ML20080E940
| ML20080E940 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 08/31/1983 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20080E931 | List: |
| References | |
| NUDOCS 8309140148 | |
| Download: ML20080E940 (181) | |
Text
{{#Wiki_filter:e :
^
1 rq LICENSING REPORT U ON : HIGH-DENSITY SPENT FUEL RACKS , FOR OYSTER CREEK NUCLEAR GENERATjNG STATION
,NFIC DOCKET NO. 50-219 I
I i GPU NUCLEAR 100 INTERPACE PARKW'AY PARSIPP ANY, NEW JEFiSEY 07054 4 AUGUST, 1983 O l
~~
pm8aaam;"g
TABLE OF CONTENTS i
- SECTION'l_- INTRODUCTION 1.0 Introduction 1-1
~
SECTION 2 - GENERAL ARRANGEMENT-
- 2.0 General Arrangement 2-1 SECTION 3'-iRACK CONSTRUCTION 3.1 Construction 3-1 3.2 Codes, Standards.and-Practices. 3-4 ;
for-the Spent Fuel Pool Modification . SECTION 4'- CRITICALITY SAFETY EVALUATION 4.1 Introduction -
. 4-1
- 4.1.1 Neutron Multiplication Factor -
4-2
- 4.1.2 Analytical Methods 4-3 4.1.3 Calculational Bias and' Uncertainty 4-5
~ 4 .1. 4 - Trend' Analysis and Comparison with 4-5 CASMO. Calculations 4.2- Input Parameters
- 4-7 14.2.1 Fuel Assembly Design. Specifications 4-7 i
4.2.2- Reference Design Spent Fuel Storage ' 4-7 Cell 4.3' ' Criticdlity Analysis 4-10 i 4.3.1 Nominal Case 4-10
~4.3.2 ~ Boron Loading Variation 4-10 i
~
4.3.3 Boraflex Width Tolerance Variations 4 ; f i' ii
-e s a,s- --,,-,p , -- .,-n v,.w .enn,,,-, ,-,,,,n,,~,,,,e -e.,,e.n.~ , , , -e ,.,--r,,, . - ,-,, . . .._-,,n.-_.,r..a., , , - ~ , - - - ,e -, . .
i TABLE OF CONTENTS f )7 I
'N_/ (Continued) l 4.3.4 Axial Cutback in Boraflex Length 4-12 4.3.5 Storage Cell Lattice Pitch Variation 4-13
. 4.3.6 Stainless-Steel Thickness variations 4-13
'4 . 3 . 7 Fuel Enrichment and Density Variation 4-13 4.3.8 Effect of Zirconium Fuel Channel 4-14 4.4 Postulated Accidents and Abnormal Conditions 4-15 4.4.1 Temperature and Water Density Effects 4-15 4.4.2 Abnormal Positioning of Fuel Assembly 4-16 Outside Storage Rack 4.4.3 Fuel Assembly Positioning in Storage 4-16 Rack 7__s
.i \
\-) 4.4.4 Effect of Zirconium Fuel Channel 4-17 Distortion 4.4.5 Dropped Fuel Assembly Accident 4-17 4.4.6 Fuel Rack Lateral Movement 4-17 4.5 Acceptance Criteria for Criticality 4-19 REFERENCES TO SECTION 4 4-20 SECTION 5_- THERMAL-HYDRAULIC CONSIDERATIONS 5-1 5.1 Basis 5-1 5.2 Model Description 5-2 5.3 Results and Discussion 5-4 REFERENCES TO SECTION 5 5-6 iii
h TABLE OF' CONTENTS (Continued)
/O,
\s / SECTION 6 - STRUCTURAL ANALYSIS 6-1 6.1 Analysis Outline 6-1 6.2 Fuel Rack - Fuel Assembly Model 3 6.2.1 Assumptions 6-3 6.2.2 Model Description 6-5
.6.2.3- Fluid Coupling. 6-6 6.2.4 Damping 6-7
- 6.2.5 Impact 6-8 6.2.6 Assembly of the Dynamic Model 6-8 6.3 Stress Analysis 6-12 6.3.1 ' Stiffness-Characteristics 6-12 6.3.2 Combined Stresses and Corner p_, Displacemer.ts 6-13
- 6.4 Time Integration of the Equations of Motion 6-14 ,
6.5 , Structural Acceptance Criteria 6-17
. 6.6 Results 6-23 REFERENCES TO SECTION 6 6-27 l
SECTION 7 - OTHER MECHANICAL LOADS 7-1 l 7.1 . Mechanical Loadings- 7-1 7.1.1 Fuel Handling 1 7.1.2 Dropped. Fuel Accident I 7-1 7.1.3 Dropped Fuel Accident II 7-1 7.2 - Local Buckling of Fuel Cell' Walls 7-2 7.3 Analysis of Welded' Joints in Rack 4 H
~ References to Section 7 7-7 SECTION 8 - SPENT FUEL POOL FLOOR STRUCTURAL ANALYSIS 8-1 f )
8-1
- 8. l' Introduction -
iv
, e ,~. y -----r - , - , . , -,-.,-,,-,w -- t . . . - ~ -,,-,-ew-w
TABLE-OF CONTENTS (Continued) -I [ ~
; 8.2 Assumptions- 8-1
- ;8.3 ' Dynamic Analysis of--Pool Floor Slab 8-2 i 8.4 Results-and Discussions 8-5 8.5- Conclusion 8-13 ;
References to Section.8 8-24
- SECTION 9 --ENVIRONMENTAL-EVALUATION 9.1 Summary 9-1 9.2 Characteristics of Stored Fuel 9-2 9.3 Related Industry Experience 9-3 9.4 . Oyster CreekLOperating Experience 9-5 9.5 . Spent Fuel. Pool Cooling and Clean-Up System (FPCC) 9-6 I/~S 9.6 Fuel Pool Radiation Levels- 9-7 b
9.7- ' Radiation Protection Plan 9-10 9.8 Re-Racking Operation 9-11 9;9- Conclusions 9-15
-REFERENCES TO SECTION 9 9-18
-SECTION 10 ' INSERVICE SURVEILLANCE PROGRAM FOR 10-1 BORAFLEX NEUTRON ABSORBING MATERIAL 10.l' Program-Intent 10-1
. 10.2 Description of Specimens 10-1 10.3 Test 10-2 10.4 Specimen Evaluation 10-2 SECTION 11 ' COST / BENEFIT ASSESSMENT 11-1 h -11.1 Specific. Needs for Spent Fuel Storage 11-1 d V. l
TABLE OF CONTENTS , (Continued) 11.2 Cost of Spent Fuel Storage. 11-2 11.3 Alternatives to Spent Fuel Storage 2 11.4- Resource Commitments 11-4 REFERENCES-TO SECTION ll. 11-5
- jiECTION 12 DESIGN CONTROL AND FABRICATION INTERFACE 12-1 12.l Introduction - 12-1 12.2- Personnel 12-1
~ SECTION~l3' OUALITY' ASSURANCE PROGRAM 13-1 13.1 Introduction - 13-1 l
13.2 General 13-1 13.3 System Highlights 13 l () 13.4 SECTION 14
- Summary PRODUCTION CONTROL 13-3 14-1
-14.1 Introduction 14-1 !
14.2 Procurement 14-1 14.3 Shop; Floor Planning 14-2 ! 14^4 . Operations Control and Coordination- 14-2 14.5 Reporting 14-2 ; SECTION 15 CONCLUSION 15-1 I l-i~ l l O . vi l l o
, . - . . . . , _ . . - - - . . . . . - . . _ - - = . . . . . . . . - _ - . . - , - .--. __.
LIST OF FIGURES
- /-~'g Page ,
\) SECTION 2 l
Fig. 2-1 Mcdule Layout 2-4
-. SECTION 3 Fig. 3-1 Array of Cells (4x4) 3-7 3-2 Elements Cross Section , 3-8 3-3 ' Angular Subelement A 3-9 l
3-4 Cruciform Element (Isometric View) 3-10 3-5 Sub-elements 3-11 3-6 Typical Cell Elevation 3-12 l 3-7(A) PlanCrossSectionofTypica5SupportLeg 3-13 i 3-7(B) Vertical Cross Section of a Support Leg 3-14 3-8 Typical Module 3-15 , r~' l ((_f SECTION 4 , 1 Fig.' 4-1 Geometric Model of Oyster Creek Spent Fuel Storage Rack Cell 4-9 2 Log-log plot' of Calculated k= values versus B-10 loading 4-11 4-3 k= of unpoisoned fuel assemblies as a function of assembly spacing and water gap 4-18 SECTION 5 Fig. 5-1 Idealization of Rack Assembly 5-7 5-2 Thermal Chimney Flow Model 5-8 SECTION 6 Fig. 6-1 Dynamic Model 6-40 6-2 Impact Springs and Fluid Dampers 6-41
/~'\ 6-3 Spring Mass Simulation for Two \m,/ Dimensional Motion 6-42 vii
LIST OF FIGURES-
, .(Continued)
.. 6-4(a) Horizontal Cross Section of-Rack- 6-43 p
6-4(b) .VerticalfCross Section of Rack 6-43
!6-5 ~ Dynamic Model 6-44 6-6 ' Stress Resultants Orientation 6-44 6 ,7 Subdivision of a Typical Rack 6-45 8 Finite Elements Model Cross Section 6-46 9 Horizontal Seismic, SSE, 4 percent-Damping 6-47 6-10 Vertical Seismic, SSE, 4 percent -
j- , -Damping 6-48 i SECTION . Fig. 7-1 Loading on Rack Wall 7-8 7-2 Welded Joint -in Rack 7-8 , SECTION 8 Fig . 8-1( A) Plan View - Reactor Pool Area 8-14
~
'8-1(B) Section through-Central N-S Girder 8-15 i
~8-1(c) Section-through E-W Girder 8 l-8-2(a) slement Locations L 8-17 L-8-2(b') Node Locations 8-18 8-3 Floor Load Sum of' Supports 8-19 ,
8-4 Floor, Load Support 1 8-20 8-5 Floor Load Support _2 8-21 p 8-6 Floor Load Support 3 8-22 p L 8 Floor Load Support 4 8-23 SECTION'9 Fig.19-l_ Existing Rack Locations , 9-14 l j -- t SECTIONL10 , ( '
- F ig . 10-1 Test' Coupon- 10-6 viii i- ' . _ _ . _ _ __..._._._._,-~..._u._.. . _ _ , . . _ . _ _ _ . . _ . _ . _ . . _ . . _ _ . . . _ , . , _ _ _ , . _ _ _ . . . . . _ _ . _ _ _ _ _ . _ _ _ _ . .
LIST OF TABLES Page b d SECTION 1
~
n Table l.l~ Oyster Creek-Fuel Discharges 1-3 SECTION 2 Table 2.1- Module Data- 2-2 . SECTION 3
. Table 3-1 Boraflex Experience List. 3-2 l
'SECTION 4 ,
j , . Table 4-1 Summary of Criticality Analysis 4-4 4-2 Fuel Assembly l Design Specifications 4-8 4-3 Effect of Temperature and Void on Reactivity of Storage Rack- 4-15 SECTION 5 f I Table.5-1 Maximum Local Pool Water Temperature & Local-Cladding Temperature 5-5 ( SECTION 6 l Table.6 Degrees of Freedom 6-5 6-2-- Numbering System for Springs, Gap-Elements, 6-10 Friction Elements . 6-3 Physical Property Data- 6-19 l. l 6-4 Cases Considered (All SSE Events) 6-26 i 6-Sa File DGPU10, - Module E (- Coef=.8, Full Rack 6-28 6-5b File'DGPU51, Module F Coef = .8, Full Rack 6-31 6-5c . File ~DGPU41, Module K Coefl= .8, Full Rack 6-34 i 6-5d File DGPU61, Module H Coef = .8, Full Rack 6-37 ix l
.--- . . . . . = . . - . . . . . - . . . . - _ . - _ _ _ - , . _ . , , . _. ,_._ _,.. .,_ ,_ -,-,--.._,... -. - .. _, -
4 LIST-OF TABLES (continued) 9 1SECTION 7 ^ Table 7-1 Compressive Stress in Rack Wall and Comparison with Local Buckling Stress for Typical Cases 7-3 SECTION 8 . Table'8-1. Floor-Slab Frequencies 8-7 8 Ultimate Moments of. Slabs and Beams 8-8 3 Shear Capacities Ve/b or Vc for plate or beam sections. 8-9' 8-4 Critical Shears and Moments for Total
- Dead -Load Analysis ~(Case 1 plus Case 2) 8-10
, 8-5 Critical-Shears and Moments for Dead Load Analysis case 2 8-11 8-6 Incremental Additive' Shear Force 8-12 8-7 Critical Pool' Floor Structural IntegrityL
() SECTION 9 Checks 8-12
. Table 9-l' Observed Radionuclide Concentrations in Spent Fuel Storage Pool Water , 9-5 9-2 Routine System Analysis- 9-8 SECTION 10 Table 10.1 Time Schedule for Removing Coupons 10-5 u
I l 9 A y x
.. .- . . ~ . . - - - _ - - _ _ _ - _ _ -
l . 0 .- Introduction Oyster Creek Nuclear Generating Station-(OCNS) is a 640MWe (net) generating unit with a' General Electric BWR-2 reactor.- The station. is owned by. Jersey-Central Power & Light Company and is operated by?GPU Nuclear . (GPUN), .a subsidiary of General Public Utilities of Parsippany, New Jersey. The plant. is located 9 mil'es s,outh of Toms River, N'ew Jersey. It has been in operation since December, 1969. The original plant constructors were Burns and Rce, Inc. and J.A. Jones' Company.- OCNS spent. fuel pool was originally. equipped with 840 storage locations. The storage require.m~ents were predicted to outstrip the original available capacity in February of 1976.' The owner sought and received NRC permission to expand storage to a maximum.of 1800 locations'using high density unpoisoned racks in
.q- the year 1977. Of six high . density racks module- (rows) licensed b in1this application, 'five rows are currently in the pool. One module containing- 220 -cells is presently stored outside the
~
pool.- A moratorium on fuel reprocessing announced by the Carter Administration, followed by repeated U.S . - government delays in establishing 'a national fuel handling program, had kept GPU's spent fuel management program in a state of uncertainty. The recent decision by Nuclear Fuel Services to return all of 224 OCNS-spent fuel assemblies held at the West Valley facility has heightened the sense of urgency for further densification of fuel storage in the spent fuel pool. As shown in Table 1.1, the full core discharge capability 'at OCNS pool will be lost in 1985 even if 'the sixth row is installed in the pool.
+f i Ev - ,
1-1 9 y e -w -
.y - ....+w--
w_ .w- e ---.,e ,,w-s acwe.-eww.--v--y - -
,- +e--^
. . . . _ . . - . . - . . +
~
Tnis. application. is for expanding the installed ' storage . capacity in the OCNS . pool to a maximum of 2645 locations, of whichi only ~ 2600 will; be utilized' for spent fuel storage. The remaining' locations will be used' for . storing miscellaneous equipment,.(i.e./ fuel channels).
- Joseph Oat. Corporation ~, a Camden, New Jersey based. supplier
~
ofLnuclear-plant' equipment, was selected.to design and fabricate
;the1 poisoned high~ density racks.
This document was prcpared ~ by Oat- in collaboration 'with .their nuclear physics consultant, Southern Science ( a division of Black and Veatch), to give an ,
~ abstract of all aspects of rack design, analysis and fabrication. Each item describe'd here has been gleaned out of detailed reports and drawings so as to present the physical facts wit,hout ' .the encumbrance of mathematical details or thousands of pages lof ~ ~ unabridged ' results. The focus .is on explaining: the-
. essentiial aspects. of ' the. design relevant to the equipment safety I
~ and, : reliability. The important features of ' the mathematical !
models to . predict the ' seismic response, h'ydrothermal behaviour and- the, effective neutron ' multiplication factor are clearly ) stated: to - show 'the built-in - conservatism and simplifications introduced into the analyses. l
-l 1Abrid'ged results' -are presented. which show that in all I analyses zacceptable margins of safety are available, and in some l cases, wide -margins. of safety. exist. In particular, the integrity of the spent fuel pool slab under the worst condition of- dead- load, temperature gradient and seismic loads is demonstrated.
l
- i. *.
1 O S
~1-2 .
Table 1.1 Oyster Creek Fue'l Discharges O O Cycle Cycle. Batch No. of Stored 'Avsilable Available Year No.- Length Size Fuel Assemblies Locations Locations ( after refueling Without After I (includes Proposed Proposed return of 224 Expansion Expansion 1 !- fuel assemblies (Capacity (Capacity stored at West -= 1800) = 2600)
. Valley, N.Y.)
l l' '10 . 14.5 200 1204 596 1396 2/83 i (Refuel- mths i
- ing !
Feb.'83 . !
-Jan.'84) i 1 l l 11 16 188 1392 408i 1208 4/85 12 18 192 1584 216 1016 1/87 13 16 180 1764 36ft 836 8/88 l 14 17 200 1964 -
636 4/90 l 15 18 212 2176 - 424T 1/92 16 18 220 2396 - 204 10/93 [ 17 . 17 , 200 2596 - 04Ti 6/95 i Full core discharge capability lost (560 assemblies) Ti Normal batch discharge capability is lost (= 200 assemblies) l 1 { 1-3 l l y _y.. -
2.0 GENERAL ARRANGEMENT The ~ high density --spent fuel racks consist of individual )' cells with a 6-inch-nominal-square cross-section, each of which-accommodates a single BWR fuel assembly. The cell walls consist of a , reutron absorber sandwiched between sheets o'f stainless steel. The cells ~are arranged in modules of varying numbers of
- cells with a 6.198" ~ (nominal) center-to-center spacing.
The high-density racks.are engineered to achieve :the dual objectives _ of maximum ' protection' against structural loadings (such as ground motion) and the maximum utilization of available storage volume. In general, a greater . width-to-height asp'ect
-ratio provides ~ greater- margin against rigid body tipping.
LHence,~ the modules are made as wide as possible within. the constraints of: transportation and- site-handling capabilities.
~
.The high-densJ.ty: spent fuel racks will be installed in the OCNS spent fuel pool.
The Oyster Creek spent fuel pool will contain ten D high-density ~ fuel racks in'different module sizes. The module ' O types are labelled A,B,C,D,E,F,G,H,J and K in Figure 2.1, which also. shows their relative placement. _There will be a total of l2645 storage locations in the spent fuel pool. i- . Table 2.1 gives the detailed module data (e.g., weight, array size, and number of storage locations). I The - spent fuel rack modules _ are free -standing , i.e. they are not anchored ' 'to the pool floor or connected to the pool walls through snubbers or lateral restraints. Baseplates of-modules E and K are raised 11.5" from the liner to' avoid physical interference-with the horizontal sparger
-lines. Other modules have.their baseplates at 6" elevation from- .
the-pool liner. 4 a 2-1
Table 2.1 Module ~ Data- ' Approximate Height of rack Weight baseplate from Tyge cells / Module Array Size 1bs/ module pool liner A. 320 20x16 3 8 ,'4 0 0 6" B 248 14x16t -29,160 6" plus 8x3 C- 300 20x15 36,000 6" l D 294 21x12t 33,960 6"
.plus 14x3 l '
E 312tt 20x16 37,920 11.5" F 315 l21x15 37,800 6" () G 192 16x12 21,120 6"
.H 176 16x11- 18,000 6" J 240 16x15 28,800 6" K 248tt 16x16 30,240 11.5" l.
l [ l l t These mo'dules are non-rectangular'in shape. it ; 8 . cells' are -lost in each of modulds E and K due to i~ interference from the vertical sparger pipe line. ( 2-2 l l L__.._.. _ . _ , _ _ . _ . _ . _ _ . . , _ _ . . _ . , _ , . . . . . . . . _ _ _ . _ _ . _ . _ . . . _ . . _ _ . _ - . . . . _ . _ . . . _ _ _ _ _ _ _ _ . . . _ _ _ _ . . . _ . .
+e-The nominal gap ~between any two modules is 1 1/2 inches along the top edge for modules in the spent fuel pool. The p/
\_- minimum acceptable gap is 1.125". The minimum gap between the spent fuel pool will and the modules 'is 1 1/2 inch. Adequate clearance from other pool resident hardware is also provided. In this manner, the possibility of inter-rack impact, or rack
. collision with other pool hardware during the postulated ground motion events is precluded. Details on rack kinematics under seismic conditions may be found in section 6.6.
l G /3 U. 2-3
-- - ,, ,.,y..,-,y . . . , , . . . _ _ . . , . . . _ . . . ,_g.,,,.,*, , , _ . . . . , , , , . . . , , , , . , _ , _ , , _ , , . _ , , , _ , , . , , _ , , , , _ , , , ,
.O O- O '
3 9 '- o " -
/3 '- 8 " _ 4 '- + " 11 '- o " _
E
'$l, ; 'lllll,/lllll,/,
/ a
,. $ $ N ^
*} \
.iw *A~ *B" =
=[ s zox/6
/f x /6 \*Ag 00 w
gx3 g% si
= B786 a
k
~
~'
N
% y
= n' N
D .g=. A w
-id N
.C.
I."
-f S d A 3.
p zouis *o" *g % # c, " *x~ d g 4 2/x/2 M $ N
/6x/2 /6 x //
,I
=\
g
'/x++ 3 4s n , ,
u-o 1r u
, s *E" f n
+p" sgar , ,, NN
$4 'l" 20 x /4- 8 ',NT 2/ x /5 /gxff
- i A i" g /6 x /6- 8
- 4 /g - - -
.- z" -
.-- z "
I % . u A u 9 11 s " u
-s,r, -
- i2+ %~ -
isa '/e~ t - 99 %~ - rr s~ _
- 19,.
! -~ .*-- z V2 " i 4 4/sLa"-. -- FIG. 2.1 MODULE LAYOUT ,
3.0 RACK CONSTRUCTION 3.1 Construction { The racks are constructed from ASTM-A240, Type 304L, austenitic steel sheet material, ASTM-A240, Type 304L'austenitic steel plate material, and ASTM-A182, Type F 304L austenitic steel forgings material. The weld filler material utilized in body welds is ASME SFA 5.9 Type 308L. The support spindle (Fig. 3.7.b) is made from SA564-Alloy'630. Boraflex, a patented brand name product of Bisco
- is the neutron absorber material. The detailed radiological properties'of Boraflex may be found in Section 4. The experience list of Boraflex is given in Table 3.1.
A typical module contains storage cells which have 6" nominal internal cross-sectional openings. These cells are straight to within 10.125" . These dimensions ensure that fuel assemblies with maximum permissible out-of-straightness can be inserted into the storage cells without interference. Figure 3.1 shows a horizontal cross-section of an array of 4x4 cells. As stated in the preceding section, the modules vary in array size from the smallest 16xil array to the largest array size of 20x16. The cells provide a smooth and continuous surface for i
. lateral contact with the fuel assembly. The construction of the rack modules may be best described by exposing the basic building blocks of this design, namely, the " cruciform", " ell" and " tee"~ elements, shown in Figure 3.2. The cruciform element is made ' of 4 angular sub-elements, "A" (Figure 3.3) with the neutron absorber material tight 1'y sandwiched between the
-
- Bisco, a Division of Brand, Inc., 1420 Renaissance Drive, Park O.V
. Ridge, Illinois
- 3-1
Table 3.1 BORAFLEX EXPERIENCE LIST f
/\ Wisconsin Electric NRC License issued (m / Point Beach l'(Docket 50-266) PWR r
Wisconsin Electric NRC License issued ' Point-Beach 2 (Docket 50-301) PWR Niagara Mohawk NRC-License ~ issued , Nine, Mile: Point.1 (Docket 50-220) BWR .l
-Niagara Mohawk' NRC License to be applied for Nine Mile Point 2 (Docket 50-410) BWR Consumer Power Company NRC License pending
. Midland-Units 1 & 2-(Dockets 50-329, 50-330) PWR TVA. . . .
NRC License pending Watts.Bar Units l'& 2 (Dockets 50-390, 50-391) PWR r Louisiana Power & Light . NRC License pending Waterfords Unit 3 (Docket 50-382) PWR O Duke Power
. Oconee Units 1 & 2 (Dockets 50-269,50-270) PWR License issued Northern States Power
[~'\ Prairie Island Units 1 & 2 (Dockets 50-282, 50-306) PWR License issued. b Detroit Edison Fermi-2 (Docket 50-341) BWR License' issued. Daltimore Gas & Electric-Calvert' Cliffs II (Docket 50-318) ~ PWR. License issued.
- Commonwealth Edison Company. -
Quad' Cities Units 1 & 2 (Dockets 50-254, 50-265) BWR License issued. Carolina Power & Light Company
'H.B. Robinson 2 (Docket 50-261) PWR License pending.
' Carolina Power & Light. Company Shearon' Harris' Unit.l (Docket 50-400) PWR, License to be applied for. '
Northeast Nuclear Energy Company
! Millstone Unit-3 (Docket 50-423) PWR, License to be applied for.
' Gulf States Utilities Company River: Bend Unit ~1.(Docket =50-458) BWR License pending.
=
4 Sacramento Municipal Utility District ; Rancho Seco Unit:1 (Docket 50-312) PWR License pending. O* BWR l& PWR- Fuel' Storage _ Capabilities . 3-2 _. .--.-_2._ ,2 . -. .. a. _ .. _ _ _ __ __ _ ___
'i
, stainless sheets. The . long edges of the cruciform are welded l using a 3/8" . thick stainless steel backing strip as shown in Figure 3 . 4 '. ' The bottomi of the cruciform assembly has 8 1/8" high s'tainless~ strips, which ensure against slippage of the I
-" poison" material downwards due to gravitational loads or operating conditions. The fabrication procedure leads to one i hundred percent surface contact (in a macroscopic sense) between
- the . poison and - the stainless sheets. The top of the cruciform is also end welded using a spa <:er strip as shown in Figure 3.4.
