ML20216J829
| ML20216J829 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 03/20/1998 |
| From: | WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | |
| Shared Package | |
| ML20216J820 | List: |
| References | |
| NUDOCS 9803240101 | |
| Download: ML20216J829 (18) | |
Text
I-Attachment IV to ET 98-0009.
Page 1-of 8 ATTACHMENT IV PROPOSED TECHNICAL SPECIFICATION CHANGES CURRENT TECHNICAL SPECIFICATIONS l
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L-i 9903240101 990320 PDR ADOCK 05000482 P l PDR
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Attachment IV to ET 98-0009 Page 2 cf 8 -
4 REFUEL OPERATIONS 3/4.9.12\SPENTFUELASSEMBLYSTORAGE LIMITING C0 T!0N FOR OPERATION
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3.9.12 Spent fu assemblies stored in Region 2 shall be subje to the following conditio s:
- a. The combin tion of initial enrichment and cumulat we exposure shall be within t accept &ble domain of Figure 3.9-1 and
- b. No spent fuel ssemblies shall be placed in gion 2. nor shall any storage locatio be changed in designation ce being in Region 1 to being in Region , while refueling operati ns are in progress.
APPLICABILITY: Whenever irr fated fuel assembli s are in the spent fuel pool.
ACTION:
- a. With the requirements of the ab e specification not satisfied, suspend all other movemen of 1 assemblies and crane operations with loads in the fuel sto areas and move the non-complying fuel assemblies to Region 1. Un 1 these requirements of the above specification are satisfi b ron concentration of the spent fuel pool shall be verified t be g star than or equal to 2000 ppe at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. -
- b. The provisions of Sp cifications 0.3 and 3.0.4 are not app 11 cable.
SURVEILLANCE REQUIREMENTS
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4.9.12 The burnup o each spent fuel assembly sto in Region 2 shall be ascertained by anal is of its burnup history and i pendently verified, prior to storage in Reg n 2. A complete record of such an lysis shall be kept for the time period t the spent fuel assembly remains th Region 2 of the spent fuel pool.
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4-WOLF CREEK - UNIT 1 3/4 9-15
Attachment IV'to ET 98-0009 l Page 3 of 8 50
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40 -
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- 30 -
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10 -
Ennerant Swme 2.1 10.572 2.8 18.000 3.1 25.14 3.8 34 001 45 43.357
! l , 1 , I . \ ! , i 1.5 2 2.5 3 3.5 4 4.5 5 ENRICHMENT (w/o)
FIGUR 3.9-1 WOLF CREEK MINIMUM REQUIRED FU L ASSEMBLY BURNUP AS A FUNCTION OF INITIAL ENRICHMENT INSEAT NN FOR STORAGE IN REGION 2 PseuRE. 3 3-1 h WOLF CREEK - UNIT 1 3/4 9-16 Amendment No. J5. 61 a
L..... ..
h Attachment IV to ET 98-0009 Page 4 of 8 INSERT A REFUELING OPEPATIONS 3/4.9.12 SPENT FUEL ASSEMBLY STORACE LIMITING CONDITION FOR OPERATION 3.9.12 Spent fuel assemblies stored in either the Spent fuel Pool or cask loading pit shall be subject to one of the following conditions:
- a. Spent fuel assemblies regardless of burnup can be placed in Region 1.
- b. Spent fuel assemblies within the burnup-enrichment range shown in Figure 3.9-1 for Region 2 fuel can be placed in Psegion 2 or 3.
- c. Spent fuel assemblies within the burnup-enrichment range shown in Figure 3.91 for Region 3 fuel shall be placed in Region 3.
APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel pool or cask loading pit.
ACTION:
- a. With the requirements of the above specification not satisfied, suspend all other movement of fuel assemblies and crane operations with loads in the fuel storage areas and move non-complying fuel assemblies to Region 1. Until the requirements of the above specification are satisfied boron concentration of the spent fuel pool shall be verified to be greater than or equal to 2000 ppm at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9,12 The burnup of each spent fuel assembly stored in Regions 2 or 3 shall be ascertained by analysis of its burnup history and independently verified prior to storage in Regions 2 or 3. A complete record of such analysis shall be kept for the time period that the spent fuel assembly remains in Regions 2 or 3 of the spent fuel pool or task loading pit.
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Attachment IV to ET 98-0009 Page 5 of 8 FIGURE 3.9-1 MINIMUM REQUIRED FUEL ASSEMBLY BURNUP AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE IN REGIONS 2 AND 3 55000; l
50000 9 : ACCEPTABLE BURNUP DOMAIN y : FOR REGIOrl 2 AND 3 STORAGE N 45000 o -
3 ~
40000 f w : $
E : p+ #
35000 p -
1 5
5 6--
5 3
30000 25000
/
/[ -
5 8'
2 -
d I
l 8 20000 4 8 g b : 4 I
W 15000
- ^(8 W -
/f /
N
-3 10000 i P/
5000
- [ UNAC CEPTAB LE BUFtNUP DOMAIN FOR ltEGION 2 OR 3 STollAGE 0 ,,,, ,,,, ,,.. .... .. ..i. . ii 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 FUEL ASSEMBLY INITIAL ENRICHMENT (w/o U-235)
l Attachment IV to ET 98-0009 Page 6 of 8 REFUELING OPERATIONS i BASES )
l 3/4.9.9 CONTAINMENT VENTILATION SYSTEM l
The OPERABILITY of this system ensures that thL containment purge penetrations will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment. l l
3/4.9.10_.3nd 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity l released from the rupture of an irradiated fuel assembly. The minimum water I depth is consistent with the assumptions of the safety analysis. ;
3/4.9.12 SPENT FUEL ASSEMBLY STORAGE The restrictie.s placed on spent- fuel asseslies stercd in l1:gi0n 2 Of the spent fuel p::1 :nsure in d/ertent criticalit' will net :::ur.
3/4.9.13 EMERGENCY EXHAUST SYSTEM - FUEL BUILDING 1
The limitations on the Emergency Exhaust System ensure that all !
l radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. Operation of the system with the heaters operating to maintain low humidity for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The OPERABILITY of this system and the resulting iodine removal .
l capacity are consistent with the assumptions of the safety analyses. !
ANSI N510-1975 and N510-1980 will be used as procedural guides for surveillance testing.
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WOLF CREEK - UNIT 1 B3/49-3 Amendment No. E, M,-69, 95
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Attechment IV to ET 98-0009 Paga 7 of 8 INSERT B 3/4.9.12 SPENT FUEL ASSEMBLY STORAGE The racks for the Spent Fuel Pool and cask loading pit are designed for storage of both new fuel and irradiated fuel. Prior to storage of fuel assemblies in either the Spent Fuel Pool or cask loading pit, overall pool storage Regions shall be prepared in accordance with administrative controls. The restrictions placed on fuel assemblies stored in the spent fuel pool and cask loading pit ensure inadvertent criticality will not occur. Region 1 is designed to accommodate new fuel with a maximum nominal initial enrichment of 5.0 weight percent U-235, or spent fuel regardless of the discharge fuel burnup. Region 2 and Region 3 are designed to accommodate fuel of various initial enrichments which have accumulated minimum burnups within the acceptable domain according to Figure 3.9-1 in the accompanying LCO.
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Attachment IV to ET 98-009 Page 8 of 8 ,
DESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with: s.o
- a. equivalent to ss than or equal to 0.95 when flooded with
-A kg , ted water, whic i includes an allowance for uncertainties as unbora described in Section o.3 of the USAR. This is based on new fuel haimm*N=) w'th.anAenrichment of!4:45-weight percent U-235 in Region 1 and on spent fuel with combination of initial enrichment and discharge l - exposures, in Figure 3.9-1, 4n Region 2. and (t3A~
7a3 l b. nomina nch center-to-center istance between fuel assemblies placed in the storage racks.
5.6.1.2 The kg for new fuel for the first core loading stored dry in the spent fuel stora,ge racks shall not exceed 0.98 when aqueous foam moderation is j assumed, l
DRAINAGE l
5.6.2 The.opent fuel storage pool is designed and shall be maintained to prevent inadverten draining of the pool below elevation 2040 feet.
l be cask l'eadih3 pit t's designed omd. s3ah be mashtairied 64h A. 1 CAPACITY storage capciH ilmitea to no mere wwn z 79 a.ssembts'es.
~ ~ -" -
5.o.3 The spent fuel storage pooT ls 3esigned and-sha1 tee maintiFned with a .
storage capacity. limited to no more th fuel assemblies.
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l l 5.7 COMP 0NENT cyclic OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Tabic 5.7-1. l n --
- c. Po.- fuel tddh nomina) enrichments greah ihan %
we'ighi percent t.L-7 35, the combinatidn o{ c.nrichment a.nd.
thtsgral -fuel barnable, olocarbers .shall be suffidient so 4 hat the I
requ.stements of Spee.16 cation S.6.t.t. a. are met. I nt43ral fuel barnab\e. absorbers are not required Gr Regidw1.
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l locabe6s on 4hc. pert g% of thc. poo) , adjacen:t
, ft> a. co nceete w all .
WOLF CREEK - UNIT 1 5-7 Amendment No. Jg,61
Attachment V to ET 98-0009 Page 1 of-10 ATTACHMENT V PROPOSED TECHNICAL SPECIFICATION CHANGES IMPROVED TECHNICAL SPECIFICATIONS i
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Attachment V to ET 98-0009 Page 2 of 10 Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTENS 3.7.17 Spent Fuel Assembly Storage the spenhfue,\ weask [ed,$ h LCO 3.7.17 The combination of ini al enrichment and burnup of each spent fuel assembly stored in =fr 2 shall be within the Acceptable Domain of Figure 3.7.171 or in accordance with Specification 4.3.1.1.
APPLICABILITY: Whenever any fuel assembly is stored in ." ;'.;; 2 :f the cask loadmg ph ACTIONS CONDITION REQUIRED ACTION CONPLETION TIE A. Requirements of the LCO A.1 .. - NOTE - -
not met. LCO 3.0.3 is not applicable.
Initiate action to move Immediately the noncomplying fuel assemb1 '= Pr;ir 2.
