ML20214Q466
| ML20214Q466 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 04/30/1986 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20214Q461 | List: |
| References | |
| 8601--84, 8601--84-DRFT, 8601-00-0084, 8601-00-0084-DRFT, NUDOCS 8609240369 | |
| Download: ML20214Q466 (78) | |
Text
{{#Wiki_filter:. b COMMONWEALTH EDISON LA SALLE STATION UNIT 1 SPENT FUEL STORAGE CAPACITY MODIFICATION SAFETY ANALYSIS REPORT APRIL 1986 8601-00-0084 9 8609240369 860520 4 DR ADOCK 0500
LA SALLE SPENT FUEL STORAGE MODIFICATION TABLE OF CONTENTS Revision 3 Pages 1.0 Introduction 1-1 2.0 Requirements for Thermal-Hydraulic Analysis 2-1 2.1 Introduction 2.2 Approach to Thermal-Hydraulic Analysis 2-6 2.3 Detailed Analysis and Results 2-7 2.4 Conclusions 2-21 3.0 Criticality Analysis 3-1 3.1 Analytical Technique 3.2 Calculational Approach , 3-5 3.3 Evaluation of Criticality Safety 3-6 3.4 Tolerances and Uncertainties ~ 3-7 3.5 Accident Analysis 3-9 3.6 Design Conservatisms 3-10 3.7 New Fuel Designs 3-11 3.8 References 3-12 to 3-28 4.0 Seismic Analysis 4-1 4.1 Introduction 4.2 Equipment Description and Material Properties 4.3 Summary of Results 4-2 4.4 References 4-3 4.5 Data Preparation 4-4
.i . 0 Mach 2nical Analygio 5-1 5.1 Summary 5.2 Description of New Spent Fuel Racks 5.3 Mechanical Analysis 5-7 5.4 References 5-18 l
- e
r 1.0 Introduction Commonwealth Edison is currently acquiring high density spent fuel racks to replace the racks supplied by the NSSS supplier of a low density design. This Safety Analysis is provided to support Commonwealth Edison's request for NRC review and approval of new spent fuel racks for La Salle County Station Unit 2. . . There are two spent fuel pools at La Salle Station, the existing racks in each of these pools have 1080 storage cells. In the 1989-90 time frame, they will no longer have Full Core Discharge Reserve. Replacingtheserackswithhighdensityracks(Figure 11)willextendthestoragecapacity ~ into the year 2000. The high density neutron absorber racks are seismically designed with no non-load bearing parasitic structures. The basic rack consists of precision made boxes welded together in rows and columns to form a highly damped honeycomb structure. The neutron absorbing material is trapped between box walls and enclosed on all sides. This Safety Analysis complies with the NRC position paper "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, dated April 14, 1978, as amended by the NRC letter dated January 18, 1979. The storage racks proposed for La Salle are similar to the Nine Mile Point 2 racks which were recently licensed. The purpose of this report is to provide a description and the technical information necessary for evaluation of the safety aspects for the proposed modification to the La Salle Unit 2 spent fuel storage pool. 1-1
T 2 N FIGURE 1.1 18 CONTROL ROD STORAGE
- l F-- :; i i I 1_=* -- -
ii i : enii_ 1 1 I 1 I 11 1 1 II I I I. I . I - I. _
.1 i I
! !: j '. i _MH
!1I i I ;
+4H--H;
-.I T I I .
- 234 l - 240 d 240 -
~
it : 217 0 +-: M [tf.lSx16) <-.-- 11 :- f I E x16-2 3 k15x16-6)E. rt f r f f rm T' - 1:}ti'l5x16)
- t. rrM*<"+1 t . i r i .t *! -*
l 1!. ii: ii? S ii ! i ii_ ' -j
> 1 1 : i .1 1. i 1 3 :1
*s
.. . . - .3- .- w. -
I'l . x I . .)iiri!i *
**tt' 'IEii i11 . i::,1
;o2 ja g(15xl' i 1 1 1: 1 1 ..
235 2; 225 "g 195 ,+ jg(15x17- 's;- t (15x15) ':n (15x12t ::::. ,y, ~
- 20) - - -- hii - -i -~E* ~ d.
g ,,
, -W *! . i ' ;;- I- k.ii!:'!I ii;-
12.:*. I eir
, 9j+ 4
.** ?t *
- i . . n- *****".-
. xiii _
-1 i? T ii .
.gp d l 4 180 = 1& Ot; I :s5 Of 160
.= a k18x12:-, '15x13) : r (1*x13) _2l 15x 12 ) .
e - - u,. ;[ ;1 ' ' l i m U g.- l 1
""""'~ . ..
, .. m M *tP****.
IiI - 1I I 1 c ~ *a
' 80 ~
195
$(15x1253 'il5x13)b],w (1 b' 5"5 "~
17 CONTR*st 13) h.
^ ' Ik*I '
RCD STORA?E l 'i: I I I I
- 1
: ;!' : ;. +
l ; 1i i1 I ; I 202 :: ~
- g,' 174 g. -
T(15x14 ,; +gg+ 202 21: j g y g 4g 145 : z 315x13- :r'~""
-- 68
- m --
- - ._[15x12 C-
. e 1 . -+- --
lo; '~ --~-~"
- - 6) 1 1 ---
A-
-6
. m. -;
5 DEFECTIVE FUEL 4073 TOTAL STORAGE SPENT FUEL STORAGE 3 CONTROL ROD !?"IDE TUBE STORAGE PLisN ARRANGEMENT OF LA SALL COUNTY STATION C'.!T 2 CO:CtOM.' ALTli ED100N CO. Cli!CAGC , ILLINOIS - 1-2
r 2.0 Requiremanta for Thermal-Hydraulic Analysic 2.1 Introduction The scope of the analysis covers aspects of the Sargent and Lundy Specification T-3758 and includes the following:
- a. Computation of decay heat loads for the spent fuel pool in accordance with the NRC's position standard APCSB 9-2.
- b. Independent verification of the adequacy of the heat exchangers supplied by Yuba Industries and furnished by Sargent and Lundy.
- c. Determination of pool bulk temperatures for design base decay heat loads and cooling conditions,
- d. Temperature changes and heat-up rates for the loss of spent fuel pool cooling accident under normal refueling and full core offload conditions.
- e. Recirculation flow characteristics in the hottest and average spent fuel assemblies to determine local coolant temperature changes and peak clad temperatures. '
A local path and an under-rack path are examined.
- f. Investigation of gamma heating in the fuel box con-taining a fuel assembly to assure adequate coolant flow exists.
- g. Conductive cooling of intercellular water gaps.
- h. Determination of the temperature distributions in the fuel box, poison,. fuel box interface.
- i. Flow blockage to hot fuel assembly.
The methods used for analyzing the thermal and hydraulic aspects of the spent fuel pool involve relatively uncomplicated correla-tions for friction factors, loss coefficients, and heat transfer coefficients that make a detailed computer analysis unnecessary.
~
i Further simplifying but conservative assemptions reduce the mathematical complexity to the point where hand calculations or programmable calculators are all that is required. 2-1
Tha design criterin ucsd for tho thermal and hydraulic analysis of the spent fuel pool for La Salle County l Station - Unit 2 are in accordance with the NRC "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications"', issued April 14, 1978. Addit-ional conditions are given by Sargent and Lundy Engineers in Addendum 2 and Specification No. T-3758 for the Spent Fuel and Special Storage Racks, July 17, 1985. Based on the NRC Position Paper and the Sargent and Lundy Specifications, the following are established as design bases or requirements:
- a. Decay heat loads for a full pool are to be determined in accordance with the NRC Branch Technical Position APCSB 9-2, " Residual Decay Energy for Light Water Reactors for Long-Term Cooling", Section 9.2.5-8a of the Standard Review Plan. (This version is superceded by ASB 9-2 but is identical in form. The initials merely reflect the branch change to Auxiliary Systems Branch. ) Full pool decay heat loads and temperatures are computed for the following cases:
- 1. A normal refueling discharge of 240 fuel assemblies cooled 7 days after reactor shutdown (DARS) @ 4FA/HR.
The remainder of the pool is filled with normal refueling discharges cooled 18 months each. The pool maximum tem-peratrue is limited to 120*F. A pool containing 3120 fuel assemblies is assumed. This allows for a final full core off-load of 764 assemblies. This leaves 189 cells empty because not enough room remains for a complete normal refueling.
- 2. An equilibrium core of 764 fuel assemblies discharged 7 DARS 9 4FA/HR and 30 days after the last refueling.
A pool containing 3884 fuel assemblies is assumed. The pool bulk temperature is limited to 150*F. l
r g -
- 3. Sama oc Coco 2, exczpt that the dicchnrga occurg 90 days after the last refueling. ,
It is normal for one spent fuel cooling loop to be in 2 operation for the design bases decay heat loads. s
- b. To ensure that adequhte time exists for an alternate s cooling method to be implemented in the event of a loss of spent fuel pool cooling capability accident. This necessitates that the heat-up rate is calculated and the time required for pool boiling to occur is determined.
- c. Coolant flow rates, temperature increases, and peak clad temperatures are determined for worst case condit-ions (i.e. high pressure drops and low heat transfer conditions for channeled or unchanneled fuel assemblies, high bundle decay heat, etc.) to verify that boiling shall not occur.
- d. The effect of gamma heating in the fuel box and intercell spaces between fuel assemblies is analyzed. Gamma heating shall not cause boiling in these positions. Adequate' flow must be established to preclude the possibility of trapp'ing air or steam anywhere in the fuel racks. 'q
- e. Coolant flow paths and sparger locations affecting the analysis shall be identified.
As noted in the design bases, conservative assumptions are employed for evaluations of all coolant and clad temperatures. Some additional assumptions used for the thermal and hydraulic analysis of the spent fuel' pool are as follows:
- a. In determining the pool bulk temperatures, only one (of two) cooling loops are assumed to be operational. A 5% heat exchanger tube blockage is also assumed.
- b. The thermal inertias of the concrete walls and the coolant and piping outside the pool boundaries are neglected in the transient heat-up analysis.
2-3
r 1
- c. Tha pool surfaco is not azgumed to mix to a lower pool bulk temperature in the heat-up analysis following the loss of spent fuel pool. cooling accident.
- d. All decay energy is assumed to be absorbed in the fuel and surrounding coolant for the hot assembly or natural cir-culation analysis. (In reality, some gamma radiation will be absorbed in the adjacent fuel boxes and poison).
- e. The gamma decay heat absorbed in the fuel box wall is taken to be proportional to the mass densities of the materials in the spent fuel pool. (In reality, most of the gamma radiation never leaves the fuel assembly due to strong uranium attenuation.) Gamma heating proportional to the mass fraction is roughly equivalent to the assumption of uniform gamma flux in the repeating unit cell.
- f. A circulation flow path from the East wall or downcomer to a position along the West wall is assumed-for the hottest assembly. This derates the flow to the hottest assembly sinco flow down the three remaining walls is also possible.
- g. Worst case combinations with the fuel assembly channeled (for high bundle pressure drop) and unchanneled (for lower clad heat transfer) are assumed in the natural circulation analysis for the hottest fuel assembly.
- h. In gamma heating of the fuel box walls, the fuel assembly is assumed to be channeled.
I
- 1. The hottest assembly is assumed to generate 7 x 10 4 BTU /hr, more than 1.5 times higher than the decay heat generated by the average spent fuel bundle in the' hot batch of spent fuel for each of the three cases analyzed. Sinusoidal heat flux distributions force the clad hot spot factor above 2.4 for all cases.
2-4
1 , L
/ ;
j, Tha domincnt.proccura drepa era ovsro,ctimated by\ fcctora < of .. e 1.5 to 2.0 in most cases. . t %, g
.s
- k. Material properties (e.g. therraal conductivities, densities,t 1
r and specific heats) are generally assumed to,ge heiindependent
- 3 of temperature and are evaluated at some spd.31'fied (averate,s inlet, or surface) temperature. (4 s g ,
i The majorf are'as of concern in the themal and hydraulic analysis i are verification that the, clad and coolant ten.peratures do nod. . become high enough to caus'e boiling. In the e,ven,t of the loss of spent fuel pool cooling accident, the heat- up rate must be slow enough to allow an alterr. ate coolant syatem to be con-i
.. s ,
nected and operating before pool boiling occurs.' '
.s si 3,
%s p '
, s A
- 5
\' .t .:
I a 4 I .I , j g\ !
, (
s, 3 1 , 3 'i , t > 7
', g 5 ,
,i
- 6, 1
, e
{ i
, e I I
\
- ! \
- v. >
\* '\g l l
+i \\
__ p\-
. t 1, .
i. g I l *
, t t -.
) i' \,' j t
. s s :( ,
.,/ ; .'s' ; A ni i '
- b s l ;r , y
\ t t t l !~
;, . \ \ Nf n, \
,., . a t *g 4
3 i j ' 3 ( ,
'-, \
) 'i 1 i
i .
. E i
f 2-5 i
- t u ,
l 1' APPrOch to Thnrmql-Hydrrulic Annlynin 2.2 l
. , :\ i < 1
; C ., 4 In this analysis, the decay heat rates for the spent fuel l
'g
pool are calculated for the normal refueling and emergency core off-load conditions. Limits on the spent fuel pool
, bulk temperatures are calculated for the specified mass flow
' rates and design bases decay heat loads for the conditions of one heat exchanger operational. The spent fuel pool heat-up rate and time until pool boiling following the loss df spent fuel pool cooling accident are then computed.
Two recirculation paths are identified. The first is a local path involving the hottest spent fuel assembly and an adja-cent long-term cooled spent fuel assembly is considered. The second is a more complex path with under-the-rack flow. The coolant temperature increases and maximum clad temperatures are calculated for the hottest fuel assemblies for normal
. refueling and full core off-load conditions.
s h s/ t~ Gamma heating of a fuel cell box containing a fuel assembly and poison " slabs" is considered. Temperature profiles in /,, , the box, poison, fuel box interface are then found.
\ ^' ,
~
Ll 1,
]
S e l 2-6 i ._ .
2.3 Datailcd Anclynin end Renultn 2.3.1 overview In this section, we present an analysis overview for the calculation summaries that follow. Decay heat fractions computed according to NRC position standards. Total heat loads for the normal refueling and full core off-load conditions are then calculated. Heat exchanger adequacy is verified and the mass flow rates and coolant temperatures are calculated for the three design cases. The thermal inertia of the spent fuel pool (SFP) is computed. Heat-up rates and the time .