~ Continuous welding of the straight segments of the top . edges ]'
j . produces a smooth lead-in ' surf ace. Ample venting is available through the roof openings of cell corners. The- " ell" 'and " tee" elements are constructed similarly using angular sub-element "B", and flat sub-elements "C" (Figure ( 3.5). Having fabricated the required quantities of the "cruciforms",." tees", and " ells", the assembly is performed in a specially designed fixture which serves the vital function. of l maintaining _ dimensional accuracy while welding all the contiguous spokes of all elements using fillet welds. Figure l 3.1'shows the fillet welds. The cells are bonded to each other along their long edges,.thus, in offect forming an " egg-crate". The - bottom ends of the cell walls are welded to the base
. plate. Machined sleeve elements are positioned concentric with l the cell center lines above the holes drilled in the base plate,
! and attached to the base plate through circular fillet welds [(Fig. 3.7.(B)]. The conical machined surface on the sleeve provides a contoured seating surface for the " nose" of the fuel
, assembly. Thus, the contact stresses at the fuel assembly nose
- j. bearing surface are minimized.
l: .The central ~ hole in the sleeve provides the coolant flow
. path.-for ' heat transport from the fuel assembly cladding.
j -. Lateral holes in . the cells walls [(Fig. 3.T.(B)] provide the l-LO . 3-3.
~
additional flew path in the unlikely event that the main coolant flow path-is clogged. Each module is supported on four " plate i type" " adjustable . supports; a sketch of a typical support -is shown in Figs.-3.7(a) and 3.7(B). The construction of -the cell' assembly .is 'an integral structure which possesses extremely high flexural rigidity. Figure 3.8 shows an elevation view for a typical module. , .3.2 CODES, STANDARDS, AND PRACTICES FOR THE SPENT FUEL PCOL , MODIFICATION ! The followin,g are the public domain codes, standards, and practices 'to which the fuel -storage racks are designed, .i ' constructed and assembled, and/or pool structure analyzed. Additional- problem-specific references related to detailed analyses are given at the end of-each-section. O I. Design Codes l (a) AISC Manual of Steel Construction, 8th edition (1980) (b) ANSI N210-1976 Design Objectives for Light Water Reactor Spent - ruel Storage Facilities at L ' Nuclear Power Stations. (c) American Society of Mechanical Engineers (ASME), Boiler & Pressure Vessel Code, Section j III,. 1980 Edition up to and including Winter !' 1982 addenda. (subsection NF) I' r O 3-4 L
.' -(d) LASNT-TC-1A' June, 1980, American Society for 3 gs Nondestructive Testing- (Recommended Practice
~ ' '
9 )L for Personnel Qualifications) (e) ANSI N412 The Determination of neutron
' reaction -rate distributions and reactivity of nuclear reactors. s
. II. Material Codes
( a). American Society for _ Testing and Materials (ASTM) Standards - A240. l (b) American Society of Mechanical: Engineers (ASME), Boiler & Pressure Vessel Code, Section II, Parts A and C, 1980 Edition up to and
- including Winter 1982-addenda.
III. Welding Codes I - J (a) ASME '- Boiler and Pressure Vessel Code, Section IX- - Welding and Brazing Qualifications, 1980 Edition up to and including Winter 1982
. 'ddenda.
a IV puality Assurance, Cleanliness,EPackaging, Shipping, Receiving, Storage, and Handling Requirements
"(a). ANSI 45.2.2, Packaging,- Shipping, Receiving, Storage-and Handling of_ Items for Nuclear Power Plants.
(b) ANSI' 45.2.1, Cleaning. of Fluid Systems- and Associated Components.During Construction Phase of Nuclear Power Plants.
,-(c) ASME Boiler and Pressure Vessel, Section V, Non-destructive Examination, 1980 Edition, I) including' Winter-1982.
.3-5 .
3 i V. Other References
- (a) . NRC Regulatory Guides, division 1, regulatory guides 1.13, 1.29, 1.61, 1.71, 1.85, 1.92, and
~ 1.124 (revisions as-applicable). (b) General Design- Criteria for Nuclear Power Plants, Code of Federal Regulations, Title 10, Part 50, Appendix A (GDC Nos. 1, 2, 61, 62, and
. . 63).
(c)' NUREG-0800,' Standard Review Plan (1981). (d) "NRC . Position for Review and Acceptance of
~
Spent Fuel Storage and Handling Applications," dated April 14, 1978, and the modifications to this document of January 18, 1979.
~.
f
'l
. l 6
O 3-6 y ,
, ,-y-- 3 ,- p- ,-y- ,,.,----,--y,,,,,-,,w~,--,.,r. --- -: , . - , ._--*e-_
M d . 1 JS./ 98 " _ - C/C TYP, l L r r' .. _ _ :._. . _ _ .
. 4' ._ __.? '
l , l r
. A T YPIC A L
" STORAGE i -
- 6 S O. -
S - LOC ATION i
' i -
! T Y P, . i , l' i J .
. .I ,
s
.m. . .
_ .. _ _ m xi ,
.; ,d , , ,
=
i
. i 8 0 R AFLEX.
SHEET
, i 9
a ' h i k
- _ ... . ,- g . _ _ _
x r
+
i
- i. .
j <
- --- -~
1 ' T-i i . , I i .! e p : i
- 7. _
.fy, _ _lf STA I N L E S S M ATERI AL
/
l i i i t' . r r- _ - . - - - - - - - i ^ ,u __ , , , , -- TYPIC A L FILLET WELD l t l FI G. 3.1 AR RAY O r CE_ _S (2 XZ-) Ln v . l I l- 3-7 L , . , , , , . . _ . . . . . - - _ _ _ , . - . ~ . . _ . . . ~ . - _ . _ _ _ _ . - _ _ . - _ _ . - ~ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ . _ _ _ _ _ - - - - - - - - - - - -
O~
. O .
'. O T .n
;END STRIPS '
6.095" ~ 6.095 " '
,l~ ,,
6.031 ( T Y P, ) -'.032 0 08 * ~ , o (6.095 ! h n 0.375TYP. h9 A x y
-1,: ,
e h o#% b.095b O.0 65T H K. S / S ' S H E E T-6.0 3 l" *p .o 6f'THK.S/S SHEET (c) T E E { l- , i . - - . n si . ~; -
,040 THK.
6,03I I + ! BOR AFLEX SHEET '
- 6.095 * ~
ALL ELEMEWTS
-~
l , , p n d-or &, s.asz~ t 0,0 63T H K S / S SH E ET A-l . OUTER ELEMENT p . 1 l (a) CRUCIFORM L --
~~i e
(b) E L L FIGURE 3.2 ELEMENTS CROSS SECTION e
I
.Q 0
0
- ,o e 3
- i t
= 6.o 7s =
/
I 1 , l, s ** 32 l -O :
*o .
S-
+p.
- /. . .* . . . .v . , ,
7 1 L 5 l FI G. 3.3 A N G U L AR SUB E L E M E N T A' O 3 9
O A p \ S. S. A N G U L A R , , 9
'SUB EL E M E N T A +
~
_ TOP END ( F I G . '3. 3' ') , _ S.S. S T R I P l
- 1
/
O ^, eoiSoN
' SHEET .
l 1 S I D E E N D.-
- S. S. ST RI P '
-~( T Y.P,)
J 9 0 f FI G. 3. 4 CR U'CI FO RM EL EMENT ( ISOMETRIC VIEW )
. -- 2, 3-10
,_-._._.._.-.__..2.._._ . . _ _ _ . _ . , _ _ . _ - . - _ . _ _ _ _ . _ . _ _ _ . _ _
i i O .,
- a 1
.063 M -
/
D. o
- . s.i 3 s
- *s , -
EDs -s,,sssss, v i 6.198" = ( a ) A N G .U L A R SUBELEMENT B' i O . m /2.3Sdi, "
/2.130 " -
- 0 6 3" vss isis ssiss/s i/////ssisiss,,
\ l l
l
, s -
( b) FLAT SU B EL E M ENT- C. l i 5 l 3 - ' (V F l G. 3. 5 SU B ELEMENTS :
~
3.- 1 1 ,
4 h,o R E F. . v 7 r-c --.- r / TOP-END M
-h / SPACER
- n. n f PI
; STRIP
. 4[ w u' I o
~ a:
CELL n F- P Of SON e ; , i I J i I '- F ! J,l a o e w : - J J o Li
,/
J 2 o y
,j /
-[ s W u) -
' . - FUEL BUNDLE
,u ---
O p g
, ' /
* ' 4 ,
L A T E R AL H O L E- .
=
A" 4 , j [ (NOMINAL . DIAM ETER 3/4"
'S N .. . e .
N js - BOTTOM EN D-SPACER STRIP
, [. .
g y , I .l - , i, BASE PLATE o ( )
\,,/
FIG.3.6 TYPICA L CELL ELEVATION i O 3-12
~
l O '
- 7e Y Floor Plate s
h d h\\\\\N i\ N \ \ \\ ' I \ x 3gP" 8
=
g s _- - -
\ g .# dg ,_
_ m v __ _ ay m
- _s 5 O
/
/ '
s [N's D / j F iA ' '
\
, y lY///// e
/////AN l l 1 U V l
s'/e" . 1 a 34" = 1 1 6 - tzVz "crxe) - i
/+ V. 4 " -
O FIG. 3.7(A) PLAN CROSS SECTION OF
. TYPICAL SIJPPORT LEG 3 13 .
-a-se l
I Lateral _ l . Flow Hole Fuel Assembly Flow-Hole Nose Seat p-
- g. f! N[
'! Cell Wall h'
'l ,s
- , I,x - i li [ Base Plate i I ' I f a y p
/ ' -> / *-- 3 /4 in c h h
T L i Q V I - i f" e
/ \ \
I V
/
- 2. /4 s / l
./
l l I A 1r J A 2 I & 2 i 8 I '2 1r 1" NY ! !- N 'A! Y b /.. . .\ ,
.. ..\ ,
I-i 'r i . ! ii I pool Liner . Adjustable Spindle 1 FIG. 3.7(B) VERTICAL CROSS SECTION OF A SUPPORT LEG O + 3-14
, ree - , - . ' , - . ,--v.--1 ,,,.,-..,---,,,,_,..,-,....,,+_.,,e'r,,,,,.__-._-...,,m,_. . - -84& , .. -,.,i-.,,.-c..,,,,..,,,..,.,m. ......,,-,e,. , _ -~ c
- +- 6./ 9g C/C !
i T Y P. _ l Il l
- j
~
v 6" S Q. CEL' T Y.P. 13 X 13 = l69
~
i CELLS ; LI I I l i l O" " ! l
)
^
5 N . a 169 CELL LGTH v M
.u . 5/8" T H K.
' A M' 'A d4 ,BA S E P L ATE O
(4) SUPPORT FIG. 3.8 TYPICAL MODULE LEGS
- i. 3-15.
~
'4.. ICRITICALITY SAFETY EVALUATION s 1-
!* 4.l' ' Introduction The- spent - fuel storage ~ racks are. designed to assure that a
- kegg equal ~ to o r. less thani 0.95 .is maintained with the racks
~ fully loaded -with fuel of the highest anticipated reactivity cand flooded with unborated water ' at a ^ temperature corresponding ~to the highest reactivity. The maximum calculated reactivity - in-cludes.a margin for uncertaihty in reactivity calculations and in
- mechanicall tolerances , statistically- combined , such ~ that ' the ' true kegg will'be equal to or less than 0.95.with a-95% probability at .
a'95%-confidence level. Applicable codes, standards. and regulations (or pertinent
. sections'thereof) include the following.-
e ' General Design - Criterion - 62 ' Prevention- of Cr i ticality
-b k/ in Fuel' Storage and Handling.
- e. NRC letter of. ' April
~
14, 1978, to all- Power Reactor Licensees - OT Position .for Review and Acceptance of
' Spent ' Fuel Storage ' and Handling Applications, including modification-letter dated January 18, 1979.
e USNRC Standard Review Plan, NUREG-0800, Section ' 9.1.2, Spent Fuel Storage. e ' Regulatory Guide 1.13., Spent Fuel Storage Facility Design Basis ~(Draft, Revision'2), December 1981. e . Regulatory Guide. 3.41, Validation of Calculational Method'
'for . Nuclear Criticality- : Safety (and related ANSI N16.9-1975).
e ~ ANSI N210-1976, Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants. e - ANSI N18.2-1973, Nuclear Safety Criteria for the Design Jof: Stationary Pressurized Water Reactor Plants. k ,, 4-1
The design basis fuel assembly is a standard 8 x 8 array of ft
' ~ '
fuel rods (BWR-type) containing UO2 clad in Zircaloy. Fixed neutron absorbing material is used in the spent fuel storage racks to maintain the required margin in subcriticality. The spent fuel racks are designed to safely accommodate fuel assem-blies with a maximum infinite multiplication factor (k,) of 1.331 i 0.002 ( lo , AMPX-KENO) as calculated for the standard core con-figuration (6-inch assembly pitch) in the cold , unrodded condi-tion, corresponding to a uniform fuel assembly enrichment of 3.01 wtt U-235. To assure the true reactivity will always be less than the calculated reactivity, the following conservative assumptions were made.
~
e- Moderator is pure, unborated water at a temperature corresponding to the highest reactivity. o Lattice of storage racks is infinite in all directions;
-l, ,) i '. e . , no credit is taken for axial or radial neutron k/ leakage, except in the evaluation of axial cutback and for certain abnormal conditions where neutron leakage is inherent.
e No credit is taken for the presence of gadolinium burnable poison or for the reduction in reactivity that accompanies fuel burnup. e Neutron absorption in minor structural members is neglected; i.e., spacers and Inconel springs are replaced by water. l e Pure zirconium is used for cladding and flow channel; l 1.e., higher neutron absorption of alloying materials in Zircaloy is neglected. 4.1.1 Neutron Multiplication Factor The nominal design case assumes . fuel of uniformly distri-buted 3.01 wt% U-235 enrichment, corresponding to 14.51 grams a U-235 per axial centimeter of fuel assembly. Fixed neutron ab-E sorber material 2 (Boraflex)' of 0.01114 g/cm boron-10 areal 4-2
. ~ . - . .
Ldensity ' .is positioned between fuel assemblies in an egg-crate structure that provides a nominal center-to-center lattice spac-d ing'of 6.198 inches for the . storage cell locations. The maximum
' infinite multiplication : factor (k,,). calculated- for the nominal design . case. :is 0.947, including all uncertainties (95% proba-bility atla 95%' confidence level) for fuel of 3.01% uniform en-r i c h m e n t ( k ,, of fuel assembly in standard. ' core geometry
- of 1.3308 i 0.0024 as calculated by AMPX-KENO).
The calculations and uncertainties supporting the critical-ity safety of the spent fuel storage racks for the Oyster Creek
. plant are summarized in Table 4-1 and described 'in Section 4.3,
.Criti'cality Analysis.
~4.1.2 Analytical' Methods The reference method.for nuclear criticality analyses'of the l .high density spent fuel storage rack is the AMPX -KENO 1 2 computer package, using the 123-group GAM-THERMOS cross-section set and the NITAWL subroutine for U-238 resonance shielding effects (Nordheim integral treatment). AMPX-KENO has been extensively benchmarked against a number of critical experiments'(e.g., Refs.
3,'4, and 5). i i l In' the standard reactor' core geometry, fuel assemblies at 39.6*F are - located on a 6.00-inch ~ center-to-center spacing, surrounded by full density (p =1.0 ) water, with all control blades removed
'and with no- credit- for the presence of gadolinium burnable
. poison.
4-3
.. ~
s 4
, , p Table.4-1.
SUMMARY
OF CRITICALITY ANALYSIS
=O , Nominal Design' Case k, in standard core: geometry 1.3308 i 0.0024-(10):
k, in spent'. fuel storage rack- 0.9295
. Calculational bias, Ak O.0036
. Uncertainties and tolerances g Calculational bias ' i0. 0028' A k
' Calculation 1(statistical). .i0.0025 A k .
?- Boraflex thickness T 0.0098 A k B-10~ concentration 70.0053 Ak Boraflex width T0.0013 Ak , Fuel enrichment *0.0043 A k l- - Fuel' density 10.0023 A k , ' Lattice pitch T0.0023 Ak * ( ))1 SS thickness- *0.0013 Ak . l Flow channel bulge 10.0038 Ak L 20.0136 Statistical' combination 10.0136 l
, Maximum k, (95% probability @ 95% confidence level) 0.9467 Abnormal / accident condition
- , . Temperature increase. negative Ak BoilingL ' negative Ak
[ -Reduced moderator density ' negative Ak
- Fuel assembly positioning negative Ak L
LA'ssemblyLoutside-rack negligible Ak i
' Dropped' fuel _ assembly negligible Ak
- O
'4-4 '
. .~ __ - _ _, . - _ . _. . .
e
.Fors ihvestigation of ' mechanical- tolerance ~ef fects, either the
- j CASMOS code 6 ~ ori a four-group diffusion / blackness theory method of
- analysis L was used to evaluate : trends and the small incremental reactivity ef fects that'. would otherwise - be lost. in the KENO -sta-ctisticalivariation.
'For two-dimensional 1 X-Y . analysis, a zero current (reflect-
- ing) boundary condition.was applied in the axial direction and at the centerline ' through' the .Boraflex absorber on all four sides of E
the cell', effectively creating an infinite array of storage cells-
-for analytical purposes.
Calculational -Bias and Uncertainty 4.1.3 Results ~ of AMPX-KENO benchmark calculations 5 on a series of
. critical' experiments indicate a calculational bias of 0, with'an uncertainty. of iO.0028--(95%- probability at a '95% confidence-
~
- level). 'In addition,- a small correction in the calculational -
- bias is- necessary to . account ~ for the slightly greater . gap thickness (1.1 inches) between fuel. assemblies ' in the Oyster Creek 1 spent - fuel rack compared 'to the - corresponding thickness p -(0.644 inch) in the ~ benchmark critical experiments. Based upon-
-the ' correlation developed' in ~ Ref. 5, the correction for water-gap
' thickness-in the Oyster Creek spent fuel' storage rack indicates a-smal1 ~ underprediction of 0.0036 A k. Thus, the net calculational !
. bias is taken as 0.0036 i 0.0028, . including the effect of the water-gap thickness .
- 4.1.4 Trend Analysis and Comparison with - CASMO Calculations
. Trend. analysis5 ; of AMPX-KENO. benchmark calculations on l critical experiments with ' varying boron content in the absorber plate between fuel assemblies indicates a tendency to overpredict k,gg with higher reactivity worth of the boron absorber.
i In the f . Oyster Creek ~ spent' fuel' rack, the boron worth is about 27% a k, 1 4-5 O e
,w.c..,s.--, ,. -- _ ,.
.,m._,,,,-o._ +-y ,..m..mm.,,.m.,-,-., -.,,.m-_,,.m.c.-w,--cm,wg.yw,,,.wm,%%w-,-rc
or. ~1.7 times ' the highest , boron worth (15.95 Ak) in the critical-N b' iexperiments analyzed in.Ref. 5. . Based upon' extrapolation of the trend c analysis ,- AMPX-KENO. calculations of the Oyster Creek ' rack would _ be expected tol overpredict ik, by an estimated 2% Ak,.in-cluding' allowance for water-gap - thickness . - . Statistically com-bining -the standard deviation of the regression analysis 5
'(10.002, lo -) and a typical._ standard deviation of the KENO variation of- the mean (10.003,;lo ), the maximum - uncertainty would L be ; iO .007, . including a one-sided tolerhnce factor 7 of 1.'92-(95%
j probability at'a 95% confidence level) for-100' generations in the reference: KENO' calculation. Thus, to - the extent extrapolation of
,- the . linear -regression analysis is valid, the AMPX-KENO calcula-tion of the . Oyster Creek rack will be high (overprediction) by O'.020 ik 0.007 A k, or a . minimum' overprediction 'of- 0.013 A k includ-ing calculational uncertainty. ,Althoug'h extrapolation of the regression trend 'much beyond the range of the measurements may be l questionable, the Janalysis does indicate that AMPX-KENO calcula-
~
~
tions 'would be , expected to overpredict k egg when strong boron
-absorbers are present.
7 Independent- calculations with the CASMO code 6 (a two-
; dimensional transport . theory code- developed by Studsvik based
.upon escape-probabilities) yielded a k, of 0.9107 for the refer-
' ence design storage rack, ~which --is. 0.0188 Ak less than the cor-
- responding AMPX-KENO value (0.9295). CASMO has been extensively benchmarked against critical experiments (Ref. 8) with good agreement. Thus, the CASMO calculation supports the expected-overprediction by AMPX-KENO.- However, no credit is taken for the expected overprediction of the reference AMPX-KENO calculation
.other than to indicate an additional level of conservatism in the criticality analysis of the Oyster Creek spent fuel storage rack.
4-t; l
r.. '~ 4.2 ' Input Parameters-O 1 25 . 4'.2.lI- Fuel.: Assemb'ly _ Design Specifications F
-Design' specifications . for the fuel assembly as used'in the g criticality. analysis are~ given in Table-4-2. Independent calcu-lations' with-~ distributed - enrichments typical of-BWRLfuel assem-
, blies confirm 'that the uniform: enrichment case yields .the higher L k, -( for the L same; average enrichment) and 'is therefore the limit-I Ling case for' criticality safety evaluations. Assemblies contain-ing ' fuel rods of- slightly different dimensions _ or configuration may.be safely accommodated in the high density spent fuel storage racks, provided the assembly;k, in -the standard core geometry is less than or equal to :1.3308 t 0.0024 as calculated by AMPX-KENO.
! 4~.2.2 ' Reference Design Spent Fuel' Storage Cell ( The nominal spent fuel storage cell model used in the criti- ' () cal'ity ' analysis. is shownLin Fig. 4.1. The-rack-is composed of Boraflex1 absorber material sandwiched between 0.063-inch stain-less-steel plates.- The fuel assemblies are centrally located in-
, each storage cell .on a nominal lattice spacing of 6.198 inches.
The'Boraflex abnorber has a nominal thickness of'0.040 inch and a nominal B-10 areal. density of 0.01114 ' gram B-10 Eper square centi-o met'er. I l l: I l 9 cO i-l L 4-7 J . .
l l Table 4-2 FUEL ASSEMBLY DESIGN . SPECIFICATIONS 1 II . l
- ". '. Al Rod Data Outside diinension, in. 0.483 Cladding thickness, in. 0.032 Cladding material Zr-2 Pellet density, g UO 2/cc 10.357 i 0.164 Pellet diameter, in. 0.410 Enrichment,-wt% U-235 (uniform) 3.01 i 0.05
' Grams U-235 per axial centimeter 14.51 Water Rod' Data Outside diameter, in. 0.591 Wall thickness 0.030 Material -
Zr-2 L - Number per assembly - 2 i-
, {v .. "uel' Assembly Data Number'of fuel rods 62
. Fuel rod pitch,-in. 0.640 Fuel channel outside dimension, in. 5.438 Fuel channel wall thickness , _ in . 0.080
_ Fuel channel material Zr-4 l i I l 4-8 _ , . - . . . _ . . - . - , __ _ ___. _ _ . - _ _ . . . . _ . - - - . _ .- .- . - - . _ _ - - - ~ - - - - - -
0.591* O.D., 0.531* l.D. WATER ROD
't h- -
~0 . 4"8 3 * 'O.D., 0.419" 8. D . CLAD .
\/ , ., . 0.410' O.D. FUEL
. 6 4 *- .
0.157' , r , , _ j j."q~~g~55~5~55j j
!i OOO~OO 00 ! ! :
i j OOOOO OO ! l ?! l 1 0000 000 l l 2? . l l O0000000 ! ! !!-
- o. ! !lO0000000 ! ! !!"
! ! OOOOOOOO ! !
!O0000000 ! !
$ sus u n n u s nuun u nun un nunsu s siin$i t
y
~
/ #
g _. ,L: ~ l ZlR C O NIUM FUEL CHANNEL 0.080* T HI C K. 5.438' OUTSIDE DIMENSION STAINLESS STEEL, 0.063* THICK l 6.198* OUTSIDE CELL DIMENSION O V. Fig. 4.1 Geometric model of Oyster Creek spent fuel storage rack cell. 4-9 ., -
. - , . ,. . . _ . . _ , , , , , .m, ...,_,,m, .-.g ,_,y , . . . . , _ _ _ . . , _ , _ , _ -
J4 '. 3 lCritica1'ity Analysis . ( . .
;J '4 ; 3 .1 - Nominal Case
~'
-Under : normal : conditions,- with. nominal dimensions, the
~
calcu1ated' k 'is 0.9295 i - 0.0014 (lo. with 250,000 histories in - 500 n generations) for o fuel of 3.01% uniform enrichment. For. a one-sided E tolerance : factor of-E1.763, corresponding to . 95% . proba-
.bility atia 95% confidence; limit for 500, generations, the maximum
- deviation of k, . is 'io .0025. With ~ the calculational bias and all
. uncertainties added,- the reactivity (kg) of the storage racks
, will x always be. less - than 0.947 with 95% probability at a 95%
confidence. level. Calculations with fuel of ' distributed enrich-
~
ments (average 3.01% enrichment)',1representat'ive of BWR-type fuel
. assemblies,- showed a lower reactivity 'than the corresponding
~ calculation-with uniform enrichment.- In. practice, neutron leak-
- age, L higher moderator temperatures 'and. the presence -of gadolinium
' burnable - poison will reduce the actual storage: rack reactivity -
below the. values!.indic,ated above.
~
- 4.3.2 : Boron Loading Variatiion The - Boraflex absorber plate is nominally 0.040 inch thick, with a B-10 areal density.of 0.01114 g/cm2 . Manufacturing toler-ance limits :. are *0.005 inch in thickness and to.00085 g/cm2 'i n' .
boron content. This assures that, at any point where'the minimum
. boron loading ~(0.01029 g B-10/cm2 ) and minimum Boraflex thickness (0.035 inch) may coincide , the boron areal density will not~be ,
less:ithan 0.009 g B-10/cm 2
~
l 2 Calculations were made . of k, f with . variations in ~ Boraflex l absorber - loading and thickness. Results of these calculations j
- (Fig. 4.2)- indicate that the k , calculated by CASMO- shows a l-4-10 f--, s ,4 y ,-
w +-l,e- , , w e-n--w-r- -w,_p-,.wn-w.m,e,,,.,,
O. 9 0 __i =_ ;u u. 7 + , i
.=.c._;_;.___.,t_._.-.._._..... . ;;=. ._. _;__-pj=___
. _ ._=._.a=.. . g .. -=.u . . - _ . ____ _ _ ,_l . _ =.;
_.p.