Region @
SURVEILLANCE REQUIREENTS SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior to-storing enrichment and burnup of the fuel assembly is the fuel in accordance with Figure 3.7.17 1 or assembly in Specification 4.3.1.1. Region 2 @
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i i Attachment V to ET 98-0009 I Page 3 of 10 l Spent Fuel Assembly Storage 3.7.17
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40 -
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30 -
- n l k
3 -
1 20 -
l l 10 -
l 1 sansevnear nume 1
11 10.572
. Es itsee i 11 25.14 I 18 MS 4.s 48J57 0 '
1.5 2 2.5 3 3.5 4 4.5 5 ENIUQO4ENT(e)
Figure 3.7.171 (page 1 of 1)
Fuel Assembly Burnup Limits in Region 2 \
-, = _ __ -
3.7 37 WCGS iTS 5/15/97 f
Attachment V to ET 98-0009 Page 4 of 10 1
FIGURE 3.7.17-1 (page 1 of 1)
MINIMUM REQUIRED FUEL ASSEMBLY BURNUP AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE IN REGIONS 2 AND 3 55000 ,
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50000 S : ACCEPTABl_E BURNUP DOMAIN g : FOR REGIOPI 2 AN D3 STORAGE N 45000 -
0 w :
3 mm
/
w : $
sM : p+ #
35000- 4 R : / #
b w
3**
/ s 4# '
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25000f 1 -
p /
20000 ~
g [
4@
y *** : /
a inn
- /g/
h 5000
- [ UNAC CEPTAB LE BURNUP DOMAIN FOR ltEGION 2 OR 3 SToltAGE o' .... .... .. . . .. .-... ii.. . . . .
1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 FUEL ASSEMBLY INITIAL ENRICHMENT (w/o U-235)
Attachment V to ET'98-0009 Page 5 of 10 Spent Fuel Assembly Storage l B 3.7.17 l
l B 3.7 PLANT SYSTEMS i B 3.7.17 ' Spent Fuel Assembly Storage l l
BASES I BACKGROUND Ir tre ",eot ;. %r.;ity L;k O'::C) i;;gr., tre ;ps,t fal ;R. ,,.
g Q,g9 _
- 1 h dhiid i-t
- t" : :r:te rd duti :t r;;i::: did, hr the prp;; Of critiality ce;.;ide. ;.ti;;.;. ;.r; ;,,7.;id;.r;d =
j
- g:r:t: p;h ("ef.1). Region 185;ith: W" f 200 ster:--
i 5.0
-. is_ designed to acconnodate new fuel with a maxialm =
enrichment of%46 wtt U 235, or s fuel regardless of the w discharge fuel burnup. Region a "ith : W- ef 11'a ste-ea-
- ord, Reg. m =3 are
- ) p;iti:::, i:* designed to accommodate fuel of various initial enrichments which have accumulated minimum burnups within the acceptable domain according to Figure 3.7.171, in the accompanying LCO. F=1 =dlin =t .xting th: crit;ri Of ri;;r: 3.7.17-1 2:11 bc :tered in e;;erder.ce 4th -
- ea ta" ' 3.1.14- Srth; i.2. Tal St;r;;;.
B 3 1-The water in the spent fuel pool nd [normally M l=h3r@
contains soluble which results in large suberiticality margins under actual operating conditions. However, the NRC guidelines, based upon the accident condition in which all soluble poison is asstmed to have been lost, specify that the limiting k,y of 0.95 be
@i +hre.Q evaluated in the absence of soluble boron. Hence, the design of 4enhegions is based on the use of unborated water, which maintains e ch r;;i:n in a suberitical condition during normal LCop-erd
- ad e lead!$ fuel peo]g operation with the regions fully loaded. The double contingency principle discussed in ANSI N 16.11975 and the April 1978 NRC letter (Ref. 2) allows credit for soluble boron under other abnormal or accident conditions, since only a single accident
- v need be considered at one timeffFo ex e, mo sev re 3 M acc 6en se ario is a ocia Reg)io1t Reg n 2, and wi t mov cide al m loa ng o af nt fue f
, ss ly nR ion . To iti te t se tula ed er tic ity
[reted cci ts, ro is d solv in I wa r, afe ra on the w h no mov t of ass lies y here re 1ev by ont ling he 1 atio of e ch a emb in g ac rda e wi the acc nyi LCO. j l
(continued)
WCGS ITS BASES B 3.7 80 5/15/97
Attachment V to ET 98-0009 Page 6 of 10 Insert B 3.7-80a The high density rack modules for the spent fuel pool and cask loading pit are designed for storage of both new fuel and spent fuel. Spent fuel storage is designated into Regions based upon administrative contro]s.
Insert B 3.7-80b Prior to storage of spent fuel assemblies in either the spent fuel pool or the cask loading pit, overall pool storage configurations are prepared in accordance with administrative controls. The pool layouts include sufficient Region 1 storage to accommodate new and discharged fuel assemblies with low burnup. Fuel storage utilizes either a Mixed Zone Three Region configuration and/or a checkerboarding configuration.
In a Mixed Zone Three Region configuration, Region 1 storage cells are only located along the outside periphery of the rack modules and must be separated by one or more Region 2 storage cells. Region 1 storage cells may be located directly across from one another when separated by a water gap. The outer rows of alternating Region 1 and 2 storage cells must be further separated from the internal Region 3 storage cells by one or more Region 2 storage cells. New fuel assemblies with enrichment greater than 4.6 wt% U-235 and less than 16 IFBA rods must be stored in any peripheral, Region 1 storage cell that is next to a concrete wall.
In the checkerboarding configuration, fuel assemblies are placed in an alternating checkerboard-style pattern with empty storage cells (i.e., fuel assemblies are surrounded on all four sides by empty storage cells, except at the checkerboard boundary). Region 1 fuel assemblies may not be located directly across from one another, even when separated by a water gap. This arrangement may be used anywhere in the spent fuel pool or cask loading pit and may be combined with the Mixed Zone Three Region configuration within a rack module, if the checkerboarding pattern is maintained in a linear array equal to or greater than 2X2. A checkerboard area may be bounded by either a water gap, empty cells, Region 2 fuel assemblies or Region 3 fuel assemblies.
Insert B 3.7-80c For example, the most severe accident scenarios, for which boron credit is taken, are:
- a. Inadvertent loading of a 5.0 wt% U-235 new fuel assembly in a Region 2 or Region 3 storage cell in a Mixed Zone Three Region configuration, or in a empty cell in a checkerboard configuration.
- b. Mis-location of a new fuel assembly in the gap between the rack modules and the concrete wall in the spent fuel pool.
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Attachment V to ET*98-0009 l Page 7 of 10 j Spent Fuel Assembly Storage B'3.7.17 ;
i BASES (continued) i
)
l APPLICABLE The hypothetical accidents can only take place during or as l SAFETY ANALYSES a result of the movement of an assembly (Ref.1). For these !
gggg"3 -
accident occurrences, the presence of soluble boron in the spent- - _-__
fuel stopogs poo1* prevents criticality int ^.:. 7;;ir;. By (d+n* P=9 i closely controlling the movement of each assembly and by identifying the fuel assembly. identification number prior to movement "': %;ia 2. it is ensured that the fuel stored in l each location in ?;;'= 2 is the fuel _that was designated for l l storage in that location. 96 span 4 61 poet or cam leedk@
The configuration of fuel assen611es in the spent fuel pool a __
i satisfies Criterion 2 of 10 CFR 50.36(c)(2)(11). (and essKle2 dig h
(.".s, iocatum 4.s.t . t - seet & 4.3)
LC0_ The restrictions ithin the
{orcasktodyp+/ spent fuel poor on the placement in accordance of Figure with fuel assemb11eps 3.7.171, in the l accompanying LCO, ensures the k,n of the spent fuel eterage pool l will always remain < 0.95, assiming the pool to be flooded with
& ratM water. ' - - ' - - - - - ~ - " - * " ' - ' * " - '
l Fi;r:
wifi s 3.7.17-1 u a 4.3.1.1t"ir.bcR str:d " 5-:-:-d ;e efS [cak\cokn3 h i
ci = 4.3. -
APPLICABILITY This LCO applies whenever any fuel assembly is stored in W ef the spent fuel poo}. [c,3k,13,3Q p.+]
ACTIONS L.1 l R(quired Action A.1 is modified by a Note indicating that l l
LCO 3.0.3 does not apply. mor gc:ask load.in3 fiO l When the configura ion of fuel assemblies stored in %;ix 2 ef !
the spent fuel pool is not in accordance with Figure 3.7.171, or I paragraph 4.3.1.1. the isenediate action is to initiate action to make the necessary fuel assembly movement (s) to bring the !
configuration into compliance with Figure 3.7.171 or Specification 4.3.1.1.
l l i (continued) i i WCGS ITS BASES B 3.7 81 5/15/97
Attachment-V to ET 98-0009 Page 8 of 10 Spent Fuel Assembly Storage B 3.7.17 BASES ACTIONS L.1 (continued)
If unable to move irradiated fuel assemblies while in MODE 5 or 6. LCO 3.0.3 would not be applicable.
If unable to move irradiated fuel assemblies while in MODE 1. 2.
- 3. or 4, the action is independent of reactor operation.
Therefore, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.
SURVEILLANCE SR 3.7.17.1 smWWng c4 fu.t 2+hin a R ibn cloes not reqde, pforrneca M this .
REQUIREENTS w - - -
This SR verifies by administrative means that the initial enrichment and burnup of the fuel assembly is in accordance with Figure 3.7.171 in the accompanying LCO. For fuel assemblies in the unacceptable range of Figure 3.7.17-1, performance of this SR y will ensure compliance with Specification 4.3.1.1. .r.
REFERENCES 1. USAR, Appendix 9.1A
- 2. Double contingency principle of ANSI N16.11975, as specified in the April 14, 1978 MtC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4. Appendix A).
' WCGS ITS BASES B 3.7 82 5/15/97
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Attachment V to ET~98-0009 '
Page 9 of 10 Design Features 4.0 4.0 ' DESIGN FEATtRES 1
4.1- Site Location 4.1.1 The WCGS r,ite is approximately 3.5 miles east of the John Redmond Reservoir in Coffey County, Kansas and is approximately 3.5 miles L northeast of the town of Burlington.
l 4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall l consist of a matrix of Zircalloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide !
(%) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved l ;
i applications of fuel . rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable imC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number l i
of lead test assent 11es that have not completed representative testing l may be placed in nonlimiting core regions.
I 4.2.2 Control Rod Assemblies l l
The reactor core shall contain 53 control rod assemblies. The control
. rod material shall be silver indita cadmiin, or hafnium metal as i l approved by the NRC._ __ _
l
@4h I w% enr'ichmerks arcater than. 4.t. nommal tac'ic)*
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L pe.vcent of u.-Z.36', the. cemios nation of enrichmevrt cw 'inteafal l
4.3 Fuel Storage fuel barnabte. absorbees shal\ be suWitsnt so%at +he reept'e.ments r cf 4.311.b. cLre. mat, inieyal ka\ burnaMs. absorloers are. nck 65 n s, o n M e.