~
taken for the pool water to reach 200*F and 212'F following a loss of spent fuel pool cooling accident are found. Make-up rates at pool boiling are also determined in this section. Natural circulation cooling analyses are performed. A 3 local recirculation path and a more complete under-rack path are considered. Clad and coolant temperature dis-tributions are specified in these worst case analyses. Gamma heating of the fuel box walls and poison " Slabs" adjacent to the hottest assembly is investigated. The temperature distributions in the stainless steel wall at the fuel box, poison fuel box interface are deter-mined. Flow blockage for the hottest fuel assembly is considered.
- 2.3.2 Decay Heat Loads For The Spent Fuel Pool 1
The NRC Branch Technical Position APCSB 9-2 (or ASB 9-2) is used to compute the decay heat fractions for the La Salle County Station - Unit 2 spent fuel pool. (SFP) . For cooling times greater than 10 7 sec. (116 days), APCSB 9-2 does not specify a fission product decay uncertainty ! factor, but SRP Section 9.1.3 recommends a value of 0.1 for timesy 10 7, secs. and is used here. 2-7
F - Based on APCSB 9-2 and the three cases outlined in the design bases, the SFP decay heat loads are as follows: Case 1: Normal Refueling, 7 days after reactor shutdown (DARS) 240 FA; 7 DARS @ 4 ASSYB/HR 10.02 x 106 BTU /hr remainder of the SFP 3.96 x 106 BTU /hr Total heat load 13.98 x 106 BTU /hr Case 2: Full Core Off-Load, 7 DARS, 30 days in-reactor 764 FA; 7 DARS @ 4 ASSYS/HR 22.82 x 106 BTU /hr 240 FA; 30 DARS 4.78 x 106 BTU /hr remainder of the SFP 4. 01 x 106 BTU /hr Total heat load , 31.61 x 106 BTU /hr Case 3: Full Core Off-Load, 7 DARS, 90 days in-reactor 764 FA; 7 DARS @ 4 ASSYS/HR 25.12 x 106 BTU /hr 240 FA; 90 DARS 2.96 x 106 BTU /hr remainder of the SFP 3. 8 7 x 106 BTU /hr Total heat load 31. 95 x 106 BTU /hr The reactor is a BWR with an eighteen month equilibrium refueling cycle (1/3 core refueling discharge) operating at a rated 3322 MWth. A burn-up of 1 MWD /MTU is approx-imately equivalent to 1 full power hour. The racks will provide storage for 4073 fuel assemblies, equivalent to 5.33 cores. Allowing for the full core off-load capacity, 13 refueling discharges (or 20 years of spent fuel) can be safely accommodated before spent fuel relocation becomes necessary. O 2-8
As applied here, a 254 margin of conservatism is expected through the use of APCSB 9-2 and the pool histories used. This margin allows for uncertainties of approximately 10% (for curve fitting), 5% (for fission yields), and 10% (for PU239 uncertainties that are likely to introduce additional credits). 2.3.3 Spent Fuel Pool Heat Exchanger Design Adequacy and Pool Bulk Temperatures. This analysis is divided into three parts. In Part 1, the Yuba Industries, Inc. . heat exchanger (HX) for the LSCS2 spent fuel pool is checked for consistency. The data sheets or specifications for the Yuba Industries, Inc. HX can be found in Reference S2. The overall heat transfer coeffic-ient, U, is given as: UClean = 553 BTU /hr ft2p U Dirty = 229 BTU /hr ft27 All fluid properties are evaluated at a mean temperature of 110*F. The mass flow rates for the one tube pass /one shell pass HX are m h = 1.5 x 10 6 lbm/hr - shell side (SFP) m h = 2.0 x 10 6 lbm/hr - tube side (CW) The HX effectiveness under these conditions is
&= 0.388 I
In Part 2, the value of U is calculated ~ explicitly by accounting for all thermal resistances expected between the shell and tube fluids. The wall resistance is determined by the tube thermal conductivity. The film coefficients at the inner and outer tube surfaces are estimated using a Dittus-Boelter correlation. 2-9
r-In Part 3, the off-design heat exchanger temperatures are determined for normal refueling and full core off-load conditions. For a conservative margin, we assume 5% of the heat exchanger tubes are plugged. It is normal for one loop to be operational. The three design cases are presented below: SPENT FUEL POOL AND HEAT EXCHANGER
, TEMPERATURE LIMITS NO. OF CONDITIONS T T T T HX's HIN HOUT HAVG COUT gy= 13.98 x 106 BTU /hr Normal Refueling 7 DARS 9 4 ASSY'S/HR 1 119.6 110.3 114.9 102.0 _
q2 = 31. 61 x 106 BTU /hr Full Core Off-Load 7 DARS 9 4 ASSY'S/HR, 30 days in reactor 1 149.4 128.3 138.8 110.8 g3 = - 31. 95 x 10 6 BTU /hr Full Core Off-Load 7 DARS 9 4 ASSY'S/HR, 90 days in reactor 1 150.0 128.6 139.3 111.0 NOTES: All temperatures in "F. T = peak pool bulk temperature (HX inlet - tubes) HIN T HAVG = pool average temperature T HOUT = sparger discharge temperature (HX outlet - tubes) T COUT = cooling water outlet temperature T = 95 F = cooling water inlet temperature CIN Under normal refueling conditions, THIN (120*F 7 DARS with
. one HX in operation. Spec. T-3758 limits this vs.lue to 120*F, in agreement with the value tabulated here. Since 5% tube plugging is assumed in all cases the heat loads 2-10
era concidarcd concarvctively high (by 25% to cover uncer-tainties), all temperatures are extreme temperatures for the given conditions of cooling time (DARS) and discharge type (i.e. normal refueling or full core off-load). With full core off-load 7 DARS, 30 days in reactor and 90 days in reactor, the peak pool bulk temperature of 149.4'F and 150'F are less than or at the spec. limit (at 7 DARS) of 150*F. 2 3.4 Pool Thermal Inertia and Heat-Up Rates. In this analysis, the SFP thermal inertia for La Salle Count Unit 2 is found. A full pool is assumed and is conservative since the UO 2 and Zircaloy in the fuel assembly have a smaller thermal inertia than the water these materials displace. For simplicity (and also con-servatism) we take pc p = 60 BTU /ft3 F for both water and stainless steel. Only the water and materials with-in the liner boundaries are considered. Materials that possess some thermal inertia that are neglected include primary and secondary water and piping exterior to the liner plates, the cask pit, the transfer canal and gates, and the concrete walls of.the spent fuel pool. The total SFP thermal inertia under these simplifying but conserva-tive assumptions is 6 p = 2.48 x 10 BTU /*F I with water and stainless steel accounting for more than 91% of this. In the heat-up analysis, the initial pool temperature is taken to be T HIN - the inlet HX temperature. Pool mixing to a lower pool temperature is neglected. Following the loss of SFP cooling, the pool heat-up rates, times to I . 2-11
reach 200'F (and 212*F), boil-off rates, and elevation changes at boiling are determined. These are tabulated below. P.ESULTS OF HEAT-UP ANALYSIS FOLLOWING THE LOSS OF SPENT FUEL POOL COOLING LA SALLE UNIT 2 B
- Conditions q ,{,
6t(to 2 'F/ A Q (GPM) h Normal Refueling 13.98x106 5.62 14.2/16.3 hrs 14,400 28.9 .21 7 DARS Full Core Off-Load 31.61x106 12.72 ,3.93/4.87 hrs 32,500 65.3 .47 7 DARS, 30 days in reactor Full Core Off-Load 31.95x106 12.8 3.90/4.84 hrs 32,900 66.0 .48 7 DARS, 90 days in reactor l Definitions: DARS = Days after reactor shutdown q = total SFP decay heat load (full pool assumed) T = heat-up rate following loss of SFP cooling at = time to reach 200*F/212*F from T HIN IO" I cooling loop 6 = boil-off rate at 212'P, 14.7 psia. Q = make-up rate (at y = 8. 3 lbm/ gal) h = rate of change of pool elevation at boiling The above it's are considered long enough for an alternate cooling source to become operational following the loss of SFP cooling.
~
2-12
2.3.5 Noturcl Circulation Cooling of the Spent Fuol. In this analysis, the natural circulation cooling of the
~
La Salle County Station Unit 2 spent fuel assemblies is considered. A local path where coolant is convectively driven up the hottest-assembly and down a " cold" assembly is studied first. A second path flowing under the spent fu,el racks, up the hot assemblies, into the mixing region above the racks, and finally down the West wall of the pool to complete the path is then modeled and analyzed. Apart from the estimation of the coolant
,. inlet temperatures.to the hot batch of spent fuel, these flow path.i are decoupled from the cooling loop (s) and SFP heat exchanger (s).
The following conservative assumptions are used in this analysis:
- 1. The hottest assembly decay heatLis taken to be 7.0 x 10 BTU 4
/hr. This hottest assembly generates no less than 1.5 times more heat than the average assembly from the most recent spent fuel discharged for the three respective cases:
Case 1: Normal refueling, 7 DARS @ 4 ASSY'S/HR Case 2: Full Core Off-Loading, 7 DARS @ 4 ASSY'S/ HR, 30 days in reactor Case 3: Full Core Off-Loading, 7 DARS @ 4 ASSY'S/ HR, 90 days in reactor A sinusoidal heat flux distribution at the clad surface is the assumed, so that the hot spot factor exceeds the average by a factor of1f/2 = 1.57.
- 2. The remainder of the pool is filled with 7.63 (full core less one) assemblies, each generating 4.5 x 104 BTU /hr.
2-13
. - - . ~ _ _ _ _ _ _
- 3. The hot batch of spent fuel is located at the West wall so that the under-rack path length and distance from the operating sparger is maximized.
- 4. The fuel assembly pressure drop (versus flow) is accu-rately correlated with available experimental data.
The data is for the channeled fuel assembly at 100"F, 34 psia and will be worst case since 100'F is a represen-tative low temperature (high viscosity) and the channeled assembly has a smaller flow area than the unchanneled assembly. As a function of the volume flow rate Q in the fuel assembly, 1.564 p (psf) = 973 Q (cfs) is the fuel assembly pressure drop equation.
- 5. For the unit cell geometry of the fuel assembly, GE 8x8, .483" rods, .640" pitch), the Nusselt number for the (laminar) flow in the spent fuel assembly is 5.7. The clad heat transfer coefficient is evaluated on the basis of an unchanneled fuel assembly (lowest h) and found to be 32 BTU /hr ft 2aF. For the channeled assembly, it would be closer to 45.
- 6. For the under-rack flow path (path 2), the losses calculated include:
- a. West wall friction losses
- b. 90' turn losses
- c. Under-rack friction losses l d. Expansion and contraction losses at the rack supports
- e. Fuel bundle losses
- f. Branching - Momentum losses i
a 2-14 I
Flow branching / momentum losses are typically small (and recoverable) in comparison to the total losses. The dominant loss (d) - the expansion and contraction losses contribute more than 50.6% to the total flow loss. These were overestimated by a f actor of 10%. _
- 7. The driving pressure (caused by water density varia-tions) is given by AP O N.
d= B 2CpQ i ge , where B=-fhp h is the thermal coefficient I of expansion for the water-only a mild function of pressure. L is the heated (active) length (12.5 f t) and q is the-decay heat rate. Fluid properties are evalu-ated at l_2 S
- F , a low temperature for the core of f-load i case. At 125'F, S = 2.6 x 10-4'F-1 and at 150*F, 8=
3.1 x 10~4'F" . The driving pressure is then d'erated
'by approximately 20%.
- 8. For the local path, the fuel assembly inlet temperature is taken to be the hottest pool bulk temperature (T HIN from Section 2.3.3). Since considerable mixing can occur l in path 2 and cold water is discharged under the rack, ,
i T IN is taken to be the pool average temperature (i.. e . T rm ection 2.3.3. HAVG A summary of significant results that apply to all three cases (for the purposes of determining peak cool' ant and l , clad temperatures) is as follows: l 2-15
l i Path 1 Path 2 Volumeflowbatein ft 8 /sec .011 .0084 the hottest assembly (GPM) (4.8) (3.5) Total pressure drop = AP d - f f 2.9- 3.5 (3.2 under rack) Coolant AT - *F 29 38 Position of clad hot spot - ft 7.8 8.2 Difference (T cladmax -TIN) - F 52 59 Following assumption 8, the peak clad and coolant temperatures T and TOUT, are as tabulated below: cladmax Path 1 Path 2 Conditions TCUT( F) T cladmax( F) TOUT ( F) Tcladmax(*F) Case 1 - Normal Refueling 148.6 171.8 158 178.8 7 DARS,4 ASSYS/HR Cane,2 - ' Full-Core Off-Load 178.6 201.6 188 208.8 I 7 DARS,,30 days
,r-ranc+m, Case 3 Full Core Qff-Load 178.6 201.6 188 208.8
^
inIbef8E8r: NOTE: T sat a 240*F at top of racks ( 25 psia) T sat a 245'F at peak clad T (~30 psia) All cases are significantly subcooled and void fractions are negligible. Since the actual flow paths will be complicated combinations of local and under-rack paths, the temperatures will not exceed those indicated. 2-16
1 2.3.6 Gamma Heating of the Fuel Box Walls, Poison, and Inter-cellular Water. In this analysis, gamma heating of the i fuel box walls, poison, and intercellular water is investigated. ' Fission product decay accounts for virtually all residual heat in the spent fuel pool with minimum cooling times t, a 7 days. At this time, a realistic but upper limit on the gamma fraction is 0.62, based on the primary reference for the NRC position standard APCSB 9-2. A typical 1 MeV electron will travel approximately 0.016"
~
in the UO fuel. Thus, all S electrons will be stopped 2 in the fuel or surrounding clad and coolant. A typical, but higher than average, 1 MeV gamma ray has a mean free path of approximately 0.56" in the UO fuel - comparable 2 to the pellet diameter .410". Therefore, the fuel will not stop all the gamma radiation emitted by the decaying fission products. It will then be conservative to assume the following when estimating the energy deposition in the fuel box, poison, and intercellular water:
- 1. The fuel box is located within an infinite array of hottest assemblies - each generating 7 x 10" BTU /hr, 62%
of which is gamma.
