ML19269B797
| ML19269B797 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 01/04/1979 |
| From: | Linder F DAIRYLAND POWER COOPERATIVE |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| LAC-6067, NUDOCS 7901150093 | |
| Download: ML19269B797 (39) | |
Text
'$
Ju DAIRYLAND POWER COOPERATIVE
& Gune, 0 % aaa 54601 January 4, 1979 In reply, please refer to LAC-6067 DOCKET NO. 50-409 Director of Nuclear Reactor Regulation U.
S. Nuclear Regulatory Commission Washington, D. C.
20555
SUBJECT:
DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACBWR)
PROVISIONAL OPERATING LICENSE NO. DPR-45 PROPOSED MODIFICATION - SPENT FUEL STORAGE
Reference:
(1) NRC Letter, Ziemann to Madgett, dated November 1, 1978.
(2) NRC Letter, Ziemann to Madgett, dated December 7, 1978.
Gentlemen:
Enclosed with this letter is additional information required for your review of the proposed modification of the LACBWR spent fuel storage pool.
This information is provided in response to those items listed in the enclosures of the above referenced letters.
1 Please contact us if additional information is required.
Very truly yours, DAIRYLAND POWER COOPERATIVE Frank'Linder, General Manager FL:NLH:af Enclosures cc:
(See attached sheet).
2199 257 7 901 15 0(Sib $+
4 Director of Nuclear Reactor Regulation LAC-6067 Washington, D. C.
20555 January 4, 1979 cc:
J. G. Keppler, Regional Director U. S. Nuclear Regulatory Commission Directorate of Regulatory Operations Region III 799 Roosevelt Road Glen Ellyn, IL 60137 Charles Bechhoefer, Esq., Chairman Atomic Safety and Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Mr. Ralph S. Decker Route 4 Dox 190D Cambridge, MD 21613 Dr. George C. Anderson Department of Oceanugraphy University of Washington Seattle, Washington 98195 C.
S. Hiestand, Jr.
Attorney at Law Morgan, Lewis & Bockius 1800 M Street, N. W.
Washington, D. C.
20036 Kevin P. Gallen Attorney at Law Morgan, Lewis & Bockius 1800 M Street, N. W.
Washington, D. C.
20036 Coulee Region Energy Coalition P. O. Box 1583 La Crosse, WI 54601 2199 258
?
s-4 Response to NRC Ouestions Submitted by NRC Letter, Ziemann to Madgett, dated November 1, 1978 Proposed Modification - LACBWR Spent Fuel Storage ITEM 1 Provide dravings shouing (1) the interlock betueen the upper
- Ser rack and lover tier rack sectione, (2) the seating surfaces and fuel assembly support plate in the upper egg crate grid of the louer tier rack, and (3) the adjustable pade, used to transmit horizontal seismic loads, and their locations on the rack structures.
DPC RESPONSE:
The requested drawings are enclosed with this submittal.
ITEM 2a Regarding the rack structural analysis provided in your August 7 submittal:
Load combination (5),
D+T+U.L, should be D+L+U.L uith a structural acceptance limit of 1.55.
Therefore, the design report, NES 81A 0546, Rev.
2, should be revised accordingly.
Indicate whether a stuck or nammed fuel assembly vill re-eult in higher atreeces than assuming the grapple gate hooked on a fuel storage cell.
A lso, justify applying the load to a louer rack grid structure versue an upper grid structure or other possible locations along the length of a storage cell.
DPC RESPONSE:
Since the live load (L) of the fuel elements negates the effects of the uplift load on the storage rack structure, load combination (5)
D+T+U.L, given in Section 5. 2 (b) of the structural analysis report (NES Report 81A0546, Rev. 2) is core critical than load combination D+L+U.L.
The load combination D+T+U.L is considered
.o be a factored load condition.
The allowable limit for the fartored load condition is taken as 1.6S in accordance with Section '.8.4.II.5 of the Standard Review Plan.
It should be noted that the stresses in the rack structure are significantly lower then rather the 1.6S or the 1.5S limit.
The weight of the fuel assembly reduces the net uplift load on the structure.
Therefore, the storage rack is subjected to higher stresses for the case of grapple being hooked onto the starage cell than for the case of a stuck or jammed fuel assembly.
2199 259 4
t-The uplift load has been applied at the critical location on the storage rack.
The critical location for the application of the up-lift load is the corner of the lower grid of an empty 4 x 10 stor-age rack, since this application point results in the lifting of the associated vertical rack support pad and maximizes the canti-lever arm between the uplift load and the nearest functioning rack support point (rack mid support pad).
Preliminary analyses were performed with the uplift load applied at the upper grid structure and at other locatione ca the storage ra.ck.
These analyses indi-cated lower stresses in the rack structure than those given in the design report, NES81A0546, Rev.
2.
ITEM 2b Regarding Locd Case 6, Accombly Drop Impact Loci:
(i)
What is the bacia for assuming a drop height of 89 inches?
DPC RESPONSE,:
The fuel assembly drop height of 89 inches represents the maximum vertical distance between tne top of the fuel storage rack and the bottom of the fuel assembly when it is lifted to the highest pos-sible elevation permitted by the vertical travel limit on tne fuel assembly grapple.
The fuel assembly drop height of 89 inches is quite conservative, since under normal fuel handling operation the fuel assembly will be approximately 36 inches above the top of the fuel storage rack.
ITEM 2b (ii)
The assembly drop impact load ehould be combined uith D+L vith the resultant groes stresa level less than 1.5S.
Local stresees may be greater in accordance uith the acceptance criteria stated on page 6-1.
DPC RESPONSE:
The effects of the fuel assembly impact load on the overall struc-tural integrity of the storage rack structure have been evaluated conservatively.
Conservative assumptions have been made with regard to the drop height, energy absorption, impact load applica-tion and distribution in the structure.
For example, in the fuel assembly drop analysis, a drop height of 89 inches has been used rather than 36 inches and it has been assumed that the fuel assembly does not absorb any kinetic energy.
In the evaluation of rack base structure, a concentrated impact load has been applied at the center of the grid intersection point rather than a distributed load over the grid structure, and the full impact load has been transmitted to one rack foot instead of being distributed to adjacent feet.
2199 260 t
i.
The impact load generated du.-ing a fuel assembly drop event is considerably larger than th. %L loads.
The maximum stresses in the storage rack and its support feet structure due to D+L loads are small.
(Maximum stress of 4.52 ksi in the rack base structure) compared with the stresses resulting from the fuel assembly drop (33.1 ksi in rack base structure).
In view of the conservative assumptions which have been made in the fuel assembly drop analysis, the relatively small effects of I'+L loads were not explicitly considered in the fuel drop analysis presented in the NES report 81A0546, Rev. 2.
The overall structural integrity of the storage rack structure has been evaluated by comparing the gross stress level in the rack structural members to the allowable value which is equal to the dynamic yield stress (Fyd).
The dynamic yield stress value (Fyd) is equal to 1.5S when S is calculated using the dynamic yield stress value (S = 0. 66 Fyd).
The use of the dynamic yield stress value is justifiable due to the short duration of the impact load.
ITEM 2.b (iii)
For drop possibility (3), atraight drop through a storage cett, it is stated that the fuel assembly uitt elightly perforate th~
ce t t bacc plate.
It is also stated that since the kinetic ene gy uitt be absorbed by bending of the base plate, ehearing of the base plate veld, and deformation of the rac.. base structure, the reaction toad transmitted to the rack base structure, rack feet an.! pool floor is tese than that for the fuet assembly drop on top of the s torage cett.
An analysis was then performed for a free fait of 9.0 inchee.
Justify your statement on page E-1?
that the entire kinetic energy of the falling fuel assembly is absorbed in chearing the veide and deforming the base plate.
Ex-plain uhy the remaining kinetic energy is sxpended deforming the L 'ee plate, i.e.,
justify the derivation of 6 as shoun on page E-17.
dicate the actual thicknees of a cett base plate.
DPC RESPONSE:
For a fuel assembly drop through the storage cell (drop possibility 3) the objectives of the analysis were to determine 1) whether the fuel assembly will perforate the cell base plate and impact the pool floor liner plate, 2) the extent of local deformation of the
- impacted region and the maximum reaction load 2 hat will be generated during the impact, and 3) the effect of the maximum reaction load on the overall structural integrity of the storage rack.
During a fuel assembly drop through the storage cell, the fuel assembly will first impact the cell base plate.
The external kinetic energy will be absorbed in varichs resisting elements 2199 261
1 a
essentially acting in series by the process of flexural and shear deformation.
The resisting elements consist ei the cell base plate and its weld to the rack grid structure. the rack base s tructt.re, the rack feet and the pool floor co ncrete directly under the rack feet.
In this series of the resisting elements, the load deformation characteristics of the cell base plate and its weld to the rack grid structure are significantly lower than those of the other elements.