4.31 Critica11t ##Y+hc4
- pool , {i n 1acant te, a. concrete. i.oan . pe pM l 4.3.1.1 The sht fuel storage racks a're designed and shall be maintained with:
--omin T a)I
- a. Fuel assemblies ha a maximum U 235 enrichment of y
-4r46 percen , 1
- b. k,n s 0.95 if fully flooded with unborated water, l which includes an allowance for uncertainties as described in Section 9.1A of the USAR:
(continued) l WCGS ITS 4.0 1 5/15/97
1 1
Attachment V to ET 98-0009 Page 10 of 10 \
Design Features 4.0 i
4.0 DESIGN FEATURES l l
Fuel Storage (continued) 4.3
- c. A nomina 4 936 inch center to center distance between fuel assemblies placed in the fuel storage hnup Domain fue F.3,an 2 ama 3 Q -
l d. ":: :hrtially spent fuel assemblies with a i discharge burnup in the */cceptable " of Figure 3.7.17-1 may be allowed unrestricted storage irf'th fri ;ters rxb; and l 3 h+accepable, New orfuel etags partially location spent _D with a fuel assemblies discharge burnup in the "/nacceptablgpenge" of
- Cfiaure 3.7.17 ?, will be stored in Region 1. -
(%rn"P Mn for Region 2 er 3 fwora]s QlocaboQ 4.3.1.2 The new fuel storage racks are designed and shall be maintained with:
- a. Fuel assemblies having a maximum U 235 enrichment of 4.45 weight percent:
1
- b. k,n s 0.95 if fully flooded with unborated water. ]
which includes an allowance for uncertainties as j described in Section 9.1 of the USAR: 1
- c. k,y s 0.98 if moderated by aqueous foam. which i includes an allowance for uncertainties as described i in Section 9.1 of the USAR: and
- d. A nominal 21 inch center to center distance between fuel assemblies placed in the storage racks.
4.3.2 Drainaae r,34.
The-spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 2040 ft. _
4.3.3 Canacity f id%
The.a.cask. lonadn3 stora$c_ ca.p,pa is designed
.c.m.g timwea +am4 shatt tha.n no more be mam%,ned zu set I ca s->e m bl ies .
The spent fuel pool is designed and shall be maintained with a storage y capacity limited to no no +34+ fuel assemblies. =
I c. 'Partialbj myent fuel assembb'es. iaWh o. dishrge. barnap k the " Acceptable. brny Domim -fby Regiin S Storou b swage. in acceptable \be.3c." storage.ofLocations Figwre. ,3.7.il-t exca.ptma3 e attowed e RegioW u,nce tric.ted z tocatidns W a Mixed Eone "Three. Region configgy, tion ; evnd.
Enclosure II to,ET 98-0009 j ENCLOSURE II NON-PROPRIETARY VERSION OF THE LICENSING REPORT FOR RERACKING OF WCGS SPENT FUEL POOL 6
l t
(
6 Holtec Center,555 Lincoln Drive West, Marlton, NJ 08053 HOLTEC lNTERNATlONAL Telephone (609) 797-0900 Fax (609) 797-0909 LICENSING REPORT FOR RERACKING OF THE l
CALLAWAY AND WOLF CREEK NUCLEAR PLANTS for UNION ELECTRIC AND WCNOC l by HOLTEC PROJECT No: 70384 & 70814 HOLTEC REPORT No: H1-971769 REPORT CATEGORY: A Report Class: SAFETY RELATED COMPANY PRIVATE This document version has all proprietary information removed and has replaced those sections, figures, and tables with highlighting and/or notes to designate the removal of such information. This document is to be used only in connection with the performance of work by Holtec International or its designated subcontractors.
Reproduction, publication or presentation, in whole or in part, for any other purpose by any party other than the Client is expressly forbidden.
qso si o npstaan
1 I
Holtec Center,555 Lincoln Drive West, Marlton, NJ 08053 .
HOLTEC tNTERN TI O N A1.
Telephone (609) 797-0900 REVIEW AND CERUFICATION IDG DOCUMENT NAME:
Licensing Report for Reracking of the Callaway and Wolf Creek Nuclear Plants HOL'IEC DOCUMENT I.D. NUMBER: HI-971769 HOLTBC PROJECT NUMBER: 70384 and 70814 CUSTOMER / CLIENT: Union Electnc and WCNOC REVISION BIDCK REVISION' AUTHOR & REVIEWER & QA APPROVED' NUMBER DATE DATE MANAGER & DATE
_ _ , r CN A & DATE . _
ORIGINAL A -
su sz-31-97 l'2br 9 7 jfs n ..?, W t5-p,[%7 e
REVISION 1 H*
w t-s-98 A s. 't/#fa , $h8 a///u REVISION 2 *
- su 1- ss 98 ^*'"^* z/isbe /hflM O W- 0-97 REVISION 3 d' #
- sw* z-tm n ^*w") dJrs ,
%bW" sue s z-1s REVISION 4 REVISION 5 REVISION 6
" Ibis documed conforms to the requirements of the design specification and the apphcable sections of the govermag codes.
Note: Signatures and pnnted names are reqmrod in the review block.
A revision of this document will be ordered by the Project Manager and carned out if any ofits contents is matenally affected during evolution of this project. The determhnation as to the need for revision will be made by the Project Manager with input from others, as deemed necessary by him.
Must be Project Manager or Ids designee.
I THE REVISION CONTROL OF THIS DOCUMENT IS BY A "
SUMMARY
OF REVISIONS LOG" PLACED BEFORE THE TEXT OF THE REPORT.
4
SUMMARY
OFREVISIONS Revislom 3 contains the following pages, including figures:
COVER PAGE 1page REVIEW AND CERTIFICATION IDG 1page
SUMMARY
OF REVISIONS 1page TABIE OF CONTENTS 4 pages 1.0 . INTRODUCTION 6 pages 2.0 OVERVIEW AND PROPOSED CAPACITY EXPANSION . 23 pages 3.0 MATERIAL, HEAVY LOAD, AND CONSTRUCTION CONSIDERATIONS 15 pages 4.0 ' CRITICALITY SAFETY ANALYSES 35 pages APPENDIX 4A - BENCHMARK CALCULATIONS 25 pages 5.0 THERMAI HYDRAULIC CONSIDERATIONS 27 pages 6.0 S'IltUCTURAI/SBSMIC CONSIDERATIONS 88 page 7.0 FUEL HANDLING AND CONSTRUCTION ACCIDENTS 22 pages 8.0 FUEL POOL STRUCTURE INTEGRITY CONSIDERATIONS 15 pages 9.0 RADIOIAGICAL EVALUATION 7 pages 10.0 INSTAILATION 11 pages 11.0 ENVIRONMENTAL COST / BENEFIT ASSESSMENT 9 pages
'IOTAL 290 pages Revision I contains changes from Wolf Creek and Callaway comment letters NE 98 0008, NED 98-007, NED 98-010, NED 98-018. The changes are discussed in Holtec response letter 70384.SP6.
Revision 2 contains changes to typographical changes on Figure 2.1-1, Tables 5.4.1 and 9.4.1, pages 3-1,4-27,5-10, 5-16, and 7-6. An additional sentence has also been added to the top of page 5-3 about emergency water makeup.
Revision 3 corrects the references of Section 3. Revision 3 also prepares two versions of the repon; the version intended for NRC review contains all information with the pur;etary information denoted by highlighting or notes, the other version intended for public viewing
. contains only highlighting with the proprietary information extracted.
Hohoc Isewustment Report m-971769
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TABLE OF CONTENTS
1.0 INTRODUCTION
........................................ 1-1 1.1 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 2.0 OVERVIEW OF PROPOSED CAPACITY EXPANSION . . . . . . . . . . . . . . . 2-1 2.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
2-1 2.2 Summary of Ps.W Design Criteria . . . . . . . . . . . . . . . . . . . . . . 2-3 2.3 Applicable Codes and Standards . . . . . . . . . . . . . . . . . . . . . . . . . . 2-4 2.4 Quality Assurance Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-11 2.5 Mechanica1 Design .................................. 2-11 2.6 Rack Fabrication . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-12 2.6.1 Fabrication Objective . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-12 2.6.2 Anatomy of the PWR Rack Module . . . . . . . . . . . . . . . . . . . . 2-12 3.0 MATRRIAI HEAVY IDAD. AND CONSTRUCTION CONRinRR ATIONS . . 3-1 3.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 1 ,
3.2 Structur:1 Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.3 Poison Material (Neutron Absorber) . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.4 Compatibility with Coolant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3 i 3.5 Heavy Imd Considerations for the Proposed Rack Installation / Removals . 3-3 3.6 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-8 4.0 CRITICAllTY SAFETY ANALYSES .......................... 4-1 ,
4.1 Design Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 1 4.2 Summary of Criticality Safety Analyses . . . . . . . . . . . . . . . . . . . . . 4-5 4.2.1 Normal Operating Conditions . . . . . . . . . . . . . . . . . . . . . . . . 4-5 4.2.2 Abnormal and Accident Conditions . . . . . . . . . . . . . . . . . . . 4-8 4.3 Reference Fuel Storage Cell . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-10 4.3.1 Reference Fuel Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . 4-10 ,
4.3.2 High Density Fuel Storage Cells . . . . . . . . . . . . . . . . . . . . . 4-11 ;
4.4 Analytical Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-12 l 4.4.1 Reference Design Calculations ...................... 4-12 4.4.2 Fuel Burnup Calculations and Uncertainties . . . . . . . . . . . . . . 4-13 4.4.3 Effect of Axial Burnup Distribution . . . . . . . . . . . . . . . . . . . 4-14 4.5 Criticality Analyses and Tolerances . . . . . . . . . . . . . . . . . . . . . . . . 4-16 4.5.1 Nominal Design Case . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-16 4.5.2 Uncertainties Due to Rack Manufacturing Tolerances . . . . . . . . 4-16 4.5.2.1 Boron Imding Tolerances . . . . . . . . . . . . . . . . 4-16 !
4.5.2.2 Boral Width Tolerance Variation . . . . . . . . . . . . 4-17 1 4.5.2.3 Tolerances in lattice Spacing . . . . . . . . . . . . . . 4-17 l 4.5.2.4 Stainless Steel Thickness Tolerances ......... 4-17 4.5.2.5 Fuel Enrichment and Density Tolerances . . . . . . . 4-17 4.5.2.6 Water Gap Spacing Between Modules ........ 4-18 Hohoc beer-e.a-al i Report HI-971769
I. I I
i 1
4.5.2.7 Eccentric Fuel Positioning . . . . . . . . . . . . . . . . 4-18 4.6 Abnormal and Accident Conditions . . . . . . . . . . . . . . . . . . . . . . . . 4-19 4.6.1 Temperature and Water Density Effects . . . . . . . . . . . . . . . . . 4-19 4.6.2 Lateral Rack Movement . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-19 4.6.3 Rack Gap Changes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-20 4.6.4 Abnormal Iecation of a Fuel Assembly . . . . . . . . . . . . . . . . . 4-20 4.6.5 Dropped Fuel Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . 4-21 4.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-24 Chapter 4 Appendix A: Benchmark Calculations 5.0 THERM AI,HYDRAUUC CONSIDER ATIONS . . . . . . . . . . . . . . . . . . . . . 5-1 5.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 !
5.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-2 1 5.3 Discharge? Cooling Alignment Scenario .....................5-3 5.4 Decay Heat Imad Limit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5 5.5 Margin Against Boiling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-8 .
