) 7 s. Allowable stress in tension on a net section = Ft = 0.6 Sy (d or Fg = (0.6) (21,300) = 12,780 psi (rack material)
Ft is equivalent to primary membrane stresses Ft- (.6) (25,000) = 15,000 psi (upper part of support feet)
= (.6) (106,300) = 63,780 psi (lower part of support feet)
- b. On the gross section, allowable strees in shear is:
Fy = .4 S y
= (.4) (21,300) = 8,520 psi (main rack body)
Fy - (.4) (25,000) = 10,000 psi (upper part of support feet)
= (.4) (106,300) = 42,520 psi (lower part of support feet)
- c. Allowable stress in compression, F,:
F, = (1-1/2a2 C~2) S y s 5 +3aBC . a3 3 e 8C2 27r2 E 1/2 where a = k1/r and Ce =
Sy k1/r for the main rack body is based on the full height and cross section of the honeycomb region. Substituting numbers, we obtain, for both support leg and honeycomb region:
F, = 12,780 psi (main rack body)
F = 15,000 psi (upper part of support feet)
= 63,780 psi (lower part of support feet)
- d. Maximum allowable bending stress at the outermost fiber due to flexure about one plane of symmetry:
Fb = 0.60 Sy = 12,780 psi (rack only)
Tb = 15,000 psi (upper part of support feet)
= 63,780 psi (lower part of support feet)
- e. Combined flexure and compression:
O f a Cmx fbx Cay fby
+ + <1 F. Dx Tbx Dy Fby 4-24 0077L/0011L
where:
f, = Direct compressive stress in the section fbx = Maximum flexural stress about x-axis iby - Maximum flexural stress about y-axis cmx - Cay = 0.85 f
a D, =1-F',,
fa Dy =1-F',y where:
12 x2 E F'er, ey =
2 kibx,y
\ rbx,y /
lb and rb indicate unbraced length and radius of gyration about the concurrent plane (x or y) and the subscripts x,y reflect the particular axis of bending.
- f. Combined flexure and compression (or tension):
fa fx b fby
__ + + < 1.0 0.6sy Fbx Fby The above requirement should be set for both the direct tension and compression case.
4.6.2.2.2 Faulted Condition (Level D Service limits) l Paragraph F-1370 (Section III, Appendix F)(2), states that the limits for the Level D condition are the minimum of 1.2 (S y t/F ) or (0.75u/Ft )
times the corresponding limits for Level A condition. Since 1.2 Sy is less than 0.7 Su for the rack material, and for the upper part of the support feet, the multiplying factor for the limits is 2.0 for the SSE condition for the upper section. Tne factor is 1.53 for the lower section under SSE conditions.
Instead of tabulating the results of these six different stresses as dimensioned values, they are presented in a dimensionless form. These so-called stress factors are defined as the ratio of the actual developed stress to its specified limiting value. With this definition, the limiting value of each stress factor is 1.0 for OBE and 2.0 or 1.53 for the SSE condition.
4-25 0077L/0011L
4.6.2.3 Material Properties The data on the physical properties of the : rack and support materials.
- p obtained from the ASME Boiler & Pressure Vessel Ode,Section III, appendices, and supplier's catalog, are listed in Tables 4 2 and 4-9. The reference design temperature for evaluation of material peaperties is 2000F.
4.6.2.4 Results for Rack Analysis Figures 4-12 through 4-14 show the pool slab ar.,: ion in horizontal z, horizontal y and vertical directions. This motion is for the SSE, Results are abstracted in Table 4-10 for modules B1, H1 and G1 (Figure 2-1).
1 A complete synopsis of the analysis of these modules subject to the SSE earthquake motions is presented in a summary Table 4-10 which gives the bounding values of stress factors Ri (i = 1,2,3,4,5,6). The stress factors are defined as:
R3= Ratio of direct tensile or compressive stress on a net section -
to its allowable value (note support feet only support compression)
R2= Ratio of gross shear on a net section to its allowable value ;
i Ratio of maximum bending stress due to bending about the x-axis j R3=
to its allowable value for the section p R4- Ratio of maximum bending stress due to bending about the y-axis j to its allowable value j Q
R5= Combined flexure and compressive factor (as defined in 4.6.2.2.le) i I
R6= Combined flexure and tension (or compression) factor (as defined in 4.6.2.2.lf)
As stated before, the allowable value of R1 (i = 1,2,3,4,5,6) is 1 for the OBE condition, (except for the lowtr section of the support where the factor is 1.53), and 2 for the SSE.
The dynamic analysis gives the maximax (maximum in time and in space) values l i
of the stress factors at critical locations in the rack module. Since these maximax values are subject to minor (under 5%) variation if the input data 1 (viz., rack baseplate height, cell inside dimension) is perturbed within the range of manufacturing tolerances, the bounding values, instead of the actual values, are presented in Table 4-10. The terms in Table 4-10 have the following meaning:
a implies Ri < 1. 0 b implies Ri < 1. 5 e implies Ri < 1.75 d implies Ri < 2.0 4-26 0077L/0011L
_ _ _ - _ _ _ _ _ _ _ _ _ _ -~
It is found that the results corresponding to SSE are sost critical when compared with the corresponding allow 1ble limits. The results given herein are for the SSE. The saximum stress factors (RI ) are below the limiting Or value for the SSE condition for all sections. It is noted that the critical load factors reported for the support feet are all for the upper sessent of the foot and are to be compared with the limiting value of 2.0.
Analyses have been carried out to show that significant margins of safety exist against local deformation of the fuel storage cell due to rattling impact of fuel assemblies and against local overstress of impact bars due to inter-rack impact.
l Analyses have also been carried out for the OBE condition to demonstrate that the stress factors are below 1.0. Results obtained for all rack sites and shapes are enveloped by the data presented herein. Overturning has also been considered for the cases where racks are adjacent to open areas.
4.6.3 Fuel Handlina Crane Uplift Analysis An analysis was performed to demonstrate that the rack can withstand an uplift load of 4,000 pounds produced by a jassed fuel assembly. This load, which exceeds the capacity of the fuel handling crane, can be applied to any point of the fuel rack without violating the criticality or structural acceptance
- criteria. Resulting stresses are within acceptable stress limits, and there is no change in rack geometry of a magnitude which causes the criticality acceptance criterion to be violated.
s 4.6.4 Impact Analyses 4.6.4.1 Impact Loading Between Fuel Assembly and Cell Wall The local stress in a cell wall is estimated from peak ispect loads obtained from the dynamic simulations. Plastic analysis is used to obtain the limiting impact load that can be tolerated. Including a safety margin of 2.0, the.
total limiting load for the number of cells is over four times the actual saximax (saximum in time and space) value.
4.6.4.2 Impacts between Adjacent Racks ,
All of the dynamic analyses assume, conservatively, that adjacent racks move completely out of phase. Thus, the highest potential for inter-rack impact is achieved. Based on the dynamic loads obtained in the gap elements simulating adjacent racks, we can study. rack integrity in the vicinity of the impact point. The use of framing material around the top of the rack allows the rack to withstand impact loads over 50,000 lbs at a corner of the rack prior to reaching the fully yie3ded state above the active fuel region. The actual rack-to-rack impact loads are less than 50,000 lbs, and thus, impacts between racks can be accommodated without violating rack integrity. It is found that pool walls are sufficiently far away from the racks such that pool wall-to-rack impact does not occur.
4.6.5 Weld Stresses O The critical weld locations under seismic loading are at the connection of the rack to the baseplate and in the support leg welds. For the rack welds, the allowable weld stress is the ASME Code value of 24,000 psi (Table NF-3324.5(a)-1, Subsection NF).
4-27 0077L/0011L
For the support logs, the allowable weld stress is govarned by the levels outlined in Section 4.6.2 (see NT-3324.5 for partial penetration velds).