- 2. The gamma energy absorbed in a unit cell comprised of one fuel assembly, one fuel box, and 4 "1/2" Poison " slabs" is proportional to a given material's mass fraction. This is roughly equivalent to the assumption of uniform y-flux since p/p is approximately constant for all materials at a given gamma energy.
2-17 9 pe
i
- 3. Water in the fuel box must remove gamma heat due to energy depositions in the 0 090" thick fuel box wall, the fuel box water itself, and the poison slabs adjacent to it.
Matching the driving head with the loss head, the flow is determined to be suffidisnt to allow adequate circul-ation. Intercellular water is free to flow between the fuel assembly channel and the fuel box boundary. The flow areas defined by two .281" diameter nozzle holes in each fuel essembly are large enough to remove heat in the 0.090" box walls and poison " slabs". 2.3.7 Stainless Steel Temperatures. In this calculation, temperature distributions in the stain-less steel box walls are determined. Heat source terms (due to y-irradiation) follow the same conservative assumptions used in the previous section 2.3.6. Channeled fuel is also assumed and the resulting geometry contains water gaps adjacent to the box wall interfaces. o 2-18'
One dimensional heat conduction in the series of " slabs" defined by the fuel assembly channels, water gaps, stainless steel box walls, and poison slabs is.modeled using the ! steady-state, 1-D, heat conduction equation h(T,-Tc) "9 II where q' is the volumetric heat generation rate in the
" slab".
Denoting the coolant temperatures in the fuel box as T c, it is found that all temperatures do not exceed T by more than 59"F. Since convection is neglected c and all q are conservatively high, the 59'F variation above T is conservative. c In Section 2.3.5, T cladmax exceeds T (i.e. TOUT) by 20*F all points in the racks are subcooled by at least 20*F, thus film boiling of the intercellular water will not occur. The maximum temperature gradient occurs across the fuel box interface and is modeled using the steady-state, 1-D, heat conduction equations stated previously. The hiaximum temperature difference is limited to less than l'F. 1 Both cases pose no thermal stress problems. 2 2.4 Conclusions The detailed thermal and hydraulic analyses described in Sections 2.3 through 2.3.7 addresses the concerns, intent, and design bases of the NRC's Position Paper "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" and the Sargent and Lundy Specification Number T-3758 plus Addendum 1. Based on these analyses, it is our professional judgment that the spent fuel pool for L'a Salle County Station - Unit 2 can be adequately cooled in accordance with the suggested regula-tory standards of the Nuclear Regulatory Commission and comply with specifications outlined by Sargent and Lundy Engineering Corporation. l l l l i l 2-2o l
i 3.0 CRITICALITY ANALYSIS The follbwing dtscussion sunmarizes the design of the spent fuel racks with respect to criticality safety. The analytical techniques described here are similar to those used to successfully license spent fuel racks for several other plants, the most recent being those for Point Beach s Units 1 and 2 and Nine Mile Point Unit 1. 3.l ANALYTICAL TECHNIOUE . The LEOPARD U) computer program was used to generate macroscopic cross sections for input to four energy group dif fusion theory calculations which are performed with the PDQ-7(2) program. LEOPARD calculates the neutron energy spectrum over the entire energy range from thermal up to 10 Mev and determines averaged cross sections over appropriate energy groups. The fundamental methods used in the LEOPARD program are those used in the MUFT(3) and SOFOCATE} programs which were developed under the Naval Reactor Program and thus are well founded and extensively tested techniques. In addition, Westinghouse Electric Corporation, the developers of the original LEOPARD program, demonstrated the accuracy of these methods by extensive analysis of measured critical assemblies consisting of slightly enriched U0 fuel rods. U) 2 In addition, Pickard, Lowe and Garrick, Inc. (PLG) has made a number of improvements to the LEOPARD program to increase its accuracy for the calculation of reactivities in systems which contain significant amounts of plutonium mixed with U0 . PLG has tested the accuracy of these 2
, modifications by analyzing a series of UO and Pu0 -UO critical 2 2 2 experiments.
These benchmarking analyses not only demonstrate the improvements obtained for the analysis of Pu0 -UO systems but also 2 2 demonstrate that these modifications have not adversely affected the accuracy of the PLG-modified LEOPARD program for calculations of slightly enriched UO . systems. 2 3-1
The UO 2 critical exp3riments chosen for benchmarking include variations in H20/UO2 volume ratios U-235 enrichments, pellet diameters and cladding materials. Although the LEOPARD model also accurately calculates the reactivity effects of soluble boron, these experiments have not been included in the LEOPARD benchmarking criticals since the spent fuel pool calculations do not involve soluble boron. Neutron leakage was represented by using measured buckling input to infinite lattice LEOPARD calculations to represent the critical assembly. A sumary of the results is shown in Table 3.1-1 for the 27 measured criticals chosen as being directly applicable for bench-marking the LEOPARD model for generating group average cross section for spent fuel rack criticality 'alculations. The average calculated k,fy is 0.9979 and the standard deviation from this average is 0.0080 Ak. Reference 5 raised questions concerning the accuracy of the measured buckling reported for the experiments number 12 through 19. If these data are excluded, the average calculated k,gg for the remaining , 19 experiments is 1.0006 with a standard deviation from this value of 0.0063 ak. In all of these experiments there are significant uncertainties in the measured bucklings which are necessary inputs to the LEOPARD analysis. These uncertainties are the same order of magnitude as the indicated errors in the LEOPARD results, and therefore a more definitive set of experimental data is used to establish the accuracy of i the combined LEOPARD /PDQ-7 model used for the criticality analysis of the spent fuel racks. The PDQ series of programs have been extensively developed and tested over a period of 25 years, and the current version, PDQ-7, is an accurate and reliable model for calculating the subcritical margin of the prcposed spent fuel rack arrangement. This code or a mathematically equivalent method is used by all the U.S. suppliers of light water reactor cores and reload fuel. In addition, this code has received extensive utilization in the U.S. Naval Reactor Program. As a specific demonstration of the accuracy of the calculational model used for the spent fuel rack calculations, the combined LEOPARD /PDQ-7 model has been used to calculate fourteen measured just critical 3-2
- - . , , , , , - - , - . . - - - . . --,,.,,,,-------,n
assemblies. The criticals are high neutron leakage systems with a large variation in H 0/UO 2 2volume ratio and include parameters in the same range as those applicable to the spent fuel rack design. Experiments including solubl'e boron are included in this demonstration since the ability of PDQ-7 to calculate neutron leakage effects is of primary interest. The use of soluble boron allows changes in the neutron leakage of the assembly while maintaining a uniform lattice and thus allows a better test of the accuracy of the model. Furthermore, it eliminates the error associated with the measured bucklings, which is inherent in the LEOPARD benchmarks, thus permitting determinations of the actual calculational uncertainty which must be accounted for in the spent fuel rack criticality analysis. These combination LEOPARD /PDQ-7 calculations result in a calculated average k,77 of 0.9928 with a standard deviation about this value of 0.0012 ak. These results, as shown in Table 3.1-2, demonstrate that the proposed LEOPARD /PDQ-7 calculational model can calculate the l reactivity of the proposed spent fuel rack arrangements with an accuracy of better than 0.010 ak at the 95 percent confidence level. The cross sections for the Boraflex* neutron absorbing material which is an integral part of the design are calculated using fundamental techniques that have been successfully applied in the past to thin heavily absorbing mediums such as control rods. N i The procedure is straightforward and is comprised of several well defined steps:
- 1. The Boraflexe sheets are associated with the stainless steel and water areas exterior to the fuel bundle to define a one-dimensional slab geometry representing the proper material volume fractions. An equivalent LEOPARD cylindrical one-dimensional geometry is used to obtain a first estimate of the spatial and energy self-shielded cross section for the B 10 in the Boraflex*.
3-3
- 2. Using th2 energy averaged cicroscopic cross sections from 1.,
integral transport theory is applied in slab geometry using They's method for calculating flux depressions and shielding factors to determine an' appropriate 8 10 number density. This approach is similar to that of Amouyal and Benoist.
- 3. The self-shielded number densities calculated in Step 2 are again input to LEOPARD to obtain corrected microscopic 8 cross sections.
- 4. Blackness theory is applied to obtain macroscopic cross sections .
which will produce the requ,1 red boundary conditions at the surface of the Boraflex* sheets. In addition to the fourteen critical assemblies in Table 3.1-2, the LEOPARD /PDQ model was used to calculate the k,ff for twelve additional critical assemblies, seven of which incorporated thin, heavily-absorbing materials for which the procedure just described was used to determine the macroscopic cross section. These twelve criticals were performed by Battelle Pacific Northwest Laboratories specifically for the purpose of providing benchmark critical experiments in support of spent fuel criticality analysis. They are described in detail in Reference 17. The results of these critical experiments are summarized in Table 3.1-3. The first seven of these ~ twelve experiments include fixed neutron poison absorber plates, and the average k,f f calculated for these just critical experiments was 0.9935, with a standard deviation around this value of 0.0007 ak. The other five critical experiments in this series do not include absorber plates and the average k,ff calculated for these just critical assemblies was O.9944, with a standard deviation around this value of 0.0007 ak. The overall average k,ff calculated for these twe1ve just critical assemblies was 0.9939, with a standard deviation around this value. of 0.0008 ok. 3-4
This ' extensive series of UO2 critical experiments further supports the applicability of the methods described above for use in calculating the subcritical margin of these fuel storage rack designs, and demonstrates that the accuracy of better than 0.010 ak at the 95 percent confidence " level established for the LEOPAR0/P0Q-7 model applies equally well to designs incorpora' ting fixed neutron absorbers for which blackness theory is used to calculate the macroscopic cross sections. As a result of this approach to separately benchmark both the cross sections and the diffusion theory calculations against applicable critical assemblies, the " transport theory correction factor" is implicitly included in the derived calculational uncertainty factor. 3.2 CALCULATIONAL APPROACH The P0Q-7 program is used in the final predictions of the reactivity of the spent fuel storage racks. The calculations are performed in four energy groups and take into account all the significant geometric details of the fuel assemblies, fuel boxes and major structural' components. The geometry used for most of the calculations is a basic cell representing one-quarter of. the area of a repeating array of stainless steel boxes. The specific geotretry of this basic cell is shown in Figure 3.2-1. The calculational approach is to use the basic cell to calculate the reactivity of an infinite array of uniform spent fuel racks and to account for any deviations of the actual spent fuel rack array from this assumed infinite array as perturbations on the calculated reactivity of l the basic cell. The effects of manufacturing tolerances, as well as thermal uncertainties, including fuel and water temperature and density variations, are also treated as perturbations on the calculated reactivity of the basic cell-. The fuel assemblies used for this analysis are the General Electric 8x8 design for which data are provided in Table 3.2-1. The fuel bundles were Aburned to be untrradiated with an enrichment of 3.416 weight percent n . - 9 3-5
~
sk
U-235 over tha entire slightly enriched section of the bundle. This is i equivalent to a loading of 16.52 gm of U-235 per axial cm of the slightly enriched section,of the fuel bundle. All of the calculations were performed at a u'niform pool temperature of 68'F, except when the : reactivity effects of pool temperature were taken into account as a perturbation on the basic cell calculations. 3.3 EVALUATION OF CRITICALITY SAFETY For the average rack cell, which is defined as the basic cell for analysis purposes, the pitch will be 6.255 inch, and the k= of this
' cell is .9179. Figure 3.3-1 presents the spent fuel storage rack reactivity as a function of fuel bundle enrichment for the basic cell geometry.
The reactivity of the basic cell as a function of 810 loading in the Boraflex* is shown in Figure 3.3-2. The 8 10 loading which was used for the criticality analysis was the minimum loading to be incorporated into the design. This corresponds to a 8 10 loading of .020 grams per square centimeter of cross sectional area in a nominal thickness of
.075 inch.
3-6
~ ' ^
,. al l
As derived in Section 3.1, the combined LEOPARD /P0Q model bias to be ' added is .0061 ak. If Zr channels are stored on the fuel bundles, the calculated k= of the rack is increased by +.0045 Ak. - As shown in Table 3.3-1, the net ef fect of all the calculational biases l is .0003 Ak, which therefore increases the basic cell k= to .9311. 3.4 TOLERANCES AND UNCERTAINTIES There are also a number of tolerances and uncertainties which result in perturbations which must be considered in the criticality analysis. The reactivity effects of all such positive perturbations are then combined statistically in accordance with Reference 18 to determine a single reactivity perturbation which is added to the calculated basic cell 4 multiplication factor (including biases) to determine the final I conservative evaluation of the spent fuel rack maximum possible multiplication factor. The tolerance on the Boraflex thickness is t .010 inch, but the minimum j , loading is .020 gm 8 10je ,2 for any thickness within this tolerance. Calculations demonstrate the k= is largest for the minimum Boraflex thickness of .068 inch, and the resulting pertubation to the basic cell is .0005 Ak. 1 The worst case in terms of manufacturing tolerances from a reactivity perspective is represented by the minimum fuel box inside dimension. The k= for this minimum inside dimension spent fuel storage rack cell at 3-7
68'F for a fuel assembly enrichment of 3.416 weight percent U-235 is
.9197. Since the k= of the basic cell is .9179, the perturbation in k= due to tolerances on fuel box cell dimensions is .0018 ak, which corresponds to a sensitivity of .072 ak/ inch.
To determine a conservative evaluation of the reactivity effects of thickness variations in the stainless steel structural materials, all stainless steel members were assumed to be at the most reactivity limiting thicknesses allowed by the tolerances. The reactivity is highest for the maximum stainless steel thickness and the k= of the resulting basic cell is .9180. Therefore, the perturbation due to stainless steel tolerances is .0001 ak. The reactivity of the spent fuel storage rack was evaluated for the effect of manufacturing tolerances on U02 density. The reference cell is based on the fuel design value of 95% theoretical density. The worst case of a UO2 density of 96% theoretical density was examined. The resulting perturbation to the basic cell was determined to be .0014 Ak due to an increase in pellet density from 95% to 96% of theoretical density.. With regard to fuel position uncertainties within the fuel boxes, calculations confirm the fact that the fuel assemblies when centered are located in their most reactive positions within the fuel boxes. The fuel bundles in a two by two array of storage locations were moved off center j in such a direction as to place all four fuel assemblies uniformly in
- closer proximity to one another. These calculations confirmed that the maximum k= is obtained with each fuel bundle centered in its storage
. position.