.. .._g ___.J.=_.:u=
...3__.,.
. _ ._. _ , _ ___,._o._..__. . _ , _ _ _ . . _ . __.. . . , . . . . _ _ .
h liPJ =i :-- +.: = I _d.=.. . =tr.: : .t=- =t= = -:1 :-
...a...
_ ___ _utii\ _.,;._. . _ . =;=
. . . =._,g_.; _ _ .=__=
.. ._ . :;~=._t= :-
, . . . . __ EIL.'.li
== m-t =__=_ ,:it -._.
= t =--
.(
y
==l=
=_.t=-._ $.1 _.._. = =_t= , . . . . . , _ . .
. _ . . . _=III=;
- t---
EE}=_= :=!== -t=
=t=:' ~- ~-~_:. _.__ .
_ ={ , _ ===. _= _ t.;
==.= :
-+-
, , -i1_=_:
J i. ! =:= E'2~-%= :===t := r h:= :.... 6 :-- . _ _ . .._&_
---x-Nf . . -0.98 . . - . . _;.n- 2. ... . n ; -a ------- . - , - . - -
--. =
_= , - - -t=
- . . - - , - --r- -
___= l E- E. --- - -
.=2=- ; t=.: ,u . - n_. i
- - - + - -
- ,--- ---+- y-. ==- _= - =r- -
\
--. : -- - =ug. 'gj=.:.:}y=.u...__t=
\.. .__ __ _.:n. ; =.~.u.-.- ;__ r- =_-- _ - - - - -i - _
_a- = -+- 9- _. 3;.g r=. .A.=.,. : =. i =jt.--+ g ;. , ;T, - +._-
.. t. =. . .:.=.._.:==. -. - =._g=- == -
. . _ _ . _ _ . = p._.--t-
----+---- -- -
1 t ._ _ _ . .g . I= :::= \.=:=\j.._.___ == C "Il ==t= . i. . .j . _ _ . _ _
==_a__.
l ___.
=_-t =:= :=_$= :i.._ . _ - _-
q, g y .=.2__
.m- _-_. Q-
_7_ .
-.3 - ---- 1 &==.
_x . ;
==\ s_-== =_y__
.:=\..g=.;
- --- ._-=:= .
-- -r== . . _ .
-+- -h \r ._.
i %-..
. _ _ . . _ _ _ _ _ _ . . . _ . ., _ . . ._9_._.
_ . i .\.. ._ . . _ . . 1 . ...g;= t _.._..:_- _
..{__ . .__ . =_ _ _ _ _ . = , . _ _
_=t_-__.. -== . _ _ + _ l .
- - -- - \=t= = __- _.,i .:\==.:-i_. ._ -- - t.- '=:
~ - - -
-- - V ' -
p 0*96 - - , - - N._ _ . * -=;\- .- _ ;___ .
- p _. __. .% ___
__._...\.___ _..-,_g
\_.__.h- .2_._a
_ . ~ _ _ ,. _ _ . .
- t. _.-
l \ __..s=t-\ e:.=.C~ES.MiOEiN~OIRVWUlliZ.E D
.___, E _ __ _ . J-~"-+: : -T----~ ~p=._y-Ey-~N'G R.EIE E RiEiN.C:E5iS.p-21:01
' ~ 0.95 -.=: A .M__ : P.T. . W__ _ ..E._N. O. _' -' _'~x. . ...".~~ -- 5 %~u_E_T. . - '- '
. Md-
\.i v
_Nt=\-- . . i X-- s\__i k Y\
\ x-~ x-t-4 -
0.94 - t _ _ . . \s % g-*
, v u . \ h 6 't. 'g , . -
t-x_j \ ! l0.93 -
- y% _
.*-_R:E EEiR:EiNic.IE_.".-----
- l. ~C_ _'s_$. ._M C_ _h__.T_ _.i._._
\\- ._._
l
\-s
~
,- Nx L A k
\=K K:
l 0.92 ! xx
. i9 A: . , -
- . _a _ _\_ -+- :\=: __
t
. x. _ w
\;
l i_l_-\ s.- g : ,_
= ; 1 V -
l 0.91 ___-
.;_g. .__,._g__
_, -s__
-_s i
x s - T-~ E- -_ 1' =.--. +-_ _ i_____N _ . _ . _ _ _=__ i.. w __ _ = _.I
==
. _ _..=_=_=_d- _
___ i k. __.s !. . _. ._._.._-- . _ _ __ __\ 0.90 : _.;;- --.- -- =r-- -- --
----t.=.=-
A-.\ _;__
=+= - ;-t-
-. __4
- = iEi= =;=a __e .
___,_=
- _ _ ___.._.r=_ . _ _ . . _,
_ _ _ _ . . ._m _; =__T,= _.r--- -. _- _ \. = _ 2--- -- \ _ . _ _ _
- - - * = ::.
! -"t=-- =t=. ---" . _.__..__.__ rh ._.._ . _ . _ . . _ . .._
=*= (*--t= _ . . _
_.1:=_==== _- _t_---
-r 0.89
__ .__ :-... -- t _y=a , =_=_:---_ __ - - - z 4 .- .
- i_..-.jEi(31!Ei =;iH-IF94 =ii ! HE [-lg.. . __~{ : q E jsE}EE'* f{=;= .~. "p~ . iEf:E-i.EI l
N 0.005 0.006 0.008 ,0.010 0.012 0.015 , ' 2 , i B-10 LOADING. 9/cm t i l Fig. 4.2 Log-log plot of calculated k ao values versus B-10 j loading. 4-11 - _ - _- .
l l l i slightly . larger. gradient than~ either- .AMPX-KENO or i
., 7 diffusion / blackness theory ~ calculations. For ~ conservatism, the.
l V incremental reactivity. values from CASMO calculations were used
.to de fi'ne . the . uncertainty .in reactivity due. to Boraflex
{ -manufacturing: ' tolerances.- -The tolerance- on limit boron ^ L
- concentrat. ion. and.'Boraflex thickness results in incremental
. r g .
reactivity; changes.of i0.0053 Ak-and 10.0098 Ak, respectively. f 4 ~.3.3. Boraflex.Widtn Tolerance Variation
@ 4dh t
- . A decrease in Boraflex r . p3 Ste _ width increases reactivity.
For.. the ~ manufactdring tolerance limit - of i0.0625 inch - on the
. width - of . - the Boraflex absorber plate, the corresponding uncer-
. tainty in . reactivity is T O .0013 ' A k, calculated by .diffu-sion/ blackness theory since the. reactivity increment is too small 4
to-ber. calculated by AMPX-KENO. l 4.3.4 ' Axial Cutback 11 Boraflex Length oO
. The axial length of the' Boraflex . absorber is less than the active- fuel length by 3- inches Eat both the ' top and bottom of the fuel' racks .
This' axial . cutback occurs in the region of' high neutron ~ 1eakage. Axial- calculations (one-dimensional) indicate
~
that the incremental reduction in' reactivity due to axial leakage-(0.0027 Ak)uis greater than the increase in reactivity due to the 1 axial' cutback - (0.0013 Ak) . .Thus, 'the .nfinite multiplication f ac tor ( k,, ) used .as the reference reactivity is the more conser-vative, condition. In addition, the upper and lower 6. inches of
~
' the: fuel ~ assembly..contain natural UO 2
, which would further reduce any reactivity effects 'of- the axial cutback in- Boraflex.
[ . 1.
., u 4-12
? i; _
- . . - - .,:.. , , . - , _ _ , - .. _.- . . ,,., ~ ...., , a .-_ _. -_.,..,- _ ._ . _-. - . . _ - _ _ , _ . . . . _ _ . - - . _ . . _ , - . _ . -
~
4.3.5- Storage ' Cell Lattice Pitch . Variation He-design' storage cell lattice spacing between fuel assem-blies -is 6.198 iL- 0.0625 inches. Increasing' the lattice pitch reduces ;/ reactivity. For the manufacturing tolerance of 10.0625
- inch, the- corresponding . maximum uncertainty. in reactivity 11s'70.0023 Ak, c'alculated ' by J CASMO, since the: reactivity. incre-Taent ..is I too small to be calculated by- AMPX-KENO. ,
i 4. 3.' 6 Stainless-Steel hickness. Variations He nomin~al stainless-steel- thickness. is 0.063 i .006
- . Linch. The - maximum 1 reactivity . effect of .the expected stainless-
.' steel J thickness tolerance variation .-(0.006 inch) was calculeed to lbe :i 0.0013 ak by diffusion / blackness theory, since the reac- a
.tivity . increment iis L too small to be calculated by AMPX-KENO.
~
4.3.7' Fuel Enrichment 'and Density . Variation ? . .
- The ~. nominal design enrichment is -3.01 wtt U-235. Calcula-Etions' .of ..the . sensitivity to small . enrichment variations by dif-fusion / blackness-theory. yielded an average coefficient :of
?0.0086 Ak per 0.11wt% U-235.- For an estimated tolerance on U-235 '
enrichment' of iO. 05% , the maximum uncertainty .is 10.0043 ak due l; to the. tolerance on fuel-enrichment. I
> Calculations Twere ' also made to determine the sensitivity to
~
-tolerances in UO 2 fuel density. These calculations indicate that the storage rack k, -is increased by 0.0023 ak: (by CASMO) for the
~
expected' tolerance: (il.5% in T.D.) in' fuel. density. A lower fuel l- ! density; results inL correspondingly: lower values of reactivity. L. - Hus , the : maximum ' uncertainty due to the tolerances on UO 2 dens-l ,' - ity is wi0.0023 Ak. In addition, for conservatism, the stack i y ' density ' was assumed equal to the - pellet density (94.5% T.D.), i-thus neglecting.any. reduction in density due to' dishing. lqt l- 4_13 , L
._ . . - , ~ _ . . _ . - _ _ _ . , . . _ - . _ _ - . _ . ~ . - . _ _ _ _ _ _ _ _ - - - _ _ _ _ _ _ _ _ _ _ - - -
M 4.3.8 Effect of .%irconium Fuel Channel
. .t Elimination of the zirconiura . fuel channel results in a small.
. " decrease in reactivity (-0.0053 Ak). 'More significant is a'small positive _ reactivity effect resulting from bulging of the . zircon- I l ium channel, . .which ; moves the channel wall outward toward the !
'Boraflex' absorber. .
For ' the maximum ~ expected bulging (to 5.93 inches outside " dimension)' uniformly - throughouti ' the assembly, an i: . incremental reactivity of +0. 0038 Ak would result,, as calculated l byJdiffusion/ blackness. theory. h ..ce actual bulging of the flow l . dannel ' would not be the maximum everywhere -in all, assemblies ,
. the reactivity -effect has been statistically - combined -with the.
j 'l
' reactivity effect of other mechanical deviations..
Fuel assembly bowing' yields.a negative reactivity-_effect and , l is treated under abnormal conditions-(Section 4.4.4 below). l l no - 1 I
'I L
l l l l. l . ! l
- i L
L u i 3 i N 4-14 i , I
- , . . . _ . ., , , < , - . , , .,,,___,,,_,_4-, ._ _ - - - . . . _ . . - _ . , -._ .- - . . . .- i...______._____----_.____
. I l
4.4 ' Postulated ~ Accidents and Abnormal Conditions i . , (O 1 j., .4.4.1 ' Temperature and-Wate.r Density Effects i The nominal criticality ~ analyses were performed for the l maximum -water. ~ density (p = 1.0) corresponding to a temperature
- As shown- in Table 3, of ~ 39'F. increasing temperature : or in-troducing void -(to simulate boiling) . decreased reactivity of the spent fuel storage rack.
L 1 I Since . the storage rack is- not intended to stiore fuel in the l
, dry condition, calculations-were not made for moderation by water of reduced . density. ~In poisoned racks, the maximum reactivity.
would occur for the fully flooded case, and the reactivity would
- be 'substantially lower for all other modei ating conditions.
Cano et al.9'have shown, in ' a parametric study, that the phe- .
. nomenon 'of a second maximum.in reactivity does not occur in a p closely-spaced lattice with a-strong neutron absorber present.
- f. i Table: 4-3 EFFECT OF TEMPERATURE AND VOID ON i l REACTIVITY OF STORAGE RACK '
I Case. Akm Comment l 39'F - (~ 4* C) Reference Maximum. water density 68'F (20*C) . -0.003 p(H O) = 0.998 2 L 104
- F ( 40
- C) - -0.007 p(H 2O) = 0.992
' 176'F- ( 80' C) -0.017~ p (H2 O) = 0.972 j 212* F ' (100
- C) -0.022- p (H2 O) = 0.958 l 212*F'with 50% void -0.156 Simulates boiling l l
4 . 4 ! l
4.4.2 Abnormal Positi6ning of Wel Assembly Outside Storage Rack (b. 8
") Since the storage rack criticality calculations were made assuming.an' infinite array of storage cells with no neutron leak-
. age , - positioning : a . fuel assembly outside and adjacent to the
^ actual finite-rack cannot add reactivity, but would, because of
. neutron leakage, result in a lower k gg e than the k ,, calculated for.the infinite array.
4.4.3 Nel Assembly Positioning in Storage Rack The fuel assembly is normally located in the center of the storage rack cell' with bottom fittings and spacers that mechan-ically. prevent lateral movement of the fuel assemblies. Never-theless, calculation: were made with adjacent fuel assemblies (each assumed to ' be located on one side of its cell with the zirconium fuel channel- touching the SS-Boraflex plate) creating an ' infinite series of two -assembly clusters separated only by the d = SS-Boraflex plate. For this case, the calculated reactivity was - sliglitly 'less than the ' nominal design case (by 0.006 A k) . Calcu-l 1ations were also made with the fuel assembly moved into the l-l - corner of the storage rack cell (four-assembly cluster at closest l approach), resulting in an even larger negative reactivity effect (calculated decrease ' in k,, of 0. 013 A k) . With the zirconium l- fuel channel removed, the reactivity effect of offset fuel assem-blies would be even more ne'gative. Thus, the nominal case, with the fuel assembly positioned in the ' center of the storage rack cell, yields the maximum reactivity. L [
^()
4-16 l i
.. , - --. - . - .~ . .
4 4.4 Effect of Zirconium Fuel Channel Distortion p. O 1 Consequences - of - bulging of the. zirconium fuel channel- are treated:as.a mechanical deviation in Section 4.3.8.above. Bowing of the .zirconiura channel -(including ; fuel rods) results in a local negative reactivity effect analogous to that of positioning the t fuel assembly <toward one side of -the. storage cell, as described
-in Section - 4. 4. 3 above. _ Thus, bowing willf result in a reduction i~ Jin reactivity.'
? l 4.4.5- Dropped Wel~ Assembly Accident l_ To. investigate the. possible reactivity ~ effeet of postulated fuel drop accidents, calculations were made for unpoisoned assem-
.blies' separated only by. water. Figure 4.3 shows the - results of
.these calcul~ations. Fro'm these data, the reactivity (k,) will be ,
iless than'_0.90 - for. .any water-gap spacing 'between unpoisoned assemblies greater than ~ 2. 5 - inches (~ 8-inch _ assembly pitch ) . ;
~ Ebr a . dropped fuel . assembly. lying ~ horizontally on top of the -
. rack,1: the minimum separation : distance is ~ 14 inches. Maximum 1 expected ' deformation under seismic- or accident conditions (see )
, :Section 6) will, not reduce the minimum l spacing to less than ~ 2.5 i l inches. Thus, _ a ~ dropped fuel assembly wil1 not constitute a I l criticality - hazard , and -the storage' rack infinite multiplication l: factor will-not be materially altered.
- 4. 4.' 6 ' Wel Rack Lateral Movement !
I
-Normally, the' individual / rack modules in the spent fuel pool 1 are; separated.by a. water gap of;several inches. For finite fuel i racks, this ' separation would reduce - the actual maximum . reactivity -
of the racks.- ,Should lateral motion of a fuel rack' occur,. clos-
- ing the gap between -racks (for whatever reason), the reactivity
- would, in-the limit, only approach the limiting reactivity of the
. reference _ infinite array.
4-17 :
. WATER-CAP DETWEEN A G O E M B l. l E S . INCHEO ;
1 2 's 4 6. e 7 23 n:: : .=- na
=lu t == -1 =~ ~q -
"t
= nl _.r;I i.
- ,/
\
< ;-m--
!--- :::In..c.I ----e--
.:.{::: :: :=t - .
uj;:=3
~~
,n -+ =:=
=rli --
---t==
-T 1*3
-:= ::- - -- =r" _
-"~_--..L..
=t..=--- =r - :
-: = - - - * - - -
._.2C ...- nri = =t'C
~ = ==- C - ~ ~ * ~
._~_~
Ms- . _.i zu t- ;=:= -----
-r-
=. :- - - - -
- .1.=- =_ l=.=. =t= =:- - - -
- ;__. n. !=n2 g =- nu t== ---t--
- t-- __ 7:r ret = :nt= <
J
.;_=_. i . n. . . . , .,. - . . -
=_. 4:_: .- -. . . . _. .:.
.,. ;=.= .,=._ :=_ =;: ..
_r_i ,=_ _=_m=;_=__
. =.
=_ ; __nn__. t.=_=.:-- t _.
2 ,=
==f=\BaEE Pa _ ==E= Ea#E 9-=iEEC2x EElis - !- r= =5=
- x. _ -
- ===w = - --- g=- : -t m= __:-,\ ::r =_ .:__= :~_ =---- :=
_ . , . _ = . r= r r}_ __,- .= _ = =. t:_--
~-. . _ .~ =.=____ _ = , . . _= _ _=
+- -t--- --+ -
1.2 =- _lE ____=- _1 _
..__s. . _ -- . -
_ _ , _ . . _.__. ,;._m_m.__,..._._. _ . _.=.z,._... . _ I
~ . . . ._ _ _ _ _ . . _ _ . . . . .
_3 __._. s i =__. __ y + .----
=-
1.~1 !'____
~
c&- - i
- _,__. gut .
- k. 4 m in.N iew----
_t 1.0 - D
-~
_ _.h D I E EV S T.O NJ.__
, t"_.\. TJH.E. 0 -R.___Y_--'---
s u _
- -\=4
\ ,
N-- ._. , y ,i 0.s - s q,,/ -
.T.
3 i\ .
~
o.a = A: __. 'N., x __x - N. x
- --i 0.7 -.___. ___
,=.=__
-4 -._.- - - -
+._
N w .
-.-- =L __ . _ . .
?
- ,_ 3__,__ ., -
0,s -_ __._< . ._ __ _
. . _ _. i._ . s :-liEim._a= .____...i--+-t__._ .
= .
_.._= 8 7 8 9 10 11 '12 CENThR-TO-CENTER A S S E M'B l. Y S P A CIN INCHES G. I A i t j- Fig. 4.3 _ km of unpoisoned fuel assemblies as a L function of assembly spacing and water gap. 4-18
4.5 . Acceptance Criteria for Criticality ~ I ' ' The USNRC . le'tter of -: April '14, 1978, ;to all Powerf Reactor - Licensees,; and the draf t revision .to Regulatory Guide 1.13 . spec- l ify! that - the neutron multiplication . factor - in spent' fuel pools-shall be:less_than'.-or equal to 0.95, including all uncertainties, when fully loaded with fuel ' of - the highest : anticipated reactiv-ity. For BWR-type-fuel with an infinite multiplication factor of. 1'.3'31 'or 31ess in the standard reactor core geometry . ( AMPX-KENO . calculation),. _ the spent -fuel storage rack described herein sat-- isfies this - criterion. l i fh l e i
')
6 0 4-19 9
,-.,-----,--,--n---
l REFERENCES l., Green, Lucious, Petrie, Ford ,- White , Wright, PSR-63/AMPX-1 (code package),- AMPX Modular Code System for Generating
. Coupled Multigroup-: Neutron-Gamma Libraries from' ENDF/B, ORNL-TM-3706, Oak' Ridge National Laboratory, March 1976.
i 2.- L". c M . t Pe trie and - N . F . Cross , - KENO-IV, An Improved Monte -
. Carlo Criticality Program, ORNL-4938,' Oak Ridge National Laboratory, November 1975.
3._ _S. R. Bierman et _ al., Critical Separation Between Subcritical Clusters of 4.~ 29 wt% U-235 Enriched .00 Rods .in
- i. Water with ~ Fixed Neutron Poisons, NUREG/CR-0073, 2Battelle
. Pacific Northwest Laboratories,~May 1978, with errata shset issued > by the USNRC August 14, 1979.
- 4. M. N. . Baldwin et al., Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, The Babcock - 6 Wilcox . Company, July 1979.
- 5. .S.'E. Turner ' and M. K. Gurley, Evaluation of AMPX-KENO Benchmark ' Calculatio~ns for High Density Spent Fuel Storage Racks, Nuclear Science and Engineering, Vol. 80, No. 2,
-pp.>230-237, February 1982.
- 6. A' . Ahlin',1 M. Edenius, H. Haggblom, ._ CASMO - A Fuel Assembly 5[V}
E Burnup ,-Program, AE-RF-76-4158, Studsvik report l l(proprietary). i
-7.- M. G. Natrella, Experimental Statistics, National Bureau of 1
. Standards,. Handbook 91, August 1963.-
- 8. E ., E. - Pilat, Methods. for the Analysis of Boiling Water Reactors,- Lattice'_ Physics, Yankee Atomic Electric-Co., YAEC 1232, December 1980.
- 9. . J. N. Cano- et .al., -Supercriticality- - Through Optimum Moderation in -Nuclear Fuel Storage, Nuclear Technology,
~ Vol.'48, pp.;252-260,.May,1980.
b l
~
+ e r
. . 9 4-20 O
<-+,,---a-evv nw~*-=,-=**"c"=~~~
'5.- THERMAL-HYDRAULIC CONSIDERATIONS
'A central. objective in the design of the high-density fuel rack is to . ensure adequate cooling of the fuel assembly cladding.-In the following, a brief synopsis of the design basis, the method of; analysis, and computed results are given. Similar analysis has been used in previous Licensing Reports ~ on High 1-Density Fuel Racks for Fermi II (Docket 50-341), Quad Cities I and ' II (Dockets 50-254 and 50-265),.~and Rancho Seco (Docket l 312)'.-
l l This section covers require.nent III.l.5(3) of the NRC "OT l Position foe Review and Acceptance of Spent Fuel Storage and
. Handling Applications" issued on April 14, 1978. Conservative methods have been used to calculate the maximum fuel cladding l .- tempe ra ture as required therein. Also, it has been determined
.that nucleate boiling or voiding of coolant on the surface of the
, ' fuel rodn does not occur.
5.1 Basis ~s , The methodology employed in this analysis may be found in
.the ASME Paper, "A Method- for ' Computing the Maximum Water Temperature in a Fuel Pool containing Spent' Nuclear Fuel", by
- K.P. Singh and A.I. Soler, '4th Nationa'l Congress, Portland, l'
Oregon, June 19-24; Paper No. 83-NE-7. In' order to determine an ' l upper bound on the maximum fuel cladding temperature, a series of conservative; assumptions are made. The most important assumptions are listed below:
- a. The pool bulk temperature is assumed to be at the
[ technical specification maximum of 125*F. I-l[ b. .As stated above, the ' fuel pool will contain spent ! fuel with varying " time-after-shutdown" (ts)*
'Since the heat emission falls off rapidly with l- increasing t, s it is obviously conservative to e
5-1
assume that all fuel assemblies are fresh (ts " 7 days), and they all have had 1430 days of continuous operating time -in the reactor. The heat emission rate of each fuel assembly is assumed to be equal to
.' the average power multiplied by the radial peaking factor (equal to 1.5 for'OCNS).
- c. As shown in Figures 2.1 in Section 2, the modules
. occupy' an irregular floor space in the pool. For l s1perposes of the hydrothermal analysis, a circle circumscribing the actual- rack floor space is drawn. It is further assumed thst the cylinder with this circle an' its base is packed with fuel assemblies at the nominal pitch of 6.198" inches
-(see Figure 5.1).-
i d. .The downcomer space around the rack module group varies, as shown in Figure 5.1. The nominal ) downconer gap (2.5") available in the pool is I assumed to be the total gap available around the
. idealized * . cylindrical rack; thus, the maximum l
- resistance to downward flow is incorporated into the ,
analysis. j !, I ! e.- No downcomer flow .is assumed to exist between the ; rack modules. 5.2 Model Description l l - 's l In this manner, a' conservative idealized model for the ' rack assemblage is devised. The water flow is axisymmetric about . the vertical axis of the. circular rack assemblage, and thus, the i flow is'two-dimensional (axisymmetric three-dimensional). Figure 5.2 shows ,a typical " flow chimney" readering of the A 5-2 I L l
~
. ? thermal-hydraulic model. The governing equation to characterize L the flow field in the - poo'1 can now be written. The resulting integral ' equation can be solved for the lower plenum velocits
- field ~(in the radial direction) and axial velocity (in-cell velocity field), by using the method of collocation. It should <
f ' l be added here that the hydrodynamic loss coefficients which enter , into the' formulation of the integral equation are also taken from l i well-recogn'ized sourcesl' and wherever, discrepancies in reported values exist, the conservative values are consistently used. 1 l l After the axial velocity field is evaluated, it is a straight-forward matter to compute the fuel assembly cladding temperature. - The ' knowledge of the overall flow field enables pinpointing the storage ' location with the minimum axial flow (i.e. maximum wa'ter outlet temperature). This is called the most l
- " choked"' location. It'is recognized that some storage locations, where rack module supports are located, have some additional i , hydraulic resist,anca not encountered'in other cells. In order to j find . an upper bound on the temperature in such a cell, it is assumed that it is located at the most " choked" location.
Knowing the global plenum velocity field, the revised axial flow , through this choked cell 'can be calculated by solving the Bernoulli's equation - 'for. the flow. circuit through this cell. Thus, 'an _ absolute upper bound on .the water exit temperature and l maximum fuel ~ cladding temperature is obtained'. It is believed that -in view of the preceding assumption, the temperatures
- calculated in this manner over-estimate the temperature rise that will actually be obtained in the pool.