5.6 Imcal Pool Water Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-9 1 5.6.1 Local Temperature Evaluation Methodology . . . . . . . . . . . . . . . . . . 5-10 ;
5.7 Fuel Rod Cladding Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . 5-13 5.8 Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 15 5.8.1 Decay Heat Imad Limits . . . . . . . . . . . . . . . . . . . . . . . . . . 5-15 5.8.2 Time to Boil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 16 i 5.8.3 local Water and Fuel Cladding Temperatures . . . . . . . . . . . . . 5-17 '
5.9 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 18 6.0 STRUCTURAIJRRISMIC CONSIDERATIONS . . . . . . . . . . . . . . . . . . . . . 6-1 6.1 Introduction . . . . .......... ................. ........ .. 6-1 6.2 Overview ofRack Structural Analysis Methodology .... . . . . . . . . . . . . 6- 1 6.2.1 Background of Analysis Methodology . . . . . . . . . . . . . . . . . . . . . 6-2 6.3 Description of Racks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-5 6.3.1 Fuel Weights . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-6 6.4 Synthetic Time-Histories . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-6 6.5 WPMR Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-7 6.5.1 Model Detalis for Spent Fuel Racks . . . . . . . . . . . . . . . . . . . . . . . . 6-8 6.5.1.1 Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-8 6.5.1.2 Element Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-9 6.5.2 Fluid Coupling Effect . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-11 6.5.2.1 Multi-Body Fluid Coupling Phenomena . . . . . . . . . . 6-12 6.5.3 Stiffness Element Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-13 6.5.4 Coefficients of Friction . . . . . . . . . . . . . .................... 6-14 j 6.5.5 Governing Equations of Motion . . . . . . . . . . . . . . . . . . . . . . . . . . 6-15 l
6.6 Structural Evaluation of Spent Fuel Rack Design . . . . . . . . . . . . . . . . .. 6-16 6.6.1 Kinematic and Stress Acceptance Criteria . . . . . . .... .... 6-16
! 6.6.2 Stress Limit Evaluations . . . . . . . .... .... ... .... 6-16 Hohec Internationni ii , Report HI-971769
6.6.3 Dimensionless Stress Factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-20 6.6.4 Loads and LW Combinations for Spent Fuel Racks . . . . . . . . . 6-21 6.7 Parametric Simulations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-22 6.8 Time History Simulation Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-23 6.8.1 Rack Displacements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-23 6.8.2 Pedestal Vertical Forces . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-26 6.8.3 Pedestal Friction Forces . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-27 6.8.4 Rack Impact Loads ...................................6-27 6.8.4.1 Rack to Rack Impacts . . . . . . . . . . . . . . . . . . . . . . . 6-28 6.8.4.2 Rack to Wall Impacts . . . . . . . . . . . . . . . . . . . . . . . 6-28 6.8.4.3 Fuel to Cell Wall Impact Loads . . . . . . . . . . . . . . . . 6-29 6.9 Rack Structural Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-30 6.9.1 Rack Stress Factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 ~ 0 6.9.2 Pedestal Thread Shear Stress . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-32 6.9.3 Local Stresses Due to Impacts . . . . . . . . . . . . . . . . . . . . . . . . . . 6-33 6.9.4 Anaeanment ofRack Fatigue Margin . . . . . . . . . . . . . . . . . . . . . 6-34 6.9.5 Weld Stresses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 6 6.9.6 Bearing Pad Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 8 6.10 Level A Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 9 6.11 Hydrodynamic Loads on Pool Walls .............................6-41 6.12 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-42 7.0 FtKI HANDLING AND CONSTRUCTION ACCIDENTS . . . . . . . . . . . . . . . . . . 7-1 7.1 Intr ~W . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7- 1 7.2 Description of Fuel Handling Accidents . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.3 Mathematical Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-3 7.4 Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-4 7.4.1 Shallow Drop Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-4 7.4.2 Deep Drop Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-4 7.5 Gate Drop Scenario . . . . . . . . . . . . . . . . . . ..... .................. 7-5 7.6 Rack Drop . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-6 7.7 Cask Drop . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-6 7.8 Closure : . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-7 7.9 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-7 8.0 FUEI POOL STRUCTURE INTEGRITY CONSIDERATIONS . . . . . . . . . . . . . . 8-1 8.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8- 1 8.2 Description ofPool Structures . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... 8-1 8.3 Definition of Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-2 8.3.1 Static Loading (Dead Loads and Live Loads) . . . . . . . . . . . . . . . . . 8-2 8.3.2 Seismic Induced Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-2 8.3.3 Thermal Loading . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-3 8.3.4 Pool Water Loading . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-3 8.4 Analysis Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-3 8.4.1 Finite Element Analysis Model . . . . . . . . . . . . . . . . . . . . . . . 8 -3 Hohec later=*w-l iii Repmt Hl.971769
i 1
8.4.2 Analysis Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-4 8.4.3 Load Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-6 8.5 Results of Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-7 ,
8.6 Pool Liner . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-8 l 8.7 Administrative Controls on Fuel Storage . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-9 8.7 Conclusions . . . . . . . . . . . . . . . . . . . . . . ....................... . 8-10 8.8 References . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8- 10 9.0 RADIOLOGICAL EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 9.1 Solid Radwaste . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9- 1 9.2 Gaseous Releases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 9.3 Personnel Exposures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 ,
9.4 Anticipated Exposure During Reracking . . . . . . . . . . . . . . . . . . . . . 9-3 i l
10.0 INSTALLATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10- 1 10.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10- 1 10.2 Rack Arrangement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-4 10.3 Pool Survey and inspection . . ........ ......................... 10-5 10.4 Pool Celiag and Purification . . . . . . . . . . . . . . . . . . . . . . ............. 10-5 10.4.1 Poo1 Cooling . . . . . . . . . . . . . . . . . . . . . . . . . .. ............. 10-5 10.4.2 Purification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... 10-6 10.5 Fuel Shuffling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-6 10.6 Removal and Decontamination of Existing Racks and Associated Structures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-6 10.7 Installation of New Racks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-8 10.8 Safety, Health Physics, and ALARA Methods . . . . . . . . . . . . . . . . . . . . . 10-9
- 10. 8.1 Safety . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-9 10.8.2 Health Physics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-9
- 10. 8.3 ALARA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10- 10 10.9 Radwaste Material Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-11 11.0 ENVIRONMENTAL COST / BENEFIT ASSESSMENT . . . . . . . . . . . . . . . . . . . . . Il-1 11.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I l -l '
11.2 Imperative for Reracking . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-1 11.3 Appraisal of Alternative Options . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-1 11.3.1 Alternative Option Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-6 l 11.4 Cost Estimate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 -6 11.5 Resource Commitment . . . . . . . .................. .............. Il-7 11.6 Environmental Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-8 11.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-8 Hohec leer ==ha-t iv Report HI-971769 l
t
1.0 INTRODUCTION
Wolf Creek and Callaway Plants were designed and constructed with nearly identical configurations a Standardized Nuclear Unit Power Plant Systems (SNUPPS). The Spent Fuel Pools (SFPs) at both plants do not offer enough storage capacity to allow continued plant operations to the end of current linenman while still maintaining ibil core ofBond capabilities *Ihis report was prepared to support a license ==and==t for raracking the pools at both plants The similarities of the two plants has allowed for ajoint esort to design, analyze, construct, and '==m11 new marin== density spent fuel storage racks with nearly identical configurations The di-a=ia== and resuhs provided herein pertain to both planen, unless otherwise noted.
The Callaway Plant (CP) is a single unit pressurized water reactor installatian located 10 miles southeast of the city of Fulton in Callaway County, Minarmri, and 80 miles west of the St. Louis metropolitan area. The CP ipeallatian is owned by the Union Electric Company in St. Louis, ;
Missouri. l l
The Callaway Final Safety Analysis Report (FSAR) was submitted to the NRC in support of the application by Union Electric for a Class 103 license to operate a nuclear plant facihty. Union Electric received a low power (< 5%) license to operate Callaway in June 1984. The full power license was issued in October 1984, and full commercial operation began in April 1985.
The Wolf Creek Generating Station (WCGS) is a single unit pressurized water reactor l
inne=Ilatian located approximately 3.5 miles northeast of the town ofBurlington, in Coffey County, Kansas. The WCGS installatian is owned by Kansas City Power and Light Company, Kanama Electric Power Cooperative Inc., and Kanaan Gas and Electric Company. The operating agent is Wolf Creek Nuclear Operating Corporation (WCNOC).
The WCGS Final Safety Analysis Report was submitted to the NRC in support of a Class 103 license to operate a nuclear plaht facihty. The WCGS Fm* al Safety Analysis Report has been revised in accordance with 10CFR50.71(c) and is now referred to as the Updated Safety Analysis Report (USAR). The Lk-z:::s received a low power (< 5%) license to operate WCGS in March HohecInternaamaal 11 Report HI-971769
1985. The full power license was issued in June 1985, and commercial operation began in September 1985.
The new inswinnun storage rack arrays proposed for the Callaway and Wolf Creek SFPs are shown in the plan views provided by Figures 1.1 and 1.2, respectively. As may be seen, only slight diferences exist in the as built pool dimensions The initial ind=Ilatian campaign wid include the fdtoen racks shown in the larger Spent Fuel Pool (SFP). The racks shown in the Cask Loading Pit will be installed at a later date, if necessary, based on storage capacity needs The new Holtec racks are C= "; and self-supponing The principal construction materials for the new racks are SA240-Type 304L stainless steel sheet and plate stock, and SA564-630 (precipitation hardened amialaan steel) for the adjustable support spindles *Ihe only non-stainless material utilized in the rack is the neutron absorber material which is a boron carbide and aluminum. composite sandwich available under the patented product name Boral .
The new Holtee racks are designed to the stress limits of, and analyzed in accordance with,Section III, Division 1, Subsection NF of the ASME Boiler and Pressure Vessel Code [1]. The material procurement, analysis, and fabrication of the rack modules conform to 10CFR50 Appendix B requirements The rack design and analysis methodologies employed in the storage capacity expansion are a direct evolution ofprevious rerack license appik dons. This Li~aalag Report documents the design and analyses performed to demonstrate that the new Holtec racks meet all governing requirements of the applicable codes and standards, in particular, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", USNRC (1978) and 1979 Addendum thereto [2].
Sections 2 and 3 of this report provide an abstract of the design and material information on the new racks Hohoc Incarnahanal 12 Report HI-971769
- l. .
The criticality safety analysis requires that the neutron multiplication factor for the stored fuel array be bounded by the USNRC h limit of 0.95 under assumptions of 95% probability and 95%
con 6dence. The criticality safety anat.s provided in Section 4 sets the requirements on the Boral panel length and the areal B-10 density for the new high density racks Thermal-hydraulic consideration requires that fuel cladding will not fait due to excessive thermal stress, and that the steady state pool bulk temperspre will remain within the limits prescribed for the spent fbel pool to satisfy the pool stmetural s'.rength, operational, and regulatory requirements. The thermal-hydraulic analyses carried out in support of this storage expansion effort are described in Section 5.