The O Weld stresses due to heating of an isolated hot cell No thermal are alma comput
temperature above the value associated with all surrounding cells.
gradient in the vertical direction is assumed so that the resulta are conservative. Using the temperatures associated with this unit, the skip welds along the entire cell length do not exceed the allowable value for a thermal loading condition.
4.6.6 Summary of Mechanical Analyses The mathematical model constructed to determine the impact velocity of falling objects is based on several conservative assumptions, such as:
- 1. The virtual mass (see Refs 8-10 for further saterial on the subject) of the body is conservatively assumed to be equal to its displaced fluid mass. Evidence in the 21terature(11),
indicates that the virtual mass can be many times higher.
- 2. The minimum frontal area is used for evaluating the drag coefficient.
1
~
- 3. The drag coefficients utilized in the analysis are the lower bound values reported in the literature (121 In particular, at the beginning of the fall when the velocity of the body is small, the corresponding Reynolds number is low, resulting in a large drag coefficient.
- 4. The falling bodies are assumed to be rigid for the purposes of impact stress calculation on the rock. The solution of the immersed body motion problem is found analytically. The impact velocity thus computed is used to determine the maximum stress generated due to stress wave propagation.
With this model, the following analyses are performed:
l
- a. Dropped Fuel Accident I j A fuel assembly (weight is conservatively analyzed as 1500 pounds with control rod assembly) is dropped from 36 inches above the module and impacts the base. The final velocity of the dropped fuel assembly (just prior to impact) is calculated and, thus, the total energy at impact is known. To study basep1*te integrity, it is assumed that this energy is all !
directed toward punching of tne baseplate in shear and thus transformed into work done by the supporting shear stresses. It !
1 is determined thtt shearing deforsction of the baseplate is less I
than the thickness of the baseplate so that we conclude that l local piercing of the baseplate will not occur. Direct impact with the pool liner does not occur. The suberiticality of the adjacent fuel assemblies is not violated.
4-28 0077L/00111
- b. Dropped Tuel Accident II One fuel assembly drops from 36 inches above the rack and hits O the top of the rack. . Permanent deformation of the rack is found to be limited to the top region such that the rack e
cross-sectional geometry at the level of the top of the active fuel (and below) is not altered. The region of local permanent l deformation does not extend ,below 6 inches from the rack top.
An energy balance approach is used here to obtain the results.
- c. Jammed Tuel Handling Equipment A 4000-pound uplift force is applied at the top of the rack at the " weakest" storage location; the force is assumed to be I applied on one wall of the storage cell boundary as an upward '
shear force. The plastic deformation is found to be limited to the region well above the top of the active fuel.
These analyses prove that the rack modules are engineered to provide maximum safety against all postulated abnormal and accident conditions.
4.6.7 Definition of Terms Used in Section 4
- SI, S2, 53, S4 Support designations P
Absolute degree-of-freedos number i i
Relative degree-of-freedos number i 91 p Coefficient of friction Pool floor slab displacement time history in ut the i-th direction z,y coordinates Horizontal direction z coordinate Vertical direction .
1 Ispact spring between fuel assemblies and cell KI 1.inear component of friction spring Kf N ,
Compression load in a support foot KR Rotational spring provided by the poni slab Subscript i When used with U or X indicated direction (i
- 1 indicates x-direction, i = 2 indicates y-direction, i= 3 indicates z-direction)
Mass Matrix (generic notation)
(M)
(Q) Generalized coordinate vector k Axial Spring of Support leg locations 4-29 0077L/0011L
l 4.6.8 Lateral Rack Movement Lateral motion of the rack modules under seismic conditions could potentially O alter the spacing between rack modules. However, girdle bars on the modules prevent closing the spacing to less than 1.50 inches, which is greater than the normal flux-trap water-gap in the Region 1 reference design. Region 2 storage cells do not use a flux-trap and the reactivity is insensitive to the spacing between modules. Furthermore, soluble poison would assure that a reactivitf less than the design limitation is maintained under all conditions.
4.7 MATERIALS, QUALITY CONTROL, AND SPECIAL CONSTRUCTION TECHNIQUES 4.7.1 Construction Materials Construction materials will conform to the requirements of the ASME Boiler and Pressure Vessel Code,Section III, Subsection NF. All the materials used in the construction are compatible with the storage pool environment and will not i contaminate the fuel assembliee or the pool water. The plates, sheets, strips, bars and structural shapes used for rack construction are Type 304L stainless steel.
4.7.2 Neutron Absorbing Material
~
The neutron absorbing material, Boraflex, used in the St Lucie spent fuel rack l construction is manufactured by Bisco and fabricated to the safety-related nuclear criteria of 10 CFR 50, Appendix B. Boraflex is a silicone-based polymer containing fine particles of boron carbide in a homogeneous, stable 7 ~p matrix. The specification for the handling and installation of the poison
(_ ) material requires that it will not be installed in a stretched condition. The specification precludes the use of adhesives in the attachment of the boraflex to the rack cell walls. TPL will require that the manufacturing process avoid techniques which could pinch the boraflex. The design of the racks requires that additional lengths of boraflex, i.e. greater than the active length of a ,
fuel assembly, be installed to account for anticipated shrinkage of the i boraflex.
4.7.3 Quality Assurance The design, procurement, and fabrication of the new high density spent fuel storage racks comply with the pertinent Quality Assurance requirements of Appendix B to 10 CFR 50 as implemented through: FPL's Topical Quality Assurance Report FPL-NQA-100Al9); the Joseph Oat Quality Assurance Plan as described in their QA Manual,; the Holtec's Nuclear Quality Assurance plan as described in their QA Manual; and the Ebasco Quality Assurance Program for Nuclear Plants, ETR-1001. All have been approved by the NRC.
4.7.4 Construction Techniques 4.7.4.1 Administrative Controls During Manufacturing and Installation The St. Lucie Unit I new spent fuel storage racks will be manufactured at the Joseph Oat Corp., Camden, New Jersey. This facility is a modern high-quality shop with extensive experience in forming, machining, welding, and assembling nuclear-grade equipment. Forming and welding equipment are specifically O% designed for fuel rack fabrication and all welders are qualified in accordance with ASME Code Section IX.
4-30 0077L/0011L
To avoid damage to the stored spent fuel during rack replacement. all work on the racks in the spent fuel pool area will be performed using written and These procedures will preclude the movement of the fuel O approved procedures. racks over the stored spent fuel assemblies.
Radiation exposures during the removal of the old racks from the pool will be controlled by procedure. For anticipated radiation doses see Table 5-5.
Water levels will be maintained to afford adequate shielding from the direct radiation of the spent fuel. Prior to rack replacement, the cleanup system will be operated to reduce the activity of the pool water to as low a level as can be practically achieved.
4.7.4.2 Procedure 4.7.4.2.1 Reinstallation The following sequence of reinstallation events is anticipated for the spent fuel storage rack replacement for Unit I.
- a. Design and fabricate new spent fuel storage racks.
- b. Prepare modification procedure.
. c. Fabricate and test all special tooling,
- d. Receive and inspect new spent fuel storage racks.
4.7.4.2.2 Installation O The final configuration of the 17 new rack modules in the spent fuel pool is shown in Figure 2-1. The installation of these racks will be accomplished in accordance with the following considerations and guidelines:
o A temporary construction crane will be installed and operated in the spent fuel pool area to move new and existing rack modules within the spent fuel pool.
o At no time will this temporary crane carry a rack module directly over another module which contains stored spent fuel.
o The temporary construction crane and all rack modules sill be installed and/or removed from the spent fuel pool area with the fuel cask cran,e.
o All load handling operations in the spent fuel pool area will be conducted in accordance with the criteria of Section 5.1.1 of NUREG-0612. " Control of Heavy Loads at Nuclear Power Plants".
o Spent fuel relocations within the pool will be performed as required to maintain separation between the stored fuel and the rerack operations.
i O ,
4-31 0077L/0011L
y '
4 i
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4.8 TESTING AND IN-SERVICE SURVEILLANCE
\
4.8.1 Program Intent A sampling program to verify the integrity of the neutron absorber saterial employed in the high density fuel racks in the St Lucie fuel pool environment is described in the following paragraphs.