The following radial'and total peaking factors 2 are used in the
,. analysis.
a ,= 1.5 (Radial peaking factor) U- . 5-3 c- -w--.w.. 'y,---,
.,- y -,3y ,..,,em .,,my,y-ww,-,*w.,., _m...w w---.%~.my- _ . - . . _ .,..w_-m -
w--<---
7 a, = 3.02 (total peaking factor) Therefore, the axial peaking factor is 2.01. l The maximum temperature rise 'of pool water in the most disadvantageously placed fuel assembly is given in Table 5.1. For sensitivity analysis purposes, pool water temperature rise for two other lower plenum heights is also given. Having determined the maximum " local" water. temperature in the pool, it is now l possible to determine the maximum fuel cladding temperature. It i is conservatively assumed .that the total peaking factor aT is 3.02. Thus, a fuel rod can produce 3.02 times the average heat I
- j. emission rate over a small length. The axial heat dissipation in
( a rod is known to reach a maximum in the central region; and taper off at its two extremities. For the sake of added conservatism it is assumed that the peak heat emission occurs at ( the top where the local water temperature also reaches its ! maximum. Furthe rmore , no credit is taken for axial conduction of
.g g heat along- the rod. The highly conservative model thus l
constructed leads to simple algebraic equations which directly give the maximum local cladding temperature, te.
'5.3 Results'and Discussion Assuming the pool bulk temperature to be at the technical specification maximum value of 125'F, Table 5.1 gives the peak values of local pool water temperature and fuel cladding temperature. Since the gap between the rack outer boundary and pool wall varies from a minimum of 1.5," to a maximum of several feet (Fig. '2-1) computations are performed for two discrete j values of rack-pool gap (downcomer width). The height of the l module base plate over the pool floor is 6" in most modules.
i However, two modules have elevated plenum heights of 11.5" (table 5.1). Parametric studies with plenum heights of 7.5" and 10" are j also performed.. l V 5-4
)
O l Table 5.1 Maximum Local Pool Water Temperature & Local Cladding Temperature Serial 4 Downcomer Plenum Pool Local Cladding Width. Height Water Temp.*F Temp. "F - 1 1 1/2" 6" 171.4 219.0 2 2 1/2" 6" 159.0 202.0 3 1 1/2" 7 1/2" 171.0 219.0 4 2 1/2" 7 1/2" 159.0 201.0 5 1 1/2" 10" 171.0 218.0
.6 2 1/2" 10" 158.0- 201.0 7 1 1/2" 11 1/2" 171.0 218.0 8 2 1/2" 11 1/2" 158.0 200.0 m
4. 5-5
. . . . ._ _- . - . . _ _ . - _ _ - - - = . _ - . . - . _ .-
REFERENCES TO SECTION 5 r< , O .
.l.. General Electric Corporation, R&D Data Books, " Heat
! Transfer and Fluid Flow", 1974 and updates. i
~
l' '2. Letter from Mr. Walter Vinpa (General Public' Utilities) , j to Mr. Raj Shah (Joseph Oat Corporation), June, 1983. I. l n O
- 5-6 -
9 0 .
l id e a liz e d Outline Ou e of i O f R a c k . A's s e m b l y
-Rack Assembly I
- l. r h ff *
~ RACK ASSEMBLY g i /
\
g3 l l' Actual Outline of Pool idealized Outline of Assumed Added ! Pool Boundary Fuel Assemblies O : FIG.:5.1 IDEALIZATION OF RACK ASSEMBLY L-.....-.--.~-.-_ ai -
O Water Assumed At The Pool Bulk Temperature I
,r - + ,
f N l I J L h
/ /
D
/ _ _/
f- l TOUTMQ
- O IN7 _'
@g "j h 2 8' -
q ] J ! O
- r y
b +-Q Heat Addition
, Y l)L 2 i\ I
_ / n l Q l
/ / IIN 4 -
i
~
(B/7/J s -
' > j
/
/
_/, t f / l FIG. 5.2: THERMAL CHIMNEY- FLOW MODEL . 5 8 ,
-. ._ - . --- - - ~ - - .- - . --. _-
l 6.. STRUCTURAL ANALYSIS The purpose : of . this section is ~ to demonstrate the structural
. adequacy of the spent fuel rack design under normal and accident loading conditions. The method of analysis presented herein is r
similar ' to that previously used in the Licensing Reports on High . Density Fuel . Racks .for Fermi II (Docket 50-341), Quad Cities I and .j II '(Dockets 50-254 and 50-265) and Rancho Seco (Docket No. 50-312). The results show that the high density spent fuel racks are structurally adequate. to resist the postulated stress ; combinations associated with normal and accident conditions. i
~ 6 .1 Analysis Outline r
The spent fuel storage racks are seismic Category I i equipment. Th.us, they are required to remain functional during and , , i after an 'SSE (Safe Shutdown Earthquake).1 As noted previously, these racks are neither anchored to the pool floor, nor are they Q attached to the side walls. 'The individual rack modnles are not
- b. interconnected.' Furthermore, a particular rack may be completely loaded with . fuel assemblies (which corresponds to greatest rack inertia), or it may be partially loaded so as to produce maximum- :
geometric eccentricity in the structure. The coefficient of between the supports and pool floor is another friction, y, indeterminate factor. According to Rabinowicz,2 the results of 199 tests performed on .austenitic stainless steel plates submerged in water show a mean value of a to be 0.503 with a standard deviation of 0.125. The ' upper and lower bounds (T2o) are - thus. 0.753 and 0.253, respectively. Two separate analyses are performed for this rack assembly with values of u equal to 0.2 (lower limit) and 0.8 (upper limit) respectively. Initially, the following four , separate analyses are performed on the largest rack module (Module A). '
- 1. Fully loaded rack (all storage lochtions occupied),
p = 0.8 ( p = coefficient of friction).
. 2.- Fully loaded rack, p = 0.2.
6-1
- 3. Empty rack, p = 0.8.
l.m 4. Empty rack, p = 0.2. The-owner intends to store fuel in the rack modules in such a manner that_ the modules are always symmetrically loaded. Therefore, asymmetric load cases reported here are strictly for informational purposes only. Based on the results of these , runs, additional analyses are performed. The actual studies performed for the different rack modules are summarized in Section 6.6. . The method of analysis employed is the time history method. The pool slab acceleration data are developed from the original i pool floor response spectra. . The object of the seismic analysis is to determine the structural response (stresses, deformation, rigid body motion,
'O etc.) due to simultaneous application of- the three orthogonal b excitations. Thus, recourse to approximate statistical summation techniques such as ' " Square-Root-of-the-Sum-of-the-Squares" method 3'is avo'ided and.the dependability of computed results is ensured.
The seismic analysis is performed in four steps; namely
- 1. Development of nonlinear dynamic model consisting of beam, gap, spring, damper and inertial coupling
,lements.
e
- 2. Derivation and computation of element stiffnesses using a sophisticated elastostatic model.
4 A Tk:51 I
- 3. Layout of the equations. of motion, and inertial
' /~l decoupling and solution of the equations using ~ the
" component element time integration" procedure 4,5 to determine nodal and element forces and displacements of nodes.
- 4. Computation of the detailed stress field in the rack
. structure, using the detailed elastostatic model, from the nodal forces calculated in Step III above.
Determine if the stress and displacement limits,given in Section 6.5, are satisfied. , A brief description'of the dynamic model follows. 6.2 Fuel Rack - Fuel Assembly Model
. 6.2.1 Assumptions I a. The fuel rack metal' structure is represented by five lumped masses connected by appropriate elastic springs as shown in Figure ' 6.1. The spring rates simulate the' elastic behavior of the fuel rack as a beamlike structure.
- b. The -fuel assembl'ies are represented by five lumped-masses located, relative to the rack, in a manner which simulates .either fully or partially loaded conditions.
- c. The local flexibility of the rack-support interface is modeled conservatively in the analysis.
- d. The rack base support.may slide or lift off the pool floor.
D 6-3
s _ _ ._ . 4
.e. The pool floor is assumed to. have a known time history .of ground accelerations along the three orthogonal~' directions.
- f. Fluid. coupling between rack and assemblies, and
-between ~ rack -and adjacent racks is simulated by
. ,. m introducing - appropriate inertial coupling into the system kinetic energy.
g.' Pctential impacts between rack and assemblies - are accounted for by appropriate spring gap connectors between masses involved. e damping between
- h. Fluid rack and assemblies, and j between rack and adjacent rack is conservatively
'., neglected.
- r-
-1. The supports are modeled as extensional elements for-
-dynamic analysis. The bottom - of a support leg is attached to a frictional spring _ as described in Section 6'.2.2. The cross section- properties of the support beams- are derived and used in the final computations to determine support leg stresses.
' j. The effect of sloshing can be shown to be negligible at the bottom of a-peol and is hence neglected.
e f a
- ^-
_7
~6-4 '
t
~
l6.2.2 Model Description t The ~abso' lute degrees of freedom associated = with eadh of the ! mass locationsLi, i*Lare as follows'(see Figure 6.1): Table.6.1- Degrees of' Freedom-i Location . . Displacement. Rotation (Node), ux. uy uz 0 x; O y Oz l- p1 p2 P3 94 95 96-1* Point is assumed. fixed to base at.X B'Y B,2=0 i . 2 p7 pg q11 ql2 l ' O 2*= p8 P10 n i o 3 .pl3 P15 '917 918 ll 3*- p14- P16 0-4 P19 P21 923 924-4* p20' P22 5- .p25 P27 . P3 2' 929 930 931
-5* - p26= P28 -
Lf O 6-5
1 Thus, there are 32 degrees' of freedom in the system. Note that elastic motion of ::w rack in extension is represented by
,' f-( _. .. neralized coordir e/4... ,3. and P32
. This is due to the
. relatively high axial rigidity of the rack. Torsional motion of the rack relative to its base.is governed by q31
-The members joining nodes 1 to 2, 2 to 3, etc., are the beam elements with deflection capability due to bending and shear (see -Reference 4,pp. . 156-161.). The elements of the stiffness
. 'watrix of these beam +,b .. re readily computed if the effective flexure modulus,-torsion modulus, etc., for the rack structure are known. These coefficier.ts follow from the elastostatic model as
-described later. The nodal points i (i = 1,2.. 5) denote the fuel rack mass at- the 5 elevations. . The node points i* (i* =
1,2.. 5) denote the cumulative mass for all the fuel ~ assemblies distributed a.t 5 elevations. The element stiffnesses of the fuel assembly are obtained from the structural properties of the OCNS fuel assemblies. The nodes i* are located at x = XBe Y
- YB
..dn: the global coordinate. syster shown in Figure 6.1.
6.2.3 Fluid Coupling An effect of some significance requiring careful modeling is the so-called " fluid coupling effect." If one body of mass mi vibrates adjacent to another body (mass m2), and both bodies are submerged in a frictionless fluid medium, then the Newton's 1
/ T 6-6
_e-r- - e - , -
t'
- i. .
. equation of motioni for the two bodies have the form
("Ij+ Mll)'X
..-I M12 X2 = applied forces on mass mi
~-M 21 X 1 + (m2 +:M22).X2 = applied forces on mass m2 I
.M1 1 ,- M12s' IM21, -and: M22 are fluid coupling coefficients which depend 'on the- shape of the two bodies, their relative-Fritz6 gives data for Mij for various body-disposition, etc.
L . shapes'and arrangements. It is to be noted -that the above equation e . indicates thatreffect of the fluid is to ~a dd a certain amocnt of
-- mass ' to the body (M11 to body ' 1) , and an external force which is proportional. to the acceleration of the adjacent body (mass m2)*
Thus, the - acceleration _of o n'e body affects the force field on another. .This force is a strong function of the interbody. gap, j reaching large : values' forms very small gaps. This inertial coupling .is called - fluid coupling. It has an important effect in rack L dynamics. . The lateral motion - of a fuel assembly inside the l\ ' storage' location'will encounter this effect. So will the motion of a rack-adjacent toanother rack. These offects are included in the equations of' motion. The fluid coupling. is between nodes i and i* (1- = 2,3.. 5) in Figure 6.1. Furthermore, nodal masses 'i contain coupling terms which model-the effect of fluid in the gaps
- between adjacent racks.
Finally,-fluid virtual mass is included in vertical direction
. vibration' equations of the rack; virtual inertia- is added to ,the governing equations: corresponding to' rotational degrees of freedom,
'such as q4, q5, 96r 911', etc.
6.2.4. Damping. In reality, damping of the_ rack motion arises from material
. hysteresis - (material damping) , rela't ive intercomponent motion in structures (structural damping), and fluid drag effects (fluid
~
-damping)'. The fluid damping acts'on the i and i* nodal masses. In 1
6-7 .
. .. ,. . . ~ . _ - _ . _ _ . . _ . . . . _ . . _ _ . _ . _ _ _ _ _ _ . _ _ . _ _ _ _ . . _ _ . _ _ _ _ . _ _ .
a
. ,Q
'g equation of motion for the two bodies have the form
.(mi +-M 11) X 1 - M12 X2 = applied forces on mass ml "M21 X1 + (m2 + M22) X2 = applied forces on mass m2 Mlla M12, M21, and M22 are fluid coupling coefficients
.which depend on the shape of the two bodies, their relative disposition, etc. Fritz6 gives data for Mij for various body shapes and arrangements. It is to be noted that the above equation indicates that effect of the fluid is to add a certain amount of mass to the body (M11 to body 1), and an external force which is proportional to the acceleration of the adjacent body (mass m2)-
Thus, the acceleration of one body affects the force field on another. This force is a strong function of the interbody gap, reaching large values forms very small gaps. This inertial t l coupling is called fluid coupling. It. has an important effect in rack dynamics. The lateral motion of a fuel assembly inside the [ ) l {V storage location will encounter this effect. So will the motion of a rack adjacent to another rack. These effects are included in the equations of motion. The fluid coupling is between nodes i and i* (i = 2,3.. 5) in Figure 6.1. Furthermore, nodal masses i contain coupling terms which model the effect of fluid in the gaps between adjacent racks. l Finally, fluid virtual mass is included in vertical direction vibration equations of the rack; virtual inertia is added to the governing equations corresponding to rotational degrees of freedom, such as q4, q5e 96 911, etc. l 6.2.4 Damping l
' In 1 reality , damping of the rack motion arises from material hysteresis .(material damping), relative intercomponent motion in structures - (structural damping), and fluid drag effects (fluid damping). The fluid damping acts on the i and i* nodal masses. In 6-7
.,, ,~~n , ,, < -,,-.-~--+,e, ,a w--n~ - -
w ~v e,,-an ,-~----,-w ---e-w4- a. - --
i the analysis, a maximum. of 4% structural damping is imposed on
% eJaments of the rack structure during SSE seismic simulations.
b .4 +ual structural damping values used in the analysis are provided in Table 6.4. This is in accordance with the FSAR and NRC guidelines 7 Material and fluid damping are conservatively neglected. 6.2.5 Impact
- w" ThvMuel assentny w w. i
- will impact , the corresponding structural mass node i. To simulate this impact, 4 impact springs around each fuel assembly node are provided (see Figure 6.2). The fluid dampers are also provided in' parallel with the springs. The spring constant of the impact springs is assumed equal to the local ;
stiffness of the vertical panel computed by evaluating the peak ! deflection of a six inch diameter circular plate subject to a l specified uniform pressure, and built in around the edge. The spring constant calculated in this manner should provide an upper r
.cabound on the-local stiffnesses of the structure.
6.2.6 Assembly of the Dynamic Model l The' dynamic model of the rack, rack base plus supports, and internal fuel assemblies, is modeled for the general three dimensional (3-D) motion simulation, by five lumped masses and inertial nodes for the rack, base, and supports, and by five lumped l masses for the - assemblage of fuel assemblies. To simulate the connectivity and the elasticity of the configuration, a total of 18 linear spring dampers, 20 nonlinear gap elements, and 16 nonlinear
]
friction elements are used. A summary of spring-damper, gap, and friction elements with their connectivity and purpose is presented in Table 6.2. l If we restrict the simulation model td two dimensions (one n horizontal motion plus vertical motion, for example) for .the
~Q pur' poses of model' clarification only, then a descriptive model of , the simulated structure which includes all necessary spring, gap, 6-8 - - - . -. - - . . - - - .
and - frictioni elements 'is shown in Figure 6.3. The beam springs, K, Oa 3 KB at each level, which represent 'a rack segment treated as structural beam,4 are located in Table 6.2 as linear springc 2, 3, 6 ', _ 7, 10, 14 , . and '15. The extensional spring, Ke, which i simulates the lowest elastic motion of the rack in extension l relative to the rack base, is given by linear spring 18 in Table
~6.2. The remaining spring-dampers _ either have zero coefficients '
. (fluid. damping ~ is neglected),- or do not enter into. the two-dimensional (2-D) motion shown in Figure 6.3. The rack mass and inertia, _ active in rack bending, is apportioned to the five levels of rack mass;- the rack mass active for vertical motions is apportioned to locations 1 and 5 in the ratio 2 to 1. The mass and inertia of the rack base and the support legs is concentrated at node 1.
The ' impacts between fuel assemblies and rack show up in the gap elements, having local stiffness K I, in Figure 6.3. In Table 6.2, these elements are gap elements 3, 4, 7, 8, 15 ,. 16, 19 and
- 20. The support leg spring rates K6 are modelled by elements- 9 and 10. in Table 6.2 for the 2-D case. Note that the local elasticity of the concrete floor is -included ' in . K6 To simulate sliding. potential, friction _ elements 2 plus 8 and 4 plus 6 (Table
!- 6.2) are shown in Figure 6.3. The local spring rates Kg reflect the' lateral elasticity of the support legs. Finally, the-support j rotational friction' springs . KR, reflect the rotational elasticity
.of the foundation. The nonlinearity of these springs (friction l-elements 9 plus 15 and 11 plus 13 in Table'6.2) reflects the. edging limitation imposed on the base of the rack support legs.
For the 3-D simulation, carried out in detail for this l. analysis, additional springs and support elements (listed in Table 6.2), 'are included in the model. Coupling between the -two horizontal seismic motions is provided by the offset ~of the fuel assembly group centroid which causes the rotation of the entire g rack. The potential' exists for the assemblage to be supported on 1 0 l 5 6-9
. ,u. _ _ . _ _ _ . _ _ ._ . _ _ _ _ _ . ., _._ __ _ _. _ _ _ __ _ _ ._. _.
3. Table 16.2 Numbering-System for Springs, Gap Elements, Friction Elements , I.- Spring Dampers (18 total) Number Node Location Description 1- 1-2 X-2 rack shear spring 2- 1-2 Y-2--rack shear 3 1-2 Y-2 rack bcnding spring ll 1-2 X-2 rack bending 5 2-3 X-2 rack shear 6 2-3 Y-2 7 2-3 Y-2 l ra'ck bending l 8- 2-3' -X-2 l 9 3 X-2 7 . rack shear L 10 3-4 Y- 2 : i- 11' 3-4 Y-2 rack bending 12 3-4 X-2 13- 4-5 .X-2 rack shear 14 4-5 Y-2 15 4-5 Y-2 l rack bending l 16 4-5 X-2 l 17 1 Rack torsion spring , l-l 18 5 -z rack extensional spring
't 6-10
- l. .
Table 6.2-(continued) O II. Nonlinear Springs (Gap Elements) (20 total) Number Node Location Description
'l 2,2* 1C rack / fuel assembly impact spring 2' 2,2* X rack / fuel assembly impact 3 ,
2,2* Y rack / fuel- assembly . impact ll 2,2* Y~ rack / fuel assembly impact
'5 3,3* X rack / fuel assembly impact 6- 3,3* X rack / fuel assembly impact 7- 3,3* Y rack / fuel assembly impact
~8 3,3* Y rack / fuel assemlby impact 9 Support S1 2 compression spring 10 Support S2 2 compression spring 11 Support S3 2 compression spring 12 Support S4 2 compression spring 13 4,4* . X rack / fuel assembly impact spring 14 4,4* X rack / fuel assembly impact spring
~15. 4 ,'4
- Y rack / fuel assembly. impact spring 16 4,4* Y rack / fuel assembly impact spring 17 5,5* X rack / fuel assembly impact spring 181 5,5* X rack / fuel assembly impact spring
- 19. :5,5* Y. rack / fuel assembly impact spring
( 20 5,5* Y rack / fuel assembly impact spring
)
III. Friction Elements (16 total)
' Number Node Location Description 1- Support Sl- X direction support friction
'2 Support S1 Y direction friction .
.3 Support S2 -X direction friction 4 Support S2 Y' direction friction 5 Support S3 1C direction friction 15 SupportLS3 Y direction friction 7- Support S4 X direction friction !
8 Support S4 Y direction friction 9 S1 X Floor Moment 10 Sl. Y Floor Moment 11- S2 X Floor Moment 12 S2 Y. Floor Moment 13 S3- X Floor Moment 14- S3 Y Floor Moment 15 S4 X Floor Moment 16 S4 Y Floor Moment r (- T l 6-11 i
. -. . .. . - - - -.--. . .. .- --..-....-. . - . ..-- .- - - - . . . . ~ - . . . - . - . - . - .
to 4 rack supports during any- instant of a complex 3-D seismic g : event. All of these potential events may be simulated during a 3-D fy) motion and 'have been observed in the results. A brief description of the elastostatic model now follows.
'This detailed model is used to obtain overall beam stiffness formulae for the rack dynamic model, and to determine detailed stress distributions in the rack from a knowledge of the results of
~
the time history analysis. 6.3 Stress Analysis 6.3.1 Stiffness Characteristics: The fuel rack is a multicell, folded-plate structure which- has . what is colloquially called a " honey-comb"
. configuration. This type of construction is very similar to the so-called " stressed-skin" construction of ribs, spars, and cover plates which -are widely used in aircraft construction. Techniques developed in the field of aircraft structural analysis are utilized herein to find the stresses and deformations in such structures.
These methods have been thoroughly tested and their reliability has been documented in a number of publications.8-12 Figure 6.4 shows two cross-sections of the fuel rack which is modeled as a' rectangular network of plates interconnected along j nodal lines shown as points in Figure 6.1. An arbitrary load with > components Fxi, Fyi, Fgi acts at an arbitrary elevation on one of the nodal _ lines. We find the displacements and stresses due to such a typical load according to the stressed-skin model as follows. The torsional deformations are solved for by using the classical theory of torsion for multicelled, thin-walled, cross sections.13 i O v , 6-12 i
_ _ ~. - .__ - __ -
.,, The bending deformation is found by using the theory of shear h flowl2 wherein all axial stresses are carried by the effective flanges' (or stringers) formed by the intersections of the plates and all transverse shears are carried by the plates modeled as shear panels.
. 'l From a knowledge of the shear flows, the bending and torsional i
. deformations, it is possible to provide a set of influence i
functions or the fol19 wing section properties for the fuel rack as a whole: (EI)eq = Bending rigidity (in two places) (GJ),q = Torsional rigidity (AE)egs = Extensional rigidity ks = Shear deformation coefficient ! Such properties are used for the dynamic analysis of seismic (] loads and serve to establish values for the spring rates of the V elastic beam elements representing each rack section. 6.3.2 Combined Stresses and Corner Displacements The cross-sectional properties and the Timoshenko shear correction factor calculated in the previous section are fed into a dynamic analysis of the system shown in Figure 6.5, with a specified ground motion simulating earthquake loading. From the dynamic analysis, the stress resultants (Fx, Fy, F 2, Mx, i lM, y Mz) act as shown in Figure 6.6 are computed for a large number of times t = at, 2At, etc., at a selected number of cross sections. The displacements (Ux, U, y Uz) at selected nodal points on the z axis are also provided by the dynamic analysis as well as' the rotations (e, x e, y .62) of the cross sections at l the nodes. i V{ l I i 6-13 l i
(N Figure 6.7 shows a typical subdivision of the structure into elements, nodes, and sections. The stresses are calculated at all sections and the displacements at all four corners of the racks are calculated at these' elevations. Since the axial stress varies linearly over the cross section and achieves ~ its extreme values at one of the four corners of the rack, .the shear stresses due to torsional loads (Mz) achieve their extreme values near the middle of each side. The shear stresses due to lateral forces (Fx, F) y will achieve their extreme values at the center of the cross section or at the middle of each side. , Thus, candidates for the most critical point on any section will be the points labelled 1 through 9.in Figure 6.8. The expression for the combined stress and ' kinematic displacement for each' of these points is written out. Similarly,' the stresses in the support legs _are evaluated. n Joseph ; Q A validated Oat Corporation proprietary program "EGELAST"i computes the stresses at the candidate points computer at each level. .It sorts out the most stressed location in space as well as time. The highest stress and maximum kinematic displacements are thus readily found. 6.4 Time Integration of the Equations of Motion Having assembled the structural model, the dynamic equations of motion corresponding to each degree of freedom can be written by using Newton's second law of motion; or using Lagrange's equation. For example, the motion of node 2 in y-direction (governed by the generalized coordinate pg) is written as follows: t This. code has been previously utilized in licensing of similar racks for Fermi II (Docket No. 50-341), Quad cities I and II (Docket Nos. 50-254 and 265), and Rancho Seco (Docket No. (r) v 50-312). 6-14 1
The inertial mass.is: m22 + A211 + B211
' where m22 is-the mass of node 2 for_y-directional motion.
A211 is the fluid coupling mass due to interaction with node 2*, and-B211 is the fluid coupling mass due to interaction of node 2 with the reference. frame (interaction between adjacent racks). Hence, Newton's law gives ('m22 + A211 + B211) ..P9 + A212 910'+ B212 u = 09 where Og represents.all the beam spring and' damper forces-on node 2, and l A212 is the cross term . fluid coupling effect of ~ node 2*;- 9 B212 is the cross term fluid coupling effect of the adjacent
-(G racks,'and u.r'epresents ground motion.