Therefore, during a fuel assembly drop through the storage cell (either at the most flexible location of the rack or at the rack support feet location) a sig-nificant portion of the external kinetic energy will be absorbed in the deformation of the cell base plate and its weld to the rack grid structure.
The analysis indicates that the thick ness of the cell base plate is greater than that required for its perforation.
The external kinetic energy is also less than that required to completely fail the weld between cell base plate and the rack base grid structure.
Consequently it has been concluded that the fuel assembly will not be able to penetrate the cell base plate and will not strike the pool floor liner plate.
In order to conservatively evaluate the effect on the pool liner plate, it has been assumed that the cell base plate to grid structure weld fails (when the cell base plate and its weld absorb all of the external kinetic energy) and the fuel assembly falls freely on the liner plate from a height of 9.0 inches (distance between the cell base plate and pool liner plate).
Analyses (performed using different methcds) indicate that the load carrying capacity of the cell base plate and its weld to the rack grid structure is in the order of 60 kips.
Therefore, the maximum reaction load that will be generated during the *lel assem-bly drop through the storage cell (drop possibility 3) is in the order of 60 kips, a value significantly less than the reaction load developed during the fuel assembly drop on top of the storage cell (94.6 kips).
The overall structural integrity of the storage rack was evaluated by applying the entire 94.6 kips reaction load at two locations:
the mid-point of the rack grid which is the most flexible location in the rack and directly over a rack support foot which is thu least flexible location in the rack.
Since the re-action load is less for the fuel assembly drop through a storage cell, the effects of the drop on the rack base structure, rack feet and pool floor will be less severe than that for the fuel assembly drop on top of the storage rack at either the most flexible rack location or least flexible location.
The derivation of 6 (as shown on page E-17) is related to the internal strain energy of the flexural deformation of the cell base plate as indicated on page E-15 of the NES report 81A0546, Rev. 2.
The thickness of the cell base plate is 0.5 inches.
2199 262 4.
ITEM 2.b (iv)
Drop possibility (3) straight drop through the storage cett, should be examined for (1) a drop at the most flexible tocation of the rack, and (2) a drop over one of the support feet.
In the latter case, the reaction load uilt be almost entirely trane-mitted through the one support foot and could be the limi ting case for bearing sizing and punching ahear strece applied to the pool floor.
Piecce provide the resulte of your examination of these tuo drop locations.
DPC RESPONSE:
Refer to answer given for Item 2.b (iii).
ITEM Rb (v)
Atao, analyses for drop possibility (3) should be done for drope assuming the fuel assembly support plate in the upper egg-crate grid of the touer tier rack is in place.
If this plate is perfor-ated, the fatting fuel assembly could strike the assembly stored in the louer tier rack section.
Picaec provide analysis for this case.
DPC RESPONSE:
The upper tier fuel assembly support
. ate is supported by the sup-port bars attached to the upper grid of the lower tier rack (NES DWG 80E1775).
The load carrying capacity of the upper tiar fuel assembly support plate is lower than that of the cell base plate.
During a fuel assembly drop through the storage cell and impact on top of the support plate, the support plate will be damaged and the weld between the support bars and the grid structure will probably shear off.
The support plate and the fuel assembly will then fall freely'for about four inches and strike the fuel assembly stored in the lower tier.
The fuel assembly stored in the lower tier will be damaged.
However, the reaction load generated during the initial impact at the support plate will be lower than that for the fuel assembly drop on top of the storage rack.
Therefore, the adverse effects on the rack structure, rack base structure, rack feet and the pool floor will be less severe for this case than that for the fuel assembly drop on top of the storage rack.
ITEM ~2b (vi)
Appendix E cites reference 1 for increase of dyncmic yield stress above static for staintese steel.
State uhat this reference is since it is not tieted on page E-19 vith the other references. 2199 263
DPC RESPONSE:
Reference 1 cited in Appendix E is given on page E-1 of Appendix E.
This teference, NES 81A046, P.ev.
1,
" Final Structural Design of a Fuel Storage Well Crash Pad for the LACBWR Nuclear Power Plant", March 31, 1976, gives equations relating stress / strain and energy in the plastic range and the dynamic stress / strain.est data for stainless steel.
ITEM 3 Page B-11 of the cask drop analysis, N.'S 81 A 05 5 0, Rev.
2, sub-mitted on September 25, indicates that the support lege can austain a maximum load of 72.72 Kipe.
Houever, page E-6 of docunent NES 81A0546, Rev.
2, submitted August 7, shove the same leg trane-mitting a reaction load of 95 Kipe.
Clarify this apparent incon-sietency.
Also, page E-7 shove a jackacreu compressive thread area of 1.405 in2 vereue the 1.757 in2 shoun on page S-11.
Please clarify this difference.
Whert is the 6" diameter base plate re-ferred to on page E-8 tocated?
Why is the thickness of 8.S" bear-ing plate multiplied by 2 in the punching strees calculation on page B-11 and not on page E-8:
Why doce acceptance criteria for concrete floor bearing strees different on pages E-8 and B-11?
What is reference 7 cited on ; age E-87 What is reference S cited on page B-11?
DPC RESPONSE:
The structural calculations given on pages B-11 and B-12 of NES report 81A0550, Rev. 2 and these given on pages E-6 through E-8 on NES Report 81A0546, Rev. 2 are being revised to correct the inconsistencies mentioned abovo and to reflect design changes which have been made to eliminate mechanical interferences between adjacent racks.
Reference 7 cited on page B-8 of NES Report 81A0546, Rev. 2 and Reference 5 cited on page B-ll 7f NES Report 81A0550, Rev.
2, is ACI318~71, " Building Code Requirements for Reinforced Concrete", American Concrete Institute.
ITEM 4 It is 3tated on page 5-2 of your.*.ugust 7 submittal, NES 81A 0546, Rev.
2, that thermal loadings are insignificant because of the clearances provided to accommodate a maximum pool temperature of 150 F.
State the maximum normal and accidental pool temperatures 0
that can be expect ~d and uhether rack expansion is provided for under the maximum accident temperature that the pool vater could reach.
Provide justification for not considering these clearances.
Although rack expansion may be provided for, indicate the maximum thermal gradient that can develop betueen adjacent fuel storage locatione, the magnitude of the reauttant etressee and justify your statement that these atreesce are not significant.
DPC RESPONSE:
The maximum normal operating and accident bulk pool temperatures are 120 F and 150 F respectively.
Clearances will be provided between the storage racks, seismic bracing and fuel pool walls at the upper, intermedirJe and lower grid levels to allow for the expected maximum thermal expansion of the stainless steel rack structures without interference.
The total thermal expansion at each bracing elevation will be calculated at the time of install-ation in the fuel pool based on overall pool dimensions, actual fuel pool temperature, an assumed maximum accident bulk pool temperature and appropriate stainless steel thermal expansions coefficients as given in Table I-5.0 of Appendix I, ASME Boiler and Pressure Vessel Code,Section III, 1977 edition.
The total calculated clearcree across the pool length and width will be distributed approximately equally among the various inter-rack and rack-to-wall clearance locations.
Since these clearance / gaps are provided to accommodate the thermal expansion effect only, impactive loadings due to these gaps have not been considered in the seismic analysis.
This is consistent with the staff position stated in Enclosure 1 of NRC letter October 4, 1977 on Docket No.
50-298.
Differential heating produced by full cells being adjacent to empty cells has small effect on the fuel storr.ge racks.
The rack base is totally unaffected by such differential heating since the base remains at the temperature of.ne pool water enter-ing the rack.
Differential heating, however-has the potential to produce different axial (Vertical) expansions of the fuel stor-age cells.
Based on the maximum temperature rise expected in a full storage cell (see NES Report 81A0548, REv. 2) the full cell potentially could have a length no more than 0.012 inches greater than an adjacent empty storage cell.
This small differential ex-pansion will result in axial stress of 3.4 ksi in the storage cell.
This small axial stress and the resulting axial load can easily be accommodated by the storage cell, storage cell welds and the storage rack grid members.
ITEM S With reference to question 4, discuse the necessity of evaluating load combinatione not addressed in the submittal, i.e.,
load combin-ations b.
(i)(e), b.
(i) ( 7 ), and b.
(i) (8) on page 3.8.4-8 of the Standard Revieu Plan.
These factored load conditione include loada due to cocident temperature effecte.
DPC RESPONSE:
As stated in answer to question 4, small clearances will be pro-vided to accommodate the thermal expansion for the maximum pool temperature condition.
Therefore, load combinations b.
(i) (6),
b.
'i) (7) & (8) on page 3.8.4-8 of the Standard Review Plan are not applicable. r 2199 263
ITEM 6 Indica te uhere the yictd stress, 30.0 kei, for staintese steel is taken from and at uhat temperature.
DPC RESPONSE:
The minimum yield strength value for type 304 stainless steel was obtained from Table I-2.2 of ASME Code Section III - " Rule for Construction of Nuclear Power Plant Components".