Rack module structural analysis requires that the primary stresses in the rack module structure I
will remain below the ASME B&PV Code (Subsection NF) [1] allowables Demonstrations of seismic and structural adequacy are presented in Section 6.0. The structum! qualification also requirer ' inst the suberiticality of the stored fuel will be maintained under all postulated accident scenarios in the FSAR and USAR. The structurai consequences of these postulated accidents are evaluated and presented in Section 7 of this report.
Section 8 contains the stmetural analysis to demonstrate the adequacy of the SFP reinforced concrete structure. A synopsis of the geometry of the reinforced concrete structure is also presented in Mm 8.
The radiological considerations are documented in Section 9.0. Sections 10, and 11, respectively, discuss the salient considerations in the installation of the new racks, and a cost / bene 6t and environmental assessment to establish the superiority of the wet storage expansion option.
All computer programs utilized to perform the analyses documented in this Licensing Report are benchmarked and vaified. These programs have been utilized by Holtec International in numerous rerack license applications over the past decade.
i i
HohecIrmernahonal 13 Report HI-971769 i
I I
1 The analyses presented herein clearly demonstrate that the rack module arrays possess wide margins of safety in respect to all considerations of safety specified in the OT Position Paper [2),
namely, =ciear suberiticality, thermal-hydraulic safety, seismic and structural adaiuscy, radiolosical compliance, and nachanical integrity. ;
, )
1.1 Rafinnoses i
l l [1] ASME Boiler & Pressure Vessel Code,Section III, Subsection NF, and Appendices l (1989).
[2] USNRC, "OT Position for Review and Accy- of Spent Fuel Storage and Handling Apphcations, April 14,1978, and Addervium dated January 18,1979.
e 0
lWiecInkrnabonal 14 Report HI-971769
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2.0 OVERVIEW OF PROPOSED CAPACITY EXPANSION 2.1 IntaduGlion All storage rack arrays will consist of Dee-standing modules, made primedly from Type 304L austenitic stainless steel containing prismatic storage cells interconnected through longitudinal welds. A panel ofBoral cermet containing a high areal loading of the B-10 intope provides appropriate neutron att==tian. between adjacent storage cells. Figure 2.1.1 provides a schematic of the typical proposed storage rack module. Data on the cross sectional 7 m gross weight and cell count for each rack module in the SFP and the Cask Loading Pit are presented in Table 2.1.1.
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1hese so-caBed Region I ceus are diaanamad in more detail in Section 4.1hs remaining storage caus have enrid=nane4m:nup restrictions.
F Enriduna=*4mrnup restrictions wiu be imposed on au fuel to be stored in these three racks, unless checiamboarding storage configurations are used.
Each asw rack module is supported by four legs which are remotely t= "1 1hus, the racks can be made vertical and the top of the racks can easily be made co-planar with each other. The rack anodule support less are engineered to =====admee undulations in the pool Soor Antaa==.
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- A bearing pad interposed between the rack pedestals and the pool liner serves to diffuse the dead i
load of the loaded racks into the reinforced concrete structure of the pool slab.
The overaH design of the racks is similar to those presently in service in the spent fuel pools at sunny other nuclear piang among them Donald C. Cook of American Electric Power, and cannacelan Yankee ofNortheast Utilities. Altogether, over 50 thousand storage cous of this design have been provided by Holtec International to various nuclear plants around the world 1
Hoheclatenshonal 22 Report HI-971769
2.2 En===ev of Princiaal Danimi Criteria h key design criteria for the new spent fuel racks are set forth in the classical USNRC -
memorandda entitled "OT Position for Review and Ac~atance of Spent Fuel Storage and Handling Applications", April 14,1978 as modified by amendment dated January 18,1979. The individual mar *iana of this report expound on the specific design bases derived from the above-mentioned "OT Position Paper". A brief summary of the design bases for the racks is listed in the fouowing:
- a. Disposition: All new rack modules are required to be free-standing l
- b. Kinanatic Stability: All free-standmg modules must be kinematically stable (against tipping or overturnirng) if a seismic event (which is 150% of the poahilatad ,
OBE or 110% of the postulated SSE) is imposed on any installed module. !
- c. Structural Congplianne: All primary stresses in the rack modules must satisfy the hmits postulated in Section III subsection NF of the 1989 ASME Boiler and Pressure Vessel Code.
- d. hrmal-Hvdrantic Camal = ace: The spatial average bulk pool temperature is required to remain under 140*F in the wake of a partial ofBond.
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- f. Radiological Compliance The raracking must not lead to a violation of the off-site dose limits, or adversely affect the area dose environment as set forth in the FSAR and USAR. The radiological imnlica' ions of the installation of the new racks also need to be ascertained and deemed to be acceptable.
- g. Pool Structure: The ability of114 reinforced concrete structure to satisfy the load combinations set forth in NUREG-0800, SRP 3.8.4 must be demonstrated.
HohecInkunstenal 2-3 Report HI-971769
- h. n mek Oveh Stress F=+i==: In addition to satisfying the primary stress criteria of q
Subsection NF, the alternating local stresses in the rack structure during a seismic event are also required to be sufficiently bounded such that the " cumulative damage factor" due to one SSE and twenty OBE events does not exceed 1.0. 1
- 1. Ijner Integritv: The integrity of the liner under cyclic in-plaae loading during a seismic event must be demonstrated.
J. Baaring Pads: The bearing pads must be =&iantly thick such that the pressure on the liner naas*=== to satisfy the ACI limits during and aAer a design basis seismic event.
- k. Accident Events: In the event ofpostulated drop events (uncontrolled lowering of a fuel assembly, for instanne), it is aana==ary to daraanstrate that the shiiics.;;if of the rack structure is not -x=gdi
- 1. Construction Events The field construction services required to be carried out for executing the reracking must be demonstrated to be within the " state of proven' '
art".' ;
The foregoing design bases are further articulated in Sections 4 through 9 of this report.
maas-a..a.
2.3 A v ac The following codes, standards and practices are used as applicable for the design, construction, and assembly of the fuel storage racks. Additional specific references related to detailed analyses are given in each section.
- a. Design Codes (1) AISC Manual of Steel Construction,1970 Edition and later.
(2) ANSI N210-1976, " Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations" (contains guidelines for fuel rack design).
(3) American Society ofMechanical Ragiaaars (ASME), Boiler and Pressure Vessel Code Section III,1989 Edition; ASME Section V,1989 edition; ASME Section VIII,1989 Edition; ASME Section IX,1989 Edition; and ASME Section XI,1989 Edition.
(4) ASNT-TC-1 A June,1980 American Society for Nondestructive Testing i (Raa-a-aaa dePractice for Personnel QualiScations).
l l
Hohec laternatmaal 2-4 Report 15 971769 t
i 1
l (5) Arnerican Concrete Institute Building Code Regt.im er.tz for Reinforced - ;
Concrete (ACD IS-63) and (ACI318-71). i l
(6) Code Pp r ti for Nuclear Safety Related Concrete Structures, ACD49-85/ACD49R-85, and ACD49.1R-80.
(7) ASME NQA-1, Quality Assurance Program Requirernants for Nuclear Facilities l (8) ASME NQA-2'-1989, Quality Assurance Requirements for Nuclear Facility i Applications.
l *
(9) ANSI Y14.5M, D'= '=' ; and Tolerancing for Engineering Drawings j and Related Documentation Practices '
1 (10) ACIDetailing Manual- 1980. '
- b. Material Codes - Standards of ASTM l
(1) E165 - Standard Methods for Uquid Penetrant Ta=aaa+%
{
l (2) A240 - Standard Eg=ciAratian for Heat-Resisting Chromium and Chrmnimn-Nickel Stainless Steel Plate, Sheet and Strip for Fusion-Weided !
Un6 red Pressure Vessels. '
l (3) A262 - Da*ar+ia: Susceptibility to Intergranular Attack in Austenitic l
Stainless Steel. l l l (4) A276 - Standard Specifa;. tion for Stainless and Heat-Resisting Steel Bars l l
and Shapes '
l i (5) A479 - Steel Bars for Bo'ders & Pressure Vessels I l (6) ASTM A564, Standard Specification for Hot-Rolled and Cold-Fm* ished Age-Hardening Stainless and Heat-Resisting Steel Bars and Shapes
]
(7) C750 - Standard Specification for Nuclear-Grade Boron Carbide Powder.
l A380 - Recommended Practice for D-:e='t g Cleaning and Marking (8)
Stainless SteelParts and EqMasa==*
l (9) C992 - Standard Speci6 cation for Boron-Based Neutron Absorbing Material Systems for Use in Nuclear Spent Fuel Storage Racks.
(10) ASTM E3, Preparation ofMetallographic Specimens.
Hohec Inner =a-1 25 Report HI-971769 l l
l
(11) ASTM E190, Guided Bend Test for Ductdity of Welds.
(12) American Society ofMechanical Engineers (ASME), Boiler and Pressure Vessel Code, Section D-Parts A and C,1989 Edition.
(13) NCA3800 - Metallic Material Mandar*nrer's and Material Supplier's Quality System Program
- c. Welding Cadas ASME Boiler and Pressure Vereel Code,Section IX - Welding and Brazing fhaldicatiaan,1989 Edition.
- d. O -84 b - =ae. r" - - - p=.i_p== e .9= Raceiving. Storage. and Handling Raquirements (1) ANSIN45.2.1 - Cleaning ofFluid Systems and Anaciatad Components during Construction Phase ofNuclear Power Plants (2) ANSI N45.2.2 - Packaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants (During the Construction Phase).
(3) ANSI - N45.2.6 - Qualdicatians ofInspection, Examination, and Testing Personnel for Nuclear Power Plants (Relawy Guide 1.58).
(4) ANSI-N45.2.8, Supplementary Quality Assurance Requirements for
'="""'
= Inspection and Testing ofMachanical Equipment and Systems for the Construction Phase ofNuclear Plants.
(5) ANSI - N45.2.11, Quality Assurance Requirements for the Design of Nuclear Power Plants (6) ANSI-N45.2.12, Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants.
(7) ANSI N45.2.13 - Quality Assurance Requirements for Control of Procurement of Equipment Materials and Services for Nuclear Power Plants (Regulatory Guide 1.123).
(8) ANSI N45.2.15 Hoisting, Rigging, and Transporting ofItems For Nuclear Power Plants.
(9) ANSI N45.2.23 - Qa=EAe=+iaa of Quality Assurance Program Audit Personnel for Nuclear Power Plants (Regulatory Guide 1.146).
(10) ASME Boiler and Pressure Vessel,Section V, Nondestructive Examinatian,1989 Edition.
i Hoher Internaha==1 26 Report HI-971769
)
-(11) ANSI - N16.9-75 Validation of Cal ~Ia% Methods for Nuclear Criticality Safety.