Let 99 " P9 - u
'910 " P10 - u l .That is, .q9.is the relative displacement of node 2 in x-direction l with ~ respect _ to the - ground. Substituting in the above equation, and rearranging, we have:
-(m22 + A211'+ B211 ) ..99 + A212 910 = 09 - (m22 +
A211 + B211 + A212 + B212) 5 fh
~
6-15
~
- A similar equation for each one of'the 32 degrees of freedom can be written.- The system of equations can be represented in matrix notation ass-(M) {'q'} = (Q) + {G}
where the vector' (Q) is a ' function of nodal displacements and velocities, and- {G} depends on the coupling inertia and the ground acceleration.. .Premultiplying above equation by - (M ]-1 renders the resulting equations uncoupled in mass.
~~
.We haver
{'q'} = - (M ]-1 (Q)4-[M]-1 {G} The generalised force - 0 9 , which contains the effects of all spring elements acting on node 2 in the " direction" of
~
coordinate' q9. (the relative displacement of node 2 in the y direction), can easily be obtained from a free body analysis of node 2. For example, in the 2-D model shown in ' Figure 6.3,
^ contributions to Og are obtained from the two shear . springs of
' the rack structure, and- the two' impact springs which couple node 2*
and node 2. Since eachl of these four spring elements contain couplings with other compo.nent deformations through the spring
- force-deformation re'lations, considerable static coupling of the complete set of equations results. The level of static coupling of
- f. the equations further' increases when 3-D motions are' considered due to the inclusion of rack torsion 'and general fuel assembly group centroid-effect.
For example, referring - to Figure 6.3, and Table 6.1, a 2-D ! simulation introduces st'atic coupling between coordinates 2,9 and
. 15 in the expression for 09 ; this coupling comes from the shear springs simulating .the ' rack elasticity which have constitutive
. relations of the form F = Ks (99 - 92) ,K s (915 - 99) . Further, the
~
. impact springs introduce two additional forces having constitutive equations of the form F = Kr (q9 - gl0) . Of course, at any . instant, . these ~ forces may be zero if the local gap is open.
The. local gap depends on the current value of q9 - q10 .
. n '
+ 4 6-16
= , y v v , e.-- y,,v+,,, .-www.r,.-w,<v.,,w -,-.. ---,.~~--.---m-..--.=,-,.-.-.-----,.v-'
It.should be noted that - in the numerical simulations run to g verify structural integrity during a seismic event all elements of b the fuel assemblies are assumed to move in phase. This will provide maximum impact force level, and hence indoce additional conservatism in the time history analysis. This. equation set is mass uncoupled, displacement coupled, and is ideally suited .for numerical solution using the central
]
difference ' scheme. The computer. program named "DYNAHIS"i, : developed by General Electric Company, performs this task in an l efficient manner. . j l Having determined the internal forces as a function of time, i c l the computer' program "EGELAST" computes the detailed stress and displacement ' fields for the rack structure as described in the preceding section. , , 6.5 Structural Acceptance. Criteria
-There are two sets of criteria to be satisfied by the rack v modules:
(a) Kinematic Criterion: This criterion seeks to ensure that adjacent recks will not impact during SSE l. (condition E'14) assuming the lower bound value of the pool floor surf ace friction coefficient. It is further L required that the factors of safety against tiltingl5 are met (1.5 for OBE, 1.1 for SSE). I I .(b) Stress Limits (1) The stress limits of the ASME Code, Section III, Subsection NF, 1980 Edition up to and including Winter l' 1982 addenda were chosen to be met, since N l ! t This code has ' been previously utilized in licensing of similar I L racks for Fermi II (Docket No. 50-341), Quad Cities I and II l [ (Docket Nos. 50-254 and 265), and Rancho Seco (Docket No. 50-312). ! I . l d' ' 6-17 l l l l l
" thio Codo providos.the most conoistent sat of limits for.
various stress types, and .various loading conditions. 7
- The following '_ loading casesl4 have been analyzed.
SRP Designation ASM'E Designation-
. (i)~ D +.L ' Level A (normal condition)
(ii) D;+ L'+ E Level B (upset condition)
.(iii) D+L+To No ASME designatio0 Primary membrane
. plus bending stress required to be
. limited to-lesser of 2 S y* and Su*
' (iv')
D+L+T+E o -No ASME designation. Stress limit same'as (iii) above. (v)- D + L + Ta + E No ASME designation. Stress limit same as above.
.-(vi) _ D + L + Ta + E' Level D (faulted condition)
-where D= ' Dead weight induced stresses L.= Liv'e load induced stresses; in this case stresses are
'f a G developed during lifting. ,
The conditions Ta and To cause local thermal stresses to
'be produced. The worst-situation will be obtained when an isolated storage location;hasca fuel assembly which 'is ' generating heat at the maximum postulated rate. The surrounding storage locations are assumed. to contain :no = fuel ~. Furthermore, the loaded storage location is . assumed to have unchanneled fuel. Thus, the heated
. water makes . unobstructed contact with the ' inside of the storage walls, thereby producing : maximum possible temperature difference between.the adjacent cells. The secondary stresses thus produced are. limited Lto the body of the l rack; that is, the support legs do not experience the secondary:-(thermal) stresses.
,p *Sy sLYield stress of the material; Su: ultimate stress.
LJ
-- 6-18
-(2) Basic ' Data: The following data ' on the physical properties- of the rack material are obtained from the ASME Codes, Section III,_ appendices.
Table 6.3 Physical Property Data
- Property Young's Yield Ultimate Allowable
. . Modulus Strength Strength Stress O 2000F 82000F '02000F @ 2000F E Sy Su S
.Value 28.3 x 106 25'KSI 71 KSI 17.8 KSI psi-Section III Table Table Table Table
- p. Reference. I-6.0 I-2.2 I-3.2 I-7.2-
- NJ
- Evaluated at 2000F. This - temperature is higher than the pool water bulk temperature under any 'of ~ the loading conditions under consideration.
L L l (3.1) Normal and upset. conditions (level A or level B): (i) Allowable stress in tension on a net section =Ft =0.6 Sy or
~Ft '=(0.6) . (25000) =15000 psi Ft is equivalent to primary membrane stresses (ii) On the gross section, allowable stress in shear is Fy = 0.4 S y
, = (0.4) (25000) = 10000 psi
- 3 i
- (%) - .
L 6-19
,,--,, ,,-,- -- - ,, ---- , , --. - - , - - , - , , - - - , ,-,.---.,,,c -
i I (iii)' Allowable: stress in compression, Fa
.(.
(1 - (S) r 2Cc 2J3 y F = [(5)'+ 3 ['3_(S)[8Cc] r - [(r
) 8C c) where 1/2 - '
Cc " ( )
. S l Y \
Substituting numbers, we obtain, for both i support leg and " honey-comb" region: - l Fa = 15000 psi l 1 (iv) Maximum bending _ stress at .the outermost l fiber due to flexure about one plane of ! symmetry: I Pb = 0.60'Sy = 15000 psi
-(v) Combined flexure and compression:
h ,C,x bx f
, C,yfby 4 y F, DF x bx DF y by where fa: Direct compressive stress in the section.
fbx: Maximum flexural stress along x-axis fby: Maximum flexural stress along y-axis Cmx = Cmy = 0.85 s v
. 20 u -. _- .. _. . . _ . . - . . _ , _ . . _ _ _ . _ . _ _ , . _ . . . , _ . _ _ - . _ _ _ _ _ . _ _ _ . _ _ _ _ . . _ _ . . . _ _ . _ _ . _ _ _ . _ _ _
=;_ _.=_.x __ f" , ; .D x =1-
- - 'F',g.
I D =1- f* Y pi ey where - 2
.p. , 12 E ex 2 l kl ,
b) l-23 (
*b l-(vi) Combined flexure and compression (or l tension) f I E a , bx + by
< l.0 0.6 S y F F bx by I- .
b The above requirement should be met for both l J direct tension'or compression case. (3.2) Faulted Co'ndition: F-1370 (Section III, - Appendix F), states that the limits for the faulted condition are 1.2 (Sy /Ft) times the corresponding limits for normal condition. Thus, the multiplication l- factor is Factor = (1.2) .5000 = 2.0 l 15000 l --. . O l 6-21 l-I t . I
, . . . - , , . . , - --..--.,-,--,-,n. - - - - - . - - . - , , - - . . , . - - , , . . . - . . - - - , , - - - - ~ . - - - - - - - - - - - - - - - - - - - , , - - - - - - - - - - - -
. . . ._ - . - _ _ - :=_ - - - -
= (3.3) Thermal Stresses: l l
There are no . stress limits for. thermal I (self-limiting) stresses in Class 3-NF Structures ' for linear-type supports. However, . thei range of primary and secondary stress
- intensity . is ~ required- to be limited to 3'Sm in the manner'. of Class 1 components; Sm is the allowable stress s intensity of the rack material I~ at the maximum operating temperature. ,
e i a . O l !~ . h. I' i l O . 6-22 L- '
+ M Y
a . 6.6'- -RESULTS A 4 bl- I The11nput - time ' history accelerations for seismic motion were '
~
developed ~by::0at -from the. Response Spectra .provided by GPU Nuclear. Typical plotscof the time history motions, utilized -in :the i present' analysisiare:shown-in in Figures 6.9.- 6.10. These plots
. correspond'to the Safe Shutdown Earthquake'(SSE).
'Since ~ there are several rack module configurations -(Figure
- 2.1)~ it iwas- decided to make an . exhaustive analysis of one rack type.- .
We ' note that rack E is an above-average, size module,,and is
. elevated 1111/2" from the pool floor'versus 6" for other modules (except type K). Therefore, rack E will, produce above-average floor reaction and -support . stress. levels. Hence rack E is chosen for
!- - performing ' extensive ~ analyses. Appropriate simulations ,are also carriediout forLother limiting rack geometries.(e.g. tipping study for ? rack ' with -low cross section to - height aspect ratio, stress
' evaluations for the heaviest . module, etc.). 'To determine 'the .;
O . magnitude _of. structural _ dampers, free lateral vibration plots of the j top. of rack E1 (in X and Y directions) for fully- loaded and emptylconditions were developed. Thefdominant natural frequency of vibration thusL evaluated
- . enables computations of the linear
. structural dampers. The : maximum percentage structural damping for
~
'SSE condition is assumed. to be '4% .and- modifications- to the st'if fness~ matrix - to incorporate 1 damping ' = is based on the dominant ifrequency~of 10' cps. Having ' determined' the damper characteristic
; data, theidynamic analysis of the rack module is performed using
. the computer; program DYNAHIS.- Two. components of the SSE horizontal acceleration are applied in two orthogonal directions concurrently
-with the : vertical seismic acceleration.
Table 6.4 lists all. cases considered for rack type E. Case 1
- (Fulli ra'ck, COF = .8, 4% damping) _ studies are also performed for rack . types F, K, and H.- Selected runs have also been made for various cases using the different- rack geI>me tries . Table 6.5
~
6-23
g n.
- illustrates selected computer outputs from among the complete sets
~
of simulations. --Table.6.5 gives the maximum excursions of'the rack n) ( in=x,_y directions. _ Note ~ that location 1 in Table 6.5 represents the top of the. rack;- while location 5' in Table 6.5 represents the base' of the rack.- The Tdisplacement. outputs are . for the critical corner pointi at _each level.- Table 6.5' also gives the maximum
, . values' . of stress factors Ri - ( l' = 1,2,3,4,5,6). The values-given in both - tables : are- maximum values in time and_ space (all sections e
of the rack). ,The time values should be divided by 100 to obtain actual elapsed time. The various stress factors ane listed below
, for convenience.
R:1 Ratio of tensile stress on a net section to its allowable OBE value
. R:2 Ratio - of maximum gross shear on a net section to its
. allowable OBE value R 3:. Ratio of maximum bending stress in one plane (x-y) to' its' allowable OBE value for the section p '
R:4 Ratio of . maximum bending - stress in one plane (y-z) to V its allowable value in OBE R:5 Combined flexure and compressive factor R:6 Combined flexure and tension.(or compression) factor
- The - . allowable value of Ri (i
= ,1,2,3,4,5,6) is l for OBE condition, and is 2'for.SSE condition'(see Section 6.5).
-The displacement - and stress tables given~ herein are for the SSE. condition.- If.necessary,_OBE studies are run to qualify the rack when the SSE simulation with . damping of _4 % does not yield stress factors ~that would clearly indicate a safe margin under.OBE conditions. ' Seismic simulations for the tipping conditions are carried.Jout 'by increasing the horizontal SSE accelerations by 50'415 These calculations indicate .that the rack remains stable, l - and ' the gross ' movement remains within the limit of small motion theory.- ,
, O 4 6-24 e .p..,,- - r. +w--y w- gu- w- vc er-,=cmv.. g e .p gw-- y-,wxe --ewgr.+rw-**-e"N8 v ?-t **T--P--'-NMm-*E**'-"*N4'*-""**"8 '#" " ' - " ' ' ' ' - - " " " " - " " - * - - ^ ~ - - ' - ~
.For. racki type- E (312 cells), the - averaged gap between adjacent structures ~is 3.41" in the x direction and 2.0" in the y O
v direction.- The - maximum displacement in either direction for the
- full'; rack ~ presented' is .125" which is less than 50% of the-
- spacing. Note that the direction along . the smallest side is the
-local x direction' of .the' rack. For - all runs, it is shown that inter-rack impacts will not occur.
~
For-rack K (248 cells), the average gap is 1.5" in the local x direction and 3.41" in the y direction. The critical deflection for the full' rai:k with .8 coefficient of friction is .079". - In all
-other runs for this rack ~ configuration, it is shown that inter-rack impacts do notLoccur.
For rack' type F (315 cells), the . averaged gap spacings are 5.375" and. 1.75" in the local x and y directions, respectively. For the- full rack with .8 coef ficient of friction, the maximum displacement of the rack top is 1.298" in the x direction. 'In all of the analyses, the -local' x , direction is taken along the smaller side of~the rack forca non-square configuration.. For rack type H, the averaged gap apacings are 2.82" and 2356",respectively. The ' maximuni local deflection is 1.46" for the full' rack. Inter-rack impact will not occur. Table 6.5 also shows the maximum values of the. stress factors
, obtained -.for - typical full racks. Of all cases simulated, the full J
' rack - _with . 8 coef ficient of friction gives the highest stress l factors.
t. Seismic simulations for the tipping ' conditions are carried out_by increasing theshorizontal SSE accelerations by 50% 15 The calculations' indicate that the racks remain stable, and that the gross movement . remains within the limit of small motion theory.
= Thus, the rack modules are seen to satisfy' both kinematic and
. stress' criteria with large margins of safety.
e. h
^
6-25
l l I Table 6.4
' Cases Considered (All SSE.' Events)
Case Number Description 1 Full rack, Damping = 4%- p= .8 12 Tipping Analysis-(1.5 SSE horizontal quake)
~4% Damping, p = .8 3 Full rack, p = .2 4% Damping
-4 Half load, Diag. Fill, y = .8 4% Damping 5' Half load, Diag. Fill, p =.2
'4% Damping -
6 Half load, Positive.X Quadrant U= .8 4% Damping. Ov_ = 7 Empty racks, y .8 4% Damping 8 Half load, Positive Y Quadrant, p = .8, 4% Damping 9 Half load, Positive _X Quadrant y= .2 4% Damping 10 Half load, Positive Y Quadrant, p= .2 - 14% Damping _ 6-26
REP'ERENCES TO SECTION 6
. [~N ; 1 .USERC Regulatory Guide 1.29, Seismic-Design Classification," ' q)
_,, .. -Rev. 3, 1978. 2.- " Friction Coefficients of Water Lubricated Stainless Steels
.for a Spent Fuel Rack Facility," by Prof. Ernest Rabinowicz, MIT, a report for Boston Edison Company, 1976.
.3. U.S. Nuclear 1 Regulatory Commission, Regulatory Guide 1.92,
" Combining Modal Responses-and Spatial Components in Seismic Response Analysis," Rev. 1, February 1976.
4[ "The Component Element Method .in Dynamics with Application to Eartuquake and Vehicle Engineering" by S. Levy and J.P.D.
-Wilkinson, McGraw Hill, 1976.
- 5. " Dynamics of-Structures" R.W. Clough & J. Penzien, McGraw Hill (1975).
- 6. R.J. Fritz, "The Effects of Liquids on the Dynamic Motions of Immersed Solids,"-Journal of Engineering for Industry,.
Trans. of the ASME, February-1972, pp. 167-172.
- 7. USNRC Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants,.1973.
') . -u e. J.T. Oden, " Mechanics of Elastic Structures," McGraw Hill,-
s/ .N.Y.,'1967.
- 9. R.M. Rivello, " Theory and Analysis of Flight Structures,"
McGraw-Hill, N.Y.,.1969.
- 10. M.F. Rubinstein, " Matrix. Computer Analysis of Structures,"
Prentice-Hall,'Englewood Cliffs, N.J., 1966.
- 11. J.S. Przemienicki, " Theory of Matrix Structural Analysis,"
McGraw-Hill, N.Y., 1966.
- 12. P. Kuhn, " Stresses in Aircraft and Shell Structures,"
McGraw-Hill, N.Y., 1956.
- 13 . S.P..Timoshenko and J.N. Goodier, " Theory of Elasticity,"
McGraw-Hill, N.Y., 1970, Chapter 10.
- 14. U.S.-Nuclear Regulatory Commission, Standard Review Plan, NUREG-0800 (1981).
-15.. ~U.S. Nuclear Regulatory Commission, Standard Review Plan, Section 3.8.5, Rev. 1, 1981.
' 16 . U.S. Nuclear Regulatory Commission, Regulatory Guide 1.124, (A-'/)' " Design Limits and Loading Cembinations for Class 1 Linear-Type Component Supports, November 1976.
6-27
---f e.- ------y 4 - -
942 [ ' FLOORILDT3~~'2
; 3 ~3 ) fM 0 5' ~~
~ ~ ~ ~ ~
943 ,J.IEPACT FORCES ( , ,) V T'4 4 ( GcP t.t#CFnr MAX,FO!CE 11%E-
#45 .1 1.995D+05 4.6530+00 i'
- 95) 2 2.0421:+ 05 7 .14 5 r,,00 s Tabie 6.5 a 947 3 2.I121+05 8, . 6 6 o 0 + 0 0 i +
945 e- 1.991D+0S 6 .4 Vll'+0p 9a 5 2. i O 2 i,+ o s s.9 60+00 950 6 2.2 dis +05 7.148D+00 F i i e- DGPU10* Moduie E* Cst 7 7.O S'JD+ u 4 5.200D+00 . 9$2 > at . 5 69 tse l'4 7.1010+00 COef= .8, te 53 9 l . 4 * /5 84 + 0 5 3.394D+n0 FuII Rack 9$4 to 1. 4'/ t see u h 4.414p+0u 455 11 I.706D+05 4.9$4h+00 9%h 12 1.9r10Defs5 S.479(s+00 937 13 2.1410+05 A.94n0+00
- 9,8 14 2.1$1D+05 6. H / S is+ 00 939 15 9.6540904 S.193D+0ft 960 tb 1.0066+05 7.0456+00 941 17 1.512n+05 6.6783*oc 902 1 is t.Sh58,+0% 1. 0 7 2 D+ fs t
- 94) 19 1 .' it 44ti+ dh 6.6T6iiWiG .
984 20 t.353b+45 4.!590+00 963 FLUID UAMPER MAX. FORCE
- 964 1 0. 0 8:4 D + 0 0 967
- 2 0.000D+00 0 9ss 3 0.opup+00
, l 9 "A 4 0.000D+00 , M 970 5 0.0000+00 CD 971- 6 0. te s!u ti,+ n o 972 7 0.000e+oo 973 e 6.000b+00 974 1 504. JU9 19. 199]t_II54 AM PMDGRAM EGELAST = 975 , 976 GPil E,C0ia.ft,SDs4% E10HZ,tsATA DGPb10,16X20 F u t.ts SSE F1hAL 4U45 971 470 ItaFur PAWA*FTEFS 9s9 no. OF 60ngs (s.unnon) = 5
- 9,0 es o , o r . r.s.E r e.. Ts t n u.s.1_t,J s 12 941 ppt 8 't uPrIri4 (10:37) s 0 942 , x-H a t.F= J DT H (A2) a 4.99tE+01 ,
9 " 3. VCA 'd.fil.LDL.I O L. m 6.234E+0t wi 4 i E f4 % . HabscLEhGr4 (kZ) a 2.I11E+01 - VoS
- 9th U S th=;55_gpd} FIDENTS Ful* kACkt 997 CFX z 5. 6 d ut.-0 3 CF Ys 5.obo6-0J CFZ z 3.H90E-03 944 CwX s 4.7706.-fib C4A y a 1.089E-04 ugg CTX a 1.673E-04 CTY a 1.R24E=04 990 0 SikkSS CS F53CIcetS FOR SUPwOPrSt 491 Cr'X7 x 6. 0 0 0 r; -'s ? C6 Y2 a 6.000F-02 CF22 s 2. 67 83 E=o 2 vf2 C3 8 t__s_g ,J e> 0 >: = 0 ) Cry 2 m o.7603-03 CT(2 3 0.000E+u0 CTy2m 0.000E+00 vo3 9 e4 4 STwESS C EFFICIErarS FON SUPP0h? h0TTOM 90s O s rp s.sh gut 6.rigJ r ers p op soPHm ts t '
v$s CF42 = 1. 3 5,.5 -o t CFY2 = t.sw r.-On CFz? = 6.0 3 0e:-02 947 CHA2
- 1.4%Ji-ut C,iY2 = 1.443F-01 CTX? z 0.000t.+00 y ut CTY2m 0.000E+00 999 : S r A IIC b rkk SS C46 t FICJ h. HTS 1900 CC X P , CC K * , CC Y e* , CC Y e' . C N S H . C YSh a 0.000F+00 0.000E+0o 0.000ec+pu 0.000E*00 tne rT L e ar ,e y r.,F 4r= 0.000E+00 0.000E+00 t o e:2 ere,rse,rvn.pn.,3 49909.0 1 ver%0 tso00ft t0900.0 ts000.0 ewooc.a wmo.O 48u00.0 tcvs .9 c r i r. n n. or unem (.next 3 = 5
f-l 'h ( \
\ l . ~
/ k h loot i's / TAbbE CF CAXIM AK imjIV ALE 47 STPESS v 1947 @ S Ff f . 1198; PO!'T CAX. SFO. OtP.ST9ESS X-9F40 STR Y-0Evn STRESXL AT.5H5; AR luo4 0 40 YLAT. SHEAR t;ET SHEAR)
*4 0. ' UD. (SFPXMX) 150) (SHX) (58Y) 1FX) 1994 (TY) (TN) 1010 1 245 9 5.847f+02 ~ 5.222E-01 -6.727E+01 8.405E*01' 1011 2 497 9 2.0m3E+02 2.oS2E+02 2.924E+02 6.670F+02 -3.230E+00 -1.lb1E+u2 -2.01hE+02 -2.921E+02 1912 3 549 9 1.0J0E+03 1.n10E+0S 3.33Si+02
-7. 4 4 5 F. + 0 0 1.6d9E+n2 -4.024E+02 -3.6o2K+02 -3.6247.+02 5.152E+02
___1033 4 54,9 3 1,13 '# E + 0 3 -4.626E+00 l 9MME+02 o -7.2}1F+02 -4.3n2E+02 1b14 5 549 -3.414E+02 0.000E+09 3 1.9177+43 -9.92nF+0ss S.101E+02 -3.059E+02 -4.302E*02 -3.414E+02 0.000F+po 1415 6 572 4 W . A f. 0 6 + u 3 -3.395t+03 3.694E+p3 ___3036
- 1. 3 0 3 5.+ o 3 2.841E+o3 -1.006E+03 1.231t+0s 7 fu3 1 1,%10L+04 -4.814h+03 -h.1hdL+u3 b. lane 403 3.93*E+03 4.op4r+03 1617 8 496 2 1.0 41 t + 0 4 - -4.550L+03 -%.p2de.+02 4.524Etut 9
-4.47AE+03 -3.44hE+0J 4. 4 9 n e;+ 0 2 2.ouSE+0J lute 549 3 1.334f+04 -5.062t+03 2.5JJF+03 -8.4 pee +ns -4.171E+o3 -3.959E+03 -1.473E+03
___1919 to 42? 1 1.e4 4 F F,+ 0 4 -7 3 4t!.F +n t 0,000F+0n 4.oonF+00 162o 493
-5.0 HSE +p3 4.091E+03 6.4h7F+03 11 2 2.7a0E+44 -1.067E*04 0.000F+no- 0.OnoE+00 1021 12 996 6 1.n70t+04
- 8 . d '8 5 E + u l -9.0496+03 1.26aE+04 1922 13
- 1. 0 *t A t + 0 4 H.000L+90 '0.04uE+00 7.811E+0J ~1.016E+03 -7.u11E+03 549 2 2,29ht+04 -1,143F+0* 0,,0 40 E + 0 0 0.000F+00 1023 1 9.420E+03 4.427E+03 -9.794E+03 COhDIT10tS aHEN X-DISPLACEFEnT OF A CuMNkk ISd A XIft A A 1024 ,
19*/S NQDE TIME W AL. COW MP C F.11HGIDAL Cp1 ks]I D_A l, TokS If W A t, 1025 NU. X-0 TSP. r-DISP. Y-DISP. AwGt,r 189 8 1 *) 01 1 2AdF-01 =1.254E-41 -1.169E-03 4. p h O r.-O d 1019 2 500 9.611E=u2 -9.601E-02 5.14SE-03 1.536E-06 1410 3 %00 6.6588-02 -e.639E-02 3.499E-43 -3.040F-06 I fdl 4 500 32 i99E-42 -3,6714-0? 1.H75E-01 -3,664h=66 1032 5 500 7.744t=03' -7.552t-03 3.092E-04 - 3.o p 0E-u 6 D 1H33 ]__Lul4 0 101s * ' D 1036 . lus7 quig!TIONS eden Y-ntSPLACEFEtt AT CDkhER ISMAXIMAX 103h 1039 NODE TIME NAX. C0hdKH CENTRO 10A1 CENTA010AL TORSIONAL ' __,1044 N0. Y-D1SP. F =f* I S P. f=OISP. A tJ GL P 1041 1 493 5.690E-02 9.227E-02 5.b50L-02 .-6.474E-06 1942 2 443 1.3knE-02 7.09pk-92 4.355E-02 -6.el4E-06 ___14*3 3 463 3.0594-02 4. 916 e. -0 2 J.u33E-0J -5.118E-06 1H44 4 4h3 1.13 J r;-0 2 J.7336-02 1.707E-02 -5.119F-u6 1945 5 1J66 4.25er-03 -2.529E-03 -2.163E-03 -4.193E-05 1046 . 1047 S TRoCTHR A1 ACCEPTANCE-ASur at - - 104A S 6.CT J n t. MuseER 1 10,4 Ria 000AT TIME $49 92a 00 HAT TIME 671 H3s 00%AT F1HE 671 1050 P4s 006Af TI.1E 245 E5m .co9A1 TIME 245 H6s 01DAT TIME 245 1051 SECTIO:4 N HE is F R 2 1952 H1z . h o re A T TIdC 549 97: .ul0AT TIME 496 R3x .01147 ?!*E 1348 1453 H4s 035Af II1E $99 F%s 021AT TIME 519 P6m 024A'r ride 518 1054 SrC1 TON hu'brk 3 1955 H!= 000AT 7 tit 5+9 R2x 015AT TIME 494 R3s 02uAT TIME 62H 1056 H4x 0354T T18E 496 kbs ,040AT F14E 497 Nhz .947A1 fl>E 497 1057 SECTION 6HMHER 4 1059 kts 001AT TIME 549 p2s .OleA1 TIME 496 R3= 0354T TI*E 564 1059 R4s 061Ar flaE 496 F5s 064AT TIME 549 N6= 075AT TI*E 549 1960 SF CT 10d 'f 0 :8HFk 5 In61 Hla ,Uu1A1 TINE 549 42n 41 HAT TTME 496 H3= 444AT Tire bn4 1062 F42 075Af TI=t 49o R$s .OnuAT T146 549 Ph= 044AT TIME 549 1063 SLCT 1*f h NUMMER 6 166a Ria .2g4 AT 11PE 344 pts .127Af TIME 572 R3m .167AT TIME 491 10ai P4= .246AT T1HE 472 H5= 51947 TIME 492 whz .571AT T14 r. 442
sL I P4 * . ( j' SECTIna CunPER- .'7 -
-( -\ ,
10$7 ' N-vu gJ27 AT 12mE 442 22n .124AT T!4E o;2 23n' .LsJa7 TI#t 404 . 11638 74a .35747 TgeE..qu2..psn 404A1-TIAE 4W3 NAs 1.CCfXf"ffEE. 443 .