The program is conducted in a sanner which allows access to the representative absorber material samples without disrupting the integrity of the entire fuel storage system. The program is tailored to evaluate the saterial in normal use mode and to forecast future changes using the data base developed.
4.8.2 Description of Specimens The absorber material used in the surveillance program, henceforth referred to as
" poison", is representative of the Boraflex saterial used within the storage It'is of the same composition, produced by the same method, and certified to the same criteria as the production lot poison. The sasple coupon is of a system.
thickness similar to the poison used within the storage systes and not less than 4 Figure 4-19 shows a typical coupon. Each poison specimen by 2 inches on a side.is encased in a stainless steel jacket of an austenitic stainless steel alloy identical to that used in the storage systes, formed so as to encase the. poison
- saterial and fix it in a position and with tolerances similar to the design used for the storage system. The jacket is closed by tack welding in such a manner as to retain its form throughout the test period and still allow rapid and easy ,
opening without causing sechanical damage to the poison specimen contained within.
The jacket permits wetting and venting of the specimen similar. to the actual rack envirennent.
4.8.3 Specimen Evaluation l After the removal of the jacketed poison specimen from the cell at a designated i l
l time, a careful evaluation of that specimen will be made to determine its actual I condition as well as its apparent durability for continued function. Immediately l
af ter the removal, the specimen and jacket section will be visually examined for any effects of environmental exposure. Specific attention will be directed to the ,
examination of the stainless steel jacket for any evidence of physical degradation. Functional evaluation of the poison material vill be accomplished by the following measurements:
o A neutron radiograph of the poison specimen aids in the determination of the maintenance cf uniformity of the boron distribution.
o Neutron attenuation sessurements will allow evaluation of the continued nuclear effectiveness of the poison. Consideration will be given in the analysis of the attenuation sensurements to the level of accuracy of such sessurements, as indicated by the degree of repeatability normally observed by the testing agency.
o A measurement of the hardness of the poison material.will establish the continuance of physical and structural durability. The hardness acceptability criterion requires that the specimen hardness will not !
be less than hardness listed in the qualifying test document for laboratory test specimen irradiated to 10 reds. 'The actual hardness O
measurement will be made af ter the specimen has been withdrawn f roc:
the pool and allowed to air dry for not less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to allow ;
for a meaningful correlation with the pre-irradiated semple.
0077L/0011L j 4-32
o Measurement of the length, tha width, cnd the svarage thickness and comparison with the pre-exposure data will indicate dimensional stability within the variation range reported in the Boraflex laboratory test reports.
e In the event of any observed deterioration of the coupon that could affect the poison function, an immediate inspection of the " poison panels" in the rack will be perf ormed. The NRC will be advised immediately if the inspection I indicates degradation of the poison material.
1
4.9 REFERENCES
FOR SECTION 4
- 1. Nuclear Regulatory Commission, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants", NUREG-0800, Revision
- 1. July 1981.
- 2. ASME Boiler & Pressure Vessel Code,Section III, Subsection NF (1983 Edition up to and including Summer 1984 Addenda).
- 3. USNRC Regulatory Guide 1.29, " Seismic Design Classification," Rev 3, 1978.
- 4. " Friction Coefficients of Water Lubricated Stainless Steels for a Spent Fuel Rack Facility," Prof. Ernest Rabinowicz, MIT, a report for Boston f
~
Edison Company, 1976.
- 5. USNRC Regulatory Guide 1.92, " Combining Modal Responses and Spatial Components in Seismic Response Analysis," Rev 1. February 1976.
O 6. "The Component Element Method in Dynamics with Application to )
V Earthquake and Vehicle Engineering," S Levy and J P D Wilkinson, McGraw Hill, 1976.
- 7. " Dynamics of Structures," R W Clough and J Penzien, McGraw Hill (1975).
- 8. " Mechanical Design of Heat Exchangers and Pressure Vessel Components,"
Chapter 16. K P Singh and A I Soler, Arcturus Publishers, Inc.,1984.
- 9. R J Fritz, "The Effects of Liquids on the Dynamic Motions of Immersed Solids," Journal of Engineering for Industry Trans. of the ASME, February 1972, pp 167-172.
- 10. " Dynamic Coupling in a Closely Spaced Two-Body System Vibrating in Liquid Medium: The Case of Fuel Racks," K P Singh and A I Soler, 3rd International Conference on Nuclear Power Safety, Keswick, England, May 1982.
- 11. " Flow Induced Vibration," R D Blevens, VonNostrand (1977). j l
- 12. " Fluid Mechanics," M C Potter and J F Foss, Ronald Press, p. 459 (1975). i
- 13. St Lucie Plant Unit I, Updated Final Safety Analysis Report, Docket No l 50-335.
O V :
4-33 0077L/0011L
Table 4-1 BORAFLEX EXPERIENCE FOR HIGH DENSITY RACKS Plant NRC Type Docket No.
Site PWR 50-226 & 301 Point Beach I & 2 BWR 50-220 Nine Mile Point 1 PWR 50-269 & 270 Oconee 1 & 2 Pkt 50-282 & 306 Prairie Island 1 & 2 PWR 50-318 Calvert Cliffs 2 BWR 50-254 & 205 Quad Cities 1 & 2 PWR 50-390 & 391 Watts Bar 1 & 2 Waterford 3 PWR 50-382 BWR 50-341 Fermi 2 H B Robinson 2 PWR 50-261 BWR 50-458 River Bend 1 Rancho Seco 1 Pkt 50-312 BWR 50-410 Nine Mile Point 2 Shearon Harris 1 PWR 50-400 PWR 50-423 Millstone 3 Grand Gulf 1 BWR 50-416 l BWR 50-219 Oyster Creek V C Summer .