- e. Governing NRC Domunents (1) 'OT Position for Review and AW.ce of Spent Fuel Storage and Handling Applications," dated April 14,1978, and the inadine=*ians to this daam==* ofJanuary 18,1979.
(2) NUREG 0612, " Control ofHeavy i nada at Nuclear Power Plants",
USNRC, Washington, D.C., July,1980.
(3) IE Informatian Notice 83-29-Fuel"Mg Caused byFuelRack Deformatinn
- f. Oehar ANEI hadards (ant Ha*M in the nWing (1) ANSFANS 8.1 (N16.1) - Nuclear Criticality Safety in Operations with l Fissionable Materials Outside Reactors 1
(2) ANSUANS 8.17, Criticality Safety Criteria for the Handling, Storage, and
~
i Transportation ofLWR Fuel Outside Reactors.
l (3) N45.2 - Quality Assurance Program Requirements for Nuclear Facilities -
1971.
(4) N45.2.9 - Requirements for CW=, Storage and Maintenance of Quahty Assurance Records for Nuclear Power Plants - 1974.
l (5) N45.2.10 - Quality Assurance T'erms and Definitions -1973.
1 (6) ANSUANS 57.2 (N210) - Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants.
(7) N14.6 - American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 pounds (4500 kg) or more for NuclearMaterials (8) ANSUASME N626-3, Qualification and Duties of Personnel Engaged in ASME Boiler and Pressure Vessel Code Section III, Div.1, Certifying Activities ,
(9) ANSI Y14.5M, Dimensioning and Tolerancing for Engineering Drawings and Related Documentation Practices.
HohocInternshonal 27 Report HI-971769
1 l
- g. eq.F wa.,al Ramdr.tiana (1) 10CFR20 - Standards for Protection Against Radiation.
(2) 10CFR21.-Reporting ofDefects and Non mu p .y j (3) 10CFR50 Appendix A - General Design Criteria for Nuclear Power Plams.
i4 (4) 10CFR50 Appendix B - Quality Assurance Criteria for Nuclear Power j Plants and FuelF-y--:"; Plants. I l
(5) 10CFR61 - Liransing Requirements for Land Disposal ofRadioactive !
Matwiel. ;
(6) 10CFR71 - Packaging and Transportation ofP hve Material
- h. Regulatorv Guides (Iatest revision unless specifically noted)
(1) RG 1.13 - Spent Fuel Storage Facility Design Basis (Revision 2 Proposed).
(2) RG 1.25 - Asm .A Used for Evaluating the Potential Radiological Ccc:r = of a Fuel Handling Accident in the Fuel Handling and Storage Facility of Boiling and Pressurized Water Reactors (3) RG 1.28 - (ANSI N45.2) - Quality Assurance Program Requirements . j (4) RG 1.29 - Seismic Design Classification (Rev. 3).
. (5) RG 1.31 - Control ofFerrite Content in Stainless Steel Weld Material.
(6) RG 1.38 - (ANSI N45.2.2) Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and Handling ofItems for Water.
Cooled Nuclear Power Plants.
(7) RG 1.44 - Control of the Use of Sensitized Stainless Steel.
(8) RG 1.58 - (ANSI N45.2.6) Qualincatian ofNudear Power Plant Inspection, Examination, and Testi ig Personnel.
(9) RG 1.61 - Damping Values for Seismic Design ofNuclear Power Plants, Rev. O,1973.
(10) RG 1.64 - (ANSI N45.2.11) Quality Assurance Requirements for the Design ofNuclear Power Plants. N Hohoc Insem*=1 2-8 Repat HI.971769
1 i
(11) RG 1.71 - Weider Quah6 cations for Areas ofLimited Accesmiility.
(12) RG 1.74 - (ANSI N45.2.10) Quality Assurance Terms and Definitions. !
-(13) RG 1.85 - Materials Code Case AWty - ASME Section 3, Div.1. i (14) RG 1.88 - (ANSI N45.2.9) 0 "- W+, Storage and Maintaamar* of j Nuclear Power Plant Quality Assurance Records (15) RG 1.92 - Combining Modal Responses and Spatial C =-;-n s in SaimmieResponse Analysis.
l (16) RG 1.122 - D: t6' ofFloor Design Response Spectra for Saimmin j Design ofFloor-Supported Equipment or Componands l
l (17) RG 1.123 - (ANSI N45.2.13) Quality Assurance Requirements for Control ,
of Procurement ofitems and Services for Nuclear Power Plants i l I l (18) RG 1.124 - Service Hmita and leading Combinations for Class 1 Linear-l Type Component Supports, Revision 1,1978. I l
(19) RG 3.4 '- Nuclear Criticality Safety in Operations with Finaianahle Materials at Fuels and Materials Facilities.
(20) RG 3.41 - Valid =*iaa of calentatianal Methods for Nuclear Criticality Safety, Revision 1,1977.
(21) RG 8.8 -Informatian Relative to Ensuring that C-+f -9 Badiatian Fyn at Nuclear Power Plants will be as Low as Reasonably Achievable (ALARA).
l (22) RG 8.38 - Control of Access to High and Very High Radiation Areas in l l NuclearPower Plants, June,1993.
- 1. Branch Technical Position 1
1 (1) CPB 9.1 Criticality in Fuel Storage Facilities (2) ASB 9 Residual Decay Energy for Light-Water Reactors for Long- 1 Term CW .
J. Standard Review Plan (1) SRP 3.2.1 - Seismic ClassiScation-Hohec Inter =n==al 2-9 Report HI 971769
(2) SRP 3.2.2 - System Quality Group Classi6 cation.
(3) SRP 3.7.1 - Seismic Design Parameters.
,(4) SRP 3.7.2 - Seismic System Analysis.
(5) SRP 3.7.3 Mamic Subsystem Analysis.
(6) SRP 3.8.4 - Other Seismic Category I Structures (including Appendix D),
Technical Position on Spent Fuel Rack.
(7) SRP 3.8.5 - Foundations for Seismic Category I Structures, Revision 1, 1981.
(8) SRP 9.1.2 - Spent Fuel Storage, Revision 3,1981.
(9) SRP 9.1.3 - Spent Fuel Pool Cooling and C6-aa. System.
(10) SRP 9.1.4 - Ught Load Handling System.
(11) SRP 9.1.5 - Heavy Imad Handling System.
(12) SRP 15.7.4 - Radiological Consequences ofFuel Handling AcWa
- k. AWS Standards (1) AWS D1.1 - Structural Welding Code, Steel.
(2) AWS DI.3 - Structure Welding Code - Sheet Steel.
(3) AWS D9.1 -Welding of Sheet Metal (4) AWS A2.4 - Standard Symbols for Welding, Brazing and Nondestructive Examination l (5) AWS A3.0 - Standard Welding Terms and Dennitions.
(6) AWS AS.12-Tungsten Arc-weldingElectrodes (7) AWS QCl - Standards and Guide for Quali6 cation and Certi6 cation of WeldingInspectors.
6 4
Hohec lasernahanal 2-10 Report HI-971769 1
1
2.4 G -Sv A===ae* Prnoram .,
The governing quality assurance requirements for fabrication of the spent fuel racks are a==4=*M in 10CFR50 Appendix B. The quality assurance program fm design of the racks are described in Holtec's Nuclear Quality Assurance Mammal, which has been reviewed and yy.ud.
by the Union Electric Company and WCNOC. This program is designed to provide a flexible but highly controued system for the design, analysis and licensing of enstami M components in accordance with various codes, speciscations, and regulatory require ===.
De -4 +-4 of the racks will be carried out by Holtec's t "E Ti mandarsurer (U.S.
Tool & Die, Inc.). The Quahty Assurance System enforced on the manufacturer's shop Soor provides for all controls necessary to fulfill all quality assurance requirements The QA system maintains sufEcient simplicity to make it functional on a day-to.< lay basis. UST&D has ;
nia-d=ctured high density racks for over 60 nuclear plants around the world. UST&D has been audited by the industry group NUPIC, and the QA branch ofNMSS with natisfac*ary results.
l The Quality Assurance System that will be used by Holtec to install the racks is also controlled by the Holtec Nuclear Quahty Asaurance Manual and by Union Electric Company and/or WCNOC's I site-specific requirements 2.5 Mechanical Design The rack modules are designed as cellular structures such that each fuel assembly has a square opening with conforming lateral support and a flat horizontal bearing surface The basic characteristics of the spent fuel racks are summarized in Tabla 2.5.1.
A central objective in the design of the new rack modules is to maximize their structural rigidity while minimizing their inertial mass. Accordingly, the rack modules have been designed to simulate multi-flange beam stnictures rm*g in excellent de-tuning characteristics with respect Hoheclaternahonal 2-11 Report HI-971769
l to the applicable seismic events The next subsection presents an item-by-item description of the anatomy of the rack modules in the context of the fabrication methodology.
2.6 Rack Fabrication The object of this section is to provide a self-contained description of rack module construction to enable an !='+; 'r appraisal of the adequacy of design. A list of applicable codes and standards were pra===*=d previously in Section 2.3 of this report.
2.6.1 Fabrication Otsactive The requirements in manidacturing the high density storage racks may be stated in four interrelated points-
- 1. The rack maailan are fabricated in such's manner that there is no weld splatter on the storege cell surfaces which would come in contact with the fuel assembly. ,
- 2. The storage locations are constructed so that redundant flow paths (i.e., located at the base of at least two cell walls and/or through the baseplate) for the coolant are availmW .
- 3. The fabrication process involves operational sequences which permit immediate verification by the inspection staff.
- 4. The storage cells are connacfad to each other by austenitic stainless steel corner welds which result in a is.gw.4, lattice construction. The extent of welding is selected to "detune" the racks from the stipulated seismic input ipeed by the time history accelerograms 2.6.2 Anatomy of the PWR Rack Module ne composite box subassembly, the bueplate, and the support le8S constituto the Principal componenen of the fuel rack modules. The following description provides details of all of the major rack components.
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Hohoclaternmanal 3-7 lbport HI 971769
3.6 References
[3.1.1] NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants," July 1980.
[3.3.1] " Nuclear Engineering International," July 1997 issue, pp 20-23.
[3.3.2] " Spent Fuel Storage Module Corrosion Report," Brooks & Peridra Report $$4, June 1,1977.
[3.3.3] " Suitability ofBrooks & Perkins Spent Fuel Storage Module for Use in PWR Storage Pools," Brooks & Perkins Report 578, July 7,1978.
[3.3.4] "Boral Neutron Absosting/ Shielding Material - Product Performance Report," Brooks
& Perkim Report 624, July 20,1982.
[3.5.1] CMMA Speci6 cation 70, " Electrical Overhead Travelling Cranes," Crane Manufacturers Association ofAmerica,Inc.,1983.
[3.5.2] ANSI N14.6-1978, Standard for Special Lifting Devices for Shipping Containers Weighing 10000 Pounds or more for Nuclear Materials," American National Standard Institute, Inc.,1978.