.SECr!OU huMHER
'l063' 8 . #' ' '
' t 073 - bla .301AT 1THE 496 >2s 15447 TIME 496 RJs .14247 TIMC 713 * .
l'8'i l H4s .299af T!r4h 496 . pts .b9pa7-TJ4E 496 Pha . 64f5T..TIAC 496 , i l in7 2.. SECTtuM NuraEn 9.- ' 075 Ris~ ,3JfAT T1*C~S 49 M7s 34bA1 TI*E $49 43s .17547 TaxE 550 7"}1074 N4s .36047 T1:e g ge s - ~w5s 7 heat T I P r! . b 4 88 - Rhm
~
.WU/4T T iiE~Sl4 1975 SECTION kH;4 HEM in ,
107A His .17waf TIMF 3 4fe 'R?s 1DIAT TJ9E 572 R iz .nnua r TIME ~0 1077 k4s 0004t ifrF 4 Ass .17ddT 'r t M F. 340 M6s .37947_f1*E . 340.
, 107d SFCFlub huPdFA 11 g.
1074 227e1~T!*t P2s .146AT TIME . 4W2 R3s Ris II7 th0 P'4 s ,~66'Eif ijlt 442
'U' 4 5 s 22/A1 T!*E , 4W2 h6s 006AT Tine
.227'A Fri.CE~ 4W)
O 1011 SEC110h huwbEN 12 ___.lj!D2 322AT TIwM , 496 . N1s , bis .2J4AT TIME 446 M2s 000AT'TtHE o 1493 M4s 000AT TIME D Hbs 214AT TIME 496 M6s _.7344fTf!ME, 496 1044 SECT 10h h0MRER '13 . 10,5 Ris .2) RAT TIME 549 #7s .14747-714E 549 . k3s '.000AT TIME 0 1065 k4s 000AT T14F. O P5s .239AT TIME S49 < Nhz ~ : .238AT TIME . 549
-- ' ~
e d 4 } \ a 'N j ' C4 33 . 4 9 A i i J l l t
,/' ~'\
ecr.rtnn> 2 5.259:et5_
*. s y , t e p.%
Cav FIA 'tl7 pCF S M43.POHCE TIME ~ (( )' a 2.st20+n5 i.2 e ^ 2 2.52su+05 5.,94n+0i. s30 00 -
.:1r a b i e - 6 . 5 b 3 1.Ph3D+45 1.105p+01 4 2.31 o+e5 5.e5:o+00 s i..si...n5 s.5+vo+no FiIe DGPU51 ha o d u i e F.
6 2. 0 31 re + 0 5 5.711D*00 $ .. 7 6.02no+o. - 9 5.4500+44 t.lo5c+0 5.NA3p+ou s -Coef : .8, Fu11 Rack 9 2, '* a
- D + 0 5 1.J95b+ul 10 2. 2 ',0 f, + 4 5 1.395D+01 il 2.l+90+H5 1.J550+01 12 2.E245555 f.165fi+ 5 - .. . jx. ,,.s ,
13 3.76/D+45 5.539n+03 . l ?; .t;; 14 1. 6 9 4_ D + 0 5 5.72586400 1% ' c ' ' .6 . 15 M.053L+44 5.542h+00 7.545D+44 16 6.55=n+00 17 2.1700+05 5.9350+00 14 3.772D+05 1. 2 7 41s + 01 , s 19 1.179fe+0S 2.346n+00 20 t_ ,9 0 P D + 0 5 1.972n+01 FLUID DANPFP PAF. theCF CD t 0.0000+90 1 2 0.0400+00 CO 3 0.0000+'10 . -s 4 0.000D+Ao $ -4 ' 5 0.fs00n+00 ? s 6 U . 0 0 0:s+ 00 7 0.000D+00 8 0.000D+00 . 1 THu, Jul. 14, 19 8 )' , 4:20 PN PROGRAM EGELAST CPU F,C0Fs.F,SDz4% at0HZ_pATA=DCPU51,15X21 f FULL RACK 4' low FEET IhPUT PANA>ETFRS h9. CF s9 PES ( se d w h00 ) a 5 h0 De e t,F
- E N T S ( N U M El.) s 12 P4IST OPTiuv (IOPT) = 0 4-patForn1H (a2) a 4.678E*01 . -
v-naLF.t01% TE2) = 6.514bT61
~ ~ ~
El.E w . HALFLEuGTH (R2) = 7.113E+03 0 $1 r, f 5 9 CrIFFYP1tWis F(iF~lis U: CFx 5.79,F-u3 rFYa 5.760E-03 CF2 z 1.920E-03 CMA a H.490E-0% CHY = 1.16ht-04 CTX w 1.616E-04 CTY s 1.92*E-04 0 STkESS COEFFICIENTS FOR SupP0kTS: rFx2 = 6.00"E-02 CFY2 a 6.000E-02 CF72 a 2.h67E-02 C 32 6.1F6T;63 CFV2 E'E.726F-oJ CTx2 = 9.000E+00. cTTz= 0. u 0 0 t:+ 0 0 , 1 STkESS COFFf1CtFNIS FOw SurPOET aOTTOM 0 S'Twed.CCfW.FFICIU'5 T FOh~50l+GwiS: CFX2 = 1.s56E-01 CFY2 s 1.356E-01 CFZ2 a 6.270E-02 CF22 = 1.4hir-4 CwY2 x 1.443E-01 CTx2 = 0.000E+00 CTY2= 0.uo0E+00 0 STA1IC STbESS C0fFFICIF4TS CC#P,CC3*,CCYP,CC)m,CX5d.CYS4m 0.000F+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 ETT, FAT,Fvt,tniz 15000.0 a sinfo.o 1C690.0 150oo.o FT8,FaP.FWb,FkBs en000.0 43000.0 32000.0 48000.0 S E CT i nt, hn. OF 9001 ( .1k noT ) = 5
,.n.. . . , - - --....a. . re .~.vm . ,. . . ., -
n O 3 0 4,ECT.
\~s -
- 71*t. Poluf . MIXDE4 TAkl.E OF-#417N42 EQUIV4bE4k wf0F23 n1F.(50)
STWss n - =Ur5TF O O' NC. MO. 40. (&EFX=X) - (SME) ' YT18 F'Ifr5TWETTUAT. (SdY).
.U EAntTA T 3 M AF3tET* 5H E'A n s~
(74)-- (TY) (TM)' - 1 %2 9 4.753F.*07 -7.4505-01 7.g7pE+an t . 7 Mu G7 4. 078iF.4C-%'O W81r FtJ1 "" 4. IMUD s
.2 1393 8 1.16 0 r.+ 01 -1.22be*ng 1.054R+0! 1.nJ7r+02 5.7ADt02 -9.9949+#9 5.1995+02 3 1393 > 1.314E+03 =t.94th+01 2 . 4 4 2 5.+ 01 A . P '8 7 E + 02 6.%43r.*b2 1.27 7t+0p -
4.
~
6.567E*02. I s 5 88 - 3 1.T)@+4 3 . - t . iiiii.+ 01 2. TTiT:+ 17 -173 W a3 -G 51aR W2 7 2 39ar402 0.000E+0v 5 1354- -3 -1.485E+03 -1.90JE+01 3.StdE+02 -1.514E+03 -6.S49E+02 -2.39aF+02
'6 1240-0.000E+00 1 9.71RF*01 -5.249E*01 1 .7 5teR+o t 2.23RE+pi 3.110s+03 .-2.S99E+03 '-3.935E+03 7 3649 2 1. friJ 6. + 4 4 -4. A fiDW78 3 -3. p lT&T+Ts 3 . 2. 41F +3 3 UtMMM3 4355ET01 -3 375Ef01 9 1445 1' 8.972F+D3 - 3. 62 2 F + te ) -2.$11E+n)J -1.375E+01 .-l.460E+04 3.71aE+o3 3.7w7F*o3' .
9 135v s 1.2%eE+04 .-6.427F.+03 1.52tE+o) -J.789E+03 - 4. 4 6 5 6:+ 0 3 -2.250E+03 ~4.466E+03 10 1290 1 2. fi 3 F. + b 4 .-l.27tE+44 o W icc+oo o.099E+4n -7.4'sl NoJ 3. n 5 TEM 1 v.42 W To 11 1969 -2 2.k12E+04 -1.130E+04 0.000E+00 ;0.000S+00 -8.344F+03 -1.007E+n4 1.2pHR+os 12 1945 1 21 004E+o4
~ -4.515E+01 0.0005.+49 0.000F+00 4.410s+03- -R.402E+03 --9.072E+03 33 1154 2 2.7 3IspiO 4 - i . 5YTT.+ 0 4 o.000E+00 0.an"F+0o
~
- 1. D OE+04 5.6ssET63 -1.137riO4 1 Cn40 t T1065 a pCN . X-DI API.ACt.*E>T 0. f a .r0RtJ e lanAXIma s 400r TIME FAA.CohMLD CFhtentDAL CFhTWOJpAL TOW 510if Al, .
*. O . X-DISP. X-OISP. Y-D I Sl*. ANGLE .
I 1373 1.299F+00 -1.291E+00 2.0Til-02 1.4 tat-05 2 1373 *.h44E-01 -9.r344-01 3.So46-02 1.M19E-OS 3 1371 6.713E-01 -6.76tE-01 1.142F=02 - 1.h19E-05 5 site.-u3 1373 3.S74E=0i -3.Sh76-on ITdTW-oS Cs . 45 . 1373 4.514E-02 -4.396F-02 1.0096-03 1.41*E-05 { (,3 >
.g. T
.N C8 sd nI T IUt85 maEN Y-018PLACEat.mf AT C0khrk IbwAAAPAX
%09F f t M r. 247 COWPF# CFNTHOtDAI. CFMTM01Dat. - 70N S I O4 A I.
- 40 T-ut8e. y-nt P. Y=fJSP. AMGLE
- 1 1069 4.049V-02 7. % e 7 F.-0 2 4.0006-02 1.0%4E-05 2 f td 4 3.077L-97 5.n47t=02 3. 0 2 3 F.-0 2 3.654E-05 -
3 1059 2.14 t t =02 4.0436-oJ 2.049t-02 1.11th-95 4 1069 1.17SF-02 2.115E-42 1.123E-02 1.111F-os
- 5 1431 2.165#-a3 -9.953E-04 1.16rE-f.) 2.175E-05 S t>UCTs'P A b ACC&.PT A'.CL- A SkF NF ,
SEC11'P. ** H:ds F F 1 als 001&T llut 13S3 WJa .ut akT Yldt hS2 k3a 0054T TI*r IO6n
#4s 017A7 Tf*> 452 W5a .utt41 TIME a52 . p6 .01347 TIMF. 952 VCT i tite shalFP 2 43u .nCtri It*F 1i43. p2s. 0194T TIME 1793 a3a 0164T T T k.E .1071 P4s .o s ).a 7 T!*t kne WSs .ss3JAT 11"E 952 p's 0394T Tt8K 452 Serif 9% rets
- F Ep 3
- 41s 002AT T!*E 1351 R/= .n22AT Tief 1391 his .6TT4Y YTRE Io70 --
#4s 059A1 TiwF 1343 95 4$2A7 TI'E 1393 E6s . 06147 1[NR 139J SFCTIUS N tJ"H t P 4 .
ela 64s
.on2AT II*E 1151 E2m 02741 TJ9r 1343 asz .ns44r Tref tE70 047AT !!*6 1393 M5s 497AT T!*tt 135H N6a . t (4 2 A T 7 t t' E ))50 .
SECTin%-%IluMP k 5 002'Af~71st itM w? kla . ti 2 2 AT 'i l * > 1393 p 's 0604T Yt*E 1670
#4s 11Aa7 TISP 1393 d5 .to7&T Tipt 135R N6s .12nAT 7tHE 1J58
f~'
) 'i re'; stt!MED o s' \ ' - ~ ~l Liu .%0 EAT Tier 139J v2; . 1;,4AT TJmE 3277 kJs .15147 TIDE 6R9 . ,
c40 .167AT T I =t' .17 7 7 - #5s 576AT T147- 1280 963 6164T TTME 12k0
.sECTink Les*cER 7 uso 19947 TI*F 1393 kas 2h Tf TI*F 1670 kJa . 219 477T79- l o 7 o C43 .196AT TIdt #3h H5a 62447 TIFF 1009 R6x 6W4 AT TIME 1069
- sfCitou 6UAREu a clo 3474T TISE 747..R2m 16747 Tike 562 R3a 170AT TJwE 562 , se aos 361AT TIRE %#7 e5s 507AT T!*F 3432 P6s .539AT Time 1432' SEf1J03 4yf 4 ER 9 ' '
Oto 42aAT TI=5 1358 R7s 21647 73u> 335H N3a .IllAT T I :46 1360 040 239AF TIet 1958 45 701A7 TIwF 1358 N6a 749AT TIttE 1358 *
.sFCTION *: 8s t' o 6
- 10 41s 249AT II=E I393 w?z 130AT T3M(~f277~N'3=.06W4T TIMF. 'O s was 06547 Tree n 95 294A1 TIhR 1393 P6 29947 1IME 1393 .
.s FrT Io re MU>ork
~
11 ' 410 .246AT T I a'F 1393 44s 17:A1 lice 1070 R3s 000AT TIME ~0 was 960AT ft*E o R5s 2kf.4T li>E 1393 P6= .2R6AT TIwg 1393 SFCTing w r!" h E R 12 W1= .255AT TIsr 747 h2m 333AT TIgE 562 R3a 000AT TTMM 0 - Res 000AT 1INE O H5= 755AT TIME 747 h6m .255AT TIME 747 '
.%FCTind MU"HER IJ Pts .315A1 TI=E 195b H2= 17267 TIME 1.834 R3s 40nAr TIME O Con 000AT TIME O R5a .315AT TIME 135R R62 .315AT TIME 1358 1
Co
.= e e e 6 l
9
CAX.FLOOP 3 2.4810 % 5 .
/ \
CAX 73PA 2CET \ / GAP t.LE . T MAa. FORCE TIMF. , , 1.94e#C+0S (.9470+00 1 2 1.a sup+o5 6.404D+04 Tabie 6.5 e i a +w, n ra m in 4 1.46ao+0s 1. a 7 s " * '" s 2. o o re n + n s 4. v s i r.+ n a Fi1e D G P U 41, Moduie K, 6 2.iiii7ITiis 5 n . ilGITlii'Du i 2 1. r i o t,+ a 4 1.to2n+01 4 S.9010+04 1.073r+ot Coef : *8* F u II Rack ' 9 l'. 2 510 +ii$ 5.W 6 eT+ ils 10 1.641D+05 4.H6 kit +00 11 1.2tip+nS 4.942n+a0
- 12 1. h fi 6
- U5 1DHOO+03 ,
13 1. 7 e A n+ n ai .6.4SSn,00 14 1.764D+05 6.79%D+00 15 si.2240,04 1. l?Qh + ee t . 16 6.950D+04 1.2twn+01 17 1.44%D+0% b.6270400 , 18 1. 4 4 8t ta + 0 5 7.1230+00 19 9.4n4D+44 3 .10 4 9e+ 41 20 9.2220+04 1.2170+01
'i e FLUID DAMPFP se A X . FOHCE 1 0.n000+00 2 n.nnnen+0A 3 0.in% D+00 I 4 es.nnoo oo
- g 5 0. n.3 0 0 + f 0 b 0.00ssts+ d4 7 n.no40+0u p O 0e100,+00
_ _ _t SAT, jut. 16, 19 k t . 11:51 P:4 PH0 GRAM EGEIA5T ., . Gail F.Cors.8,5034% ' A t t>H7,o AT A spi e U41,16X16 FU f,t. R A CK 4' FEET I*. POT PAWAut.TEFS e.e s . nF 6014.8 ( viv *stin ) a 5 r' n . t if F. L.6 .' E sv 7 S ( 'T d E f. ) s 12 F P I
- 7 O P T T Ws (f7pf) s 4 X-H At ra rD tp (A2) a 4.991E+01 Y $eal.sTiiie4 ( t.'} ) = 4.441 961 ELrw. HaLFf,FaGTH (PZ) a 2.111F+01 0 STR*:SS CCFFFICIFNTS F04 RifM f* F x . m 7. fi 4 0 e'-0 3 Cp in 7.a%8sF=0 ) CF7. m 2.350E-0.4 cax s 3.34h6-04 0%y a 1.3e6E=04 CTX z 7.422Ea04 CTY = 2.422E-04 O STR F.SS Cri>_'F F I Ci t'*l f $ FOR SilPPnPTSt Crir 2 s 4.000F=n2 CF Y 2 = b . 0't4 E= P 2 CF22 = 2.667F-02 CdR2 s 6.7%or-0) CI'72 = h .' 7 E WR-Ts ) C rX 2 = 0.5unE+oo CTY2s p.000E+00 n STseFSS Cast:rFICI E*f 73 FuW SUP90HT AUT TU ht 0 S1p>65 C00 FIC1F4fS FOR SitPP0HTSI .
CFX2 a 1.356F-pl CFY2 = 1.156E-01 CFZ2 s h.030F=02 C'a t 2 = t.4h1F-01 CMV2 a 1. 4 56 3 F = 01 CTI? s 0.000E+00 CTY2s 0.000E+00 0 STsTIC Sinf5% CDFFFICIF*.T5 CCEP2 C_Ca",CC1Pg CC f S ,C K S60,CYS$ts 0.000E+00 0. 0 0 0 **
- n 0 0.000F+00 0.000E+00 0.000E+00 0.000E+00 e t1, F A T . I' d f , F al s 15 0 t'0. 0 1 S o 'J 0. 0 100nd.0 15 H @ . 4 Fra,Faw,Fvn, Fens 4st000.0 4 Hon 0.o 42000.0 4Ho00.0 S5 CTI'm No. OF 001T (JpOnt) a 5
. . . - . , - ... .,e, , . , . . . . s .- r e ,a
f re - G s ( \ 1
\
l \ v/
@ SECT. Itet 60tdf 444 TAkIA Or FAX 1 MAX EQUIVAf# e/DtCES4 S *.W. 01 w.S rRF.SS O' ro. no. NO. (SFWAMX)
-x=,JMD STw t .pl.3TITt11'TIETY.~5EF A a IL A r.sHT2T.
wet sptAns (50) (SRX) .(SBY) (TX) (TY) (TN) 1 246 9 5.793F+02 5.63%E-01 -0.390C+01 2 1974 9 9.546E+02 8.thAE+01 2.016r+02 2.Gi3E+o2 1.471F+04 -1.013F+02 2.31tE*02 ).540E*02
- f. R 97E + 0 2 3 495 9 1.03eE+03 ~7.6446+0C 2.32nE+07 -2.499E+02 2.402E+0? 4.27eg,o2 4 1974 1 1.231E+0A 1. ti t 7 % + 0 9 -5.h64M+07
~ -2.1ht>.+42 5.179F+02 5 ANN 4. f4TiG62 ~ -4.734F+C27. M G7sul 5~7 MFi7i2 -S.75U TO,e A
1 1.51tv+03 -7.177E-01 -6.2d%E+02 8.b21E+02 4.J37E+0? 1.uR2F+07
- 44) 4 4.h40F+0S -2.hti%E
- 0 % 3.004E*01 1.0t#E+03 0.00hF+00 7 4,3 s .Y b r + 0i 7.N66F+0? -7.119E+03 -1.077F+41 u
1
-4T1FI RO3 -4.(J Y+Tb t 4TS S D +D 3 375f4FFG3 4e9 2 7.SaoF+01 - 2 . h 5 4 ** + 0 3 -1.05ag+03 3.591E+03-~~~4.562E+GI 9 -3.29dE+44 '-2.54%F+03 4.10nE+07 12d2 3 1.20aE+04 -4.3tSE+63 1.24HEt03 ~4. 4 31E+ u 4 -3.472F+0) -2.534E+01 1.157t603 to e77 o 1.219*:+04 - W.M + 0 4 ii!6MRTG i G.0 @ ri39 -%.476Uas -5.h9sE+02 11 4 68 .4 7 2. J 6 7.:+ 0 4 -9.714E+D) 1 5117.+05 s.470nu s 0.000 Coon 0.000v+00 -7.447F603 af. 497E+o3 12 499 6 1.320E+64 1.134b+04 13
-6.464E+01 0.000F+00 0.000r+00 5.752F+03 -1. 4 3 t F.+ 01 ~5.757E+03 1262 2 2.10d.+94 -4.~157F+03 0.008)EFTO_t o.000F+04 7.7DW53 1 C;wnITIOa8 % HEN 1-DI S P t. A CE a t:4T OV A COHbEH ISMAXTwaX 5.73M +03 -9.5DETG1 N.Ji'E Tire > A 1. 00h 'e t.R Ce.NTNO40AL C F tYa L 0 A l, T0w3 t el'4 4 L hD. X-DISP. X-DISP. Y-DISP. ANGI,E 1 4P6 7.466R-02 7.d33E-02 6.94tF-02 2 496 7.0616-Oh .
6.06)(-07 6.0?H6-02 5.121F=07 7.06tE-06 - 1 48o 4.226E-02 4.1946:-02 3.693E-02 6.442E-06 i 107 ~02 4 496 2. 3 WGh 2. GT6FT62 n.4Ti2E-on 5 496 5.706E-03 9.3h7F-03 4.32EE-01 6.402E-06 C g 03 * (y3 Coe 01 tIDAS wHEN y-51SPLACOE*aT KT~CHW4EH IS*AUPAX MOGE TIME MAX. C0Yl4 ER CF#TRilfDAL CETTPOTn&L 70RSTONAL A9 V-0JSP. X-O!3p. Y-DISP. 41GLS , 1 4d4 7.591F-a2 7.602/-n2 7.5e0E-02 6.t!5E-06 2 444 5.847t.-02 5.wJ%E-07 5.90tr-97 6.115E-06 3 424 4.0 3 b 662 4.os/Fu2 4.0lV IU2 4 . 3 7 Tie.- n a 4 4H4 2. 25 5 F:-0 7 2.270R-02
% da4 2.23]E-02 4.320E-06
%.00g5.-03 5.156E-03 4.790E-03 4.120E-06 STkuCTURAI. ACCEPTANCF-AS*E NF
.* (C 710 *4 hU'8EP 1 ar t : 000AT HME 4836 P2= 997XtT!WE T672 P T= .ocuT Tien 24a Ras 006AT TIME 1072 R5s 009Ar RIME 246 R5a 01147 T t rt E 746 SFCTION WHuRER 2 Plc 000AT TI-E 490 R73 .017AT TIMF 1074 RJ2 R4= 014Af TIun 1114 0194f 71%E 718 RSs 02347 7t*P. 1073 P6 .027AT T[MF 1473 SFCTION t: H W E A 3 ans 000At'TisE 495 was .O rot F 'lI *E G8 Roa 03?AT TISE 714 RSz .038AT Ti>F Pl= .GTTET TnMTh 4H6 P6= .014Af TIME 4N4 SFC T Ite m hU'BFR 4 Ris 000AT TIPE 490 R7= 0198T T Tr E 693 93: .049AT T14E 1171 Eds 654A1 T14C 657 R5s 069AT TINF 4Hk p6: 081 AT T T'4E 4HH S*CTION NU>eEP 5 te l : 004As 1t*E 494 eya .dt447 Yl#F.