The stress value of 30.0 ksi represnets the minimum yield strength at the normal 0
operating temperature condition
(<
100 F) in the spent fuel pool.
ITEM 7 It is stated on page 3-2 of NES 81A0546, Rev.
2, submitted August 7, that the fuel storage racks and associatei aeismic bracing are fabricated at Type 304 s tainlece atec t.
- Houever, on page 8-12, there te a note referencing 17-4 PH etcintese steel.
Indicate uhether this material is being utilized.
If s o, state uhere it is being used, uhere the yield stress is taken from and at uhat temperature.
Alco provide the heat treatment temperature, and specify that the pieces uitt be hardness tested to verify heat treatment and either pickled or gri)? blasted to remove the surface film reauttirg from the heat treatment.
DPC RESPONSE:
17-4 PH stainless steel will be used for the 1 " diameter threaded rod and associated spherical washer in each fuel rack support leg.
The design yield tress is taken from the "ASME Boiler and Pressure Vessel Code", ANSI /ASME BPV-III-I-A, 1977 and is conservatively assumed to be 106.3 ksi at a maxinum design temperature of 100 F.
The fabrication specification for the fuel racks specifies a heat treatment temperaturo for 0
the 17-4 PH material of 1100 F (593 C) and requires hardness testing to verify a minimum hardness of 32 Rockwell C or 311 Brinell after heat treatment.
The specification also requires pickling of the heat treated components to remove surface film resulting from the heat treatment.
ITEM B If any materiate other than type 304 of 17-4 PH etainlece steele are being used, tivt the materiale along uith their yield strength, uhere the yield strength is taken from and at uhat temperature.
DPC RESPONSE:
No other structural materials are being used for the LACBWR fuel racks.
2199 266
ITEM 9 Drovide the basis for the acceptance criteria, O.SFy (shcar) and 0.'"y, given on page 6-1 of NES 81AOS46, Rev.
2.
DPC RESPONSE:
Load combinations 4 and 5 of NES 81A0546, Rev.
2, represent the applicable factored load conditions specified in the NRD Standard Review Plan, Section 3.8.4.
The Standard Review Plan defines an allowable limit of 1.6 S to constitute the structural acceptance criteria for these load combinations.
In certain situations, this acceptance criteria develops an allowable stress limit greater than the minimum yield strength of the material.
Therefore, NES conservatively assumes an upper bound of 90% of yield for the allowable bending stress criteria and 50% of yield for the shear stress criteria.
The shear stress criteria is determined assuming shear failure to occur when the shear stresses become equal to 1/2 the minimum tensile strength of the material.
ITEM 10 Provide the acceptance criteria for veld stresses.
DPC RESPONSE:
The following allowable stress limits constitute the acceptance criteria for weld stresses used for each of the loading combin-ations presented in Section 5.2 of NES 81A0546, Rev. 2:
LOAD COMBINATIONS LIMIT 1,2,3 S
la, 3a 1.5S 4,5 1.6S Where S is the permissible stress limit defined in the AISC " Spec-ification for the Design, Fabrication and Erection of Structural Steel for Buildings", February 12, 1969 (Table 1.5.2.1).
The permissible stress limits of Table 1.5.2.1 are reduced by the pro-portion of the yield stress value Fy for stainless steel (30.0 ksi) to the yield stress value Fy for A36 carbon steel (36.0 ksi).
ITEM 11 Provide the cater chemistry uhich uilt be maintained in the apent fuel pool.
Include the boron concentration, p H,
- chloride, fluoride and any heavy metal concentratione.
DPC RESPONSE:
The following water chemistry limits are specified for the LACBWR spent fuel pool:
pH 4.5-8 Chlorides (Volhard Method)
< 0. 5 ppm Conductivity 10 pmhos/cm max.
Typical water chemistry analysis results for the fuel pool water are pH 5-6, chlorides < 0.02 ppm and conductivity 1-2 umhos/cm.
Doron is not used in the fuel cool or reactor refueling cavity and therefore no limits are specified.
Analysis for heavy metal concentrations is not performed.
ITEM 12 Discues hou the effects of fuet assembly " rattling" are accounted
- for, i.e.,
indicate if the generated loads and reauttant stresses are added to the other stresses in att the combinations that include seismic t oads, uhether att storage cette are accumed to contain fuel and if att the fuel assemblies are assumed to move in phase.
A ta o, indicate the groce and locat stressee in the racke due to this
" rattling" and demonstrate that the fuel assemblies themselves uitt retain their structural integrity and uiti not suffer cladding damage as a result of impacting the storage ceit.
DPC RESPONSE:
Clearances are provided between the fuel assembly and the storage cell to avoid interferences during fuel storage and removal oper-ations.
The storage cell / fuel assembly clearance (or gap) results in the " rattling" fuel assembly impacting the storage cell during a seismic event.
The LACBWR fuel storage racks have been analyzed using the linear response spectrum modal superposition method of dynamic analysis with the effect of impacting masses conservatively accounted for by imposing the following assumptions:
1)
All storage cells contain the fuel assemblies 2)
All fuel assemblies simultaneously impact the storage cells 2199 268 3)
Adjacent racks move in phase 4)
The effect of fuel assembly impact is a two-fold increase in the seismic intertia loadings produced by the impacting fuel assemblies mass.
The inpact and seismic inertia loads of the impacting masses are added to the seismic inertia loads of the non-impacting masses.
The generated loads and resultant stresses due to the impact and seismic inertia loads are added to the stresses resulting from other loads as given in the load combination (NES 81A0546, Rev.
2, Section 5.0).
The maximum stresses in the storage cell and storage rack structure due to the fuel assembly impact (rattling), seismic loads and various other loads combined in accordance with the Standard Review Plan 3.8.4 are given in Table 8.2 of NES Report 81A0546, Rev.
2.
During a seismic event the fuel assembly will be subjected to a maximum impact load of 936 pounds.
Based on the fact that the load carrying capacity of similarly configured fuel assemblies is three to four times greater than the maximum impact load value (936 pounds), it is reasonable to assume that the LACBWR fur &
assemblies will accommodate the maximum impact load without suffer-ing any damage.
ITEM 13 Provide details of the pool modification phases and indicate uhe ther all racka vitt be seismically supported during att phases.
DPC RESPONSE:
Installation plans for the fuel racks are contained in Atinchment "A"
to DPC letter LAC-5498, Madgett to Director of Nuclear Reactor Regulation dated October 16, 1978.
The racks cannot be seismically supported during all phases of the modification.
However, the installation will be completed in a timely manner to minimize the time during which the racks will be in a configuration not analyzed for a seismic event.
Steps which will be taken to minimize install-ation time include pceparation of detailed work procedure (s),
advance planning and scheduling of necessary manpower and equipment and advance design and fabrication of special tooling.
Additionally, installation of the new racks will not commence until all rack sections and associated corponents are on-site and ready for in-sta11ation. 2199 269
ITEM 14 The structural analysis report, NES 81A 0095, Rev.
1, submitted October 4, does not include all the load combinations found on page 3. 8. 4-7 of ths Standard Revieu Plan.
Specifically, provide justification for not considering load combinatione alii) (2b '),
a.(ii) (3b') and b.(4) through b.(8).
Also indicate uhether both casca of L having its full value or being completely aboent vere checkec DPC RESPONSE:
In evaluating the load combinations for concrete structures in accordance with Section a of the Standard Review Plan, load combin-ation a. (ii)- (2) {l.4D + 1.7L + 1.9E} is considered to be more critical than the load combination a (ii) (2b') {l.2D + 1.9E}.
Therefore, load combination a(ii) (2) has been examined as load combination (2) in NES Report 81A0095, Rev. 1.
Load combinations
- a. (ii) (3b') {1.2D + 1.7w) is not applicable since there are no wind loadings on the pool structure.
Factored load combinations b. (4) and b. (7) (Section b of Standard Review Plan) were examined as load combination (2) and (5) in NES 81A0095, Rev.
1.
In the analysis, a linear thermal gradient of 0
80 F across the pool walls was considered to be applicable for both the normal and accident conditions.
Factored load combination b(5) is not applicable since there are no wind loadings on the pool structure.
Load combination b (6) is less critical than load combination b (4).
Load combination b(8) was not considered since the probability of a cask drop accident occurring simultaneously with a safe shutdown earthquake is considered unlikely.
A review of the results for load combination (5) of NES 81A0095, Rev. 1, indicate that the design of the pool structure is adequate to withstand the loadings of factored load combination b(8).
ITFM 15 Load combination (5), given on pages S-2 and 8-7 of NES 81A0096, Rev.
1, does not agree.
Indicate which combination was examined.
DPC RESPONSE:
Load combination (5) given in Section 5.2 (page 5-2) of NES 81A0095, Rev. 1, should be corrected to read (D + L + 1. 2 5E + I. L. ) as indi-cated on page 8-7 of the same report. 2199 270
ITEM 16 Indicate uhere the value for the compreceive strength of ocncrete vae obtained?