[3.5.3] ANSI /ASME B30.2, " Overhead and Gantry Cranes, (Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist)," American Society ofMechanical Engineers,1983.
[3.5.4] ANSI /ASME B30.20, "Below-the-Hook Lifting Devices," American Society of '
Mechanical Engineers,1993.
/
Hoheclaternahonal 3-8 Report HI 971769
Tatde 3.3.1 BORAL EXPERIENCE LIST-FWRs Plant Utility Docket Mfg.
No. Year
) 1 maser Yamisse nessee Yanime AnouncPower 50-309 1977 Donald C. Cook immens s Musamma Decenc 5G-315816 1979 Seemorah 1.2 Tamasesse Valley Autanty 50-3275 28 1979 Selsen 1.2 Pubhc ServiceElecenc & Gas 50 272/311 1980 l Zica1.2 -
Pa==a=weakh FA== Co. 50-295/304 1980 Baussante 1.2 Tammessee Valley Ausbanty 50-438/439 1981 Yankee Rowe Yambec AtonnePower 50-29 1964/1983 lases Point 3 NY Power /atbanty 50-286 1987
)
Byrom 1.2 r===a= wealth FA== Co. 50-454/455 1988 Brandwood 1.2 ca==a= wealth FAnan Co. 50-456/457 1988 Yankee Rowe Yambee Atomic Power 50-29 1988 "Ilirse h4ileIsland I GPU Nuclear 50-289 1990 Seenoyah(rerack) T- Valley Authonty 50-327 1992 Donald C. Cook (rarack) Amences Electne Power 50-315/316 1992 Besvar ValleyUnit 1 Duquesne Lisht Company 50-334 1993 Fort Calhoun Omaha Pubhc PowerDuenot 50-285 1993 Zion 1 & 2(serack) Ca==anwealth E& ace Co. 50 295/304 1993 SaleenUnits 1 & 2(rerack) Pubhc Gas andElectne Company 50-272/311 1995 Haddam Neck cannarvia* Yankee Atoauc Power Company 50-213 1996 Gosana KernkraRwerk Gossm-Deniken AG (Switzerland) - 1984 Kosbers 1.2 ESCOM(South Aince) - 1985 Bauman 1.2 Noniostschweeenache KreRwerke AG (Switaarland) - 1985 12 vanousPlants Elemencate deFrance(France) - 1986 Ulobin Unit i Kores Elecenc Power ra-pany(Kores) - 1995 U$lunUnit 2 Korus Electne Power Company (Kores) - 1996 Kon-4 KoreaElecencPower Company (Korea) - 1996 Yommawaas1.2 Kores Electne Power Company (Kores) - - 1996 SeewellB Nedeer Eleotnc, plc (United 16narinm) - 1997
- a1 Furnas C-*ans F'-ln, s SA (BraziD
. - 1997 Hohec laternaha==I 3-9 Report HI.971769
i Table 3.3.2 BORAL EXPERIENCE LIST- BWRs N c: UGty Dmlet No. Mfs. Year fanpar, ._,,__ , , _ , _ , _ Nebraska Public Power _ _ 50-298 _ 19L J.A. FitzPatrick NY Power Authonty 50-333 1978 Duane Arnold Iowa Electric Lisht & Power 50-331 1579 Browns Ferry 1,2,3 Temmessee Valley Authonty 50- 1980.
259/260/2 %
Bmnswuk1,2 CarchnaPower& Lisht 50-3245 25 1981 Choson flhanse Power 50-461/462 ' 1981 Dresden 2,3 en==nt=weshh Edison Company 50-237/249 1981 E.I. Haka 1,2 Gearspa Power 50-321G66 1981 Hope Creek Pubhc Serwoe Electnc & Gas 50-354/355 1985 Humboldt Bay Pacific Gas & Electnc Company 50-133 1985 Lacrosse Dairyland Power 50-409 1976 l
Lonerick 1,2 P61adelphia Electnc Company 50-352G53 1980 Manteella Northen States Power 50-263 1978 Pescabosom 2,3 P61.,lelpius Electnc 50-277/278 1980 Peny 1,2 Cleveland Electne Ill>=aiadias 50-440/441 1979 Pilgrim Boston Edison Company 50-293 l'978 Susquahanna 1,2 Pennsylvania Power & Light 50 387,388 1979 Vermont Yankee Vermont Yankee Anneme Power 50-271 1978/1986 Hope Creek Public SerwoeElecenc& Gas 50 354055 1989 Shearon Hams Pool B Carolina Power & Light 50-401 1991 Duane Arnold lows Eledne Light & Power 50-331 1993 Pilsnm Boston Edison Company 50-293 1993 f asalle 1 enmenanwealth Fehann Company 50-373 1992 Mal *=a Unit 1 Northeast Utilities 50-245 1989 James A.FitzPatrick NY Power Authority 50-333 1990 Hope Cred Public Serwoe Elechc & Gas Company 50-354 1991 Duane ArnoldEnergy Iows Elceic' Power Company 50-331 1994 Center Hohec Internatxmal 3-10 Report HI-971769
Table 3.3.2 BORAL EXPERIENCE IJST-BWRs Plant Utility Du+ad_ No. Mfg. Year isnarkt Ud.: 1,2 PECOimargy 50-352/50- 1994 353 i ShenronHama PoolW Camhna Power & T i* Cm
_ 50 401 1996 Chin =han 1,2 Taiwan Power Company (Taiwan) - 1986 Y=ashamw 1,2 Taiwan Power Cm(Taiwan) - 1991 IAgene Verde 1,2 Counnon FederaldeElectricidad - 1991 (Memoo) i I
l l
4 HoltecInternanonal 3-11 Report HI-971769
2- . _ . . _z __ _ ,-_
. Table 3.3.3 1100 alloy ALUMINUM FHYSICAL CHARACIERISTIO Density 0.098lbref 2.713 deaf Melting Range 1190'F. 1215'F 643' . 657'C nennel Conductivity (77'F) 128 BTUAr/A8/F/A 0.53 cal /sec/an2 /'C/cm Camdkind ofDermalEwomasian 13.1 x 104 inrn>*F (6C'F . 212'F) 23.6 x 104cm/cm *C __ ,
Speedic Heat (221'F) 0.22 BTU /lb/*F 0.23 cal /s/*C Modulus of Bastusty -
10 x 10'asi Tensile Strength (75'F) 13,000 psi (annealed) 18,000 psi (m rolled)
Yisid Strength (75'F) 5,000 psi (annealed 17,000 psi (as rolled)
Elongation (75'F) 35-45%(annealed) 9-20%(as rolled)
Hanhems(Bnnell) 23 (annealed) 32(as rolled)
AansahosTamparature 650'F 343*C :
Hobsclaternshonal 3 12 Report HI 971769 I
i
i I .
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Tatde 3.3.4 l
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CHEMICAL COMPOSITION - ALUMINUM i I
(1100 ALLOY)
~
99.00% min. Aluminum 1.00% max. Silicone and Iron 0.05-0.20% max. Copper 0.05% max. Manganese 0.10% max. Zinc 0.15% max. Other O
h i
HohecInternahonal 3-13 Report HI-971769
r ]
l l
Tshk 3A5 1
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CHDHCAL COMPOSITION AND PHYSICAL PROPERTIES
! OF BORON CARBIDE i
! c, I
CHEMICAL COMPOSITION (WEIGHT PERCENT) l l Total boron 70.0 min.
l B" isotopic content in natural boron 18.0 t
l Boric oxide 3.0 max. ,_
l Iron 2.0 max.
l Total baron plus total carbon 94.0 min.
PHYSICAL PROPERTIES
! Chemical formula B,C Baron content (weight percent) 78.28 %
Carbon content (weight percent) 21.72 %
Crystal structure rhombohedral l Density 0.0907 Wm' 2.51 g/cm8 Melting Point 4442"F 2450'C Boiling Point 6332'F 3500*C l
l
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Hobsc Inser==hanal 3-14 Report HI 971769 l
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O Hdtoc hdernahonal 3-15 Report 15971769
i i
4.0 CRITICALITY SAFETY ANALYSES 4.1 DESIGN BASES The high density spctit fWI storage racks for the Cillsway asl Wolf CNek Nuclear P6wei Plants are designed to assure that the effective neutmn multiplication factor (k,) in the spent nuclear fuel pool and cask loading pit is equal to or less than 0.95 with the racks fully loaded with fuel of the highest anticipated recctivity, and flooded with unborated water at the temperature within the operating range curisponding to the highest reactivity. The spent fuel storage racks are designed to acconanodate any and all of the following Westinghouse fuel assembly types:
T 17x17 OFA ,17x17 Standard, and 17x17 Vantage SH (V5H), with a maximum nominal initial enrichment of 5.0 wt% 2"U and a minimum of 16 Integral Fuel Burnable Absorber (IFBA)
(1.5x) rods. Additional restrictions are specified to allow the storage of the aforementioned l
fuel assembly types without IFBA rods. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including mechanical tolerarces. All uncerta'mties are s tatistically combined, such that the final k, will be equal to or less than 0.95 'with a 95 %
probability at a 95% confidence level. Enrichments less than 5.0 wt% 2"U are also evaluated, and soluble boron concentrations necessary to protect against postulated accidents are determined.
l Applicable codes, standards, and regulations or pcdkut sections thereof, include the following: ;
- General Design Criteria 62, Prevention of Criticality in Fuel Storage and Handling.
- USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage, ;
Rev. 3 - July 1981
- USNRC letter of April 14,1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, including '
modification letter dated January 18,1979. !
T The OFA designation is used generically throughout this section and includes V-5 and V+ fuel.
Holtec International 4-1 Report HI-971769
l I
USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2 (proposed), December 1981.
ANSI ANS-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuct Out:!de Ret.ctors.
USNRC guidelines and the applicable ANSI standards specify that the maximum effective multiplication factor, k, , including bias, uncertainties, and calculational statistics, shall be less than or equal to 0.95, with 95% probability at the 95% confidence level.
1 To assure that the true reactivity will always be less than the calculated maximum reactivity, the following conservative assumptions were made: 1 Moderator is unborated water at a temperature that results in the highest reactivity l (4 C, corresponding to the maximum possible moderator density).
No soluble poison or control rods are assumed to be present for normal operations, although the additional margin due to the presence of soluble boron is identified.
The effective multiplication factor of an infinite radial array of fuel assemblies was used except for the assessment of peripheral effects and certain abnormal / accident conditions where neutron leakage is inherent.
Neutron absorption in minor structural members is conservatively neglected, i.e.,
spacer grids are replaced by water.
Depletion calculations assume conservative operating conditions; highest fuel and moderator temperature and an allowance for the soluble boron concentrations during in< ore operation.