R4= 67f P3= 0CITT Tt C f21 06 EAT Tier 657 P%: .OR6At TIME 488 R6s 301AT Tide 448 S5c1104 f:UwnEP 6 413 721AT ll E Sol W2s 340AT TIME 677 RJa pas 794AT Tf*F .200AT Tlav 497 677 RSz 4thAT TJ4F 500 ,Na= 451 AT T T?a$ SOH .
e o e f
~
i
.N
. cy- . -
a I
' . - I
[' 3 I6 12 o7 n8 o0
,2 T1 E1 7R T2 b 8 0
~01 4
,4 1,
$ 4 5 2 4 ~
EE W* T1 6E kV IT M1 Mm Ii r EE ar I ! EE
- M EE' MM I1 eM E
9 I I II f I , , s TT TT TT TY TT TT ~T T . TT TT r TT TT TT ~TT TA e 1A 6 F0 M2A A4 O5 A4 n1 4A 09 lA d1 50 W% 05 T1 i 39 74 o8
'4 9 34 54 01 1 01 42
. n. 2 . ? t e
so ss ss 2 8 ss ss ss Sa t. 6 J6 36 t 6 t6 J6 .'
#C RR EN 4P p4 PP P9 .
e h3 99 6h l9 O4 9 9 44 82 12 9 77 70 65 7U n4 90 90 11 29 45 1 2 j I 11 i 11 fE EE Er EM 6 E EE FE k M vd MM
- iI fT 1 i TT F!* II TT 1!
TT W d JI
- m
!J g
4 TT TT
~
fT AA TT AA 7T TT ?T T7 fF Tr A4 aA 44 A A A% 09 T7 l5 sM 31 0S 41 81 '1 3 2 6S 14 14 W7 0. l 41 12 0% 01 41 12 7 9 9 t 1 2 1 1 1 1 1 s3 ss sz ss sa ss 2 = 25 2% 2s 2% 2$ 25 75 CC P9 #R RW RR HR 9R P R u P R a~ r E E E F . F3 54b M59M 046 09 4 17M7O8 99M 0 s0 W 8 R 4 0 0O4 1 0 d ue4U 54U24U 4 U 4 U 5 EW'2 h % $ 1 n :
? N N 1 N J. E?4 EEJ 7
- er4 E E 9 E 'E N EEN :t r 0Fm0>rDw 1Ti8 Iif 1 l1 a re s. PO III h m O M S0 *e 1i T II1 ! j T7 0 TTT'TTTTT1 r TTrTTV C C C TT TT11 C C TT Aa F r~T T El T T '. 9 TT 25 AA3AAk AA8 AASAA s5 C0 pt 60 10 S0 10 1 77 95 50 10 S0 1 0 23 27 23 10 20 10 20
>o ss ss ss ss su ss 9t a l 4 te l4 14 l 4 i 4 CC O2 eR aR 4R a4 Hk 9 g
Q c)
3 7Q L MAY.FL W C4Da 27663O+05
- I ~~ ~~ ~
~ " ~ ' ~ ~
943 Q ; MAX.I4 PACT FIPCES' - \, / \^ 944 GAP ELEMENT CAX.FOPCF. 717E - 945 1- 1'.2040I05
~
"E.76f6i63
.946- 2 :1.2470+05' 947 7.3050+00 Tabie 3 7.364n+04 7.660n+00 .-
- 6. 5 d 948 . 4 6.493b+04 ,1.0760+01 -
5.25 yo* 951
+ .
;g ;, : . p .i l e' D G P U 6.1 , Module H.
7 8.5220+03 4.1190+00 952 R 1.10tD+04 1.2220+01
-953
~~5S4 9 1.248D+05- 1.2790+01 *COef -
""
- 8 ', FuII Rack 10 -1.i3SO+05 , -6.3430460 -
955 11- 1.17vn+05 n. 956 6.7220+00 .* 12 1.1990+05 5.4970+00 957 13 6.4390+04 959 7.0830+00 . 14 6.543Dv04 7.268D+00 i ___9 5 9 15 1.627D+04 4.124n+00 960- 16- 2.7760+03 961 _' 1.085D+01 17 1.057D+05 - 1.012n+01 -
,__962 1.8 1.467D+05 '
+-
2.566D+00 "4- '
... 4-963 19 5.922D+04 964 2.349n+00 20 8.2890+04 1.084D+01
( ___9 6. 5 FLU 1D DAMPER . MAX. FORCF
- 966 1 0.000p+00 ,_____
967 .. a + e. 2 0.0000+00 , ,
~'
963 3 0.0000+00 - *
- 969 4 0.0000+00 970 5 0.000D+00 * ' '
l CD' 71 6 0.000D+00 ^
- 972
-D' s -
I 7 0.000D+00 ' 973 8 0.000D+00 CO . _ q ._.9 7 4 1 SUN,.JUL 24, 198__3, 2 06 PM - PROGRAM EGELAST , 976 CPU H,COF=.8,50s4% 410HZ,DATAzDGPU61,11X16 FULL RACK 4 LOW FT. 977 l 978 INPUT PARAMETEPS 979 NO. OF NODES (NUMNOD) = 5 920 NO._OF E LE,M EN TS ( N_ U M E L) = 12
- 921 PR147 O*T109 (IOPT) = 0 982
- X-HALFWTDTH (A2) s 3.434Er01 ~
1 ___993 Y-HALFWIDTH (82) a 4.991E+01 i
- 984 ELEd. HALFLENGTH (RZ) e 2.113E+01 s
985 I 986 0 STRESS COEFFICIENTS FOR RACKt 067 CFX .a 1.013E-02 CFYz 1.013E-02 959 CFZ z 3.300E-03 CMX = 1.934E-04 CFY a 2.730E-04 CTX = 4.042E-04 .CTY z 4.651E-04 l 964 990 0 STRESS COEFFICIENTS.FOR SUPPORTS ! 991 CFX2 = 6.000E-02 992 CTY2 a 6.000E-02 CFZ2 = 2.667E-02 ) CMX2 a 6.760E-03" CMY2 m 6.760E-03 CTX2 = 0.000E+00 CTY2s 0.000E+00
- 993 j 944 0 SfPESS COEFFICIENTS FOR SUPPORT BOTTOM i __995 0 STHESS COgrF1CIEgTS FOR_SUPPORTSt '
4 996_ CFX2 = 1.356E-01 CFY2 = 1;356E-01___CFt2 E~T;17Dtr02 ~ , 997 CMX2 = 3.483E-01 CMY2 a 1.483E-01 l 999 CTX2 a 0.000E+00 CTY2m 0.000E+00 j 939 0 STATIC STRESS COEPFlClENTS
~
l 2000 CCXP,CCX4,CCYp,CCYh,CXSM,CYSHz 0.000E+00 0.000E+00 1001 FTTjFAT,FVT,FBTa 15000.0 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1 ' 15000.0 10000.0 15000.0 1002 FT8, FAB,FVB/ Fed ="'48000.0'~~ . 48006;0 i 1003 12660.0 480000 SECTION NO. OF ROOT (JR0nT) = 5
2005 ' - . 1006 1 -TABLE OF MAXIMAY E4HIVALENT STREAS 1907 0 TIME POINT SECT. M4X. SEO.
~~ 'DIR.8 TRESS X-RK40 STR Y= PEND STRE8XLAT.8MEAR TLAT.8 HEAR' NET SHEAR)-
1008 0 NO. NO. NO. (SkNXMX) ~(30) ( ABX T "( 55Y F' -"~l T X )~ ' '"lTY)' (TN) - 1009 1610 1 942 6 7.246E+02 =5.959E-02 9.178E-02 -2.014E+02. -3.622E*02 ' =6.771E-02 3.623E+02 1011 2 262 9 7.637EIO2 ~=I.f24sF60 7 70h6t+0 M ;664E+62 7 '267EiO2 . =U 9*ME4 2 3.~918E 42 1012 3 1159 6 8.936E402' =1.A76F+00 1.210E+02 -3.951F+02 =4.35RE+02 5.965E+00 4.374E+02 1013 4 1290 4 1.272r+03 =2.440E+00 5.h00E+02 6.994E+02 2.340E+02 ~3.533E+02 0.000E+00 1914 5 1200 4 1.548E+03 - 2. 8 4 0 E+ 0 d "~~ 7.115 E+ 0 2'~~4. 3 3 4 E + 0 2 ~~~ 2. 3 4 0 E + 0 2 ~~e 3. 5 3 3 E+ 0 2 -~ 0. 0 0 0 E + 0 0' 1015 6 1280 4 5.175E+03 -3.320E+03 9.694E+02 8.822E+02 1.306E+03 -1.434E+03' ~4.5A4E+01 1016 7 A41' 2 6.230E+03 -2.856E+03 =1.60RF+03 1.026E+03 1.517E+03 2.378t+03 =2.59sE+03 1017 8 1377 7 5.624E+03 .=2.160E+01 =1.'2AE+63 ~5 .=57497EiO2 =9'.124E+02 2.258E+03 -2.258E+03-late 9 552 2 5.45SF+03 -3.086F+03' p.572E+02 -1.265E+03 =1.872F+03 =1.268F+03 2.094F+03 1019 10 1280 1 1.173E+04 -7.906E+03 0.000E+00 0.000E+00 ' ~2.951E+03 3.241E+n3 4.377E+03
- 1020 11 841 2 1.414E+04 -6.715E+03 6.000E+00 7 0.000E+00 7 3'.429E+03 75.374E+01 1021 6.224E+03 -
12 1377 5 1.210E+04 .=6.489E+03 0.000E+00 '0.000E+00 1.972E+03 .=5.104E+03 ~5.104E+03 1027 13 552 2 1.238E404 ~7.254E+03 0.000E+00 0.000E+00 4.231E+03 2.865E+03 ~5.017E+03 1023 1 CONDITIONS WHEN X-DESPLACEMEhf~DF A CORNER ISMAXIMAX 1024
' ' j025 NODE TIME MAX.CORMER CENTRO!DAL CENTROIDAL TO,RSIO4AL
__1026 NO. X-DISP. X-DISP. Y-DISP. ANGLE 1027 1029 1 781 1.461E+n0__=1,461E+00 5.948E;03 4.150E-06 1029 2 781 1.10CE+00 -1.108E+00 4.823E-03 4.150E-06 1010 3 781 -7.551E-01 -7.549E-01 3.475E-03' 4.224E-06 103.1 4 781 4 t021E-01 -4.019E ~ 1.807E-03 4.224t=06 1032 5 781 5.003E-02 -4.9425J02 3.614f;04 1033 ' 4.214E-06 ' ' 1034 1035 * ~ ~ 1036 1037 CONDITIONS WHEN Y-DISPLACEMENT AT CORNER TSMAXIMAX 1A38 s 1019 NODE TIME MAX. CORNER CENTRO 1DAL CENTRot0AL TORSIONAL 1040 N O, Y-DISP. X-DISP. Y= DISP.. 1041 aNGL{ 1 841 3.652E-02 7.264E-01 3.640E-02 3.626E-66 1042 2 841 2.766E-02 5.510E-01 2.754E-02 3. 62 6 F.-0 6 ___[n43 3 1378 1.929E-02 ~7 t814E-02 1.R06E-02 3.331E-05 1044 4 1378 1.099E-02 -4.088E-02 9.843E-03 3.331E-05 1045 5 1379 3.425E-03 -1.204E-03 1.902E-03 4.434E-05 1046 - - e-e s_ _ + _ - . _
- s r [g N
9 4 i 1047 STRUCTUR AL ACCEPTANCf;A5ME~df ~ ~ ~ ~ ~ ~ ~ 1048 SECTION WUM4'ER 1 1049 R!s 000AT TIME 806 R2m .0I2AT TIME 942 R3s 006AT TIPC 1094 1650 R4s 014AT TIME 942 R5s .012AT. TIME ~~3Tl R6s 015 AT TfNE~~381 0 1051 SECTIOM MUMBER 2
~~~1052 R_1 s .,0014T TI"E _806 R2s 012AT TIME 734 R3s'~~~~~ 01647 TIME 1210 - ! 1053 R4s . 03147 TIME. 942 "R5= 032AT TIME 277 R6s 037k?"TIMd'~274-i 1054 SECTION NUMRER 3
- i 1055 pts 001AT TIME RCA R2s 015AT TIME 1159 R)s 027AT TIME 1210 10 6 R4m 645AT" TIM I 8df~s5s
~
04'7AT fiME 2T1 R6s
~
05IKT 'Tli4E' 2K1 1057 SECTION NUMBER' 4 1956 Ris ,002AT TIME 406 R2= 01647 TIME 880 R3s~~.049AT TIME 1281
~~~1059 R4s 064AT TIME 807 R5s 072AT TIME 1280~~R6s 045AT' TIME'i286 j 1060 SECTION NUMBER 5 .
~~~jo61 Ris 002AT TIME 406 R2s .01647 TIME- ORO R3s 061AT TIME 1241
- 1062 P43 070AT TIME- 807 R5m . 08847,TIMi~T290 R6s .103A7 TIME ~1280 j 1063 SECTION MUMBER 6 J
___1064 Ris .228AT 7tHE_1290 R2s .076AT TIME 12R1 R3s 07 7 A T_T I M E_12 91 d 106% P4s 0624T TIME 1383- R5s .32647 TIME 1280 R6s 345AT TIME 1280 1066 SECTION NUMBER. 7 1067 Ris 192AT TI4E 635 t IO6AT TIME 841 R3s 1_07AT_ TIME 841 10Ae R4s 068AT TIME 841 . R5s R2s .34047 TIME 841 R6s .366AT TIME 841 1 1069 SECTION'NUwRER 8 l 00___1u70 Ris__.204AT TIME 613 R2= .100AT TIME 1377 R3s 7 102 AT_TI M E,,1377 11 1071 R4s 056AT TIME 1296 R$m .305AT TIME 1378 R6s 326AT TIME 1378 CO 1072 SECTION NUMRER 9 Pts 212AT TIME .087AT TIME 1282 -R3s 0RRAT TIME-1282 U3~~~Jn73 1074 R4s 094AT TIME 552 ~~551 ~"N2= R5s .32847 TIME 551 R6= .349AT TIME 551 1075 SECTION NUMBER to 1C76 Ris .16347 TIME 1280 R2s ,06147 TI4E 12R1 R3s O
~~~1077 000AT T14E
~
.000A7 TIME ~
R4s 0 R5= .i63AT T145~1200 R6= .1'63AT'~ TIME TF06 1078 SECTION NUMBER 11 1079 Ris ,141AT TIME 635 R2s .0844T TIME 841 R3s 000AT TIME
~~~
1080 R4s 000AT TIME ~ -0
- O R5s .141AT TIME-~635 R6s .'141 A T T I' N $~~~6 3 5 1 1081. SECTION NUMBER 12. .
t 1_002 _R l a .150AT TIME 673 P2s 000AT TIME 1377- R3s 000AT TIME o j 1083 R4s 000AT TIME O R5s . 150AT TIME 673 .R6s .150AT TIME 6Y)
- 1 1084 SECTION NUMBER 13
- l 1065 Ris .156AT_ TIME 551 R2s 069AT TIME 1282 Ris' .000AT TIME "
1086 0
- R4s 000AT TIME O 'R5s .156AT TIME 551 R6s 156AT TIME 551 l ~ ~~
i 4 i
O Z. "a 3
. n ) ,
/
C O U PLIN G ELEMENTS l 4 TYPICAL FU EL ASS EM B LY 3 GROUP M AS S H TYPICAL FUEL RACK MAS S 2 FUEL R ACK B A SE 2 l
= AY p :
/ \ /
1, Ax 1 l Ip - Ys 7 Y A+-y,xa 4 jja i
, ! t
, i 4 m i h
,5 $ '
FUEL R ACK SUPPORT l x XB, YB - LOCATION OF CE N TROI D O F FU EL ROD GROUP M ASSES - REL ATIVE TO
] CENTER OF FU EL R A C K Di = UNIT VECTORS FIG. 6.1 D Y N A M I C MODEL
. O l , U l l L L 1
' .y .
IMPACT J L SPRIN G S 1
, L t
? 'f I.H MASS W 5"' ~
l ! .E F LU ID DA M P E R S l \ l RIGID FR A M E X O Fl G. 6.2 l M PA C T S P R I N G S AND FLUID D A MPERS l .-._ _ . . - . . - . . . 6-41 --._-.-:..-...-...--..--.
5 ' . O %@W
) 5 l ~J -%. . k (Typ.) \
\
4 . l Seismic Monens K E h WM Z N \ l Fuel Assembly
' ' . Group Lumped Mass (Typ.)
l q
~
u
- 3 g
a:+ . 3 Rock Lumped. 3 Mass & Inertia - For Horizontal L Motions (Typ.) ( Y f
, 2 K,(Typ.)
l t 1 6 m6 ._ Ng a ," A
/
wn
. _r-xi
, 1 A, >
Figure 6.3 Spring Mass Simulation For Two Dimensional Motion
*'t 8-42
O i j'fY YJ l B L
; ; B , .
u Fx l "X i (a) TOP V EW j i zo . l. 3 ,
,,Fz .
p , l O T =F x-(b) AX AL CROSS ,
///// / / ///////
S ECTlON ( B-B :) FIG. 6.4 (a) F ORIZONTAL CROSS . SECTION OF RACK ( 3) VERTICAL CROSS SECTION OF RACK O I 6-43
CELL Z(W) O we us " ffC
^
t ^l/ EAss'
'/ nvIAAAX)"c n(VAAAW a
b= NyC a=NxC7" C A g, y(V) A RIGID PLATE B /
= X (u)
I I f ~- Y (+f*7}
-4 e A=-4/r^ ,
SUPPORTS
= G. 6.5 ovnamic uoeei dMz 1 a jMy fzY " ,
-d / /!
B
=~M*
Fx C
.A 8
. O F G. 6.6 s tress aesuitants o rie n t a tio n l 6 44
d O . l l hZ
. NODE I
/g E L.I ,
SEC.I ~ ---- N O D E 2 -* / E L.2 SEC.2 - - N O D 5 3 -.- E L,3 S E C. 3 ~ -- t
~n U NODE4 ---- -
a
- E L,4 .
f- y S E C 4 -.- -
'/ ,
E L.5 ; SEC,5 - ' ^* s '_ w'g. NODE 5
,- ,- e A S E'C,6
,c.'_ yp
- E L,8
- EL,7- ' .
- f. ROOT OF RACK ,
? S E C. 8 ' S E C,9
. N O. O F E L E M E N T S = 8 F
N O. OF S EC TI ON S =9
- N O.~O F N O'D E S =5 O F G. 6.7 E
SUBDIVISION OF A TYPIC AL R ACK 6-45 .
e . - - 4 -
- O
.=
4 40 g y .
~
b 0 y LO g . 4 0 = I e e G. 6.8 FINITE ELEMENTS MODEL CROSS-SECTION O
, 6-46
( HORIZONTRL. SEISMIC,SSE,4 PERCENT DRMPING Figuro 6.9 i O, q ,30 m ! O i w 0,25 g l O . 0,20 # 'l fr .j W l
.2 i
t 0,15 W l g O 0,10 I
@ 0,05 ,
l1 tQ !k G J . j M 2 0,00-3l -
- LJ i a ,
i ita O -0,05 ' ' c I I i , l -0,10 '
-0,15 i
L l ':
--0,20
, O 25
; i
? . %) i _ l l 0.0 200.0 400.0 600.0 800.0 1000.0 1200.0 1100.0 1600.0 RRr. rLJLJ51stParas w
VEliTICAL SEISNIC,SSE,4 PERCENT DRMPING Figuro 6.10 n
/ 3 O 16 .n 0
0,14- 1-l 2 0,12 - 0 l-0.10- I m J
<0,05-
- O <
o , i ; O,06 j
,.i
; i ,
s . 0,04- , jl , l t (, d, j
.(i
- q ,,
. t t , ' i. .
CD i. j g j t ' j.,5 g j , ' i
. i i li ,!
l ....I .n. l [I: i ; 0,02 '3 q
+ .. '- -
# .i g' p
1 _, n ) 'l Ir. l }r f.C o ;: i' ' it "9,' t :=# n,} ]' j W 's->r L ::H l
- 1 :
i g{- j !
'p
. , !i j
t.!
- lll,'/[vj '),
! l l-o , j I <
; I I-'
, 1l '
; ji ., l,
-'.---W)j,!-'t-1-
C~ 0,00 -
-W'l i 4 ii j q ;l ,i l !
j I ;. i.
--4 i .t lh i
9 0:: !dl'i ; l M 1l] n. , . - l i l i ',: - ' 1
!L T; i';
i l . I I '
! l l d ,1 j i J
M -0,02 l s L
" ' 1 ,
r ,. I'. I; i I,j !' }p
- i. " - t o l g l' i j j .t i l F2 i
"llk, a l O <
l h :l ' i ' 4 C ' jf j 'l
-0.01 ,
j ~ j' ['[,,
- 0 n
l
-0,06 'i l
b l ' l
- l,
-0,08 '
l l
-0,10
-0,12- b-1
) .m
\
, ' (%. l. 4 I
..p y , ,
l 0.0 200.0 100.0 600.0 800.0 1000.0 1200.0 1400.0 1600.0 ! TIMF ( _01 SFC) 6-48
'7. OTHER MECHANICAL LOADS n
) 7.1 Mechanical Loadings 7.1.1 Fuel Handling In addition to the Seismic analysis presented in Section 6,
'the-racks were also analyzed for the mechanical loading condition specified in' the GPU Nuclear Design Specification [1] .
Reference 1~ specifies that the fucl handling bridge crane _is capable of exerting a 900 lb. downward thrust and a 1700 lb. upward thrust on_a fuel assembly during fuel manipulation. The 900 lb. downwards thrust, together with 800 lb. fuel assembly weight, produces a net downwards load of 1700 lbs. on the rack. Calculations show that a load of 1700 lb. (in upwards or downwards direction) applied on a 1" characteristic dimension of the rack will produce .i local stress of approximately 14000 psi. Since the. yield stress c:f the. ASTM 240-304L material is 25000 psi, even local plasticity is precluded. 7.1.2 Dropped Fuel Accident.I A fuel assembly (weight - 600 pounds) is dropped from 36 inches above a storage location and impacts the base. Local failure of the baseplate is acceptable; however, the rack design should preclude impact with the pool liner. The subcriticality of the adjacent fuel assemblies is not to be violated. Calculated results show that the baseplate is not pierced and the rack feet loading on the liner'is well below those-caused by seismic loads. The- maximum depth of baseplate penetration is conservatively estimated to be 0.446" (vs. 0.625" baseplate nominal thickness). 7.1.3 Dropped Fuel ~ Accident II i n One fuel assembly dropping from 36 inches above the rack and hitting the top of the rack. Permanent deformation of the rack is
. 7-1
acceptable, 'but is required to be limited to the top region such s that the rack cross-sectional geometry at the level of the top of the active fuel (and below) is not altered. Analysis. dictates that
~
the maximum-local stress at the top of the rack is limited to 21000 psi which is- less- than material yield point. Thus,- - the functionality of-the' rack is not affected. I 7.2' Local Buckling of Fuel Cell Walls ! The allowable local. buckling stresses in the fuel cell walls are obtained by using classical plate buckling analysis. The following formula for the critical stress has been uso [a]. 2 8v Et 2 (1)
,cr , 12 b 2 ( y_ y2)
- where E= 27x10 8 psi, t = . 062" , b = 6.0". the factor S is i
suggested ? in Ref. 1 to be 4.0 for a long panel loaded as shown in Fig. 7.1. For the given data' i a = 10423 psi cr It~ should be noted that this calculation is based on' the applied
- l. stress being uniform along the entire length of the cell wall. In the actual fuel- rack, the compressive stress comes from consideration 'of overall bending - of the rack structure during a seismic. event and as such is neglible at the rack top and maximum at the rack bottom. It is conservative to apply eq. (1) to the 4
[a] Strength of Materials,-S.P. Timoshenko, 3rd Edition, 1956, Part II, .pp. 194-197. 7-2
-m - E, --6 w n, m ,,my-- ,w++w--,--~,,ea-- e,+,mwe --,-,---.,e,se-. ,w, - - - ee,,rw~,,-+,,--,-,-em, r o w, ---o,,,--,e,e--m o-,-w,---m---www.m,ew,--,r---,
rack cell wall if - we compare acr .with the maximum compressive stress anywhere in the cell wall. O The output of the dynamic analysis programs DYNAHIS and EGELAST provides the time history of critical stresses at various levels in the rack and in the rack' support feet._ In particular, an output is
. provided for the maximum direct plus total bending stress in the -
outermost cell wall at the bottom of the rack. This translates into a maximum compressive stress in the cell wall at'some critical time in the . seismic event. The output is given in terms of the stress factor R5 or R 6 .- As defined, the stress factor is the ratio of the actual stress to ,the allowable value (allowable = 15000 psi for2304 S/S based on yield and/or overall column buckling). Therefore, the maximum compressive stress in the cell wall is given: as - o=R6 x 15000 psi Table'7.1 below shows critical values of R6 for different cases and l- .the corresponding local compressive stress. l C Table 7.1 Compressive Stress in Rack Wall and l ! Comparison with Local Buckling Stress for Typical Cases
' Load Case R6 o (981) c er/*
i l v Full rack-E . 094 1410 7.4 L< [. Full Rack F .126 1890 5.5 !~ Full Rack K- .101- 1515 6.9 k s.- 3 l l
. - - . - - . . - . . . - . . . - . _ . . - . . - - . . -.. -.- _=.... - -- - -..
r t t The values' given in Table 7.1 are obtained from the detailed 1 4 outputs at the root.of the honeycomb (Point 5 in Table 6.5). O ~ In the above, note that all results are obtained for SSE j conditions.- The lowest safety margin- acr/a is 5.5. The calculation is based on a plate simply supported on four edges. In. the- actual ~ structure, the boundary condition- is probably more ,
- nearly. clamped on'at least two opposite edges. This.would increase the-factor 8 to a value near-7.0 in accordance with Timoshenko's
- . <
- analysis.
t 7.3 Analysis of We'lded Joints in Rack Welded joints are examined under the loading conditions i arising.from thermal effects due to an isolated hot shell, and due
' to seismic loadings. Under.both sets of, load conditions, the weld stresses are.found to be. bel'ow the allowable value of 24000 psi in f shear .that is _given in Table NF329.1-1 of ASME Section III,
}
. Division 1,-Subsection NF,.1980.
l ~ ! A.' A thermal. gradient between cells will. develop when an isolated i storage location contains a - fuel assembly emitting maximum ! '~ postulated heat, while the surrounding locations are empty. We can obtain a conservative estimate of weld stresses along the. length of an-isolated hot cell by considering a beam strip ,; uniformly heated by 20*F, and restrained from growth 'along one long edge. The configuration is shown in Fig. 7.2.