DPC RESPONSE:
The compressive strength of concrete used in the evaluation of the spent fuel pool structure to withstand the revised floor and wall loadings associated with the high density fuel storage rack has been taken as 3500 psi at 28 days.
This minimum concrete strength was obtained from the following Specification:
1)
" Building Work - La Crosse Boiling Water Reactor -
Reactor Plant", Specification No. AC 41-561, including S&L No. W-1743, Rev. 15, August 1966, and 2)
" Specification for Concrete Chimney - La Crosse Boiling Water Reactor - Reactor Plant", Specification AC 41-560 including SYL No. W-1742, Standard Specification form 1715-P, May 1963.
ITEM 17 Since the uater levet in the spent fuel pool must nov be main-tained above the refueling canal gate, indicate whether gate seat integrity can be maintained under seismic conditions. If not, discuse the consequences.
DPC RESPONSE:
The refueling canal gate isolates the reactor cavity from the spent fuel storage pool.
The canal gate seal is glued and bolted to the canal gate.
After the canal gate is installed, the sealing gasket is compressed to produce the required leak-tightness by tightening the canal gate bolts.
Additional compression of the gate snal due to the hydrostatic pressure of water and the hydro-dynamic pressures generated by the lateral seismic inertia loads and water sloshing effects of pool water will increase the effective-ness of the seal.
Therefore, seismic conditions will only improve the leak tight integrity of the gate seal.
ITEM 18 The loading due to uater closhing during a seismic event must be included in the analysie of the pool structure.
Indicate hou these loads have been accounted for.
DPC RESPONSE:
The hydrodynamic loading due to water sloshing effect during a seismic event is in the order of 6.9 kips applied locally to the upper 5 feet of the pool walls.
The water sloshing load of 6.9 kips is small compared with the total seismic inertia load of the constrained water mass (117.2 kips).
Furthermore, the dimensions of the pool structure (11' x 11' x 40') and its con-struction (pool walls supported by adjoining walls of the shield building) are such that the water sloshing loads will be trans-ferred directly to the adjoining walls of the shield building.
In addition the water sloshing loads will have insignificant effect on the sections of the pool walls which are subjected to the seismic effects of the storage racks and the constrained water mass (including the hydrostatic pressure loads)since these sections are located well below the sloshing water mass region.
ITEM 19 Based on discussions uith your staff it is our understanding that the spent fuel pool drain line is not Seismic Category I piping.
Therefore, unless modifications are made, e.g.,
rerouting the line and permanently seating the drain or seismicatty qualifying the piping, it must be accumed the piping faite during an earth-q ua ke.
Please provide detailed drawings of your planned modifi-cations and associated seismic analyses.
DPC RESPONSE:
In order to eliminate the possibility of loss of spent fuel storage pool cooling water due to a break in the fuel storage pool drain line, tua isolation check valvaswill be installed approx-imately 12 feet downstream of the storage pool drain connection.
The check valves serve to isolate the pool drain line from any possible adverse effects downstream of the valve.
The rigid anchor (approximately 7 feet downstream of check valve) at the wall penetration in the forced circulchion pump cubicle concrete wall serves to seismically decouple this portion of the drainline from the remainder of the piping system.
The drain line between the pool drain connection and the concrete wall penetration in-cluding the isolation check valve has been analyzed as a Seismic Category I System.
Drawings of the proposed modification and the results of the seismic and stress analysis performed in accordance with the design requirements for Class 1 piping components of the ASME Boiler and Pressure Vessel Code Section III, Division I,
" Nuclear Power Plant Components", 1974 are given in NES report 81A0032. 2199 272
ITEM 20 Scotion G of the Caek Drop Analysia, NES 81A 05 50, Rev.
2, submitted September 25, presente the structural acceptance criteria for the three toad cases c:amined.
If U is the strees limit for the concrate, explain the remaining limite given on page 6-1.
Also clarify the acceptance criteria for Load Case 3.
DPC RESPONSE:
The structural acceptance criteria limit of 1.6S presented in Section 6 of NES 81A0550, Rev. 2, is in accordance with the acceptance criteria given in Standard Review Plan Section 3.8.4 for steel structure subjected to factored load conditions.
The stress limit values of.5Fy for shear stress and.9Fy for tensile or compressive stress are explained in the NES respense to Question 9.
The ultimate strain limit of 100% represents the allowable strain limit in the tensile energy absorbing modules of the crash pad.
The acceptance criteria for load case 3 is that the cask drop accident event will not result in the total collapse of the storage rack so as to adversely affect the value of keff.
The leak-tightness integrity of the liner plate and the fuel storage pool floor should also be maintained.
ITEM 22 Discuse the possibility of the cask dropping auch that it directly impacte or is deflected onto the refueling canal gate.
If the gate may be struck by the cack, and since the poot uater levet vitt nou be above the gate, indicate whether gate and/or seat in-tagrity uitt be maintained.
If not, discues the consequences with regard to (t) spent fuel and (2) plant equipment in areaa that may roccive the leaking pool uater.
DPC RESPONSE:
The transfer gate seal is a one-piece, 3/8 inch thick x 3-3/8 inch wide 30-40 durometer rubber gasket which extends along both sides and the bottom of the canal gate.
The gasket is glued and bolted to the side of the gate opposite the fuel pool and seals between the canal gate and the canal gate housing.
The gasket is compressed by bolts which, when tightened, press the gate against the housing.
Additional gasket compression results from hydrostatic pressure as the fuel pool water level is raised above the bottom of the canal gate.
During the cask handling operations, the shipping cask will not be brought directly over the top of the canal gate.
Even if the cask were brought over the top of the canal gate and the cask drops, the difference in the elevation between the top of the pool walls (Elev. 701' 3") and the top of the canal gate (Elevation 701'-1),
the presence of the canal plug (Elevation 701'-1"), and the geonetry of the canal gate area and the cask (Figure 1) will essentially prevent a direct impact of the cask on the canal gate and the seal.
Even if, it is conservatively assumed that the cask impacts the top of the canal gate, it will cause local damage to the canal gate and the seal for the upper few feet of the canal gate may become ineffective.
However, due to the close proximity of the canal plug and the canal gate, it is unlikely that the latter will undergo extensive damage.
In the event of a cask drop on top of the bolt (which compress the gate against the housing to make the seal effective), some of the bolts will be damaged and the cask will then tip over.
It is unlikely that most of the bolts will damage.
The effectiveness of the seal at the location of the damaged bolts will be maintained by the hydrostatic pressure on the canal gate.
It it is conserv-atively assumed that the hydrostatic pressure is not sufficient enough to maintain the full seal, then some of the water will be lost by slow leakage in the damaged area of the bolts.
If the cask drops such that it is deflected onto the refueling canal gate, it will damage the impacted bolts and/or liner plate; however, the layout of the canal gate and pool liner plate / concrete wall is (Figure 1) such that the 50.5 inches diameter cask (Vandenburg cask) will not be able to impact the canal gate.
Therefore the leaktight integrity of the canal gate will be main-tained.
Even if it is assumed that the diameter of the cask is less than 50.5 inches, a light impact of the cask onto the canal gate will increase the compression on the gasket, thereby increas-ing its leaktight effectiveness.
A hard impact of the cask onto the canal gate will deform the canal gate, but the canal plug will prevent the unseating of the gasket / canal gate from its support points.
Therefore, no gross failure or leakage will occur.
It can, therefore, be concluded that the cask drop event onto the canal gate may cause local damage to the canal gate and could re-sult in slow leakage.
However, the overall structural integrity of the canal gate will be maintained.
Any minor leakage into the upper vessel cavity area will be handled by the upper cavity drains and no plant equipment will be damaged.
ITEM 22 It appeare that none of the caek drop analysee performed examined the possibility of striking along an edge or corner of the caek.
Thcee types of drope may lead to more severe damage of the crash pad, poot liner and floor, or racks.
Provide analysea considering these types of drope or a detailed ductification for not doing so.
DPC RESPONSE:
Due to the limited size of the pool, the presence of fuel stturage racks and seismic bracing around the crash pad / pool, and the uniform geometry and even distribution of cask mass throughout its height, it is very unlikely that the cask will drop with its axis significantly off vertical, thareby striking the crash pad 2}99 27k
[1]
M R
M T
T ii
- R 8
b
%m
.h
'L y
Q X
~
', h b
\\1 t
I i
. x lo 7
. g 0
3
~Jt o
W. y Q
4 gin i
y
+
3 1
3*
- p, e
1 3
e 3
M:3 m
.< y t
N
- k N
tu g
0$
i.
=
.x i.
.2 c
TYr 48 b8 Ni
&!4 L -7 ls
}Q' 1
~
'i df-2199 275 l'
FIGURE 1 si 16a -
11 4
or the fuel storage racks along its edge.
In the original design of the crash pad (NES Report 81A0426, Rev.
1,
" Final Structural Design of a Fuel Storage Well Crash Pad for the LACBWR Nuclear Power Plant" (March 31, 1976), cask drop analyses were performed considering the possibility of striking along an edge or corner of the cask.