Based on the results of the calculational model benchmarking described in Section 3.1, the ko uncertainty in the model, which corresponds to a 95/95 confidence statement, is .0022 ak. 3-8
The reactivity of the basic cell as a function of temperature is shown in Figure 3.3-3. With a maximum pool temperature of 200*F, the k= is less than the v(lue at 68'F by .0215 Ak. This is to be expected for this design which incorporates a heavy loading of 810 as a neutron poison l and indicates that the lower temperature conditions produce the higher ' spent fuel storage rack k=. Since the reference case was based on a temperature of 68'F, which is clearly conservative, no additional reactivity ef fect needs to be added to account for temperature.
~
The sensitivity of the spent fuel rack multiplication factor to variations in the water density throughout the pool is illustrated in Figure 3.3-4. Again, the effect.of the heavy 810 loading is to produce the most reactive conditions at full water density. A summary of the perturbations to.the basic cell reactivity calculations is shown in Table 3.3-1. As shown in this tab;e, the total reactivity perturbation to be added to the biased basic : ell reactivity to account for tolerances and uncertainties is .0032 ak. Therefore, the conservatively calculated reactivity of the spent fuel rack f ully loaded with unirradiated bundles with 3.416 weight percent U-235 and .no burnable poison is .9343 for a pool temperature of 68'F including conservative allowances for manufacturing and calculational uncertainties. l 3.5 ACCIDENT ANALYSIS lhe fuel racks are designed to prevent a dropped fuel bundle f rom penetrating and occupying a position other than a normal fuel storage location. The only positive effect of such a bundle on the reactivity of the rack would be by virtue of a reduction in axial neutron leakage f rom the rack. Since there is approximately 13 inches between the top of the active fuel and the top of the boxes comprising the rack, a dropped fuel bundle will be neutronicly decoupled from the fuel in the rack and would not have any measurable effect on the reported maximum possible reactivity of the spent fuel storage rack. 3-9 ( - . . _ _ -
The lattico of the fuel bundles results in an undermoderated ccnfiguration; so that any crushing or compaction of the fuel bundigs would tend to reduce the neutron multiplication factor of the spent fuel pool. Therefore(deformationsresultingfromthedroppingofheavy objects into the fuel pool or from the effects of earthquakes or tornadoes will not produce a criticality accident. The reactivity effect of a fresh fuel assembly located adjacent to the fully loaded spent fuel storage rack has been evaluated for all postulated locations other than normal fuel storage locations. The model used to evaluate the maximum effect of a fuel bundle, which is accidentally mislocated directly. adjacent to the outer row of fuel storage racks, consisted of a 3 box by 6 box section of the fuel rack with the extra bundle located directly on the centerline of the 3 box array. It is obvious that the centerline location selected, which is directly in line with the centerwine of the stored fuel bundle, is the position which results in maximum reactivity. The maximum per Jrbation associated with such an accident has been calculated to be .0101 Ak. For this accident, even assuming the worst case k= of .9444, the spent fuel storage rack design assures that the multiplication factor is less than 0.95. Because of the well founded, conservative technique used for determination of the infinite multiplication factor, there is more than reasonable assurance that this spent fuel rack design will not cause a significent hazard to the public health and safety resulting from criticality considerations.
, 3.6 OESIGN CONSERVATISMS Best Edimate. Calculations When the fuel assembly is represented by an explicit fuel pin distribution of selected U-235 enrichments typical of the General
- Electric Company's intra-assembly fuel pin arrangement (which produces a bundle slightly enriched section enrichment of 3.416 weight percent U-235), the k is calculated to be .9054. This is less than the k=
of the basic cell which utilizes a single average enrichment in all fuel l l
/
3-10 l l
pins, and therefore, the perturbation to be applied to account for the more realistic explicit multi-enrichment fuel pin distribution is
.0125 ak. .
Doubling the number of mesh points used to represent the basic cell geometry results in a reduction in k= of .0004 as shown in Table 3.3-1. As discussed previously, the basic cell calculations make the conservative assumption that all fuel bundles are unirradiated and contain no burnable poisons. Calculations, verified by reactor operation, show that with the burnable poison loadings required for a fuel bundle initial enrichment of 3.416 w/o, the maximum possible bundle k= is at least 0.030 ak less than the initial k= of the bundle with t 3 - f D. l
no burnable poisen.- In addition, the spacer grids in the fuel bundle rsduce the calculated k= by 0.029 ak. These effects would reduce the calculated k= of,the basic cell by at least 0.030 ak. Thus, the actual maximum possible multiplication factor of the spent fuel racks when completely filled with unirradiated fuel of 3.416 w/o U-235 and under the worst accident condition is less than .8986. 3.7 MEW FUEL DESIGNS It is anticipated that in the future new reload fuel designs will be developed, and the initial enrichments of these designs will be greater than the 3.416 w/o loading analyzed herein. However, the maximum reactivity of such higher enrichment designs is limited by the existing reactor control systen.s which must provide assurance that the specified rcactor shutdown margin will be maintained at all times for all reload fuel designs. As a practical matter Gd23 0 burnable poison containing fuel rods are incorporated in all BWR reload fuel designs, and for the higher enrichment designs, increased Gd 230 concentrations and/or larger numbers of burnable poison containing fuel rods are incorporated in such i designs. In order to provide a criterion to establish that it is safe to store higher enrichment designs, a generic calculation model was evaluated. An average assembly enrichment of 4.25 w/o U-235 was . selected for this generic evaluation which correspands to a maximum average planar enrichment of 4.558 w/o U-235 in the fuel assembly. Using this
. enrichment, cell calculations were performed corresponding to different concentrations of Gd23 0 in burnable poison containing fuel rods. The i results, shcwn in Figure 3.3-5, allow the determination of the storage rack k= as a function of the fuel assembly k= which was varied by changing the burnable poison concentration.
3-11 n- -
-,v.,--- ,--
---,r -- - - - - - , - _____w _ --_ _ , _ _ _ _ _ , _ , , , , , _
, ,- y 7
~s t,. c F
,;'% /1 3
-! g g j '
,l ,
3.8 REFERENCES
,( y ,
) ;
./
>- ,J' <
N ~ ll . t
. s ~
i; , i ( N ,,'; 1 -
) ,
b , 3 f L< ,
- 1. R. F. Barry,' ' LEOPARD--A Sphtrum DependenTNon-Spatial Deple ion '
Code for the IBH-7094," WGi-3259, Septembdr 1963. y
} a 7,1;. c
- 2. W.R.Caldwell,'P0Q-7ReferenceManual,"WAPL-lfA78,.j January 1967. 3
- 3. H.Bohl,E.GelbardandG.Ryan,"MUFT#$-FastNeutronSpectrt$Cbdee t / 5 i f or the IBM-740," . WASP-TM-77, ' July 195 r. , c l '
T>
, *)
r. 4.
~t A \
H. Amster and FT. , Suarez, "The Calcalation\ of H ,! e 's Averaged Over a'Wigner-Wilkips Flux Spectrumd. ermabConstants -
.Oc<.cription:.of, the .
SOFOCATE Code," WAPD-TM-39,CJanuary 1957. j I i > .
*I ,
+. .
\ ; c,
- 5. L. , E,.
- Strawbridge i and R. F. Barry, " Criticality w l!talculations I
for Uniform he.ter-Moderated Lattices," Nuclear tu Scier.ce and Engineering, 23, 58, 1965. *
\
, )A,! /
6.
'f "Large Closed-Cy~c'.e Water Reactor Research and Development Program Progress Report for' the Firpd~ January 1 - March 31,31965,"
s WCAP-3269-12. / < 3 a f a, 'N c g(s g4 y ., - A,
- 7. " List of Equipment and App:iratJs at h7EC," Westinghouse Reactor Evaluation Centet February 1967.
- 8. W. L. Orr, H. I 1Sternberg, P. Deramaix, R. H. Chastain, L. Binder and A. J. Impink, "Saxton P,lutoniur's, Program, Nuclear Design of the
, Saxton Partial, PNtonium Core," WCAP-3385-51, December 1965. (Also EURAEC-1490) 1'
- 9. R. D. Leamer, W.el. Orr, R. L. Stover, E. G. Taylor, J. P. Tobin and A. Bukmir, "Pu0 -UO2 Fueled Critical Experiments," WCAP-3726-1, July 1967.
I .. T i (' 3-12 9 i
- 10. A. F. Henry, "A Thetretical Method for Determining the Worth of Control Rods,'WAPD-218, August 1959,
- 11. P. W. Davis [n, et al., " Yankee Critical Experiments Measurements on Lattices of Stainless Steel Clad Slightly Enriched Uranium Dioxide Fuel Rods in Light Water," YAEC-94, Westinghouse Atomic Power Division (1959).
- 12. V. E. Grob and P. W. Davison, et al., " Multi-Region Reactor Lattice Studies - Results of Critical Experiments in Loose Lattices of U0
- 2 Rods in H20," WCAP-1412, Westinghouse Atomic Power Division (1960).
- 13. W. J. Eich and W. P. Kovacik, " Reactivity and Neutron Flux Studies in Multiregion loaded Core," WCAP-1433, Westinghouse Atomic Power Division (1961).
- 14. W. J. Eich, Personal Communication (1963).
- 15. T. C. Engelder, et al., Measurement and Analysis of Uniform Lattices-of Slightly Enriched 00 2Moderated by D 0-H Mixtures "
~ 2 2 BAW-1273, the Babcock & Wilcox Company (1963).
it. A. L. MacKinney and R. M. Ball, " Reactivity Measurements on Unperturbed, Slightly Enriched Uranium Dioxide Lattices," BAW-1199, the Babcock & Wilcox Company (1960).
- 17. Battelle Pacific Northwest Laboratories, " Critical Separation Between Subcritical Clusters of 2.35 WtX 235-U Enriched UO2 Rods in Water with Fixed Neutron Poisons," PNL-2438.
- 18. "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," U.S. NRC, April 14, 1978.
3-13
TABLE 3.1-1. SUMP.ARY OF LEOPARD RESULTS FOR MEASURED CRITICAL 5 Case ** Reference Enrichment H 0/U n ty Number Number (atom %) V Iume D ter Di er Th ck e s tch Buc W ng Calculate ggjc,3) (cm) (cm) (cm) -2 (cm) m eff 1 11 2.734 2.18 10.18 0.7620 0.8594 0.04085 1.0287 2 2.734 40.75 1.0015 11 2.93 10.18 0.7620 0.8594 0.04085 1.1049 53.23 3 11 2.734 3.80 10.18 1.0052 0.7620 0.8594 0.04085 1.1938 63.28 1.0043 4 12 2.734 7.02 10.18 0.7620 0.8594 0.04085 1.4554 65.64 1.0098 5 12 2.734 8.49 10.18 0.7620 0.8594 0.04085 6 1.5621 60.07 1.0118 12 2.734 10.13 10.18 0.7620 0.8594 0.04085 7 1.6891 52.92 1.0072 13 2.734 2.50 10.18 0.7620 0.8594 0.04085 8 1.0617 47.5 1.0008 13 2.734 4.51 10.18 0.7620 0.8594 0.04085 1.2522 9 3.745 68'.8 O.9987 13 2.50 10.37 0.7544 0.8600 0.0406 1.0617 68.3 10 13 3.745 4.51 1.0010 10.37 0.7544 0.8600 0.0406 1.2522 95.1 y 11 14 3.745 4.51 1.0025 10.37 0.7544 0.8600 0.0406 1.2522 95.68 H 12 15 4.099 2.55 9.46 1.0009
^ 1.1278 1.2090 0.0406 1.5113 88.0 0.9889 13 15 4.099 2.14 9.46 1.1278 1.2090 0.0406 1.450 79.0 0.9830 14 16 4.099 2.59 9.45 1.1268 1.2701 0.07163 1.555 15 16 69.25 0.9999 4.069 3.53 9.45 1.1268 1.2701 0.07163 1.684 16 16 4.069 85.52 0.9958 8.02 9.45 1.1268 1.2701 0.07163 2.198 92.84 17 16 4.069 9.90 1.0040 9.45 1.1268 1.2701 0.07163 2.381 91. 79 18 16 3.037 2.64 0.9872 9.28 1.1268 1.2701 0.07163 1.555 50.75 19 16 3.037 8.10 9.28 0.9946 1.1268 1.2701 0.07163 2.198 68.81 0.9809 20 8 0.714* 1.68 9.52 0.8570 0.9931 0.0592 1.3208 108.8 0.9912 21 8 0.714* 2.17 9.52 0.8570 0.9931 0.0592 .1.4224 121.5 22 8 0.714* 4.70 9.52 1.0029 0.8570 0.9931 0.0592 1.8669 159.6 0.9944 23 6 0.714* 10.76 9.52~ 0.8570 0.9931 0.0592 2.6416 128.4 24 9 0.729* 1.0008 1.11 9.35 1.2827 1.4427 0.0800 1.7526 89.1 25 9 0.729* 3.49 0.9902 9.35 1.2827 1.4427 0.0800 2.4785 104.72 1.0055 26 9 0.729* 3.49 9.35 1.2827 1.4427 0.0800 2.4785 79.5 27 9 0.729* 1.54 0.9948 9.35 1.2827 1.4427 0.0800 1.9050 90.0 0.9878 Thesa are Pu02 in Natural UO2 - -
- Cases 1 though 19 are with stainless steel clad. Cases 20 through 27 are rircaloy.