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4.2
SUMMARY
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4.3 REFERENCE FUEL STORAGE CELLS 4.3.1 Reference Fuel Acaembly I l
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4.4 ANALYTICAL METHODOLOGY 4.4.1 Reference Design Calculations 4 L
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1 1 l 4.5 CRITICALITY ANALYSES AND TOLERANCES i 4.5.1 Nominal Desien 1 i
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1 4.5.2.2 Boral Width Tolerance
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Ilottec Intemational 4-17 Report HI-971769
4.5.3 Water-Gap Spacing Between Modules
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4.6 ABNORMAL AND ACCIDENT CONDITIONS 4.6.1 Temnerature mM Water Denairy Effacte
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4.7 REFERENCES
I 4.4.1 L.M. Petrie and N.F. Landers, " KENO Va - An Improved Monte Carlo Criticality ! Program with Supergrouping," Volume 2, Section Fil from " SCALE: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation" NUREG/CR 0200, Rev. 4, January 1990. / L 4.4.2 J.F. Briesmeister, Editor, "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A," LA-12625, Los Alamos National Laboratory (1993).
'4.4.'3 N.M. Greene, L.M. Petrie and R.M. Westfall, "NITAWL-II: Scale System Module for Performing Shielding and Working Library Production," Volume 1, Section F1 from
" SCALE: A Modular System for Performing Standardized Computer Analysis for l Licensing Evaluation" NUREG/CR-0200, Rev. 4, January 1990.
1 4.4.4 M.G. Natrella Experimental Statistics. National Bureau of Standards Handbook 91, l August 1%3. 4.4.5 A. Ahlin, M. Edenius, and H. Haggblom, "CASMO - A Fuel Assembly Burnup l Program", AERF-76-4158, Studsvik mport. 4.4.6 A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion Code for LWR Analysis", ANS Transactions, Vol. 26, p. 604,1977. 4.4.7 M. Edenius, A. Ahlin and B.H. Forssen, "CASMO-3 A Fuel Assembly Burnup l Program User's Manual," Studsvik/NFA-86/7, Studsvik Energitechnik AB, November 1986. 4.4.8 S.E. Turner, "Waterford Criticality Analsysis," HI-961562,1996. I
. 4.4.9 M. Edenius and A. Ahlin,."CASMO-3: New Features, Benchmarking, and Advanced l Applications," Nucl. Sci. Eng., 100 (1988)
{ i 4.4.10 S.E. Turner, " Uncertainty Analysis - Burnup Distributions", presented at the i ! DOE /SANDIA Technical Meeting on Fuel Burnup Credit, Special Session, ANS/ ENS l Conference, Washington, D.C., November 2,1988 l 1 Holtec International - 4-23 Report HI-971769
i i l Table 4.2.1 Summary of the Criticality Safety Analyses for the MZTR Storage Configuration u (.. c
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Table 4.2.2 Summary of the Criticality Safety Analyses for the Interim Checkerboard Storage Configuration
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i Table 4.2.3 Reactivity Effects of Abnormal and Accident Conditions
' Abnormal / Accident Conditions Reactivi Effect i
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i l Table 4.3.1 i Design Basis Fuel Assembly Specifications l Fuel Rod Data Assembi OFA Standard Vanta e-5H l I i Fuel Assembi Data llottec International 4-27 Report HI-971769
Table 4.5.1 Reactivity Effects of Manufacturing Tolerances Tolerance Reactivi Effect!Ak E Holtec Intemational 4-28 Repon 111-971769
Table 4.6.1 Reactivity Effects of Temperature and Void i Reactivi Effect, Ak l l Holtec International 4-29 Report HI-971769
9 6 7 K 1 E 7 9 E I R H C F L ' OE WG A DR NO N AT O S I Y T AN A WO M AI R LG . O LE F AR N C I E RE Y OR R FH A T T T UE E ON I R YO P AZ O L D R LE P OX OI PM 1 2 4 E R U G _ I F _ 4
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5.0 THERMAL-HYDRAULIC CONSIDERATIONS 5.1 lataduction This section provides a summary of the M-:-is, models, analyses, and numerical results to demonstrate the compliance of the reracked twin site Spent Fuel Pool and Cask Loading Pit and the Spent Fuel Pool CW and Cleanup System (SFPCCS) with the provisions of Section1II of the USNRC "OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling l l Applications", (April 14,1978). Similar methods of thermal-hydraulic analysis have been used in ' other reracklicensing pi@. The thermal-hydraulic qualification analyses for the rack arrays may be broken down into the following categories:
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As stated previously in this licensing report, the Callaway and Wolf Creek spent fuel pools and cask loading pits are nominally identical. A survey of the pools by Holtec personnel showed that the actual pool cavity dimensions of the twa plants deviates slightly (less than fourm ' ches). Accordingly, to provide bounding thermal-hydraulic calculations, the pool water volume is conservatively based on the minimum east-west and north-south dimensions of the two pools. Hohec h l - 5-1 Repost HI-971769 i j
Thus, a lower bound thermal inertia and outer periphery downcomer dimension is used in the l thermal-hydraulic calculations , The following sections present the plant system description, analysis assumptions, a synopsis of l the analysis methods employed, and final results. Hereinafter, the tenn " plant" used in this section refers to both Callaway and Wolf Creek.- / l 5.2 System Description The Spent Fuel Pool Cooling and Cleanup System (SFPCCS) at Callaway and Wolf Creek consist of two cooling trains, a cleanup loop, and a surface skimmer loop. The fuel pool cooling system consists of two 100% capacity cooling trains for the removal of decay heat generated by irradiated fuel stored in the spent fuel pool and cask loading pit. Each train consists of a horizontal centrifugal pump, a shell, and U-tube heat excha a, a strainer, manual valves, and the instrumentation required for system operation. The decay heat generated by the stored fuel in the pool is transferred from the fuel pool cooling system through the fuel pool cooling heat exchangers. The fuel pool cooling heat exchangers are serviced by the component cooling water system on the shell side with remote manual-operated isolation valves provided. During normal system operation, one fuel pool cooling pump takes suction from the spent fuel pool and transfers the pool water through a fuel pool cooling heat exchanger back to the spent fuel pool. The fuel pool cooling pump suction is protected by a permanent strainer located at the terminal end of the suction piping within the spent fuel pool. The pump suction line penetrates the spent fuel pool wall, near the normal spent fuel pool water level. The return line terminates at the bottom of the spent fuel pool. In order to prevent the draining of the spent fuel pool by siphoning action, an antisiphon hole is located in each return line, near the surface of the pool water. I l Hohec Inese-a-1 5-2 Report HI-971769 l
Nonnal makeup water to the spent fuel pool is supplied by the reactor makeup water system. An ahernate source ormakeup water is the RWST via the fuel pool cleanup pumps. Emergency makeup water is supplied from the Easential Service Water system Boron addition to the spent fuel pool is normally accomphshed by supplying borated water from the boric acid tanks via the boric acid blending tee. Boron may also be added by using the RWST as the source of makeup water to the spent fuel pool. Isolation of non-safety related portions of the SFPCCS is a ma6ual action. The fuel pool cleanup system contains two inhne centrifugal pumps and two filters in parallel, a mixed bed demineralizer, and a wye-type strainer. The pumps and filters are designed for fifty-percent of the system capacity, and the demineralizer and strainer are designed for one-hundred percent system capacity. The demier zer removes ionic corrosion impurities and fission products. 'the filters are provided nove particulate matter which would have othenvise entered the domineralizer, and the wye strainer downstream of the domineralizer removes resin fines which may be released from the resin bed. The fuel pool cleanup system provides the capability for purification of the water in the spent fuel pool, the cask loading pit, the transfer canal, the refueling pool, and the RWST. The cleanup system is an essential adjunct to the SFPCCS system to maintain clarity and water chemistry controlin the spent fuel pool. 5.3 Discharge /Coohng Alignment Scenario Consistent with the current plant practice, two discharge scenarios are postulated. They are:
- 1. partial core ofBond ii. full-core ofBond Hohne bese==*-1 5-3 Report HI-971769
/
l L . In lieu of prescribing a batch size and cooling period for partial core ofBoad, the twin plants seek to determine the maximum pool heat load resulting in a steady state bulk pool temperature limit of 140*F under this scenario with only one cooling train operating. Similarly, the fbil core of5ond scenario is required to be executed such that the maximum pool heat load will not allow for bulk pool boiling at the end of a postulated 2 hour lou of forced 4 cooling transient which occurs immediately after the full core ofBond. More speci6cally, the bulk water temperature is sought to be limited to 207'F (which *meludes 5'F of raargin) after two hours of pool heat-up in the absence of all forced cooling paths. Evaluation of these two scenarios will allow maximum flexibility in batch sizes and cooling periods prior to ofBond into the pool. In both scenarios, the component cooling water used to remove heat from the spent fuel cooler is assumed to be at its maximum design temperature. Durir6 the partial core ofBond scenario CCW flow is assumed to be at its nominal rate. During Full Core OfBoad conditions CCW flow is assumed to be at its design basis flow rate. With the thermal effectiveness of the spent fuel pool cooler thus fixed, the requirement of the ceiling on the bulk pool temperature essentially translates into a limit on the total heat generation rate in the pool. An additional evaluation is performed for a loss of cooling accident occurring some time after restart. This evaluation considers a four hour long loss of forced cooling in the SFPCCS followed by a twenty hour long period with cooling provided at one-half the normal coolant flow rate. It j I must be demonstrated that the spent fuel pool does not reach the bulk boiling temperature during the 24 hour period. For this evaluation, the component cooling water to the heat exchanger is l assumed to be at an elevated temperature and reduced flow rate. l l l
- 1. Hohse h l 5-4 Report HI-971769 I
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5.5 Margin Against BoilinF l As stated previously, it is necessary to ensure that the pool bulk temperature will remain less than 207'F (i.e., adequate margin against boiling) if: (1) all forced cooling paths are lost following a l full core ofBond and cooling is not restored for two hours, and (2) a los of coolant accidenf 1 occurs after restart and partial cooling is restored after 4 hours. The SFPCCS system has two independent trains, both of which are seismically qualified and safety-related, so a complete loss i of forced cooling is not possible under single failure criteria. Regardless of this fact, these evaluations are performed for postulated non-mechanistic loss of forced cooling accidents. ! The following conservatisms are applied in the heat-up calculations.
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maximize the pool temperature increase rate and minimize the corresponding time-to-boil. As a transient decay heat load would necessitate a reduced average net heat load, the steady-state assumptions are indeed conservative.
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This differential equation is solved using a numerical solution technique to obtain the bulk pool temperature as a function of time. The major input values for these analysis are sun-wa in Table 5.5.1. 5.6 I-1 Pool Wa*- T-a-ture In this section, a summary of the methodology for evaluating the local pool water temperature is presented. A single conservative evaluation for a bounding arnalgam of conditions is performed. The result of this single evaluation is a bounding temperature difference between the maximum local water temperature and the bulk pool temperature. In order to determine an upper bound on the maximum local water temperature, a series of conservative assumptions are made. The most important of these assumptions are:
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