These analyses indicated the following:
a.
If the cask strikes the crash pad at an angle, its velocity of strike will be less than for a vertical fall of the cask.
This is because the drag forces opposing the motion of the cask will be larger for a cask falling at an angle than for a vertical fall.
b.
If the center of gravity of the cask hitting the crash pad at an angle does not lie in the vertical plane passing through the point of strike, the cask will topple over as soon as it strikes the crash pad and the reaction transmitted by the crash pad to the floor will be less than for a vertical fall of the cask.
While toppling over, the cask will hit the fuel racks or the pool wall.
The local deformation of the crash pad and the cask will be more for a cask striking at its edge but the reaction loads transmitted to the floor will be less.
c.
If the cask strikes the crash pad at an angle such that the center of gravity of the cask and the point of strike are in the same vertical plane, it will cause more local damage and will have a tendency to penetrate through the crash pad and at the same time bring itself to a vertical position or fall on the side.
It is difficult to imagine that throughout the penetration into the crash pad, the point of strike and the C.G. of the cask will remain in the same vertical plane.
The number of impacted tensile modules will be less while the local deformation of the crash pad will be greater, the maximum reaction loads transmitted to the pool floor will be less.
On the basis of these results and the consideration that the overall structural integrity of the storage rack and the leak tightness integrity of the pool liner plate depends primarily on the maximum reaction load that is generated during the impact, the cask drop analyses for the revised crash pad (NES Report 81A0550, Rev. 1) were limited to cask drops in which the cask axis was vertical.
ITEM 23 For the 3/8" staintese eteel barrier plate that is to be provided under the otorage racke, clarify the locatione in the poot uhere 2199 276
the plate vill be provided, if the plate is onc piccc or acueral smaller plates, and the exact ma terial of the plate.
DPC RESPONSE As mentioned in NES 81A0550, Rev.
2, for ease in installation, the 3/8" stainless steel barrier plate could be made up of a number of smaller size (2' x 2',
etc.) plates.
These plates will be placed under all of the storage rack structures and will cover a signifi-cant part of the projected area of the storage racks to eliminate the possibility of the storage rack bottom grid impacting the liner plate, The detail design and spacing of the barrier plates have not been presently established.
The barrier plate will be fabricated from Type 304 stainless steel material.
- TEM 24 Justify the assumption that only 52 otorage colla can be impacted by a dropped cask, since the cask vill not necessarily strike the rack in a vertical position.
DPC RESPONSE For a cask dropping with its axis vertical, the 52 impacted storage cells represent the tributory number of storage cells under the projected (impacted) area of the cask.
For the case of a cask dropping with its axis off-vertical, the number of storage cells that will be impacted simultaneously will be less that 52 (N27 cells).
While the resulting local damage to these impacted storage cells will be greater, the maximum reaction load that will be generated for this case will be smaller.
Therefore, the resulting damage to the overall rack structure and the pool floor liner plate will be smaller for this case than that for the case of a cask dropping with its axis vertical.
ITEM 2S The cash drop analysia, NES BlA OSSO, Rev.
2, submitted September 25, references NES document 8tA0426, Rev.
2, dated March 31, l976.
Was this document ever submitted to the NRC and, if 00, uhen?
DPC RESPONSE Revision 0 of the above document was submitted to the NRC by DPC letter, Madgett to Goller, LAC-3482, dated November 5, 1975.
We have no record of submitting Revision 1 and therefore are submitting 15 copies of the referenced document with this submittal.
2199 277 ITEM 2C In A ttachment "A " to DPC le t ter LA C-5 49 8, dated October 26, 1979, it is stated that the poison material celected for uce in the neu racks is a B C Composite manufactured by the Carborundum Company.
4 Please provide the follouing information:
temperature of the B C Composite material?
a)
What is the melting 4
DPC RESPONSE The B C Composite material can operate continuously at temperatures 4
up to 350 P.
While the materials within the Composite system do 0
not melt, the most temperature sensitive component begins convert-ing to carbon at 4500F.
During residence in the spent fuel pool, the maximum temperature 0
at which the Composite material will operate is well below 350 F (see Response to 26c).
During fabrication, the fuel storage cell will be assembled by welding with the Composite material in place.
Consequently, the storage cell has been designed to preclude the Composite material from reaching an excassive temperature during 0
fabrication (more than 450 F).
This is accomplished by:
1) providing a suitable distance between the weld regions and the boron containing material 2) providing significant heat sinks in the vicinity of the welds (e.g. storage cell angles), and 3) selecting weld processes which minimize the heat input into the weld region (e.g. resistance weld. )
ITEM 28b What vill the maximum integrated neutron and gamma flux be in the boron containing material over the life time of the rache?
What spent fuet assembly power density and burnup, and uhat rack life vere assumed in calculating these maximum integrated fluxes?
DPC RESPONSE An analysis was performed to establish the maximum integrated neutron and gamma dose to which the B C Composite material would 4
be subjected over the assumed lifetime of the racks (25 years).
The calculations were performed using the ANISN computer program which modeled an infinite array of LACBWR fuel assemblies using cylindrical geometry.
The following parameters were assumed for the fuel assembly array:
19 -
2199 278
Fuel aszembly power density 1.63 Int / assembly (average for 3 years)
Fuel assembly burnup 16,500 MWD /MTU (for 3 years)
Gamma spectrum Used averaged gamma energy of 0.8 MEV Integrated exposure time From 1 second up to 25 years The results of the analysis indicate that the contribution of neutron irradiation to the total dose received by the Composite material is negligible compared with the dose resulting from gamma irradiation.
The following gamma irradiation data summarizes the results of the analysis for the principal exposure times:
Dese at 1 year exposure 3.2 x 109 Rad Time required to reach
>25 years 10 Rad Dose at 25 years exposure 8.4 x 109 Rad The short-term phase of the Carborundum qualification test program has exposed the Composite material to 1010 Rad gamma with the results showing that the material retains acceptable properties (see Response 26f).
The long-term phase is scheduled to expose the material to 10 11 Rad gamma.
As can be seen, the results of the short-term program exceed the exposure anticipated for the life of the fuel racks and the long-term program irradiation will greatly exceed the anticipated life exposure (10 11 Rad vs. 8.4 x 109 Rad).
The listed exposure values are predicated on the reason-able assumption that a fuel assembly placed in a storage cell will remain in that position until it is shipped from the site.
A more conservative exposure basis is to assume that fuel assemblies will be removed from storage locations every five (5) years during the 25 year life of the rack and replaced with freshly discharged assemblies.
This 5-year period is concistent with the cooling times that are being mentioned as prerequisites for shipment to Away From Reactor (AFR) Storage Facilities.
Using the assumption that a freshly discharged fuel assembly is placed in every storage location every 5 years, the resulting total exposure of the Compos-ite material is calculated to be 2.9 x 10 10 Rad for the 25-year rack life.
This value is substantially less than the total exposure scheduled for the long-term phase of the Carborundum qualification test program (10 11 Rad).
The most conservative exposure basis for any boron-containing material is to assume that spent fuel assemblies will be removed from the storage locations every refueling outage and replaced with 2199 279
_ 20 _
freshly discharged assemblies.
This exposure basis can occur consistently only if the current LACBWR spent fuel handling procedures are modified to force the yearly replacement.
A spent fuel handling procedure which tends to maximize fuel handling will not be permitted by DPC because it is contrary to DPC's objectives to 1) reduce the possibility of fuel damage during handling and, therefore, the possible release of radio-active material to the pool environment and 2) reduce the exposure of plant personnel to the radiation environment existing at the surface of the spent fuel pool during fuel handling. Con-sequently, DPC concludes that the application of this exposure basis to the Composite material is not justifiable.
Even if it is assumed that an accidental replacement of spent fuel will occur on a yearly basis, the composite material would still not reach an 11 Rads during the 25-year rack life.
exposure limit of 10 ITEM 2Ge What vitt the maximum temperature be in the center of the boron material, accuming the highest neutron and gamma flux and the vorst accident conditions?
DPC RESPONSE:
The interior temperature of the B C Composite is determined by 4
the temperature rise in the material due to gamma density (the neutron flux levels are insignificant and produce negligible heating) and the temperature of the spent fuel pool water which is in contact with the storage cell walls.
The maximum temperature rise in the Composite material occurs when freshly discharged fuel is placed in the storage cell '(and the adjacent storage cells).
The maximum dose rate associated with 6 Rad /hr. using freshly discharged fuel has been calculated to be 10 the calculational model described in Response 26b and assuming 5 days of cooling.
The temperature rise for this dose rate will be less than 2 F.
This value is based on temperature test data ob-tained in the Carborundum qualification test progran (see figure on page E-2 of Reference 1).