TABLE 3.1-2. IdESTINGHOUSE UO2 2R-4 CLAD CYLINORICAL CORE CRITICAL EXPERIMENTS PITCH CRITICAL NO. EXPERIMENT CONCENTRATION SUCKING FOR k
, ,p FUEL REGION (ppm) LEOPARD CM-2 (on) 1 0.600 0 .008793 489.4 19.021 2 0.690 0 .009725 0.9912 317.0 17.605 0.9941 3 0.848 0 .008637 251.6 19.276 4 0.976 0 0.9927
.006458 293.0 23.935 0.9935 5 0.600 306. .007177 659.9 22.088 6 0.600 536.4 0.9927
.006244 807.2 24.429 0.9937 7 0.600 727.7 .005572 950.2 26.504 8 0.600 104. 0.9940
.009165 546.3 20.097 0.9919 9 0.600 218. .007599 607.1 21.186 10 0.600 330. 0.9917
.007601 669.5 22.248 0.9916 11 0.600 446. .006661 735.3 02 0.600 23.315 0.9909 657.1 .005809 895.3 25.727 13 0.848 104 0.9944
.007320 321.0 21.772 0.9938 14 0.848 218. .006073 420.5 24.919 0.9925 0.9928 Mean 0.0012 Std NOTES:
(a) Fuel Region Data Enrichment = 2.719 w/o U-235 (b) Thickness of water reflector is that required to attain Fuel Density = 10.41 g/cm3 total radius of 50 cm for model. Pellet Density = 0.20 in C?ad IR = 0.2027 in Clad OR = 0.23415 in - (c) 8 = .00527 an-2
\
3-15
TA81.E 3.1-3. SATTELLE FIXE 0 NEUTRON POISON CRITICALSUI) Length No. of'
- Case Absorber Absorber #
Times Assenbiles * "' k err Ididth* Type Thickness $'E' '" In Array Cluster of Clusters LEOPAP.D/PM 020 20 17 3 Boral .713 cm 017 22.21:168 * .645 cm 6.34 ca 0.9932 3 " " 002 20:18.88+ 1 *
- 5.22 cm 0.9944 2.732 cm 0.9925 028 20:16 3 5.5. 485 cm .645 cm 6.88 ca 027 20x16 3 5.5. .302 on " 0.9946 7.43 ca 0.9935 032 20x17 3 5.5. 1.1w/o 8 038 .298 cm .645 cm 7.56 cm 0.9933 20:17 3 5.5. 1.6w/o 8 *
- 7.36 on 0.9931 0028 20:18.075 1 None - -
015 20:17 3 "
- 0.9956 0 13 20 16 3
- 11.92 cm 0.9942 022 20:15 3 "
8.39 on 0.9945 021 20 14 3 " 6.39 cm 0.9933 4.46 cm 0.9946 N t f 3-16
s TABLE 3.1-3 (Continued) BATTELLE FIXE 0 NEUTRON POISON CRITICALSIIII STATISTICAL
SUMMARY
Series Number Mean kegg a Boral 3 0.9934 0.0008 S.S. 2 0.9941 0.0006 S.S. (8 orated) 2 0.9932 0.0001 Fixed Poison Total 7 0.9935 0.0007 Non-Poison Total 5 0.9944 0.0007 Overall 12 0.9939 0.0008
*This is in units of pitch (Pitch = 2.032 cm).
x center assembly was 20x16 and the other two were elongated at 22.21x16.
+ 20x18.88 was one assembly with a boral sheet on two sides.
Fuel region data: Enrichment = 2.35 w/o, pellet radius = 0.5588 cm, Clad OR = .635 cm, Wall thickness l = .0762 cm, Pitch = 2.032 cm. ( 1 i 3-17
TABLE 3.2-1. FUEL BUNDLE CHARACTERISTICS **
-r Fuel Bundle ,
Geometry 8x8 Active Fuel Height (in.) 150. Rod Pitch (in.) 0.64 Assembly Cross Section* (in.) 5.12 Fuel Rods Material sintered U02 Pellet Diameter (in.) 0.410 Pellet Immersion Density (%TD) 95.0 Claddina Material Zr-2 Outside Diameter (in.) 0.483 Thickness 0.032 i Water Rod Material Zr-2 Outside Diameter (in.) 0.591 Thickness 0.030 Spacers ' Material Zr-4 with Inconel X-750 Springs Number per bundle 7 Fuel Channel Material Zr-4 Inside Dimension (in.) 5 278 Wall Thickness (in.) 0.10
- Fuel assembly is comprised of 64 - 0.64 in. x 0.64 in. cells: 62' fuel rod cells and two water rod cells.
** Licensing Topical Report, General Electric Standard Application for Reactor Fuel," NEDD-240ll-A-4, 82NED057, Class I, January 1982.
8 3-18
Figure 3.2-1 BASIC RACK CELL GEOMETRY
~
AND DIMENSIONS (ALL DIMENSIONS IN INCHES) A a A n A r / M
. , 14 C ti %
9 9 k *0 M
/ N U ~
w m /
@/ z y
_ 1
'N'N'N'N
/ / / /W / / /l @
MATERIALS
- 1. Homogenized Fuel Pin Cells
- 2. Explicit Water
- 3. Explicit Stainless Steel 4.
Homogenized Water Fod Cell S. Explicit Boraflex'
/
3-19
TABLE 3.3-1.
SUMMARY
OF PERTURBATIONS TO THE MULTIPLICATION FACTOR OF THE BASIC CELL Description ak Effect k= Basic Cell at 68'F, 3.60 w/o U-235 in enriched .9179 center section of bundle, .028 gm B-10 cm2 in 0.106 inch Boraflex Calculational Biases Axial Leakage net effect +.0026 LEOPARD /P00 Model Bias +.0061 Zr Channel Stored on Fuel Assembly +.0045 Basic Cell including Biases .931) Tolerances and Uncertainties Minimum Boraflex thickness ,.0006 Minimum pitch (i.e., box to box) +.0018 Tolerance on SS box thickness +.0001 Maximum pellet density +.0014 Fuel position uncertainty 0 Calculational uncertainty (ko) +.0022 l l Total Uncertainty (statistical) +.0032 Extra Assemb}y Accide.wt Onut.t.o cacID
- oogs 5!keE' k"N1- c.Yme'nt5*p k di((cibidion ,8$
hesYgn@@rIatism
.0300 Spacer Grids
.0029 Maximum, including design conservatism and .8983 spacer grids 3-20
. e
s' TABLE 3.2-2. FUEL BUNDLE VOLUME FRACTIONS (68F) 002 PELLETS = .3123 Zr-2 CLADDING = .1113 He = .0138 HO 2 = .5626 I 1 I 3-21
\
Figure 3.2-2 BASIC RACK CELL GEOMETRY AND DIMENSIONS (All Dimensions in Inches) Zr Channel on Fuel Bundle A J n n 4 4 r S x z ou Db-12
$' N xN 9l D
e k@ Ui(Y CY S
/
w a N \ /
@/ N NZ ,e l
v xN'NNNNN'@NN (8 \Q ir
\
,r
///$//// $
MATERIALS
- 1. Homogenized Fuel Pin Cells
- 2. Explicit Water
- 3. Explicit Stainless Steel
- 4. Homogenized Water Rod Cell S. Explicit Boraflex 6.
Zr-H2 O (1.29/1.00) 3-22 I
l l l 1 FIGURE 3.3-1 PRELIMINARY ~ INFINifE~IIdli~1 PLICATION FACTOR I VS 1 FUEL BUtlDLE INITIAL ENRICHMElli . (slightly enriched central section of bundle)
'- !!dT ! J-!iNI4:i: ' lM - - l .. .
-l- ! I i.
.i -l: ik:.:.l-I' .- ;-l ~ . ; l I
. 95 -l.. -l j... l l- . p .j
. _ . g .g ._ .. .. . p . . ..
.g.
; j g.
..,,l. ;. <
., i
- MiliSL:W=
* =It- -c.+1=_, l== - ii i/: -lO-i,:1
;+ning p.=b i--l. e-l.:
- i= 1
- . p. e . . . = p +i. i . . . - :-l i .
i.. i- -i . ij
+ :- t=id=lPRELIMIN ARY l-
- i / : -
! eli
} -i.a.n. _j- .i - h:.i :;.j -- :l l- .l
' :-_:! "i ri:7?;- l5.d E
/ i : .. : l:
'l- ~l I" *'l- -
li
..; . . l .
. c g . .. :: ~_. g . . . . . . , . l
} j.- g .g l- - l. in I-.
- 1. . I ! : I i t I 90 ' ' '" ! I -'
- .q -~ I" ' - I
--...i.'pIli-- '"
~ ' '- I "'# -- ! ~ ' .'"..' tI. - '".~. :'ll. '
i - i . .- l . 1 - l l- i i
- i. ! ' -i- l - -l
~ ! ! I
! l !
{ .F .:$ - :.;. ! j- ! g ! - gggg'gyy pQ l
.l.
..l- J.
i 3
,f g b b ,a c y s 3 - -
- g 9 ;
" - l . i .. ) .i... l - I - 1 I l ! -
- . .u ni. =. ; ; i
) -lf. 'j;; i a.l" . ; ;-
l.
- l i : :;fsl -- r I -
t .
.j ;... j.: :jE: [
l j j- l -t
- - i --i : [ tu-y : r-i j
l -(
- l. i l l- : :.-l m:;
- i. ..g -
I 85 - --
.- l i l .
l l j
. , . .. l . ,. . ..i. . 3. .._ ;..
9 j '. g- .l . - i --i-- }- - : - l !- ! j -- :
- _ . L .. : _ ..p ;. . j . ._ ir4 . .: . ! - . '
; l ar -l - - - i :-- l i' l )- i l t i .
1 - l J
! i i t I .
i i i [ i ! I i =i ; i i l i I i
.[
! I i ' [
i I 80 I t [ i i I i I s ; 2.0 3.0 4.0 I:llT!AL E aICitMElli (w/o) 3-23 N
i FIGURE 3.3-2
. P,ui_IMINnRY INFINITE MULTIPLICATION FACTOR VS BORAFLEX B10 LOADIttG
, _.__..,.._.:m...,.......,._..
.7 . . _ . , . . .
.93 - ~ I" = i - 'i*I *~ '~'~E ^I~
. : _- :- - - " "I--~ ~ ~I ~ " ~ i
_ i. . : : :. . l - g::= :27 := ::u_ l .l l l . . ,. . . I :
..:::-[- :- z ;- ! ;.l. l:
.i
- -- . -- i : g,_. ,. , r cri=I. . 2:9 . . .
. i-l . . .. j
. =: . . . ._._x- -g
, . ' . . . .. ... .. ..- . :-i ; I '. ul - ? *
' a-i r h.:. ; }
. . , . ::=. . . - :_: . ; . . . .
nr . ::.: . - 25 53'I#E::-h ~ !I.l-3!ii : N:.- l i- . . .
<.i .~l N =is;i s.--j .! rI 7 '
Jicii 'j :\ -
. l pg ggg t" .;
- =. _ . -s .c. l.. : : 4.. . :. .. .
. - 1. -.
. . ~'
- i. [ : ..l j
-ii-[@:. [i-i -fikei.5 f i-hi.- l.. .
t: . .i :;
- l:. -l' .I-sp@.. p::l.3-):f.hi:;d .-j .::; : '
l i .j;- j.r-- . j.
.925 -- ..l.r. i:;;).;~5i-4:ni -i f. ;.e l l j.
. . uMn- i.,:b,:h;:- 4:~yt.- l.,.g- j :
} l [
. i- i l2 ~ , j .; . . l
*~l2-! I-l~ 6i> b Y il M ih1LIRV :I -I
*:4':2 -I
...[. ..:. p.. ,... ;.J..4.1.m.*
...;. t**'. ? " ' ' ' ' ' . ;" . . } .
ir; li :;i.l:: r :l f .7 l :i , l : . l .', .
....r .T k= -l
-l . .l -
;:.i' i
.l. 4; . . j . i. . j .
l_ ; o
=: . _2 l ~.. . /J. I-i..: . . . . - u.-l. . : i
.. i j ! .
. l' *_ . .
. l- .:- t.. p n-l;. . .:.. [ . .j i j ;j .): .;
. . :.:-. l .:. en: j.: . r .l . .. .:- . 9. . -
j . t . l .l -l --: -l ._ ._l - i. . l ...i.: .;. :: . . I:....
-\.. t : -l : - I . ~ .-l . . : -. l :
- l. ::l_
.92 __ . i ::: . . ;.:.. .. h 12;.l :.:. . t..;; .
r-
. .l ..! . -- h_ . . l-
.: 1_.:: lu. :..:::, .:;: -- --- i:: .. .. l ._ . -
I.
. . . l . .
- ]. . . -- g~
l- ::- ).-- 1:- j p .i;-ir ::---l. . .- - rj -
~
-[.,- . M. :5l.ciGy F..: .' ; f.
; j .l -l - q
- l
- i
.l : g. -. : . . p . . . g . . . g.,
. i --[. :. ] . W . .
, ,l g g. . . l . [
_ -l :--l . i -l - , -j- . -. j j., j
.l :l + j.-r.l; i . - l- !- '
i j -i. ) .I l' : l-- J _ .l . .
) j.., ; .; .; j...j
. i. . . , . . . ,
.i. r
, t
. i. .
l I e i i i l i i- I; 2 i l } l i I >
.915 . e
! ! i
! -i i !
.028 .
.029 .030 B10 LOADING Ill BORAFLEX (gm/cm )
3-24
6 FIGURE 3.3-3 r nEumliuR ( IrlFlfilTE flULTIPLICATI0ft FACTOR VS TEMPERATURE PREllMINARY
- :I We .: l !. - .
l 1. - i i fi i-l :::: .33 ., i. -
!- . i:. .I
- l.: . -l l l. . ) :. i. ..;
*95
* '4I '
; p~ . ~' T- ---%. ::,,M@i:[.h i"i:
E'^ I
- " I" "I'F- I" '** I
.p:, :::{d::;:.:.'}:12; .. lf. p:-li::ds:l- : ;:l
[ @: M i-5!!5lE I::'I9 e:: =c :==qa:g :- l:ig:: ' i+.i il ! i:i- il:-!:*t :it. l" 14l: !+:( : :l
- -l l l
- ! . . i . :. - -l
- ! :!:=-i=iI:. vic : - : -) 1 .i !
I !: - i
. u_ l i:.i.:.p:++1- -i:- i :l -
94 -- l . , . , i
-l .i
; i..: i. f +: : l ..; - l , .. I si. .-l-
.p ;. :- f.l : -- l .
1 i : i- ; : . l: .
.l i i. i i
- I . i: -I , -
l .1 I i l l-i 1- 1
. i':. }a..:: .Li. ... ..i: j. . ; ..j ). -
.!- ! :. <-. . l.: q .:
;. l ..l .' .: l .. . ij k .
- l. .j
- e. *93-- - . ,
l . l .
- l. j - tl
.i . - .)::iit . . : .i . . l l- I, l l l i i- : 1- i . : i .!. i : ! .. i i !