PWR 50-395 Diablo Canyon 1 & 2 PWR 50-275 & 323 Byron Units 1 & 2 PWR 50-454 & 455 l
O 4-34 0077L/0011L
o 7 7 5 8 4 0 9 0 5 2 9 0 0 0 F 1 3 1 1 2 S 1 1 s
s e t t t t t r t t e s e a a t e e e l 1 l t t S t L l 4 e0 l
5 e0 c8 7 e8 e0 e2 eL e6 0 8 1 8 8 8 5 0 rrA 5 5 1 8 1 9 7# 7 5f1 a cW 661 1e2 1e1 492 1e2 en h o 5 4 6 5 2 SC 8 7 0 1 5 8 0 6 4
ef F 1 1 1 1 1 1 1 noS o
t t t t t t n t t e a s e e e e e l l l l l M i c 5 e l
7 e e e 7 e e T 0 0 6 1 5 6 A 3 1 1 165 8e5 695 197 769 M 8e8 195 7 5 1 9 7 1 0 6 5 5
) 4 1 9 6
) s F 1 I
4 6 8 3 5 7 S s S 1 P e
( r t
Y S RH t t t t t t t A C n e e e e 0e e s MN 3l 4l 9l 2l 41 MI oeL 8l 3 e6 0e6 3 e6 l
4 e6 0e6 2 e6 2 e2 U it L 5 6 4 3 1 6 7 6 5 0 s eA 3 5 9 6 SE 92 - 92 91 O4 2 s rW - 91 92 6#2 - 91 R
SA ec SU rn 9 7 7 9 1 po 6 4 9 E EO mC 4 2 0 6 7 7 L t 1 S o F B T Cf S 6 4 6 5 3 3 6 A SRF o T MF m t t t t t t t U t e e e e e e 9 e e MS t 6l 8l 3l 1l 6l 4l 6l I B 5 e 0 e 5 e 0 e 7 e XL x T 1 e 3 e 6 7 9 1 5 a A 6 1 9 - 99 - 95 4
M M 88 - 99 - e9 - 95 199 M= '
T 3 I
8 3 3 2 0 1 N 1 5 9 9 1 4 8 U F
( S 4 1 1 1 1 1 2 t 9t 6t 3t 5t 6t 2t 0e 4 e 4 e 4 e 1 s 8 s 4 e s L 1l 5, i 6, e0 l 7, e8 l 7, e4 l 4, e4 l 7, e4 l s
e L
A 6, e 6 0 3 e4 8 8 3 8 8 2 3 5 8 2 0 1G1 192 192 291 1e2 r W 8e2 2e1 t r S a 1 8 6 4 3 b 1 0 5 ee 8 4 5 9 7 7, nRF 2 2 1 1 1 1 e S 1 t f so 9t 3t 3t 3t 5t 0t a 7 t 5 e 0e 7 e 0e M 3 e 7 e 3 e 9, l 9, e l 3, e l 1, e l 4, e l 3, e l 8, e l T e 8 0 9 3 5 0 A 9 1 4 4 M 1e8 1e3 1e5 195 2%7 292 2#5 e
e )
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TABLE 4-3 O ,
STRESS / STRAIN
SUMMARY
FOR LINER AND ANCHORS Strain or Disp 1 Stress or Force Remark Actual . Allowable SF Actual Allowable SF
-11.1 No Buckling --- -3.847x10-4 -2.0x10-3 5.20 Constraint-Liner Induced-Load kai due to in/in in/in.
hydrostatic pressure.
Evaluation not required --- 0.025 in 0.0625 in 2.5 ' constraint-Liner Induced Load
. Anchor by Code. .
2.03 1.33 Evaluation not required --- Horizontal 1.525 Seismic Load kips /in kips /in 'by Code.
(N-S)
NOTE: Actual stress and strain are based on Factored!(abnormal)
Load condition and allowables are based on Service (normal)
Load condition where these terms are as defined in Paragraph CC-3220 of ACI-ASME Section III Division'2.
)
(s / !
4~36 0077L/0011L-
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t TABLE 4-4 '
SOIL BLARING STRESSES (KST) .
j Corner 1 Location. D + L + SSE
'f N-I 4.1 '
S-E -1.7 N-W 12.0-S-W 7.9
)
NOTE: Allowable soil bearing stress is 12.0 KSF
~
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4-37 0077L/0011L
TABLE 4-5 STABILITY SATETY TACTORS Earthquake ' Overturning Sliding Loading Type D + L~+ SSE D + L + SSE
'SSE(N-5) + Vert. Up 7.82 3*10 SSE(I-W) + Vert. Up 4.59 3,31 0
)
1 0 ;
'-38 00771/0011L i
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Table 4-6 O DEGREES OF PREEDOM Displacement Rotation Location ux uy uz 6x gy Sz (Node) 1 p1 p2 p3 q4 q5 q6 1* Point 18 is assumed fixed to base at XB , Yg, Z-0 2 Point 2 is assumed attached to rigid rack at the top most point.
Pi = qi (t) + Ug(t)
Other , p10 Rattling
.,' p9, pil, p12 Node points 3*, 48, 5*
Masses I pl3, p14 0
4 0
0 4-39 0077L/0011L
m l
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NUMBERING SYSTEM FOR gap ELEMENTS AND TRICTION ELEMENTS
- 1. Nonlinear Springs (Cap Elements) (60 total)
Number Node Location Description 1 Support S1 Z compression only element i 2 Support 52 Z compression only element 3 Support 53 Z compression only element 4 Support 54 2 compression only element 5 2,28 X rack / fuel assembly impact element 6 2,2* X rack / fuel assembly impact element 7 2,2* Y rack / fuel assembly impact element 8 2,2* Y rack / fuel assembly impact element i
9-20 Other rattling masses 21-40 Bottom cross-section Inter-rack impact elements of rack (around edge) 41-60 Top cross-section Inter-rack impact elements i of rack (around edge) Inter-rack impact elements [
II. Friction Elements (16 total)
Node Location Description ]
Number 1 Support 51 X direction friction N* 2 Support 51 Y direction friction
\ 3 Support 52 X direction friction 4 Support S2 Y direction friction 5 Support S3 X direction friction 6 Support S3 Y direction friction 7 Support S4 X direction friction 8 Support S4 Y direction friction 9 51 X Slab moment 10 S1 Y Slab soment i 11 S2 X Slab moment 12 S2 Y Slab moment 13 S3 K Slab soment 14 S3 Y Slab soment 15 S4 X Slab soment 16 S4 . Y Slab soment 4-40 0077L/0011L
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Table 4-8 O< Young's RACK MATERIAL DATA Yield Ultimate Modulus Strength Strength Material E (psi) Sy (psi) Su (psi) 27.6 x 106 21300 68100 304 Stainless-Steel ASME Table Table Table Section III I-3.2 Reference I-6.0 I-2.2 l
I
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4-41 oo77t/co12t ]
- - _ - - - - - _ - - - - - - - - a
Table 4-9 ADJUSTABLE HEIGHT SUPPORT MATERIAL DATA O (Reference Temperature = 200er) ,
1 J
Young's Yield Ultimate Modulus x 10-6 Strength Strength Material (psi) ksi kai i
27.9 25.0 70 Upper part (Female)
SA-351-CF3 Lower Part 27.9 106.3 140 (Male)
SA-564-630 (age hardened to 11000F)
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O 4-42 co77L/co11t
TABLE 4-10 BOUNDING VALUES FOR STRESS FACTORS O Stress Factors
- R1 R2 R3 R4 R5 16 Run No. (Upper values for rack base - lower values for upper part of the support feet) a a a a a a SSE p= .8, full a a b b b I Module B1 a a a a a a SSE a a a a
- = . 2 , f 411 a a Module B1 a a a a a a SSE p= .8, empty i I I I I I Module B1 g
- The terms a, b, e and d imply'the stress factors Rg (i = 1, 2 .. 6) s ,/ are bounded by the following limiting values
- l a: 1.0 b: 1.5 c: 1.75 d: 2 I
. I O
4-43 0977t/co11L
- i i
TABLE 4-10 (Cont'd) t BOUNDING VALUES FOR STRESS FACTORS O Stress Factors
- R2 R3 R4 R5 16 R1 Run No.
(Upper values for rock base - lower values for upper part of the support feet) ;
a 3 a a )
SSE i i b b a a a F = .8, a Module H1 full SSI i i a a 3 '
i a b b a a a
- = .8 '
Module H1 1/2 full in positive z-half a a a a l SSE a a !
i i i a
- = .2, I I Module H1 l 1/2 full in positive z-half I
)
O 4-44 0077L/0011L
TABLE 4-10 (Cont'd)
BOUNDING VALUES FOR STRESS FACTORS O i Stress Factors
- R1 R2 R3 R4 R5 R6 Run lio. (Upper values for rack base - lower values for i upper part of the support feet) .