0 The 2 F temperature is conservative for LACBWR since the test data was obtained for the relatively thick D4C Plate material (0.210 inches thick) placed dry in the stainless steel sample holders and in casual contact with the bottle walls while the thinner Composite material in the LACBWR racks is immersed in the pool water and is in intimate contact with the storage cell walls (0.095 inches thick).
The maximum temperature of spent fuel pool water in contact with the storage cell walls is primarily a function of the maximum bulk water temperature of the LACBWR spent fuel pool (150 F).
Thermal hydraulic analyses performed for the LACBWR fuel racks indicate a maximum temperature rise (<within the two-tier storage cells) of 2199 280
- 21 _
0 25.8 F.
Consequently the maximum temperature of pool water con-tacting the storage cell walls is 175.8 F.
Combining the maximum temperature rise (2 F) and the maximum pool water temperature (175.8 F) results in a maximum temperature of 178 F for the Composite material.
This temperature is well below 0
s the gteady state operating temperature of the Composite material (350 F).
ITEH 26d What will the chemical composition of the boron containing material be af.se receiving the design dose of irradiation?
DPC RESPONSE:
The chemical coruposition of the B C Composite materiel af ter 4
10 Rad gamma is analytically the same as the chem-exposure to 10 ical composition prior to irradiation with an estimat:d 0.2 w/4 loss of hydrogen.
It is expected that the chemical composition after 10 11 Rad exposure will also be essentially the same with, however, a proportionately larger loss of hydrogen.
ITEM 28e Provide the acceptance criteria for mechanical strength of the poison plates, including the basis for the criteria, and specify the minimum (or maximum) acceptable values for the modulus of rupture, modulus of elasticity, and ultimate tensile scrength.
DPC RESPONSE:
The acceptance criteria for the mechanical strength of the Composite material selected for use in the LACBWR fuel storage racks are as follows:
Tensile Strength OBE Value 0.2 UTS minimum
- Compressive Strength OBE value Use Tensile Strength Value**
SSE Value Use Tensile Strength Value**
Modulus of Elt.sticity MOE maximum ***
Modulus of Rupture This material property is not applicable to the Composite material because of its inherent flexibility.
2199 281
- where UTS minimum is the minimum Ultimate Tensile Strength value permitted by the Procurement Specification for material exposed to prototypical spent fuel pool environment including 11 Rad.
The irradiation up to and including gamma doses of 10 Procurement Specification permits the minimum UTS value to be 25% of the UTS value for dry unirradiated material.
The mini-mum UTS value for the LACBWR material will be N 2200 psi.
- The compressive strength values were not determined in the Carborundum test program for the Composite material.
- However, it is evident from the nature of the material and from com-pressive strength data available for the Plate material (which has the same binder as the Composite) that the com-pressive strength value will be superior to the tensile strength value (UTS values).
Therefore, the compressive strength acceptance criterion has been conservatively set at the tensile strength value.
- where MOE maximum is the maximum Modulus of Elasticity value permitted by the Procurement Specification for exposed mater-ial.
The Procurement Specification permits the maximum MOE value to be 10 x 10 5 psi.
The basis for these acceptance criteria is to assure that signifi-cant safety margins exist between the conservatively calculated stress values to which the material will be subjected and the minimum permitted ultimate strength values.
For the LACBWR fuel racks, the maximum bending stress in the Composite material during the SSE event is s 620 psi as compared with the acceptance criteria value of 704 psi and the minimum permitted UTS value of 2200 psi.
As can be seen, the safety margin is acceptable.
The compressive loading imposed on the Composite material by fuel ast embly impact is negligible because of the design of the fuel storage in which all fuel assembly impact loads are transmitted directly to the corner angles of the storage cells.
ITEM 2Cf Submit the results of testing uhich shou that the composite poison plates uilt retain acceptable levele of mechanicci strength, in accordance with the criteria discussed above, throughout their service lifetime uhen exposed to the maximum expected radiation doce lave t, dose rate, and pool vater environment.
DPC RESPONSE:
A principal objective of the qualification test program initiated by the Carborundum Company is to establish the mechanical and physical properties of the D C Composite material proposed 4
for the LACBWR racks as a function of the gamma irradiation dose with simultaneous exposure to prototypical pool water environments.
2199 282 evaluated for gamma irradiation doses up to 10perties have been At this point, the mechanical and physical pro 0 nads with simul-taneous immersion in demineralized (D.I.) water and in borated water (2500 PPM).
In addition the properties have been deter-I0 Rads only anc 2) exposure to mined for:
- 1) exposure to 10 deminoralized or borated water only.
A test to establish the 11 Rads mechat'ical and physical properties for exposure to 10 with simultaneous immersion in demineralized water has been initia-ted by Carborundum and is scheduled to be completed by mid-February.
As indicated in Response 26b, the available test data _
exceeds the gamma exposure anticipated for the life of the LACBWR fuel racks (10 10 Rads vs. 8.4 x 109 Rads) and that the data being 11 Rads will easily bound exposure levels assoc-developed for 10 lated with the fuel stored in each storage location periodically replaced by freshly discharged fuel at a frequency in excess of any operationally justifiable value.
Table 1 presents the summary results of the qualification test program to date.
Specifically the tabla presents the principal mechanical and physical properties evaluated during the test program and the changes in those properties that resulted from combinations of exposure to gamma radiation and prototypical pool water environments.
The following observations and conclusions can be made from Table 1:
a.
The changes in mechanical properties are more pronounced for the Composite material immersed in D.I. water than in borated water.
Consequently 11 Rads is being the long term testing to 10 carried out in D.I. water.
b.
The change in the principal mechanical properties (UTS) of the Composite material immersed in D.I./
borated water only are similar to the changes observed for the material immersed in D.I./ borated water and 10 Rads.
simultaneously exposed to 10 c.
The mechanical properties of dry Composite material irradiated to 10 10 Rad are essentially the same as the properties of the unirradiated baseline material.
10 d.
Considering items b and c, it is evident that up to 10 Rad, the decrease in mechanical property values for the Composite material simultaneously exposed to radiation and water is due primarily to immersion of the material in either the D.I. or borated water.
With respect to water immersion, the tests indicate that essentially all of the observed change in mechanical properties due to immersion occurs over the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of immersion.
After that time, the rate at which additional 2199 283 TABLE 1 COMPOSITE MATERIAL TEST DATA BASELINE WATER SHORT TERM E
S SPECIFICATION C'.MMA DATA IMMERSION TEST PROGRAM DATA DATA VALUE
~
EXPOSURE (Rad) 0 0
0 109 RADS 10 RADS 10 Id1 10 l1 WATER EXPOSURE
- DRY DI BORATED DI DORATEE DRY DI
- ORATEI' DI D'I Mechanical UTS (psi) 8838 7300 7969 8457 8160 9406 6591~
7695 4500 2200
% of Baseline 100 82.6%
90s 95.7 92.3 106.4 74.6 87.0 50.9 25 t
1 (psi x 10-5) 2.24 2.28 2.27 1,64 1.67 2.26 1.73 1.75 10 MOE y
1 of Baseline 100 101.8 101.3 73.2 74.5 100.9 77.2 78.1 1
Physical Length (m) 3.004 3.005 2.999 2.999
% Change
-0.5
+0.2 0
+0.1 Width. (m) 0.538 0.531 0.533 0.522
% chango
+0.9
+0.2 0
0 Thickness (m) 0.048 0.048 0.049 0.048 1 Chango
+2.1 0
4 0
g Y
Weight (gm) 1.757 1.773 1.767 1.733
% Change
-0.23
-0<2
-1.1
-0.3 N
CD b
property changes occur becomes nagligible so that no additional significant changes in the mechanical or physical properties due to immersion in the pool water are expected over the assumed 25 year life of the LhCBWR fuel racks.
The changes in the physical properties of the Composite e.
10 Rads, material are negligible up to 10 f.
The changes in the mechanical properties are modest up to 1010 Rad with simultaneous immersion in either D.I.
or borated water (25% or less of the baseline values).
The mechanical property values remain well above the requirements specified in the Composite material purchase specification.
The minimum cpecified UTS (Ultimate Tensile Strength) value is significantly greater than the maximum stresses that the Composite material will exper-ience during a SSE event (see Response 26e).
11 Rad with simultaneous g.
The predicted UTS value for 10 exposure to D.I. water (4500 psi) is also significantly greater than the minimum specified UTS value (2200 psi) and the maximum tensile stress that the Composite material will experience.
The predicted value was conservatively 9 and 10 10 Rad gamma developed from a comparison of the 10 radiation test data with data obtained from electron beam 11 Rad) performed pre-irradiation tests (up to 2 x 10 viously for the composite materials.
Compressive strergth values were not determined in the Carborundum test pro-gram for the Composite material.
However, it is evident from the nature of the material and from compressive strength data available for the Plate material (a boron containing material which has the same binder as the Composite) that the compressive strength will be equal or superior to the tensile strength (UTS values).
ITEM 2Ga Submit reeutte of testing and/or analysee which indicate that these composite poison plates vill maintain their structural integrity during a vibratory environment auch as can be expected during an SSE.