= !: :. ! = 1. ! : i
-- PRELIMINARYi:-i
,92 _--
i i >\_i-l 1 I i
. :j- :i.;p
~~'
! I
..i IY I I ~l 4
.z_ , , . .j =: . j i: i. .. }.
. -l - -[:+.- !- .
I i t. .i..j-; I : l-- R REL lMINARY -
~ l' I i l- .l.. .~l 1 : i i , i I. i i
, 91 _ i
* .i+ i .- - . 1" 4 a .. i
. , n :. j .
. i i i j..
l N.-i-! i
, .N , _.. j
.-: i .L-- ' -I ; .i . J. i j. ;N
.i i :- i- l 1 l l
.i i-li t. i i ;
, I ! l
! i.i i i 1
.90 ' I i - l l g., : I
-e' i . .
0 , 100 200 TEMPERATURE (F) 3-25
- p. .
FIGURE 3.3-4 PRELIMINARY INFIlllTE MULTIPLICATI0ll FACTOR VS RELATIVE WATER Dell 5ITY
] ::.i :: l :- i;i:. !- H1 .I j 1.0 ii.i-j ' q g-ij i: i . :-j:. :p. .1.
} .
.}. l .l - l- l
- ) .
;; :{_: .: -i :. ".. i :.
l. 1 -l H :-l - l g). . l . ; .: l .. - l - !.- .j.. . :... l.; Tslfi u E'E.* ?.'il?
- :jfj _;.i.- : fii!i-
' fi"I 7 i \ ! 1 l:::*!" *i-hTl:i :
.i@i: .$.. l, !.j.. l ,,
- =-
.}:h.i::. ! : ;: .l_.. g.i j .2:t_;_
2...--, :. . - : I;i ; .: i i:Wlr.Pi l - i .;l- : 9' 0*9 5-. r : F:H .l T - l .t :-l -- l - i -l - l9!-l2 "- i
-l W:m . If
.: 2 W i5."i :: .: le t.E! - -; i.
i
--:l : l
* : If' i .:1
- - l-
.l bp_gtyfijgfRy1: l i
.. i
. ... : .f..:. l.p.i .
i i I t; i I q ;- i . l : .i
. :-l :=i::H.. - .I l/ ; ; ;-
.l 0.,<, 1 1 ,
-i:!.-- l a jt-1 f! 1 1 -i i
g ,,= g , rg r g. , _; - , i-
.) :) i j l=> l l
..,. ;j l .
i l' I I ! I 1 i i 1 ..l I
..I I - i. i i
I . I i 2 L. _ l I i k 07- -- i ! i...i ; trHt:LIMINhK Y ..I l i 1 i i i t . i. .' ;-- i l l Ji i - i i ! I
! i iz -l
~
! t/ i : ;
! ! I !
l :.;, I i
/ - I ! I i -
i i I. . i/ l .i :! .F I I i i i I
.i .j- .. l
! 1. . .I : 1. ti
- 0. 6 - i. i . , ; .; j
~
~g ,
; , . ... g, g , ,
l ..! .}...j
; .g
,g.3 3 l j i -
l : I ! I I I
- -l i
i :; -i i ; : ' i I !. l i :._ t l i j :.J i . I
/ ; .- l - ; i : i I : il i 0.5 - l- -
;. ; ...j .i j.
l t : :-: --l .- i., i,; l l i } ; i l i i
! ..l ..j .
.[
i 1iI i ! - l ! - i i I t ? I i t - t . j
. I l !
I
- i
! I _i !
I 1 0.4 - .i .- - -
-I
- ( ' t i
i t 0 . T-
.2 .4 .6 .8 1.0 RELATIVE WATER CEttSITY 3-26
mili: t itl l I i.la - li: iir l i i i:1 ! FIGUEE-3,3-5 *
- i :. ":t;
- 4d
.." l': FI"
'j. :i
'ti aj.' -- "
1 ..j :l-I' !' FLEb-ASSE/4Bl:V- k-) 4:lMRS-R4M
- it .l:
!= j. ~1N
.].. ;.jr I LIAL- EN'RICHMENLOF_4.55Lw'/ok235 l
I' "I : I' I d l l i [ '
, i l: ; l l -
l-l: l I I I ' i: l ' l i j l j. . - l'- : 1.: . ;- l - i i i. l : i. -l L -!- i i i- J i: I !- l ! 11,00 __ !:: +- t.l : I: !' i .j . I ~ !' 1 -i" 'l+ 3 j.
; j ;
l -j. :.1 i l- f p l-
- - - i j j i- '
i n.
- l. i- ..!-
i
.t .
l l i
- l- !! '*!
- i. .i.. .. ,
,c .
1 ..:
. . i . -- 1 o
.i i: . w. ...
. j. :
i- ;gis ..l.;. m i :
- 1:. ;. p t;. I' l-puse e
l l :l. ;: :4p .i:- .j..-i'- ~i l: = i. t 9
- j. [- i
- t .-
i j l- .!- ' i.
- f. l : . t I
i i
- l.: l l-i ! I t- r! f i- !z !,
i q<_ 1 I
! I i Ef 'i1 l <i ;
; i '~~_i l- l-i i
Il /! ' i i ~f~
! t- 11 ..
! i. ip _i
! i: i ! 1 _ -i- i i i: i i
! i i I l i If i !,
-T !: ). . l l l l .ii lPf I.. i .
i I
.!r R:
i
- - .. t- ~ ,
_.a
* ! i l- i i i ! '! i -
g ! i - . i i- M. . I i ! t-i i I i i ! L' !
-i :_j i ! l: l .! l i /! ii ~
7 i l-i ! i 1. i . I is . l -l ! l i / !i !- f. -! .J l- !
~ . .i i. I l- -l i ! : :
/. !
-! . i i l
; g__! :l- l- t i i
!/ -
$ J. i ;
i- ' l .
.! I ii
i;{ 3, i i
! I
! I I
i
/
/ t i
i i i-t- 1 1 I-I I- j i ; : - -
} i
~
{- t ,
! ! l . .
$: i
! I I'~R
, i t ! ! l ! l ., !
i i;- i ! i- i
! ! ! l ! i !
' ~~
i ! - I : i i ; i l 'T' i i ! I i i ; ' 1 .! l j i ; ; ! j . .
- l j i ! i
-l i 8
- -- -{i
, i i
,3 1 -
- , I , . .
. i. .
+
.s I ; i '
i
. . l- i t Il i ,
1 ! i
'0 1 i /
!- ; I .
il1 ! i I j.!.
. :._. } . - _.
!. i
. ,i !_ ; !-
i !
- r
[
=1 i :
1 - - L.._, ' l l , l g ; l i i l ; ; j i i i s '
} i
.L . , _.._.L ? ' l:
e i . .i i ;
-. _ .!, . ._ j '
_ - i - i 1 L__I
! _. I i i ! I i i
- ~
i j ,1 ( l' I i I . i , t i i l ! ' I , I [ f
._J_1!. ? ' i -
-I i
. a 1.3 ._ l 1", 4 ; -- '
l ! I i ' .l . . _ L ' ! : Li ! I ! ' FlJEL ASSEMGLY km I~ t. I ! 1 ii i I i _i_ . ._' .. . { . iii i i ii __l j i i ) i i I !
;; i i i l i i
i i l 3-27
em t' - ' 1
,j
,4m a, \
.ghM ! 7, .
l j FIGURE 3.3-6. BASIC REFERENCE CELL i With Explicit Enrichment Distribution This distribution was obtained by increasing an enrig distribution similar to that shown in Figure 2-2.26atgspent i by the ratio of 3.416/3.016. Such a distribution is typical and certainly conse,rvative'since no credit was taken for the large negative reactivity ef fect of the gadolinia burnable poison which would necessarily be present in such a design.
/d 2 / /. 4- /
4I
/ 2 3 4-
/ /
/ 3 4 5 '
/
4 4 5 7 l
/
2 9
'N
\
//AV//// 2 MATERIALS
- 1. Fuel Pin Cell, 4.474 w/o U-235
- 2. Fuel Pin Cell, 3.738 w/o U-235
- 3. Fuel Pin Cell, 3.398 w/o U-235
- 4. Fuel Pin Cell, 3.058 w/o U-235
- 5. Fuel Pin Cell, 2.718 w/o U-235
- 7. Fuel Pin Cell, 1.925 w/o U-235
- 8. Explicit Water
- 9. Explicit Stainless Steel
- 10. Water Rod Cell
- 11. Explicit Boraflex (1) " Licensing Topical Report, General Electric Standard Application for Reactor Fuel," NEDO-24011-A-4, 82NED057, Class I, January 1982. .
3-28
p .4.0 Dynamic Annlyais 4.1 Introduction The spent fuel storage racks are classified and designed 4 as Seismic Category 1 Structures in accordance with USNRC Regulatory Guide 1.29. A non-linear time history analysis is performed in order to provide sliding, and tigjsng ~ loads. The non-linearites inherent in the analysis are: l
- 1. Fuel to Cell Wall. Impacts
- 2. Rack Sliding
- 3. Rack Tipping j -
In-house computer program called RACKOE, was developed by i Professor Stokey, one of the co-authors of The Shock and Vibration Handbook by Crede. This program was based on the i three-dimensional analysis performed on the SNUPPS Project Spent . Fuel-Racks using the ANSYS computer code. The RACKOE code was later verified against'ANSYS by Stone and Webster Engineering Co. 4.2 Equipment Description and Material Properties i 4 General arrangement of spent fuel storage racks for the i La Salle County Station - Unit 2, is shown in Figure 1-1 The array is composed of 20 modules designed to hold 4073 fuel assemblies varying from 180 to-240 each. ) Each module is composed of vertical stainless steel thin-j wall rectangular boxes welded together with local fusion welds.
- All boxes are about 6" x 6" x 168" long. Each box contains one j
fuel assembly. A fuel support plate is welded to the bottom of each fuel box. This plate has a chamfered hole which provides lateral location for the bottom of the fuel assembly and provides i the coolant passage for the fuel. Each rack is supported on flya vertically adjustable pedestals. i The new spent fuel racks.are fabricated from Type.304 stainless steel. The 304 stainless steel rack material proper-
, ties used in the seismic analysis are: Reference #8.
j Density = 501.0 PCF Young's Modulus = 28.0E06 PSI Shear Modulus = 10.7E06 PSI The fuel assemblies contain clad and channels constructed of zircaloy whose properties are: Reference #9. Density = 409.0 PCF Young's Modulus = 13.0E06 PSI Shear Modulus = 5.0E06 PSI 4-1
1 1 Other densities used in the analysis: Water = 62.4 PCF UO2 = 643.0 PCF 4.3 Summary of Results Table 4.3-1 shows the resultant vertical and horizontal dynamI1c loads per pedestal for a 240 cell rack. These loads are used in the mechanical analysis to determine the stresses. TABLE 4.3-1 The maximum forces on one pedestal for the seismic disturbances are: 2 2 2 Vertical = submerged wt + /F NS +F EW
+
VT Horizontal = +F EW NS 240 - Cell Vertical, (LBS) Horizontal, (LBS) OBE 431,700 92,520 SSE 497,700 118,700 4 6 9 4-2
4.4 Rafarencen
- 1. Specification T-3758, Spent Fuel and Special Storage Racks, La Salle County Station - Unit 1 and 2, Common-wealth Edison Company, Project No. 7043-73.
- 2. U.S. Nuclear Regulatory Commission, Standard Review Plan 3.7.2 " Seismic System Analysis," Revision 1, July, 1981.
- 3. Pritz, R. J., "The Effects of Liquids on the Dynamic Motions of Immersed Solids," ASME February 1972.
- 4. Dong, R. G., " Effective Mass and Damping of Submerged Structures," UCRL-52342, L.L.L., April 1978.
- 5. Stokey, W. J., Scavuzzo, R. J. and Radke, E. E.,
" Dynamic Fluid Structure Coupling of Rectangular Modules in Rectangular Pools," ASME Special Publi-cation PVP-39, 3.979.
- 6. Regulatory Guide 1.61, " Damping Values for Seismic Design of Nuclear Power Plants," October, 1973.
- 7. Rabinowicz, E., " Friction Coefficients of Water-Lubricated Stainless Steels for a Spent Fuel Rack Facility," Study performed for Boston Edison Co.,
November, 1976.
- 8. ASME Boiler and Pressure Vessels, NUCLEAR VESSELS, Section III, 1980 ed.
- 9. G. E. Technical Paper 22A5833, Rev. Dec. 26, 1979, Appendix II, FUEL ASSEMBLY STRUCTURAL CIIARACTERISTICS.
- 10. R. D. Blevins, Ph.D., FORMULAS FOR NATURAL FREQUENCY AND MODE SHAFE, Van Nostrand Reinhold Co., N. Y.,
N. Y.,.1979.
- 11. S. Timoshenko, S. Woinowsky-krieger, THEORY OF PLATES AND SHELLS, McGraw-Hill Book Co., Inc., N. Y., 1959.
- 12. R. J. Roark, W. C. Young, FORMULAS FOR STRESS AND STRAIN, McGrow-Hill Book Co., N. Y., 5th Ed. 1975.
i ( e 0 4-3
4.5 Data Preparation The analysis is for the fuel rack (shown in Figure 4.1) which can be placed in any position in the spent fuel pool. The finite element representation of the defective fuel storage rack is shown on Figure 4.2. The symbolic representation is: MASS ELEMENTS l-6 Mass of Rack (horizontal) 7-11 Mass of Fuel Assemblies (horizontal) ' 12 Rotary Inertia of Rack and Fuel FLEXIBLE ELEMENTS 1-5 Rack 6-10 Fuel Assembly 11 Horizontal Restraint 12,13 Vertical Restraint FRICTION element attached to MASS #1 MHrw HYDRODYNAMIC MASS (rack to wall) MHrt HYDRODYNAMIC MASS (rack to fuel) REDUNDANTS i
=
4-4
.. .__ - - - - - - .- - . - - 2__ - _
O l
,/h DI R EC TIC H CF LOAD / {
'l 5
10 l 5 10 % 9 l 4 9 4 MHrw' MHr f 8 i.. 3 7 g _ _ *_ - ' . horez 7 h/W(1) e
/
W k 'i h [iB 31
~ jrvere S@ 14 NG 3
i
- i; ,.