)
a a a a a SSE a_
a a a a a a p = .2, empty Module B1 a a a a a a SSE a i I 5 5 E 4 = .8, 1/2 full Pos. X Quadrant Module B1 SSE a a a a a a a a a
- = .2, 1/2 Full a a a Pos. X Quadrant Module B1 O SSE a a a a a a M = .8, Full a a 5 5 c I Module G1
- 4-45 0077L/0011L
TABLE 4 (Cont'd)
BOUNDING VALUES FOR STRESS FACTORS O
Stress Factort*
R1 R2 R3 R4 15 R6 Run No. (Upper values for rack base - lower values f or upper part of the support feet) a a a a a a SSE I I I I I I F = .2 Module G1 Full SSE a, a, a a, a_ a_
a b b c c p=.2 a Module G1 Positive x half loaded SSE a a a a a a l
Module G1 I a i a i i
- = .2, Positive x O half loaded U'
SSE a a a a a a V = .2 a a a a 5 E Hodule G1 Empty l
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4-46 0077L/0011L
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5.0 COST / BENEFIT AND ENVIRONMENTAL ASSESSMENT 5.1 COST / BENEFIT AND THERMAL ASSESSMENT The cost / benefit of the chosen reracking alteration is demonstrated in the following sections. {
l 5.1.1 Need for Increased Storage Capacity
- a. FPL currently has no contractual arrangements with any fuel {
reprocessing facilities. j FPL executed three contracts with the Department of Energy (DOE) on j June 16, 1983 pursuant to the Nuclear Waste Policy Act of 1982, but ]
the disposal facilities are not expected to be available for spent I fuel any earlier than 2003.
f
- b. Table 5-1 includes a proposed refueling schedule for St. Lucie Unit i 1 and the expected number of fuel assemblies that will be l transferred into the spent fuel pool at each refueling until the j total existing capacity is reached and it becomes impossible to conduct a full refueling past cycle 9-10 shutdown in 1991.. At present the licensed capacity of Unit 1 is 728 storage cells. All !
I calculations in the table for loss of full core reserve (FCR) are I
~
based on the number of licensed total cells in the pool. The table is then continued assuming the installation of 1706 replacement cells and loss of full core reserve is projected in the year 2009.
l
- c. The St. Lucie Unit 1 spent fuel pool is expected to contain 529 p spent fuel assemblies at the time of raracking.
- d. Adoption of this proposed spent fuel storage expansion would not necessarily extend the time period that spent fuel assemblies would be stored on site. Spent fuel will be sent offsite for final disposition under existing legislation, but the government facility is not expected to be available until after 2003.
- e. The estimated date when the spent fuel pool will be filled with the proposed increase in storage capacity is provided in Table 5-1. In addition to the fuel assemblies, six storage locations are occupied by non-fuel equipment (e.g. dummy fuel assembly, trash basket, etc.).
5.1.2 Estimated Costs Total construction cost associated with the proposed modification is 8 million dollars. This figure includes the cost of designing and fabricating the spent fuel racks; engineering costs; and installation and support costs at the site.
O 5-1 0078L/0011L
5.1.3 consideration of Alternatives
- a. There are no operational commercial reprocessing facilities available for FPL's needs in the United States, nor are there expected to be any in the foreseeable future.
l b. At the present time, there are no existing available independent spent fuel storage facilities. While plans are being formulated by DOE for construction of a spent fuel repository per the Nuclear Waste policy Act of 1982, this facility is not expected to be available to accept spent fuel any earlier than 2003.
- c. At present, FPL has no license to transship fuel between facilities.
St. Lucie Unit 1 lost full core reserve capacity upon startup of cycle 7 in 1937. Permanent transfer of St. Lucie Unit 1 spent fuel to other facilities would only compound storage problems there and is not a viable option.
- d. Estimates for costs of replacement power were calculated based on the last official rate of return. The assumption was made that the unit could be operated without maintaining full core reserve, thus cycle 09-10 in 1990 would be the last refueling possible with existing storage capacity. Table 5-2 indicates the average yearly fuel cost increases for St. Lucie Unit 1 after three years of
- shutdown. Plant shutdown would place a heavy financial burden on Florida residents within FPL's service area and cannot be justified.
5.1.4 Resources Committed Reracking of the spent fuel pools will not result in any irreversible and irretrievable commitments of water, land, and air resources. The land area now used for the spent fuel pools will be used more efficiently by safely increasing the density of fuel storage.
The materials used for new rack fabrication are discussed in Section 4.7.1.
These materials are not expected to significantly foreclose alternatives available with respect to any other licensing actions designed to improve the possible shortage of spent fuel storage capacity.
5.1.5 Thermal Iepact on the Environment Section 3 2 considered the following: the additional heat load and the l'
anticipated maximum temperature of water in the SFP that would result from the proposed expansion, the resulting increase in evaporation rates, the additional heat load on component and/or plant cooling water systems, and whether there will be any significant increase in the amount of heat released to the environment. As discussed in Section 3.2, the proposed increase in storage capacity will result in an insignificant impact on the environment.
O 5-2 0078L/0011L
l 1
5.2 RADIOLOGICAL EVALUATION i
5.2.1 Solid Radwaste Current
() Currently, resins are generated by the SFP purification system.No significant frequency of resia change out is approximately once per year.
increase in volume of solid radioactive wastes is expected due to the new racks. It is estimated that a minimal amount of additional resins will be generated by the spent fuel pool cleanup system during reracking. The most recent isotopic analysis of the spent fuel pool resin is present in Table 5-7.
Operating plant experience with high density fuel storage has not indicated l
any noticeable increase in the solid radioactive wastes generated by the l increased fuel storage capability. l 5.2.2 Gaseous Releases Table 5-3 summarizes the FHB Gaseous releases in 1985 and 1986. No significant increases are expected as a result of the reracking.
5.2.3 Personnel Exposure
- a. The range of values for recent (October 1985 refueling) samma 1 isotopic analyses of spent fuel pool water is shown on Table 5-4.
- b. Operating experience shows dose rates of less than 10 arem/ hour either at the edge or above the center of the spent fuel pools regardless of the quantity of fuel stored. This is not expected to change with the proposed reracking because radiation levels above the pool are due primarily to radioactivity in the water,
( which experience shows to return to a level of equilibrium.
Stored spent fuel is so well shielded by the water above the fuel that dose rates at the top of the pool from this source are l
negligible. l
- c. There have been negligible concentrations of airborne radioactivity from the spent fuel pools. Operating plant experience with dense fuel storage has shown no noticeable increases in airborne radioactivity above the spent fuel pool or l
at the site boundary. Recent spent fuel pool airborne radioactivity is depicted in Table 5-8. No significant increases are expected from more dense storage.
- d. As stated in Section 5.2.1, reracking and utilization of the new racks will resolt in no significant increase in the radwaste generated by the spent fuel pool cleanup system. This is because operating experience has shown that with high density storage racks, there is no significant increase in the radioactivity levels in the spent fuel pool water, and no significant increase in the annual m ?-rem due to the increased fuel storage, including the changing of spent fuel pool cooling system resins and filters.
ID
, U 0078L/0011L l
5-3
- e. Most of the corrosion products " crud" associated with spent fuel storage is released soon after fuel is' removed from the reactor. Once fuel is placed into the pool storage positions, O addL ional crud contribution is minimal.
The highest possible water level is maintained in the spent fuel pool to keep exposure as low as reasonably achievable. Should crud buildup ever be detected on the spent fuel pool walls around the pool edge, it could easily be washed down.
- f. There is no' access underneath the spent fuel pool. During normal operation, the radistion dose rate'around the outside of the pool could increase locally up to .53 ares per hour should freshly discharged fuel be located in the cells adjacent to the pool liner. ~This dose rate will decrease to below .25 area per hour after approximately 25 days. The depth of the water above the fuel is sufficient so there will be no seasurable increase in dose rates above the pool due to radiation emitted directly from the fuel. _
Operating experience has shovu a negligible increase in man-rem due to the increased fuel storage with high density racks. ~Therefore, a negligible increase in the annual man-res is expected at St. Lucie as a result of the increased storage capacity of the spent fuel pools with the higher density storage racks.
The existing St. Lucie or Turkey Point health physics program did not have to
-be modified as a result of the previous increase in storage of spent fuel. It 0 is not anticipated that the health physics program will need to be sodified for this increase in storage capability.