DPC RESPONSE:
The Composite material proposed for the LACBWR fuel racks was developed to provide inherently greater structural integrity particularly in a vibratory environment such as can be expected during an SSE.
This was accomplished by adding a reinforcing material to the basic B4C binder system.
This addition makes the Composite r.taterial a flexible material which can easily 2199 285 accommodate a vibratory environment and will maintain structural integrity provided the stress levels in the material during the SSE event are less than the allowable value (see Response 25e).
For the LACBWR fuel racks, the maximum stress in the Composite material during the SSE event will be 620 psi.
This is well below the minimum UTS value of 2200 psi as specified in the Composite material procurement specification and obviously sig-nificantly less than the UTS value of 4500 psi conservatively estimated for the Composite material after 10 11 Rad exposure.
It should be pointed out that the storage cells have been designed to preclude any significant loss of Composite material or change in the Composite mate lal configuration even in the event that the Composite plate or sheet breaks into more than one piece or tends to crumble (see Response 261).
ITEM 26h_
State your plans for periodic monitoring of the composite of the poison plate material to ensure test reautta acrrelate uith actual La Croe.? spent fuel pool conditions uith regard to possible corross in and mechanical strength deterioration.
DPC RESPONSE:
Periodic nonitoring of the properties of the Composite material proposed for LACBWR will be performed as part of the surveillance progran described in Response 261.
ITEM 26i_
Provide data to ehnu that, under the combined effects of irradiation and immersion in fuel pool water, the teachability of the boron vill not be synergistically enhanced over the life of the high density storage racks.
DPC RESPONSE:
The rates at which boron is leached from B C Plate material (a 4
boron containing material similar to the B C Composite proposed 4
for LACBWR) were established for two environmental conditions:
- 1) immersion in demineralized water with no exposure to radiation and 2). immersion in demineralized water with simultaneous exposure to gamma radiation.
The results of the leaching tests for the two environments are presented graphically on page H-5 of Reference 1.
The test data indicate that the leaching behavior for the two environments are sufficiently similar to conclude that boron leachability is not synergistically enhanced by the presence of gamma radiation. 2199 286
An analysis of the test data obtained with simultaneous irrad-iation indicates that approximately 0.14% of the available boron will be leached from the tested coron containing material over a 25 year period if the leaching rate that exists at 19 days is maintained over the 25 year period.
It is evident from the test data and the analysis that a further reduction in the leaching rate is expected as the immersion time increases beyond 19 days.
Since the Composite material has similar proportions of B C and 4
binder the leaching rate obtained from the Plate material test data is concluded to be directly applicable to the Composite material.
The minimum B10 concentration specified for LACBWR is 0.024 gms of B10/cm.
This concentration will be supplied by Carborundum 2
10 contained in (actually exceeded) without taking credit for the B either the B 03 impurity in the B C powder or the boron constituent 2
4 in the reinforcing material.
If it is assumed that all of the leachable boron comes from the B C material, the reduction in the 4
minimum B10 concentration will be negligible (from 0.024 to 0.0239 gms B 10/cm ).
The criticality margin for the LACBWR fuel racks 2
was determined for a B10 concentration of 0.022 gms B10/cm.
It can, 2
therefore, be concluded that the leaching of boron from the Compos-ite material will not adversely affect the calculated criticality margin even if significantly higher leaching rates are assumed.
ITEM 265 Provide data that shove that the high dose rates used for accelerated irradiation teste have the same effect on the boron plates as the toucr Jose rates that vitt be received in the opent fuel pool.
DPC RESPONSE:
Information available in the literature cautioned against the use of excessively high irradiation rates since data indicated that the associated irradiation times were too short to permit secon-dary processes to occur in organics which could influence the degradative process (see Reference 2).
Consequently the irradiation dose rate as well as the type of radiation were major considerations in developing and implementing the Carborundum test program to qualify the LACBWR B C Composite 4
material.
For the program, a gamma source (spent fuel from a test reactor) was selected which had gamma dose rates sufficiently high to accelerate the test program but not so much greater than the prototypical..diation levels associated with freshly discharged spent fuel as to invalidate the test results (N 5 x 10 7 Rad /hr or less vs s 106 Rad /hr).
The test program was structured to include a gas generation test for the B C Plate material which has the 4
same bind as the Composite material.
The results of the gas generation test show that the rate of noncondensible gas generation was essentially the same as observed at Haddam Neck where B C Plates 4
2199 287 were irradiated by spent fuel over N 9 months to a level of s 3 x 109 Rad.
In addition, the gas species and the associated volume magnitudes in the gas generation test were very similar to those determined for Haddam Neck.
Since it is believed that gas generation is a sensitive indicator of the operational behavior of the binder, it has been concluded by Dairyland Power Cooperative that the gamma dose ratet utilized in the Carborundum qualification test program have the same effect on the composite material as the lower dose rates that will be received in the LACBWR spent fuel pool.
ITEM 26k Provide assurance that adequate venting at the ends of the plates uitt prevent sueiling due to trapped gas in the central portion of the composite plates.
DPC RESPONSE:
Swelling of the fuel storage cell walls due to the pressure buildup of trapped gasses will not occur for the following reasons:
a.
Each Composite material compartment contains a large vent hole (1/2 inch diameter) placed at the top of the compartment.
(The hole also permits visual verification of the presence of Composite material within the compartment after cell fabrication).
b.
The Composite material compartments are intermittently welded along the periphery of the compartment.
There-fore venting can occur at any location along the length of the storage cell.
c.
The low rate of gas ger.eration combined with the venting provisions assure that a pressure buildup within the compartment due to " trapped" gas is not feasible.
Furthermore, the principal gas species generated is hydrogen which is extremely difficult to contain even in sealed systems.
d.
The Composite material has not exhibited any significant dimensional changes during exposure to gamma radiation up to 10 10 Rads and is expected to exhibit the same behavior 11 Rads.
Consequently, the gas generated by the up to 10 Composite material does not cause the material itself to swell. 2199 288
ITEM 262 Describe the surveillance program that vill be performed to shou the ccntinued presence of the boron in att of the boron plates over the complete life of the storage rache, and ateo describe what action vould be taken if a decrease of boron in the plates is de tected.
DPC RESPONSE:
The surveillance program for LACBWR is designed to permit samples of the B C Composite material to be periodically removed from the 4
spent fuel pool and examined for corrosion and changes in the properties of the material, Stability of the physical properties (i.e., no significant changes in sample dimensions and weight) is indicative of a negligible change in the boron content of the Composite material.
Accelerated exposure to gamma radiation vill be achieved by pro-viding a sample holder which can be moved each outage to a new storage location surrounded by freshly discharged fuel.
The sample holder will be similar in size and basic shape to a fuel assembly with the samples located around the periphery of the holder.
The sample holder will be moved from storage location to location using available fuel handling equipment.
The Composite material samples will be placed between stainless steel sheet and exposed to the pool water to simulate the geometry that exists in the fuel racks.
Table 2 shows that the sample exposure level after five years will be at least a factor of 2 greater than the levels associated with fuel storage locations.
The table also shows that an exposure level of 10 11 Rad will not be reached for either the surveillance samples or the fuel rack material during the 25 year life of the fuel rack.
The purpose of the surveillance program is to detect any unan-ticipated changes in the properties of the Composite material before such changes progress to the point where they adversely affect the boron content of the material and, hence, the criticality margin.
Two mechanisms can be postulated for the loss of boron from the Composite material:
- 1) boron being leached from the material, or 2) boron being removed in the form of grains of B C as the 4
result of the loss of binder and/or backing integrity.
The results of the leaching tests (see Response 26i) indicate that the loss of boron due to leaching will be negligible over the life of the racks.
The results of the mechanical properties tests as well as the suc-cessful long-term commercial application of binder / reinforcing materials in aqueous environments indicate that the loss of binder and/or reinforcing integrity is extremly unlikely (see Response 26f).
An additional safeguard has been taken in the design of the LACBWR racks to minimize the effects of the Composite material becoming granular:
the Composite material compartments in the storage cells 2199 289 30 -
are sufficiently tight to preclude the significant loss of B C 4
grains even though the compartments are not watertight; the void regions in the poison compartments have been minimized to prevent a significant reduction in the level of the B C even if the loose 4
B C grains fill the void regions.
4 In the unexpected event that the LACBWR surveillance program detected a loss of boron from the Composite material, Dairyland Power Cooperative would develop a course of action which would prevent any condition resulting in an unacceptable criticality.
The specific course of action would depend on the mechanism by which the boron is being lost, the rate at which it is being lost, the quantity of spent fuel stored in the spent fuel pools and the alternative storage options that would axist at the time the loss of boron was projected to become substantial.