F bcriz hl_ I '
. Fvert FIGURE 4.1 i
- I 4-5 VW 0 *- C A, _
,a 'u a. - - - .
O FIGURE 4.2
* *w r
i J
~~ , x , ;
^
^'Auc I #
hhh , 4 7 4 I g
"2 .<i..i Sa , / sac. s WIMN T
,i 4
N ( )/)ii f -- n e-J r- , , . ~2z I }%I (0 34 y5 go - 2
.N 4 -.-
_7_ f O% ,
-'~
fMl (7 -v
/ 6 /
d A
// -
W i ca ( 2 - d }r 4-6
5.0 Mechanical Analysis , 5.1 Summary The new spent fuel racks are free standing poison wall design. These spent fuel stcrage racks provide smooth full length square storage cells of stainless steel in a welded honeycomb structure. Each storage cell, except on the periphery of the complete array, is bordered on all four sides by Boraflex neutron absorbing poison sheets sandwiched between adjacent cell walls. Each rack is supported to the pool floor on flys pedestal structures welded to the bottom of the rack at the four corners. A screw adjustable pad is provided in this structure to be used for rack leveling. UST&D provides the appropriate tool to make
! these adjustments from the surface through the cells over the 1
pedestals. The male thread is chrome plated to Federal Speci-fication QQ-C-320B Class 2a (.0002 to .0005 thickness). Our experience shows that this is the protection needed to prevent thread seizure during rack leveling, especially since the rack weight must be supported on the installation crane while pedestal 4 adjustments are being made. After leveling operations are com-plete, the pedestal thread is no longer a moving part. The height of the bottom of the rack above the pool floor, resulting i from the necessary vertical dimension of the pedestal structure, provides adequate underneath space for cooling water flow, i The individual racks are sized to fit into the space avail-able in the pool. .W here rack pedestals coincide with leak chan-i nels or floor liner seams,-there will be floor plates under these pedestals which are grooved on the underside to span the trenches or floor liner seams and transfer loads to acceptable floor areas. The rack sizes are dimensionally compatible with overlar.d trucking constraints on width and height. 5.2 Description of New Spent Fuel Racks 5.2.1 Module Construction
- . The-La Salle rack design is a honeycomb configuration of identical stainless steel boxes with sheet Boraflex poison material captured between the side walls of all adjacent boxes.
To provide the space for the poison sheet between boxes, a double row of matching flat round raised areas are coined into the side walls of all boxes. The raised dimension of these locally formed areas on each box wall is half the thickness of the poison sheet. The boxes are fused together at all these local areas. The poison sheets are scalloped along their edges to clear these raised areas. 5-1
The poison sheets are axially centered on the active fuel region of the stored spent fuel assemblies. They ar'e physically captured, as stated above, between adjacent box walls and within the double row of raised areas on the box walls. In addition each is effectively contained axially by a narrow sheet of stainless steel positioned at the bottom of-the poison and welded across one of the two adjacent box walls.- This sheet is the same thickness or less than the poison material. Between adjacent racks each poison sheet is held in place on each outside box of one of the racks by a thin cover of stainless steel which is welded intermittently, all around, to the box wall. All of the poison in this design concept is unsealed. Each fuel storage cell has a welded-in bottom plate to. support the stored fuel assembly. It has a chamfered centra 1 hole to accept the tapered nose of the fuel assembly and provide for the cooling water flow. The storage cell material is .090" thick stainless steel. The sheet Boraflex po
.075"thickwithaprobableB{gonmaterialisnominally loading of .020 gm. per sq. cm. The thin stainless steel cover sheets are .025" thick.
MATERIALS
- 1. Poison Design
- a. Spent Fuel-Boxes, Channels, Stainless Steel Poison Cladding, Bottom Plates, Type 304 Pedestals, Shims, Lead-In Guides, and Rack Appurtenances
- b. Poison Material Boraflex (See Appendix B)
- 2. Special Tools l a. Rack Lifting Stainless Steel i
1 Type 304
- b. Spreader Bars for Rack Stainless Steel
, Horizontal Lift and Rack Type 304 Upending
- c. Vertical Spreader Stainless Steel Type 304
- d. Cable Slings and Straps Commercial j as required i
l 1 i
- 5-2
5.2.2 Rack Fabrication The fabrication and assembly of the Spent Fuel Storage Rack is manufactured in strict adherence to in-house controls and tight tolerance, as well as those described within the client specification requirements. Our expertise, as well as experience, has enabled us to control growth, shrinkage and distortions by means of close tolerance manufacturing and special fixturing. A majority of our parts and sub-assemblies are machined to tight tolerance, thereby preventing, wherever possible, undesirable accumulation of tolerance and unacceptable deviation from required alignment. Finally, our experience has proven that the added time incurred by close tolerance machining is offset by the relative ease in assembly and final function testing. The following are typical requirements used in fabrication and assembly, as well as general information used in our manu-facturing: The fuel boxes and water boxes are formed in channel sections, fusion welded continuously-full length, and constructed out of two (2) full length sheets. Each box . (fuel, water) is constructed out of two (2) sheets (to form channel section) continuous over their total height. All interior surfaces of the fuel boxes shall be smooth (125 AA microinches) . All weld beads shall be flush with interior surface. There shall be no protrusions inside the fuel boxes that will interfere with insertion or withdrawal of fuel assemblies or to result in marring, scoring, or other damage to the fuel assemblies. Racks shall be fabricated by welding unless otherwise agreed by Commonwealth Edison. The fabricated racks shall comply with all requirements of the Specification and the CECO. approved drawings. Racks shall be fully assembled into their final module size and checked for compliance with dimen-sional requirements, alignment, and clear-ance in our shop prior to shipment to the jobsite. Any errors or discrepancies dis-covered shall be corrected prior to shipment. Fabrication and installation of the racks shall conform to the general requirements of Sub-section NF of Section III, Division I, of the ASME Boiler and Pressure Vessel Code for Class 3 component supports. 5-3
Austenitic stainless steel shall be supplied in the solution-annealed condition. Except as required and allowed during welding, austenitic stainless steel material shall not be heated above 350' F unless it is subsequently given a' full solution annealing as the material Supplier's recom-mended temperature and holding period followed by water-quenching from the annealing temper-ature. Cutting, forming, welding ahd. handling of materials shall be performed under the supervision of experienced. personnel quali-fied to work with the s'pecified materials. Flame cutting is not permitted. - Contact with dissimilar metals, low-melting point metals, chlorides, halogens, sulfur, phosphorus, and other'potentially harmful materials shall be avoided. Drawing dimensions, geometric requirements, ' and tolerances shall be interpreted and defined in accordance with ANSI Y14.'3. Projections such as weld reinforements will not be permitted on the inside surfaces or envelope of the storage cell. Corners shall be rounded. Sharp edges shall be deburred or chamfered. Weld splatter, chips, burrs, and other foreign matter shall be removed from the base metal in their entirety and the affected surface restored to a 125-micro-inch (maximum) finish. Abrasive blast-cleaning is not permitted. The inside of fuel boxes shall have a commercial 2B finish, except where the fusion welding of the longitudinal seam weld is applied. Reworked interior surfaces that may come in contact with the fuel assemblies shall have a surface roughness exceeding 125 AA microinches. (Reworked interior surfaces for the failed-fuel and miscellaneous parts racks may have roughnessos not exceeding 250 AA microinches). The interface between the lead-in guides and the fuel storage boxes shall be blended to provide a smooth entrance and exit for the fuel assemblies. 5-4
- = - --
8-+ 1 All interior edges of the fuel boxes, lateral and vertical, which may contact the fuel assemblies, shall be finished to a 1/8" minimum radius and chamfer. If they are chamfered, blending of the 4 intersection edges is required. Threads shall be clean with all burrs and ragged edges removed. All male threads will be chrome plated. Each spent fuel storage rack assembly shall be prominently marked by means of Vibratool or stainless steel identifi-cation plate welded to the rack. All welding procedures and welder quali-fications shall be in accordance with the ASME Boiler and Pressure Vessel Code, , Section IX, and/or applicable sectionslof l AWS Section Dl-1, Structural Welding Code. 5.2.3 Quality Assurance UST&D will provide documentary evidence of material trace-ability, inspections, and' tests for each rack / shipping piece in 1- accordance with our Approved Quality Assurance Program. This documentation will be submitted with each shipment, of finished product, to the job site. l Quality Assurance Plan Outline l This document describes the Quality Assurance Program Plan which U. S. Tool & Die, Inc. (UST&D) follows to control the activities required for the design, procurement, fabrication, inspection, packaging and shipping of nuclear power plant struct-ural components involving Seismic Category I structures, systems, i and cca.gonents, including their foundations and supports. , Classification of structures, systems, and components ! shall be as follows: Modification to existing structures, l systems and components shall be designated the same seismic classification as the existing system. New structures, systems and components shall be designated 1 a seismic classification in accordance with the guidelines in the current edition of the USNRC Regulation Guide 1.29. l Defined are the management systems, procedures, and controls l established to assure conformance with applicable section of Nuclear Regulatory Commission Regulation 10CFR50, Appendix B, ' Quality Assurance Criteria ~for Nuclear Plants; American National 1 l 1 } 5-5
,. .i Standards Institute N45.2, " Quality Assurance Program for Nuclear Powar Plants"; American Society of Mechanical Engineers; Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components and? applicable Nuclear Regulatory Commission Guides.
This plan is subject to continuous review and shall be revised when, required to reflect latest changes in Quality Assurance Pclicy and Planning for fulfillment of Contract Requirements.
. '. Definitions of terms used in this plan are consistent with those in ANSI N45.2, Section 1.4.
The UST&D Quality Assurance Program main objectives are: To establish and implement methods for controlling design and procurement activities, and analytical stpdies for assurance-that requirements for design, safety, and materials, and fabrica-tion are defined and correctly translated into design documents, procedures and instructions. To establish and employ selection, classification, and identification practices for materials, components, parts, and processes to be used for controlling and verifying quality of items and services throughout the project and specify traco-ability, quality, procurement, and certification required for the components covered by this document. To establish'and implement planning and control procedures required to assure that design, fabrication, processing, inspection and test activities conform to the latest drawings, specifications, proceduros and instructions' approved by Client, if applicable, or Engineeriag as required. It is the endeaver of UST&D.to design, procure, fabricate and deliver quality equipment in accordance with contract require-ments by systematic planning, controlled execution, and effective quality assurance supervision. The desired quality will be achieved by providing clearly defined design documents, thorcugh checking, complete analysis, adequate inspection coverage for proper process control, proper auditing of design and fabrication adequacy, and adequate records to document product quality. - Control of quality shall be implemented by adherence to policies and procedures described in this Quality Assurance Program Plan. ItST&D recognizes the Client's right to audit this Quality Assurance Program as necessary to comply with the Client's obligation under the applicable codes, standards, and regulations. Necessary documentation and records shall be furnished the Client on a timely basis. The Client shall normally schedule audits 30 days in advance or in accordance with contractual requirements. 5-6
UST&D is to design, procure, fabricate, certify, and deliver nuclear power plant structural components in accordance with the contract ordering data. The principal function of Quality Assurance is to assure , that an effective Quality Assurance Program is working. Quality Assurance (QA) monitors, by review and audit, the task of each project. The significant tasks selected for moni-toring are based upon QA assessment of program work scope and contract objectives. QA reserves audit authority to propose and implement hold points. Aud'iting is the technique utilized on all projects covered by this QA Program Plan. QA has authority to identify problem areas, initiate and recommend corrective or preventative action for resolution of problems at all levels of Project Management, Engineering, Procurement, Manufacturing, and Quality Control organizations. The Quality Assurance organization has the vested authority to prevent non-conforming work or components from being incorporated into the final product.- QA will be implemented by all Managers. Assistance will be provided by QA to consult, train and indoctrinate personnel, as required. Significant failures and malfunctions will be reported to the Client through Project Management. Thorough analysis will be conducted to determine the cause and appropriate action required to prevent recurrence. The failure analysis will be formally reported to the Client with pertinent information pro-vided as required. Action will be implemented expeditiously to resolve all problem areas. Documentation of all procurement, quality control, and quality assurance records will be kept on file with trace-ability to components being reviewed or inspected. 5.3 Mechanical Analysis 5.3.1 Basis for Analysis The effect of earthquake on the free standing Spent Fuel Storage Racks is addressed in the Seismic Analysis (Ref. 2). Section 4.0 explains the method and the computer program which requires earthquake input in the form of acceleration time histories. SRV and chugging forces are included. Loads and displacement values, horizontal and vertical, are generated. The loads from the Seismic Analysis are used in the Mechanical Analysis for rack stress calculations. These loads are also used to determine the acceptability of the pool floor loadings which in turn are to be used in the determination of the pool structural integrity. 5-7
5.3.2 Dropped Fuel Bundle Analyses , These analyses consider both the straight drop and slant drop of a fuel bundle. These analyses are supported by actual physical tests performed on the rack sections. The analytical results were confirmed by the tests. 5.3.3 Results of Dropped Fuel Bundle Analyses The results of these analyses confirm that the suberit-icality array is maintained in the event of this accident. 5.3.4 Summary of Results The detailed structural analysis shows that the spent fuel storage equipment meets the requirements of the speci-fication. Stresses in the internal rack welds are computed due to both deadweight and seismic loads. Seismic loads are the peak reactions from the seismic analysis. Seismic stresses are calculated using the SRSS method and are combined with the deadweight stresses. The analysis of any external loads such as the drop of a fuel bundle are supported by physical testing. The result of any of these accidents would not be catastrophic and would not destroy the subcritical array in the pool. Any resulting damage to the racks would .be minimal. The maximum stresses for the rack and rack components are listed in Table 5.3.4.-l along with the allowables, and factors of safety. The factors of safety were computed assuming that the Level B allowable stresses were equal to the Level A allowable stresses and the Level D allowable stresses were approximately 60% greater than the Level A allowable stresses. This is very conservative since, in fact, the Level B allowable stresses are 77% greater than the Level A allowable stresses and the Level D allowable stresses are 167% greater than the Level A allowable stresses. 5.3.4.1 Methods of An_alysis The analyses use the lo, ads developed in the Seismic Analysis Report (Reference 2) and applies them to high , stress portions of the rack and supporting structure. Stresses (axial, shear, bending and torsion) are then evaluated and compared to allowables to determine structural adequacy. 5-8
4 , TABLE 5. 3. 4.-l .