5.2.4 Radiation Protection During Re-Rack Activities 5.2.4.1 General Description of Protective Measures The radiation protection aspects of the spent fuel pool modification are the responsibility of the Plant Health Physicist, who is assisted by his staff, with the support of the Corporate Health Physicist and his staff. Gamma ,
radiation levels in the pool area are constantly monitored by the station Area Radiation Monitoring System, which has a high level alors feature.
Additionally, periodic radiation and contamination surveys are conducted in work areas as necessary. Where there is a potential for significant airborne radionuclides concentrations, continuous air samplers can be used.in addition to periodic grab sampling . Personnel working in radiologically controlled areas will wear protective clothing and when required by work area conditions .,
respiratory protective equipment, as required by the applicable Radiation Work Permit (RWP). Personnel monitoring equipment is assigned to and worn by all personnel in the work area. At a minimus, this equipment consists of a thermoluminescent dosimeter (TLD) and self-reading pocket dosimeter.
Additional personnel sonitoring equipment, such as extremity badges, are utilized as required.
Contamination control measures are used to protect persons from internal-
. exposures to radioactive material and to prevent the spread of contamination.
Work, personnel traffic, and the movement of material and equipment in and out of the area are controlled so as to minimize contamination problems. Material 5-4 0078L/0011L
and equipment will be monitored and appropriately decontaminated and/or wrapped prior to removal from the spent fuel pool area. The station radiation p protection staff will clorely monitor and control all aspects of the work so
( that personnel exposures, both internal and external, are maintained as low as reasonably achievable (ALARA).
Water levels in the spent fuel pool will be maintained to provide adequate !
shielding from the direct radiation of the spent fuel. Prior .to rack replacement, the spent fuel pool cleanup system will be operated to reduce the activity of the pool water to as low a level as can be practically achieved.
5.2.4.2 Anticipated Exposures During Reracking Table 5-5 is a summary of expected exposures for each phase of the Unit I reracking operation. These estimates are made based on the proposed installation plan, including fuel transfers, the use of long aandled tools, and the onsite decontamination cleanup and packaging of the old storage racks. Also, current pool radioactivity levels were conservatively increased in calculating these exposures. The total occupational erposure for the Unit i reracking operation is conservatively estimated to be between 10 and 15 pe. son-rea. See Table 5-5.
5.2.5 Rack Disposal l The spent fuel storage rack modules th.st will be removed from the spent fuel pool weigh between 32,200 and 45,000 pounds each. The total weight of these racks is approximately 223 tons and the racks occupy a volume of 910 ft3 . )
They will be cleaned of loose contamination, packaged and shipped to a l licensed radioactive waste processing facility. !
Shipping containers will meet the requirements of DOT regulations pertaining to radioactive waste shipments, including limitations with respect to the waste surface dose and radionuclides activity distribution. Shipping containers will be certified to meet all requirements for a strong tight package. The maximum weight of a loaded shipping container will be in accordance with the American Association of State Highway and Transportation Officials (AASHTO). Trucks and drivers used for rack and waste transportation will have all permits and qualifications required by the Federal DOT and the DOT for each State through which the truck will pass.
At the waste processing facility, the racks will be decontaminated to the maximum extent possible. Remaining portions of the racks and contaminated waste generated from decontamination will be buried at a licensed radioactive waste burial site. In preparing non-decontaminable waste for shipment and subsequent burial, volume reduction methodologies will be employed such as compaction, combining metallic materials with " soft waste" to minimize void space, and super compaction where feasible.
O .
1 5-5 0078L/0011L
5.3 ACCIDENT ITALUATION 5.3.1 Spent Fuel Handling Accidents 5.3.1.1 Fuel Assembly Drop Analysis l
For a drop on top of the rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum separation distance from the fuel of more than 12 inches, sufficient to preclude neutron coupling. Maximum expected deformation under seismic or accident conditions will not reduce the l minimum spacing between the dropped assembly and the stored fuel assemblies to less than 12 inches. Consequently, fuel assembly drop accidents will not result in a significant increase in reactivity due to the separation j distance. Furthermore, soluble boron in the pool water would substantially {'
reduce the reactivity and assure that the true reactivity is always les:: than the limiting value for any conceivable dropped fuel accident.
As discussed in Section 4.1.2.2, the proposed spent fuel pool m Wifications will not increase the radiological consequences of fJe1 handling accidents previously evaluated in Section 9.1.4.3 of the St Lucie Unit 1 Updated FSAR(2).
5.3.1.2 Cask Drop Analysis 5.3.1.2.1 Cask Handling .
As discussed in Subsection 9.1.4.3 of the St Lucie Unit 1 Updated FSAR(2),
limit switches prevent movement of the cask beyond the spent fuel cask storage area in the northeast corner of the fuel pool; prevent interference of the
{
,[gN cask crane bridge, trolley, and hoist with fuel racks or building structures; d and restrict vertical lift of the cask to an elevation sufficient to gain entry to the Fuel Handling Building. The Terack program does not alter the l cask. handling procedures de6cribed in Updated FSAR Section 9.1. The cask handling crane meets the design and operational requirements of Section 5.1.1 !
of NUREG-0612. " Control of Heavy loads at Nuclear Power Plants"(3). ]
5.3.1.2.2 Radiological Consequences For the calculation of radiological consequences potentially resulting from a cask drop accident, two cases were evaluated regarding the number of fuel assemblies that are assumed to suffer a loss of integrity:
Case 1: One-third of a core is placed in the spent fuel pool each year during refueling for the next 23 years, until the pool is filled.' The number of assemblies damaged is equal to the number offloaded during a normal refueling plus the remainder of the pool filled with discharged assemblies from previous refuelings.
O S-6 0078L/0011L
Case III. Ont-third of a cora'is placed in th2 spsnt fual pool each year during refueling for the next 20 years. Following the 21st year of operation, the entire core is removed from the reactor aLJ.placed into the pool, which fills the pool. The -
number of assemblies damaged is equal to a full-core offload plus the remainder of the pool filled with discharged assemblies from previous refuelings.
The model for calculating the thyroid and mole-body esclusion area boundary doses incorporates the conservative assumptions specified in Standard Review Plan (SRP) Section 15.7.5(4) and Regulatory Guide 1.25(5) with the exception that a 1.0 Radial Peaking Factor (RPF) is utilized. An RPF of 1.65
~
as specified in Regulatory Guide 1.25 is intended to represent the highest-burnup fuel assembly. While this value may be appropriate for the analysis of a postulated accident involving a single assembly, it is grossly overconservative when applied to an analysis of a normai refueling batch or a full core whose fuel assemblies have various exposure histories. An RPF of 1.0 has been selected as being more representative for the off-load of one or i more regions from the core and has been applied to each assembly in the present analysis. The use of a 1.0 RPF for the calculation of cask drop radiological consequences has been previously submitted to the NRC for FPL's St. Lucie Unit 1 plant.(6)
The core inventory used in the analysis of the dropped spent fuel cask is given by the St. Lucie Unit 1 Updated FSAR Table 15.4.1-1c. As indicated in the St. Lucie Unit 1 Updated FSAR Table 15.4.1-4 and the St. Lucie Unit 1 Updated FSAR Subsection 2.3.4.3 the 0-2 hour exclusion area boundary (EAB)
X/Q value of 8.55 x 10-5 sec/m3,is used for the analysis. The results of the analysis demonstrate that by retaining the required decay time of spent fuel in the pool to be the minimum times-imposed in Technical Specification
, Q_ 3.9.14 prior to s.oving a spent fuel cask into the spent fuel pool, the potential offsite doses are less than 10 percent of 10 CFR Part 100'11mits even if all the assemblies in a full pool are damaged and no credit is taken for filtration. For a decay time in the spent fuel pool of either 1180 hours0.0137 days <br />0.328 hours <br />0.00195 weeks <br />4.4899e-4 months <br /> (Case I described above) or a decay time of 1490 hours0.0172 days <br />0.414 hours <br />0.00246 weeks <br />5.66945e-4 months <br /> (Case II), the EAB thyroid dose, which is governing, is approximately 15 res. The SRP 15.7.5 l acceptance criterion for these analyses (25 percent of 10 CFR 100) is 75 res. 1 The whole-body doses calculated for Case I and Case II for the corresponding 1
. decay times are less than 0.1 res, compared to the SRP 15.7.5 seceptance ;
criterion of 6 rea'. Accordingly, the Technical Specification 3.9.14 decay time requirements prior to cask handling operations are acceptable. This is conservative, since not all spent fuel storage modules located in the pool are susceptible to impact from any single cask drop. Thus, the proposed spent fuel pool modifications do not increase the radiological consequences of the cask drop accident previousfy evaluated.