It should be emphasized that a significant quantity of boron can be lost from the Composite material before the criticality criterion (0.95 or less) is reached for spent fuel stored in the racks since the racks were designed to provide an acceptable criticality margin (0.9275 maximum value) with fresh fuel and with the minimum IO content.
specified B TABLE 2 GAMMA EXPOSURE LEVELS Fresh Fuel Fresh Fuel Estimated Stored for Replacement Sample 25 Years Every 5 Years Exposure 0
0 0
0 1
3.2 x 109 3.2 x 109 2 x 109 2
4.5 x 109 4.5 x 109 5 x 109 3
5 x 109 5 x 109 7 x 109 5
6 x 109 6 x 109 1.2 x 10 10 10 7 x 109 1.2 x 10 10 2.5 x 10 10 15 7.5 x 109 1.8 x 10 10 3.7 x 10 10 20 8 x 109 2.4 x 10 10 5 x 10 10 25 8,4 x 109 2.9 x 10 10 6 x 10 10
References:
1.
Carborundum Report No. CBO-78-299, dated October, 1978.
This proprietary report has been submitted to NRC previously on November 3, 1970 by Northeast Utilities for Dockets 50-213 and 50-245.
.2.
Effects of Radiation on Materials and Components, Kircher and Bowman, Reinhold Publishing Corporation.
]1{}
[]),
\\, -
l' i
s s
n i
NO 4Y2 -
a
} VEL D j -
, = =. =
't 9
d 4
'N v
y# ~
g V (rye 2199 291 wrmoce urna
\\
l..........
AlorE :
/. All MATER /Al rfyE 3 0 4 SE G74.
t/PPER EACK a
N I 1
tCWER 24c<
=
2199 292 7M76RLOCK;BETWEEN THE UPPER 7/ER RACK AWD LOW /PRActSrcr/oN' DA/RYLAND IPOWER COOPERA7/p'E LA CR0SSE, W/SCONSIH se. I of 3 geggg ~
.Dert /-3-79 Dwc.
ao.
22 54 l'
~
rp.
SrficM-/
ca.m / #k
5
/
-m N
r' N
l/ -
x as s
/
'4'
/
N f
m/
N l
> ',/
N ems
/
l I
(
s s
l h
}
I x
'N/
\\
\\
o her s/
/
/
f3 j-
/
90, l
[TYP N
N.
N
)
\\l w
\\
N u
/
- _p*
2199 293
//07E:
/, du /fMERML i
1,x e a eo-stes too.4-ti
ptN.
r-i SUPPORT PLATE
/
/
(x id
~
/%'$snSS8 l
&'SSh*
V i
~ gLCCK/NG COLL AR u
SEC7/CM *A - A "
'\\
UPPER GR/D AA/D SuPPORrB M 2199 294 wtrAssins. saspoer sisrs-DA/A?YL AND POWER COOPERA T/VE
'LA CROSSE,. Wisc&ws/H SM' 2 N Vf 304 CE 371.
SCALE ~
my;g p n~a.
no.
ry, ZKETCH 2_
cx, App) 1j'r h !
=
\\.
~
I' i
i
/
i
' ^
c r-
,, n r, r
i 3
C I
i 3
c t
3 E
[
3
[
I
}
J i
i c
t 3
C
[
3 C
E 3
"a m'
m m
m m
m m b
TYP/ cal fB0 0 0 0
[ D 0 00O J
lJ-eppte ace
=
f
[00 0 0 0 C D 0 00D
]
(
[QD D D D
[ O O D00]
/
4 xQ h L O WER R GCAf
=
'A a
100 000
[ 0 0 00D
)
i sy; _
=
rYP/CM 2199 295
\\
,/' V8 " l/ ft " 3 '
g/gyg,.
~/
.L
/, AlL MA7ff/A L TH0f 304 JI~571..
47%
/
U
~
~
t cain
/Yz'
- i. &
s 7
I BRAC/NG 2 -
N g-L
-ano 2199 296
%*ifle*x/fz/G.
l
/l(g'V POO4 FUfL S7O' RAGf AACM CfMM/C y
DMER sencista g
DA /RYL AND POWfR C00Pf2A 7/t/6 LA CRossf, W/Sco^fS/Af M3 of 3 Scx[
~
DR7E
/-J-7 9 Dwn' A/o.
BRA C/A/G pg, de
79, 3KETC//. 3 cw, m.f.,/5/44
4, Renconse to NRC Ouestions Submitted by c,.
NRC Let.er, Ziemann to fladgett, dated December 7, 1978 Proposed Modification - LACBWR Spent Fuel Storace NRC ITEM NO. 2 Your response to question 13 (forwarded by letter dated September 28, '978) did not indicate uhether the doec rate of 10-30 mrem /hr 2ith the cater level in the Fuel Element Storage Well (FESW) at the 680 foot elevation, was as lov as reasonably achievable (t LA RA ).
If you have determined that this dose rate is ALARA, p ease provide the basis for your finding.
The basis should include an estimate of the incremental decrease in doce increases in FESW vater level above the 680 foot rate ve.
elevation and a detailed discussion of the reasons uhy (including the FESW level should quantitative cost effectiveness criteria) not be maximized (i.e. maintained at or near the 700 foot elevation) ao that the dose rate vill be minimized.
If you have determined that the dose rate is not ALARA and there-fore that an increase in FESW level is needed to attain an ALARA please propose a change to your Technical Specifications dose rate, that appropriately reflects the proper level requiremente.
DPC RESPONSE:
There is no significant decrease in dose rate as the water level is raised from the 680 foot elevation to the 700 foot elevation (see attached survey summary).
It appears that as long as there is at least 15 (13 feet) tenth value layers (TVL) of water over the spent fuel that exposure levels are minimal.
It is emphasized that the dose rates of 10-30 mrem /hr discussed in the answer to Question 13 are very conservative.
The 30 mrem /hr dose rate is directly over the center of the fuel pool'at 100% power, while most exposure levels encountered during routine operations on the 701' level are normally less than 5 mrem /hr.
Based on this information, it is concluded that -he dose rates are ALARA, since a further increase in water depth (i. e., more than 15 TVL's above the spent fuel elements) does not significantly reduce the dose rates encountered on the 701' level.
2199 297 L ACBWR 701' L E VE L.
DATE:
JANUARY 4, 1979 TIME:
0830 SURVEY MAP POWER LEVEL:
49%
PERFOIU1ED BY:
R.
PRINCE l
i 1
2 8
\\
4 10 FESW 12 3
5 6
7 O
Survey Meter:
RO-3
~
S/N:
- 171, s
Dose Rate Levels - mrem /hr WATER Location 1
2 3
4 5
6 7
8 9
10 11 12 680 4
4 4
4 5
3 5
5 9
5 8
9 690 4
4 4
4 5
3 5
5 9
5 7
7 700 4
4 4
4 5
3 5
5 7
5 7
7 2199 298 NRC ITEM NO. 2 In your response to our question 4 (forvarded by letter dated September 28, 1978) you stated that the average accumulated data for the year of 1978 to date, shoved that the FESW pool concen-tration for radionuclides vaa about 3:10~3 pc/ml uith a resulting doce rate of approximately G. 0 mrem /hr.. Is this dose rate typical of what is c=pected with expanded storage capacity?
Indicate whether or not the expected dose rate is ALARA, and pro-vide the basis for your conclusion.
Your response should contain a discussion of the ability of the purification system to minimize the concentration of radionuclides in the vater, including a description of the purification system, average yearly operating time, flou rate, minimum decontamination factor (DF), capacity of the demineratiner resin bed, and criteria for resin replacement.
In particular, justify uhy (including quantitative cos t effective-neca factors) an increase in system flou and/or minimum DF (eg.
DF = 10) should not be utilized to further reduce the present dose rate resulting from the concentration of radionuclides in the pool vater.
DPC RESPONSE:
The storage well purification system consists of a filter assembly and an ion exchanger with a six cubic foot capacity.
The filter is a Commercial Filter Corporation Fulflow Filter with a pore size of 10 microns.
A complete filter assembly consists of 30 filters.
The resin bed operates continuously except during replacement periods which amount to a few hours per year.
Filters are in line for approximately 85% of the time.
Normal storage well flow rate will vary from 260 gpm down, depend-ing on the cooling requirements.
Presently, the flow rate is 190 gpm.
The filter assembly receives normal pool flow.
Upon existing from the filter assembly, a partial flow of approximately 20-25 gpm proceeds to the resin bed while the remaining flow is circulated back through the FESW system.
The resin bed is replaced when the decontamination factor drops to one.
additional information was provided previously in applicant's re-sponse to CREC's first set of interrogatories dated October 5, 1978 and in applicant's repsonse to CREC's second, third and fourth sets of interrogatories dated October 16, 1978.
Interroga-tories, Set One, Questions 17 and 18, and set Two, Questions 4, 6
and 21.
In the response to these intervenors questions, we examined data from FESW samples taken during 1977 and 1978.
This data was tabulated and reviewed.
Based on this data, it was concluded that 6.0 mrem /hr could be considered a typical dose rate.
There was no indication of an upward trend in activity associated with increased pool inventory. 2199 299