, STRESS
SUMMARY
l ALLOW-STRESS STRESS ABLE SAFETY COMPONENT TYPE LOAD (KSI) (KSI) FACTOR RACK WELDS - 4 240 B-B OBE 14.65 24.0 1.64 INTERNAL RACK WELD SHEAR SSE 28.77 36.0 1.25 ' (Fusion) . 240 C-C - INTERNAL RACK WELD SHEAR SSE 25.69 36.0 1.40 (Fusion) 4 i 180 B-B OBE 18.40 24.0 1.30 i INTERNAL' RACK WELD SHEAR (Fusion) PEDESTALS I RACK BOTTOM PLATE TO BOX OBE 16.95 24.0 1.41 WALL (0.09" WeJd) SHEAR SSE 31.52 36.0 1.15 PEDESTAL TOP PLATE TO RACK OBE 2.42 24.0 9.80 j BOTTOM PLATE-(3/16" Weld) SHEAR SSE 3.87 36.0 9.29 i I 1" THICK PLATE l BENDING SSE 7.00 18.0 2.57 l THREAD (EXTERNAL) OBE 6.50 9.6 1.47 SHEAR SSE 12.13 14.4 1.18 i i THREAD (INTERNAL) OBE 5.51 9.6 1.74 14.4 SHEAR SSE 9.93 1.45 POOL FLOOR LOADING OBE 3.72 4.17 1.12 BEARING SSE 3.91 4.17 1.06 5-9 .
5.3.4.2 Danign Crit rim according to the equipment specification, Reference 1, the design and analysis of the spent fuel and special storage racks shall be performed in .accordance with the requirements of the ASME Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NF (Ref.9). The racks shall be classified as ANS Safety Class 3, ASME Code Class 3 component supports, and Seismic Category I structures. The analysis is to be performed to include anticipated loadings such as seismic forces, pool hydrodynamic forces (SRV, CHUGGING, CONDENSATION, OSCILLATION), pool slosh forces, impact forces f rom accident and abnormal occurrences, thermal forces, applied loads and deadloads. The accident and abnormal forces shall include cases (a), (b), (d), (f), (h), and (i) of paraqraph 6. 4. 2.1. 3 of AN3I/ANS 57.2 (Ref. 10). These cases are: (a) Tipping or falling of a spent fuel assembly, , (d) Fuel drop accidents, (f) Horizontal movement of fuel before complete removal f rom rack , (h) Objects _that may fall onto the stored assemblies, and (i) Missibs generated by f ailure of rotating machinery or generated by natural phenomena as described in facility SAR. Per paragraph 305.12 of Reference 1, the loads are to com-bined in accordance with Appendix D of Section 3.8.4-II.3 of tha Standard Eeview Plan (Reference 11) with certain modifications. This appendix stipulates that spent fuel pool racks are to be designed to criteria for Subsection NF tor Class 3 component supports. The load combinations are listed in Table 5.3.4.2-1 The abbreviations and their signif~icance in the analysis of spent fuel storage racks are as follows:
- a. D -- Dead Load including the rack' deadweight and the entrained pool water.
- b. L -- Live Loads including fuel assemblies and fuel channels.
! c. T o-- Temperature effects and loads during normal
" operating or shutdown conditions. The Thermal-Hydraulic Analysis, Reference 5, concludes that the temperature gradient across the rack structure, due to differential heating between a full and an empty cell, is negligible, as is the temperature gradient through the thickness of the cell walls (less
-y?than2*F). Additionally, the pool tempera-ture after a full core discharge is shown to be approximately 120*F(with heat exchangers operational). This temperature (120'F) will, therefore,-be used for evaluating material properties for the load combinations con-taining this term.
i TABLE 5. 3. 4. 2-1
, LOAD COMBINATION ACCEPTANCE LIMIT D+L Level A service limits D+L+To ,
D+L+To+E Level B service limits D+L+Ta+E D+L+To+Pg D + L + T, + E' Level D service limits a D+L+F The functional capability d of the racks should be demonstrated. NOTE: The provisions of NF-3231.1 (Reference 9) of the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Division 1, shall be amended by the requirements of Paragraphs c.2.3 and 4 of Regulatory Guide 1.124 (Reference 12) entitled " Design Limits and Load Combinations for Class 1 Linear-Type Component Supports". 5-11
- d. Ta-- Tcmparaturo offects ct the highest temperature accociated with the postulated abnormal design conditions. In the case of a spent fuel.
storage pool, this would consist of a full core discharge with the heat exchanger inoperable. The Thermal-F.ydraulic Analysis concludes that the La Salle County Station - Unit 2 Pool will be heated to the boiling point in ev4 hours. Therefore, the maximum temperature of the racks will be assumed to be 240*F, the saturation temperature under 23' of water. e. E -- Operating basis earthquake (OBE), replaced by UCBV per 305.12a of Reference 1 where UCBV is the loads generated by the service level B artificial acceleration time history.
- f. E'-- Safe shutdown earthquake (SSE), replaced by FCBV per 305.12b of Reference 1 where FCBV is the loads generated by service level C artificial acceleration time history.
- g. P g-- Upward force on rack caused by postulated structural fuel assembly. Reference Page 3.8 of the Specification, " External Loadings" gives 1200 lb. uplift force due to fuel handling equipment jamming or maloperation.
- h. Fd-- L ads due to accidental drop of heaviest load from maximum possible height.
From NF-1201 (Reference 9), the racks and pedestals would be considered as linear-type supports. NF-3143 states that linear-type supports shall be analyzed based cn the maximum stress theory when elastic analysis is performed in accordance with the rules of NP-3300. This section states that when linear elastic analysis is performed, the allowable stresses are deter-mined in NF-3320 which provides stress limits for design and Level A service limits in Section NF-3322. For Level B and Level D, stress limits are to be increased by one-third over the i factors shown in Table NF-3523.2-1. Bearing type stress limits are excluded from rules for Level D service. The Level A service stress limits are summarized in Table 5.3.4.2-2 The factors that need to be applied to the Level A t limits to obtain the Level B and Level D service limits are given in Table 5.3.4.2-3. 5-12 l
TABLE 5.3.4.2-2 l LEVEL A SERVICE LIMITS PER SUBSECTION NF STRESS CONDITION ALLOWABLE REFERENCE STRESS * -NF-
- a. Tension on net section 0.6S 3322.1 (a) (1)
(except for pin Y connected members) -
- b. Tension on net section 0.45S Y
3322.1 (a) (2) of pin connected members
- c. ' Shear on gross section 0.4S 3322.1 (b) (1)
- d. Shear on effective throat 24 KSI (120*F) Table NF-3324.5 (a)-1 of fillet welds 21 KSI (240 F)
- e. Bending of solid round 0.75S 3322.1 (d) (3) and square bars'and Y solid rectangular sections bent about their weaker axis
- f. Tension on extreme fibers 0.60s 3322.1 (d) (5) (a) of hot rolled or built-up Y members in bending
- g. Compression bearing on 0.9S 3322.1 (f) (1) milled surfaces and pins Y
- h. Compression bearing on 3322.1 (f) (3) bolt projected area 1.5S" t
[ S and S, at temperature l l l 5-13 -
TABLE 5. 3. 4. 2-3 FOR CLASS 3 ELASTIC ANALYSIS STRESS CATEGOR:ES AND STRESS LIMIT FACT LINEAR TYPE SUPPORTS DESIGNED BY ANALYGIS COMPO Stress test r xtors for Lsadine) Categcries lNete d)l Service Levet service Level Stress Category A Service Level 5 Note (2)) 0 (Note Oil Petmary Stre ses lNote .5)] = 1.0 f, = f*77 K. c, =2 66 K, 1.o K, 3 /.77. Note fait (, .2 66 INote (4:1 K., - 1.o K., . i,7 7 K,, . 2 6 6 but stress s
'/, of critical but stress s bucktseg stress of critical buckhng stress Peimary Plus Nondary Stresses (Note (611 EvaluJte for is reewred this evaluation. for critical tukt.ng for all le ming categones. The equarterent cf this Sucer Peak Stresses Evaluation not tauered.
NOMENCLA TURE: K, = stress 1.m4 factor acot.c2b8e to the Cesari allowabse tem.6e and bend.nq stresses K, = stress Ivn t factor apol<atte to the Design allowable 9*Jr stresws K.,
= stress lemst factor 2cCl<2oW to the Ces qn alloweete comgressive As al and bredang i u strestes
.eg br9.t to determ re b cht NOTES:
(1) consafered Coctrcl ofsesarate'y. debrmate es not msured by these stress limit 'xt~t .WFen req .rNcec.f .. ty Ce gn 3 rJt on, <1e#ctN*Jtion Coetrol must (2) K, K, and K., = 1.0 for deson of wi:ttes O) Stress shall not esceed C 75 (4) K, shall not esceed 0.425 . ,. ($) For Semce Leve: S As primary 'tres5ss. A. 8. C, anJ 0, stresses ind.4ed ty mtraint et free eN 1 s.1 ccwt Jed wror met.ons of.- cising irati be ec ib) Thermal strnses eithen t** sccort as 1ehnas by NF 312111 streites iPall be femated to J raage 3f 2$, Jr $, at te'*g#ratar*r*ed net. te es suJted. For $*rv<e Leve's A ard 8 4Pecrever s P**L
. a cr-e Jry ci v er1-I 5 14
5.3.4.4 Material Properties . The storage racks and their supports are fabricated from stainless steel, ASTM-A240, Type 304 plate. Properties are evaluated at both T (120'F) and T (240*F). Values are taken from Tables I 2.2 and I-3.2 of ASME, Section III, Division 1, Appendix I. 120*F 240*F* Yield Stress 28.8 KSI 24.0 KSI Ultimate Stress 74.0 KSI 69.2 KSI Modulus of Elasticity 28.0 x 10 PSI 6 27.4 x 10 pg; Shear Modulus 11.0 x 10 6 PSI 10.6 x 10 PSI Density 500 PCF 499 PCF Values of 120*F and 240*F are linearly interpolated between the values given for 100*F and 200*F, and 300*F.
- Based on no boiling in FA's under 23' water.
5.3.4.5 Allowable Stresses for Service Load conditions The allowable stresses for the loading conditions frcm Table 5.3.4.2-1 are calculated using the limits specified in Table 5.3.4. 2-2 and the material properties from Section 5.3.4.4. Table 5.3.4.5-1 gives the allowable stresses for 2 categories. (1) Level A - normal, and (2) Level D - faulted. Both use the material properties at 240*F. This is slightly conserva-tive but experience has shown that these combinaticns are less severe and never control the design. Note that Level D allowable stresses are assumed to be 50% and 60% greater than the Level A. allowable shear and normal stresses, respectively.
- This is extremely conservative since Table 5.3.4.5-2 indicates that
! the ASME Code, Subsection NF (Reference 9) allowables for Level D are 167% greater than the Level A allowable stresses. Also very conservative is the fact that it was assumed that the Level B allowable stresses were equal to the Level A allowable stresses, whereas in fact the Level B allowable stresses are 77% greater than the Level A allowable stresses (see Table 5. 3. 4. 2-3).
\
5-15
Y - TABLE 5. 3. 4. 5-1 ALLOWADLE STRESSES IN KSI - 2 (1000 lb./in y 240'F TYPE OF STRESS LEVEL A LEVEL D *
- a. Tension - membrane 14.4 23.0
- b. Tension - net section at holcs 10.8 17.3
- c. Shear gross section 9.6 14.4
- d. Shear - weld base material 9.6 14.4
- e. Shear - fillet weld throat 24.0 36.0
- f. Shear - fusion weld 24.0 36.0
- g. Bending - tensicn/ compression 18.0 28.8 on solid sections
- h. Bending - tension / compression 15.0 24.0 on rolled or built up sections
- i. Compression - bearing 21.6 N/A
- j. Compression - bolt bearing N/A
- Level D allowables are taken as 50% and 60% greater than Level A allowables for shear and normal stresses, respectively.
5-16
r 5.3.4.6 Equipment Des _cription , The Spent Fuel Storage equipment consists of 20, poison wall design racks. The complete pool layout is shown in Figure 1.1. . The rack array is free standing. There is no gap between the racks as they butt against each other in all four directions. A 5.3. inch gap is present between the West wall of the racks and the West wall of the pool and a 3.0 inch gap is present between the South wall of the racks and the South . wall of the pool. All rack components and appurtenances are fabricated from 304 stainless steel. : The pool layout of fuel storage and box dimensions are all in agreement with the criticality analysis. ' Each rack is provided with five screw adjustable pedestals welded to the bottom of the rack. These pedestals transmit the rack weight and vertical seismic 1 cads to the pool floor.
.5.3.4.7 Conclusion As a result of the Structural Analysis the following conclusions result:
5.1 The spent fuel storage racks are structurally safe and will maintain a subcritical array during all credible storage conditions. 5.2 The racks are structurally safe for full pool storage of spent fuel subassemblies. 5.3 The rack fusion weld stresses are conservatively analyzed and acceptable, based on adequate calculated factors of safety, the smallest of which is 1.14. 5.4 The suberitical array is maintained for all specified external loading conditions. These include straight and inclined drop of a fuel bundle. i i 5-17
i 5.4 References 1 Specification for Fuel Storage Racks, La Salle County Station - Unit 2, Commonwealth Edison Specification No. T-3758, Dated 7-17-85.
- 2. Seismic Phase I, Analysis of the La Salle County Station -
3. Criticality Analysis for La Salle County Station - Phase I, Spent Fuel Storage Racks. (8601-00-0007 April 1986)
- 4. Spent Fuel Storage Racks Fuel Box Crush Tests MaximumReport an Density Rack Design Typical PWR Fuel.
90 7F 17. Spent Fuel Storage
- 5. Thermal and Hydraulic Analysis Report,2. (8601-00-0083 April 198' Pool, La Salle County Station - Unit Lincoln
- 6. Blodgett, " Design of Welded Structures," The James F.
Arc Welding Foundation, 1966.
- 7. Rabinowicz, E., " Friction Coef ficients of Water-Lubricated Study Stainless Steels for a Spent Fuel Rack Facility" .
performed for Boston Edison- Co. , November 1976. Roarke & Young,
" Formulas for Stress and Strain," Fifth 8.
Edition, McGraw-Hill Book Co., 1975. s 5-18 __ -- .}}