5.3.1.2.3 overhead Cranes Except for the area described in Section 5.3.1.2.1, the spent fuel cask crane is not capable of traveling over or into the vicinity of the spent fuel pool.
A complete cask crane component description, cask handling description, and cask crane design evaluation are provided in Updated FSAR Section 9.1 and will not be affected as a result of the rerack program.
5-7 0078L/0011L
q l
l 5.3.1.2.4 Acceptability The accident aspects of review establish acceptability with respect to Sections 5.3.1.2.1 and 5.3.1.2.2 of this report.
Technical Specification requirements for spent fuel decay time prior to moving :
a spent fuel cask into the spent fuel pool containing freshly-discharged fuel assemblies result in potential offsite doses less than 10 percent of 10 CTR Part 100 limits should a dropped cask strike the stored fuel assemblies.
5.3.1.3 Abnormal Location of a Fuel Assembly The abnormal location of a fresh unirradiated fuel assembly of 4.5% enrichment {
could, in the absence of soluble poison, result in exceeding the design reactivity limitation (keff of 0.95). This could occur if the essembly were to be eitner positioned outside and adjacent to a storage rack module or j' inadvertently loaded into a Region 2 storage cell, with the latter condition producing the larger positive reactivity increment. Soluble poison, however, is present in the spent fuel pool water (for which credit is permitted under I these conditions) and would maintain the reactivity substantially less than the design limitation.
The largest r;eactivity increase occurs for accidentally placing a new fuel assembly into a Region 2 storage cell with all other cells fully loaded.
Under this condition, the presence of 500 ppa soluble boron assures that the infinite multiplication factor would not exceed the design basis reactivity.
With the normal concentration of soluble poison present (1720 ppm boron), k .
is normally .less than 0.80 and will not be critical even if Region 2 were to be fully loaded with fresh fuel.of 4.5% enrichment. Administrative procedures
(,5 will be used to confirm and assure the continued presence of soluble poison in the spent fuel pool water.
5.3.1.4 New Fuel 5torage in Region 2 ,
l In a confirming calculation, it was determined that a checkerboard storage l pattern in Region 2 would allow new fuel assemblies of 4.5% enrichment to be j safely accommodated without exceeding the limiting 0.95 keff value. In this ;
checkerboard loading pattern, the fuel assemblies are located on a diagonal array with alternate storage cells empty of any fuel.
The Monte Carlo calculation (AMPX-KF.NO) resulted in a k . of 0.6084 + 0.0085.
With a one sided K-factor (13) for 95% probability at a 95% confidence level and a 6 k of 0.0125 for uncertainties (Table 3-1 for Region 2), the maximus k . is 0.857, which is substantially less than the 0.95 limiting value. Thus, Region 2 may be safely used for.the temporary storage of new fuel assemblies provided the storage configuration is restricted to the checkerboard pattern with alternate storage locations empty of fuel.
5.3.2 Fuel Decay Technical Specification 3.9.14 requires decay time for freshly discharged fuel prior to movement of the cask into the pool. As a result, with the increased storage capacity, the radiological consequences of a cask drop are less than 10 percent of the requirements of 10 CFR Part 100.
5-8 0078L/0011L
5.3.3 Loads Over Spant Fusi Administrative procedures and Tech Spec 3.9.7 licite the nazicus weight of loads that may be transported over spent fuel.
5.3.4 Temperature and Water Density Effects The moderator temperature coefficient of reactivity in both regions is negative; a moderator temperature of 40C, with a water density of 1.0 g/cm3, was assumed for the reference designs, which assures that the true reactivity will always be lower, regardless of temperature.
i j
Temperature effects on reactivity have been calculated and the results are shown in Table 5-6. Introducing voids in the water internal to the storage cell (to simulate boiling) decreased reactivity. Since, at saturation j temperature, there is no significant thermal driving force, voids due to boiling will not occur in the outer (flus-trap) water region of Region 1.
5.3.5 Conclusions Since the spent fuel cask will not be handled over or in the vicinity of spent fuel as discussed in Section 5.3.1.2.1, the proposed modification does not result in a significant increase in the probability of the cask drop accident previously evaluated in the St Lucie Updated FSAR or Safety Evaluation Report (9). Furthermore, as shown in Section 5.3.1.2.2, by requiring a minimum decay time for spent fuel prior to moving a spent fuel cask into the spent fuel pool, the potential offsite doses are less than 10 percent of 10 CFR Fart 100 limits should a dropped cask strike the stored fuel assemblies.
The proposed spent fuel pool modifications do not increase the radiological ,,
consequences of a cask drop accident previously evaluate 6.
I Since there will be a negligible change in radiological conditions due to the increased storage capacity of the spent fuel pool, no change is anticipated in the radiation protection program. In addition, the environmental consequences of a postulated fuel handling accident in the spent fuel pool, described in Updated FSAR Section 15.0, remain unchanged. Therefore, there will be no '
change or impact to any previous determinations of the Final Environment statement (8). Based on the foregoing, the proposed amendments will not significantly affect the quality of the human environment; therefore, under 10 l
CFR 51, issuance of a negative declaration is appropriate.
o O 0078L/0011L 5-9
5.4 REFERENCES
FOR SECTION 5
- 1. PROMOD III Computer Code, Version 22.8, Energy Managenznt Associates.
- 2. St Lucie Plant - Unit 1. Updated Final Safety Analysis Report, f--)
g ,j Docket No. 50-335.
- 3. Nuclear Regulatory Commission, " Control of Heavy Loads at Nuclear Power Plants", NUREG-0612, July 1980.
- 4. Nuclear Regulatory Comnission, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, NUREG-0800, Revision 1. July 1981.
- 5. Nuclear Regulatory Commission, " Assumption Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," Regulatory Guide 1.25, March 1972.
- 6. St Lucie Plant Unit 1. Final Safety Analysis Report, Section 9.1, Docket No. 50-335.
- 7. St Lucie Plant - Unit 1. Technical Specifications, Facility Operating License DPR-67.
- 8. NUREG 0575, " Final Environmental Impact Statement on Handling and Storage of Spent Light Wa'.er Power Reactor Fuel," Vol 1-3 USNRC August 1979.
- 9. St Lucie Plant - Unit 1, Safety Evaluation Report. Licenses DPR-Docket Nos. 50-335.
- 10. St Lucie Plant - Unit 1 Final Environmental Statement, Docket No.
50-335.
'- 5-10 0078L/0011L
TABLE 5-1 NUCLEAR FUEL DISCHARGE INFORMATION ST LUCIE UNIT 1 Cumulative Total Number of of Spent Fuel Cycle Shutdown Assemblies Assemblies No Dates Discharged in the pool 3/28/78 60 60 01 3/31/79 68 128 02 3/15/80 88 216 03 9/9/81 64 280 04 2/26/83 92 372 05 10/21/85 73(1) 445 06 2/7/87 84 529*
07 728 CURRENTLY INSTALLED / USABLE CELLS ACTUAL CYCLE INFORMATION THROUGH CYCLE SEVEN, PROJECTED THEREAFTER 08 10/8/88 72 601 '
09 3/16/90 80 681 10 10/1/91 64 745 I 11 3/15/93 68 813 -
l 12 10/1/94 64 877 13 3/15/96 68 945 .'
l 14 10/1/97 64 1009 15 3/15/99 68 1077 16 10/1/2000 64 1141 1 1
17 3/15/02 68 1209 18 10/1/03 64 1273 19 3/15/05 68 1341 20 10/1/06 64 1405 '
21 3/15/08 68 1473 '
22 10/1/09 64 1537(2) 23 3/15/10** 68 1605 '
24 10/1/11 64 1669 25 3/15/13 ,
68 1737 26 10/1/14 64 1801 END OF LIFE- 3/1/16 217 Final Offload 2018
- FULL CORE RESERVE (FCR) LOST AT 511 CELLS WITH CURRENT RACKS; RERACK REQUIRED TO REGAIN FCR
- CURRENT END OF LIFE = 7/1/2010, CURRENT APPLICATION PENDING TO EXTEND LIFE TO MARCH 1, 2016 O (1) INCLUDES FUEL ASSEMBLY H406 WHICH WAS PART OF A RECONSTITUTION.
b (2) FCR LOST AT 1489 CELLS WITH POISONED RERACK (ASSUMES 1706 AVAILABLE STORAGE LOCATIONS) 5-11 0078L/0011L
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TABLE 5-3 GASEOUS RELEASES O FROM FUEL HANDLING BUILDING l 1985 Radionuclides Curies Xe-133 64.9 Xe-135 9.9 I-131 2.74 E-4 Kr-85m 1.0 ,
Kr-87 7.94 E-1 1 Kr-88 1.5 1986 Radionuelide Curies Xe-133 20.1 Xe-135 7.5 I-131 1.45 E-4 Kr-85m 1.0 0, Kr-87 Kr-88 1.3 1.3 1-133 9.88 E-5 1 O
5-13 0078L/0011L
TABLE 5-4 GAMMA ISOTOPIC ANALYSIS SPENT FUEL POOL WATER RADION1)CLIDE ACTIVITY l Co-58 4.90 E-4 $ /d -
Co-60 6.70 E-4 pCi/s1 Cs-134 5.80 E-4 yCi/s1 Cs-137 9.32 E-4 pCi/a1 7.70 E-2 pCi/s1 H-3 i
~
O 1
O 5-14 0078L/0011L
l TABLE 5-5 ANTICIPATED DOSES DURING RERACKING
'u Project Dose (san-res)
Spent Fuel Movement 0.5 Install Rack Removal Crane 0.1 q and Remove Crane l l
Decontaminate Racks Underwater 2.5 and Remove from Pool Decontamination of Racks in 2.0 Cask Washdown Bldg and Crate for Shipment Install New Racks 2.5 Repair Equipment 0.5 Measurement of New Racks and 1.0 Dummy Assembly Testing q
- 0. Health Physics Coverage and Surveys 2.0 General Entry and Inspection 1.0 Fuel Pool Floor vacuum and 2.0 Filter Disposal TOTAL 14.1 l -
O 5-15 0078L/0011L
a.
TABLE 5-6 O
EFFECT OF TEMPERATURE AND VOID ON CALCULATED REACTIVITY OF STORAGE RACK Case Incremental Reactivity Change, dk Region 1 Region 2 40C Reference Reference 200C -0.0014 -0.0009 500C -0.0054 -0.0032 800C -0.0109 -0.0062 1200C -0.0216 -0.0117 1200 + 20% void -0.0767 -0.0446 0 a O
5-16 0078L/0011L
TABLE 5-7 i
SPENT FUEL POOL PURIFICATION SYSTEM O RADIOFUCLIDE ANALYSIS REPORT RESIN ACTIVITY l
f RADIONUCLIDES ACTIVITY NON-TRANSURANIC #Ci/cm3 Co-58 56.93 Cs-137 23.25 Cs-134 17.06 j Co-60 6.63 i I-131 3.37 Cs-136 1.09 Mn-54 0.44 i 6.63E-3 j C-14 1.63E-4 j Tc-99 I-129 1.95E-6 I H-3 7.6E-2 Sr-90 3.95E-3 Ni-63 1.39 Fe-55 9.68E-2 TRANSURANIC nCi/gm Pu-239, 240 1.39E-5 Pu-241 2.52E-3 l Cm-242 9.28E-6 )
TRU* 3.71E-5 Resin Volume = 35 ft3 or 0.991 m3
- Other alpha-emitting transuranic nuclides with half-lives greater than 5 years.
O 5-17 0078L/0011L
TABLE 5-8 l SPENT FUEL POOL AIRBORNE ACTIVITY p RADIONUCLIDES ANALYSIS REPORT ,
\. *l CONCENTRATION RADIONUCLIDES yCi/mi Ag-110m < 3.0E-11 Ar-41 = 9.SE-11 Ba-139 = 1.2E-10 Ba-140 < 7.8E-11 l Ba-141
- 8.5E-11 l l Ce-139 =c2.5E-11 l Ce-144 < 1.5E-10 Ce-141 < 3.6E-11 {
co-57 <1.7E-11 Co-58 < 3.6E-11 l l Co-60 <4.4E-11 ,
j Cr-51 < 2.0E-10 l Cs-134 < 2.7E-11 I. Cs-137 = 3.7E-11 -
Cs-138 =c2.8E-11 l Fe-59 < 4.0E-11 l
V I-134 = 2.7E-11 1-135 < 8.0E-11 Kr-85 < 5.2E-9 Kr-85m <2 7E-11 Kr-87 <4.3E-11 La-140 <2.0E-11 l La-142 < 9.9E-11 Mn-54 <2.5E-11 l Mo-99 <1.4E-10 l Nb-95 ac 2. 4E-11 l Nb-97 < 3. 4 E-11 l
" < " = Lower level of Detection and was not detected in the sample.
- I-131 and Xe-133 were detected in the sample. Fuel movement and fuel reconstitution was in progress during the sample collection. 'Ihese are normally not detected and their lower levels of detection are I-131 == 3. 0E-11 pCi/ml and Xe-133 m: 9.0E-11 FCi/al.
l O
5-18 0078L/0011L O-
q
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TABLE 5-8 (Cont'd ACTIVITY Q
L/
FUEL NALYSISPOOL AIRBORNE .,... _._._ .,. . REPORT SPENTRADIONUCLIDE A CONCENTRATION pCi/mi RADIONUCLIDES < 6 4E-11 *
- 1.9E-10 .
e 6 7E-11 Np-239 ac2 4E-11 Rb-88 =c 1 8E-10 D /
Rb-89 < 2 6E-11
- Ru-103 = 5.SE-11 Ru-106 =c 1 7E-11 Sb-124 =2 3E-11 Sb-125 <7 1E-11 5n-113 <1 7E-11 y Sr-85 <2 2E-11 / '
Sr-91 =c2 3E-11 Sr-92 =cl.1E 3 Te-99m 3 4E 10 i Te-132 ac2 0E-10 / !
Xe-131m .c1 3E-10 lY /
' exe-133 =cl.0E-10 /
Xe-133m ac2 3E-11
();
f" 5 Xe-135 ac3 7E-11
/
Xe-138 <5 4E-11 Y-88
/
not detected in the fuel /
Fuel movement andThese'are and was tion I
/j wer level of Detec he sample. collection.are
- < " = Lo during the sampleof detection
/ 3 waswere detected in progress lower levels in t9.OE-11 pCi/ml.
- I-131 and Xe-13 reconstitution detected and theirpCi/mi aod .7e-133 sw mi normally not1-131 = 3 0E-11 i
0078L/0011L O 3-19