.Using shear beam. theory, and subjecting the strip to a uniform temperature rise AT = 20
- F , we can calculate an estimate of L the maximum value of the average shear stress in the strip. -
The : strip is subjected to the following boun'dary conditions.
- l. s l
O L 7-4 L s . , , -.,,..-..-._.___E.._...-.,_._-.-_._._______.-.--_ _ _ . - - - _ _ .
- a. Displacement Ux(x,y) = 0 at~ x = 0, all y and at y= ;
w/2, all x. n T * ~ .l. . .
- b. Average force Nx, acting on the cross section Hxt =0 '
at x = L, all y. The final result for wall shear stress, maximum at x = L, is found to be given as: T a.
.E a A T
. s_
.931 ;
where E = 28 x'10 6 psi, a = 9.5 x 10-6 in/in *F and AT = 20*F. Therefore, we obtain an estimate of maximum weld shear stress in an isolated hot cell, due to thermal gradient, as T = 5775 psi MAX l. l Q B. The critical weld locations, when the loading is seismic,'are !- at the bottom of the rack-(at the connection to the baseplate) and ~ in the welds on the support legs. The results from the i ' dynamic analyses using DYNAHIS and EGELAST are surveyed and L the maximum . loading' used to qualify the welds in these locations. The ASME code allowable value of 24000 psi is used i on' allowable weld stress. All welds are qualified using SSE seismic results. l i The welds at the rack base are 1/8" fillet welds. The shear l stress T in the weld throat induced by a normal stress o in 1 the rack-is given as: 1 T = ct/1.414h (h = weld size) (2) (t = skin thickness) f <v
'7-5 L
..w-r
-w--- ,,- o,,,,v. ~,or,w - . , , , ,- w ,e ,s - -..,-n , ,,&n,,,,.,, , , -
m e- n e e w em- -n,ww,-,. -,
4 G 7 The above result- assumes a continuous weld at the -critical location and .a uniform o along the length of the weld. For the weld between the rack base and th'e cell walls, we have t= .12" h = .125" . Examination of~ the structural acceptance factors from the transient ~ analysis outputs of all of the simulations yields a typical value for R6 at'the rack base-as , R6 = .2 Since R6 = a/15000, we have, at the weld in question:
-o = .2 x 15000 = 3000 psi
- Therefore, using eq. (2), and doubling the stress to account
~- for, skip welding,.we obtain y = 6000 x .120 = 4074 psi rack base 1.414 .125 l
l i I h l l l i-7-6 l.
n -
~
' References - Section-7
\
[1] " Technical Specification for' Procurement Specification for Spent Fuel Rack Assemblies", Oyster Creek' Nuclear Station,
- - Spec. No. 1302-12-007, Rev. 3, GPUN (1981).
. l n *
' - T i I e ! e 4 i f e S l' 9 l l l l: 4 e G h i i' . L f 7-7 o e
+e ---=-, '- ,-,e.,w. e.n*-m-, ..-+---we,._.e-.....-,e,,.-=-.-=-.,-.
I O l c a 1 1 - d 4 l I 8 l
* + b
+ a b)4 I= y -
=
l FIG. 7.1 LO ADING ON R ACK W ALL l O
- i --
et Heated d Cell Wall l =x H vassmassagazzas If L '
~
Weld Line if Y FIG. 7.2 ' WELDED JOINT IN R ACK O . 7-8
---------..r- ,.,-a--.,w.v,-u-, --,-.-s -m-v = - -- , ~ -ew--- - - , - , - - - - - - - - - - - - - - - - - - -
. . _ . . . -- . . - . - - __ - .- =_ . - - _ .
8.0 SPENT FUEL POOL' FLOOR STRUCTURAL ANALYSIS-4 8.1 Introduction : The high density rack modules for long term fuel storage . described in Section 2 are located in the reactor building fuel pool structure. In essence, the Oyster Creek fuel pool slab is .' treated as a- reinforced- concrete plate type structure with additional reinforced concrete beams. Figures 8.l(a), (b), and (c) show plan and cross sectiorial views-of the pool slab. In this section, pertinent results of pool slab analysis ; are presented to demonstrate its structural integrity for all postulated loading conditions. In particular, compliance with ACI-349 III.and NUREG-0800 [2] is shown.
, 8.2 Assumptions I
seismic qualification of the Fuel Pool Floor is carried out and analysis ! using the following conservative assumptions
- ~ -
techniques: i
- 1. For_the purposes of finite element modeling, the pool-floor .is analyzed. as a composite concrete structure.
l The floor-slab is modelled with plate elements, while the reinforced concrete beams are modeled by beam [ ' elements. Appropriate Young's moduli and inertia properties are used to account for reinforcing. The floor slab is assumed to be clamped at the reactor I wall and simply supported around the -remaining walls
.and at attachments to the reinforced concrete beams.
- 2. Calculation of the stif fness and strength properties for- the concrete 'is based on the- assumption of i'
complete cracking of the concrete in tension over the I ( entire floor plan area. ' l: p G n .
- 8-1 t
. , . . , . - . - - - . - - . . ~ --- -- - . - - . .- _-_- _.-
.. .- - - . .- .. _. ~. . - .- .
. 3. - ' ' The loading used to quali.fy the pool floor assumes that: all racks are fully loaded with channelled fuel
, assemblies.
4.. The ANSYS finite element code (31 is used to determine the -static stress state under dead weight gravitational loading (including 40' of water) and under concentrated- loads at appropria'e c locations wh'ich represent the dead weight of loaded racks. Figures 8.2a and-b.show-node and element descriptions for the ANSYS Model.
- 5. - The dynamic mass used in the floor slab analysis includes the concrete and reinforcement mass of all excited material modeled, and the virtual mass of l water set in motion by the pool floor.
- 6. The' ANSYS ' finite element- code is also used to
( (k determine' the- natural frequencles,- the normalized; L mode shapes, and the participation factors for the floor slab model. .Nine master degrees of freedom are ! - used in the modal' analysis. i , 7 .- The Joseph Oat proprietary computer code DYNAHIS is o . used to obtain a transient dynamic analysis of the
- floor slab und.er direct seismic excitation and under
- impact loadingLfrom rack analysis.
8.3 Dynamic Analysis of' Pool Floor Slab ; The _ pool floor and associated reinforced concrete beam structure isJ modeled as a linear sy' stem subjected to direct seismic input and to impact loading from the racks at selected
. load points. The governing equations for the system have the
- matrix form:
O
. 8-2 . ;
,- -.. ,wsw.. , . , , . , - - . .,-,,.,-,,.yr. ..,w_._,-- --,_..,,,,__,,...,,,m ,,--...,.,,--,_.,,,..n---.. - - - -
f a
~ [M] {x} + [K) {x} = - {P(t) } -' [M] {I}a( t)
.c
, , 'In'the above equation, {x} is the matrix. of selected floor displacements, {P(t) } :is the matrix of impact loads, input at the location of :the floor degrees 'of freedom, and a(t) is the vertical: seismic acceleration history applied'to the floor. [M], -
[K]' .are the mass and stiffness matrices. of the ' m o d e.1 , y ( respectively,:and {I} is_the unit column matrix. The ANSYS mode-frequency analysis of the freely vibrating
' slab model yields natural frequencies u n (n=1,2,..N),. and corresponding mode shapes { n}. Following standard linear systems analysis, the transient solution for {x(t) } under _ the lo'adings-{P(t}} and a(t) can be obtained in the form:
{x(t}} =-[{Wi}{W2} **** {WN }]{2(t)} i,
.where, for n =,1,2,...N, 2n satisfies l
=
{Wn } {P(t)} - {W n M]{I a(t) zn (t) + "n 2-(t) {Wn } IMI $Nn} ' In the above equations { n}T represents' the transpose of a modal : column matrix. The output of the ANSYS free vibration analysis ~ gives each- '{ n} in a form such that
. {Wn}T[M] { n} = 1, and 'gives a value for each participation
! - f actor-- , Pn" n} [M] {I}. l 8-3 i e 1' l1 x
^ = -
i-Therefore,: the~ above equations for ' Zn(t) can - be solved for any
-input . load history LP( t), and for any seismic history -a(t) . The actualj time.. nistory analysis . of the _ pol floor is carried out using :' the Joseph Oat . proprietary computer. code DYNAHIS. The input.' loading P(t) for dynamic analysis of th'e: pool floor slab
~
is obtained from the results of detailed dynamic analysis of a
. . single representative fuel rack. Output floor loads, obtained
- from a; time history analysis of an individual rack, are
- . converted to a floor load time history matrix P(t) . The input loading la(t) is the actual vertical time history seismic loading applied to'the pool. ,
- The results of .the transient analysis of the pool floor slab ' yield - the time history of the degrees of freedom 'x(t) selected as master degrees of freedom. Nine lateral -
displacements of 'the floor slab are chosen to model the dynamic ; behavior. The nine, locations: used correspond, as nearly as
.possible, to the locations of rack support locations on the floor slab. The DYNAHIS code tracks the maximum displacements O* of each master degree of freedom over the total time of the seismic event. By equating these maximum' displacements (obtained
~
j first under P(t) loading, and .then under a(t) loading) with the' finite-element solution for the static displace'ments at the
.same locations , for the similar event, we obtain a conservative
~
estimate of amplification factors that need to be applied to.the detailed 'ANSYS static results ~ in order to model the dynamic I effects.. These amplified static loadings, representing dynamic effects over the total time of the event, can then be used,
- together with .the detailed dead load results, in a standard
. strength- qualification of the pool floor as outlined in SRP
'3.8.4.
The dynamic load histories P(t) are obtained from the
-results'of dynamic analyses of the most signilicant fully loaded rack. The resulting input loadings represent the average time 8-4
9a
-history' of one foot of the rack. That is, the- time history of each foot is added algebraically at each time point-and divided
[ by four to obtain the input. foot history. Figures 8.3-8.7 show the total floor load (sum of 4 supports) and each support time history used. Structural damping, based on, the lowest calculated pool
. floor natural frequency is incorporated into the- model by, modification of the structural stiffness matrix. 4% damping is used for siab analysis ' involving OBE events, and 7% ' damping is used for slab analysis involving SSE events. For the Oys,ter
_ Creek - slab, these damping parameters are based on the lowest l frequency 28 cps-obtained from the ANSYS analysis. 8.4 Results and Discussions - + A survey of the -floor slab critical loadings suggests that we must consider the following loadings: O D = dead load -
'To.= thermal loading due to slab temperature gradient E' = SSE seismic load'
_E = OBE seismic load The critical load. combinations that must be considered for the floor slab are: a) 1.4D + 1.9E b)- .75 x'(1.4D + 1.4To)- c) .75 (1.4D t 1.4To i 1.9E) d) -D i To i E' The thermal -gradient across the floor slab induces compressive stress on the water side and tensile stress on the i fair side of the floor slab.- For the purpose-of computing the To loading, we use'a-temperature drop _of 21*F across the slab and conservatively assume that all floor. curvature is suppressed. 9 8-5
i' Then 'the thermal moment due ' to a temperature gradient AT = 21*F l through the. concrete is AT H 2-aE MT" I"~I/I" 12.(1-ve) l-where H is the - thickr ess of-
. th'e floor slab and E*
c is the l ,
. effective Young's Modulus of an equivalent homogeneous slab
( (after accounting for reinforcement). Using values appropriate to the Oyster Creek-pool floor slab, we obtain , M T = 36.566 KIP-in/in f To obtain the moments and shear force distribution in' the' floor slab under dead loading, the following ANSYS static analyses
; were' carried out using the discretized model of. Figs. 8.2(a)
L and 8.2(b).- Dead Load Case 1: Gravitational loading including concrete and
- reinforcement, virtual mass of 40' of water, and weight - of the north, - east, and west walls applied to the nearly rigid edge
. beams.
Dead Load Case 2: Concentrated loads applied at appropriate node points to simulate' the dead ; weight of fully loaded racks applied to the pool floor.through the rack feet. Also included in load case was a slab' pressure loading of 161 psi applied.to
- elements 17 and 26 to simulate an additional stored cask. The points of load application are shown in Fig. 8.2b.
l:
- To obtain- the mode-frequency characteristics of the floor
. slab, nine master degrees of. freedom are chosen at the points of concentrated load application' and a frequency analysis performed
- on a slightly different slab model (clamped south wall, zero 10 8-6
-$' 9% v -- % ery---- -+=-,w,#., s.w,g, ,,,--,,,_w,e.g-w-.6,,,,,..-,,,-,wg. ,---y ,,..,v,.- .m--,...e,,w,+.e,..- -,,-.s--.-%-,-,--,=-w-----=-w..-
width of beams and girders) using ANSYS. Table 8.1 gives the frequencies obtained. The effect of'40' of water vibrating with ( the pool floor is included. Table 8.1 Floor Slab Frequencies
. Mode Frequency (H2) 1 28.3 2 . 38.9 3 56.9 4 69.8 5 74.3 6 101.4 7 102.9 8
135.'7 9 186.0 Using the mode frequency information, a nine degree of freedom slab. model has been simulated on the . Joseph Oat ! transient analysis program DYNAHIS for the purpose of determining amplification factors for equivalent static loadings that would simulate the - ef fect of either direct seismic floor loading or the effect of the impacting rack feet during a seismic event. The results indicated that to account for direct l seismic load on the floor, it is sufficient to use the results of static load case 1 above with a multiplying factor .005. Similarly, to . account for rack impact loads during a seismic [ event with the pool full of loaded racks, it is sufficient to use the results of static-load case 2 above with a multiplying factor 0.918. j'^s i I 8-7 l .
. Table 8.2 gives - the ultimate moment capacities for the concrete sections involved in the analysis.
These ultimace moments are calculated using standard formulas used in
-reinfo'rc'ed concrete design.
Table 8.2' Ultimate Moments of Slabs-and Beams Location Mu i Floor Slab (plate) 317430 in.6/in. (East / West face) Water side [
'in tension i Floor Slab.(plate) 304950 in.9/in.
[ , (North / South. face)' Water side
- in tension Floor Slab - East, West, & South 48350 in.#/in.
edges;-Water side in compression
~
N-S Concrete-Beam under pool floor 39.506 x 10 7 in.#
}E-W Concrete. Beam under pool floor 18.5 x 107 in.#
N-S Concrete-Beam outboard. 23.45 x 10 7 in.# of pool floor N-S' concrete Beam under pool floor 52.06 x 10 7 in.#
-(water side in tension) .
E-W concrete beam outboard 6.58 x 10 7 -in.# j of pool floor J L l
' N s/ 8-8 L
O
,1
- g
l Table 8.3 below gives the concrete shear strength of plate i and .: beam sections away from concentrated load points and near concentrated load points. For plate sections, near concentrated load points, where the load application area is essentially a squaro ' patch, the shear strength is increased by a factor of , l two. 1 l Table 8.3 I l Shear capacities V c /D #Y c f r P ate l or beam sections Lo. cation Vc /b or V e
-Floor slab-(E-W face) 5304 t/in Floor slab (N-S face) .
5100 t/in l Floor. slab.(near concentrated load points) 10200 t/in i N-S. concrete beam under slab 3.310 x~lo s, E-W Concrete Beam under slab ~ 2.148 x 10 6 , j N-S Beam outboard of slab 1.389 x 10 6 E-W concrete beam outboard of slab 1 171 x 10 # l O l Referring- to Figures 8.1, 8.2 and the ANSYS results, Tables 8.4, 8.5 can be'.' constructed. These tables show typical critical locations for shear forces and bending moments in the plate and in the beam elements. The outputs have been converted to plate shear.and moment resultants where appropriate by division of the appropriate side length. The x-y directions refer to the global coordinate system.
.B-9 O~ .
e
+ -.n._
Table 8.4 Critical Shears and Moments for Total Dead Load Analysis (Case 1 plus Case 2) Plate Element Node - Q( t/ir ) M y(in.5/in)' Mg (int /in) 98 247 3355
- 63. 183 .039' x 108 19-28 127 ~1575. .0352 x10 6 60 192 -
998.6 .0316 x10 6 69-78 205 1978.6 .056-x 10 6 78-87 218 1725.96 .0488.x 10 6 i Beam -- Node -Q(t) -M (int) Y M x(in.4) Element-
~ (1) ,
.46 gy) 166 .03596'x 10 6 1.483 x 10 6 ,
I fN 49 169 .8713 x 10 6 32.43 x 10 6 57 177 .0152 x 10 6 2.509 x 10 6
)
116 g4) 268 .1604 x 10 6 10.1 x 10 6 316 .0494 x 10 6 2.115 x 10 6 130'('5) # 102 251- .4856 x 10 6 21.51 x 10 6 _ Column 108 2.941 x lo st (Vertical Load) (1) N-S beam-(olements 46-55 in figures) (2) N-S beam (elements 56-57)
~(3) N-S beam (elements 115-116)
(4) E-W beam (element.130) (5) E-W beam (elements under floor slab) [ g
.. 8-10 ,
. . . - - - . , - - . . - - . . . . - - . . :....-.--.-.--.- ~_- - . . - - - -
. _ . . ~ . . _ . _ _ _ -_ . _ _ -. -
r Table 8.5
~
~ Critical' Shears and-Moments for Dead Load Analysis Case 2
( -- Plate , . Element Node O(#/in) M y(in.9/in) M x II"I/i"I . 98- 247. 972.3 . 63 183 ' .0126 x 10 6 19-28 127 15 .0331 x 10 4 6 60-69 192 - 108.6 .-.0102 x 10 69-78 205 390.7 .0191 x 106 I 6 78-87 218 304.4 .0174 x 10 i Beam Node Q(#) M - (in.9) - M (in.9)
- Element' y. x i
l (1) i 46.y) g 166 .0692.x 10 6 2.698 x 10 6 49 169 .2546 x 10' 8.93 x 10 6 L 57 177 .0119 x 10 6 1.71 x 10 6 116 g4)_ 268 .003 x 10 6 .394 x 10 6 L 316 .0008'x 10 6 .058 x 10 6 130(5)- . , ! 102 251 .2009 x 10 6~ 9.01 x 10 6
~ Column
. - 108 .605 x 10 6 (Vertical Load) l-l Since the rack concentrated loads adjacent to the fixed
-South wall are not included in the ANSYS Model, the localizec'
' shear forces 'are directly added to the ' results in Tables 8.4 and 8 . 5 .- Based on a static ~ weight of 742 pounds per loaded cell, and assuming that each-localized load is reacted entirely.
- j. .by_ ' the adjacent slab sections yields the following incremental
. shear loadings-to be added to Tables 8.4 and 8.5.
l i-8-11 y , er W 9 g- ytw* 9e9-,,.g e p,,, y g-. w-r
Table 8.6 Incremental Additive Shear Force Node AO (#/in.)
-140 1226 218 1450.6 The following Table 8.7 shows a. sampling of comparisons of actual and . allowable loadings at various locations in the pool floor.
Table 8.7 Critical Pool Floor Structural Integrity Checks Item- Condition Actual Allowable Load Load s Slab bending 1.4D+1.9E . 100x10 6 in.#/in. .305x10 6 in.=#/in.
.g- -
D+T 0 +E' . 104x10 in.8/in. .305x10 6 in.#/in. Beam bending Not critical since allowables are an order
.of magnitude greater than applied loadings.
Beam shear 1.4D + 1.9E 1.52 x 10 6# 3.31 x 10 ,6 (N-S beam) Slab shear 1.4D + 1.9E 3223 t/in 5103 t/in (away from concentrated load. point) Slab' shear 1.4D + 1.9 E 6483'#/in 10200 t/in (near concen-trated load point along
- South wall) -
W l 2
.-+m# (, . -..-,s ,,,,,-r,- - . , ,es., - , ,- , . . - . . . - - i ---w,- -m,, . - - - , - . . - % +-r=r-*-i---,-
l I l 8.5 Conclusion O The pool floor has been shown to meet' all structural i
. acceptance requirements when analyzed as a concrete slab-beam >
s
. configuration. . Conservatism has been built into the analysis by t
assuming that the slab cannot transmit moments to the east, - w est, or north walls of the pool structure. In addition, no l L credit has been taken for the effect of hydrostatic loading on the walls which causes a load reducing uplif t on the pool floor
- . slab. .
. l t
I l O l
. I L
l 1 I . 1O 4 L_ 8-13
~
i O . I i O p' ' - O?.- i y i ' iy ll - tr M - Il I'..._'( I?F 1l ll ~ ~ l
's"' - --
7 '. 1 l j
,_ _ _l ' t
..li 1 l i
' 8 l
3 I I .i N.. - t_ g i. . ; l ie ; I
/ ' -
t } l I*! ' '
_- __' ! i i l l ~,., i i .
g
, co n I I , [I 3, ,i NN 4
I i ' i N l' I - i l :
,____LJ _ A-Lj'----6',
i_L---= _/ - M._wgiif.J- -- l
-RE
, ,- SPENT FUY 3 _('i T
- Girder} -
I F-' - g 3 POOLi g i I g- .,
) i i : -
6-0 + h 1
/_ _ } .
( REACTOR i lI i . l *- G l
, --_+u f _irder 1.
! =, 7 _ _i __ __ q_{___._ 1. ..
-i r -po A
1 _ ___ _ _ .
"e , i p ,-i- T u ---
q ,;
. ( 8 I is I Il -
iil i, =
' l i
.'h I
[Whils I l -* ll i N 4'~ 1 p i i I i <ii; ! I
\
i i s 8
/ I g 80 "
I l lt
--L _ 'Mr _ _ _ _l g - L _ _ __ _I '
~__cr__/
- 1_c ___ ___LU__ _ __A _
m _ . i: -
- - RC
__ _ _ t __ . ti,,i i
,i ii :.__
u > y _ _ _ _J i i 4 , L J i,i I ii ' ,t _.i r - - - - J - i - - -- J l il l!- _ _ i l.I
"- J
( -- - ' L .J +. R3 R4 R5 R6 R7 FIG 8.1(A) PLAN VIEW - REACTOR POOL AREA
1 O O O . i I
~
l' SPENT FUEL " 6 .
. STOR AG E . POOL ~ ~
STN STL LINER. ~
- ~'
l i Pool. Wall EL,79-Id'
- I
- 'y I ,
!" Ea st-Westj i t~ ~~~~~~~~
i " Girder 'l 'I ~~~~', -1 m .I ' ! ., t_. ._J g 3_-/' 9-0 .' a i
~
l f P y 8 l
- .e .s +-+-
i . l SECTION A-A FIG. 8.1(B) SECTION THROUGH CENTRAL N-S GIRDER
O O - O i i i
, s' d' ,
^
g ^- l , . . Y~ i i, E L.7 9-id' i i ' * ! . e a i a i # a
! I I !
i i
- J l ,
r-3 g- - - - q p--- = i i u
- p . ,
I i r- - ' 'O l 7g l g- - -
- - - _o,_ _ _ _ _ , .
i 1 l i u) , i <r , 4
D N-S Girders l
RD RE -*+- RC RF i l SECTION E-E
- 4 i
! FIG. 8.1(C) SECTION THROUGH E-W GIRDER
i t i19 C Mb
*X o n
U . 11 9 12 0 121 I?? 123 l'24 12 5 IE8 12 6 12 7 iia _c _ ie _ 7 77 y 7 7777( 96 97- 98 99 10 0 101 0 103 104 87: 88_ 89 90 91 92 9 95 9 5 78 7.9 80 81 82- 83 85 8 6 69 70 71 72 73 74 7 76 7 7 60 _ 61' 62 63 64 65 6 67 68 7XXXXX X X W ee _e e, .
~ ~ ~ ~
f37 38 39 40 41 42 4 44 4 5 28 29 30 31 32 33 35 3 6 19 20 21 '- 22 23 24 2 26 2 7
.c ..
- /^&
. . v o
- CROSS H ATC HI N G INDICATES GIRDERS i O FIG. 8.2 a ELEMENT LOC ATIONS 8-17
.si s e t s . s >
m y i L 7303 18 g-
.O .
290 o 270 RZ3 2,25 P.77 - 279
'"*~
eY XXXXXXXX . . XX XX XX X X X
~
23, 234 // . 2,8 a2, >3 / / n.
~
20s 1 2e. //
. , 92 ,,, //
o ',- - 'c- 'e' // - ', X X XX XX X X'X ,, s _,,e m XX XX XX XXX
~
4Te
~
,e2 14 0 , ,14 3 ,, ,, .14 9 127 .13 0 13 6
.11 4 .il7 12 3 s -u 1 roi toa io3,,lo4 ios,,ios
/ /
107,los 109 iso 0~ - -
. (s.s > J ,,
, '267 . .,
e 27 ; = 18 - l CROSS- H ATCHING lN DIC ATE S GIRDERS O- POINTS OF CONCENTR ATED LO AD APPLIC ATION FI G. 8. 2 b NODE LOCATIONS 8-18
~ ~
~
FIG. 8.3 FLOOR LOAD g SUM OF.SUPPO.RTS a
., 8,
'l l
I
)
t o g l l
? N i , , I o
.sg Jg .,
E $_ 8' ' Eg N '
'0.00 2'.00 4'.00 e'. oo a 00 io.co i2.00 .i4.oc
0 0
'4 (j i 0
0
'2 i
0 0
'0 D '
i A O L 0 0 R 8 O , O . L 0 F 1 0
'6 O 4 T
R 8 O 0 P 0
'P '4 G U I S Y ll F 0 0
'2 0
0
@? k 9
!$ '0 "a
- r c
O 4
O. O - ~ G FIG.. 8.5 FLOOR LOAD SUPPORT 2
~
S
.o "o
7'
) I.
i j
* ; W .
S g l m ' i k l I I ' to
-' , _j o
4 3- i l ' db.- , t O i 1 o8 J i o
. CS_
-os
/ o ! _J i iS o'
- '0.00 2'.00 4'.00 6'.00 8'.00 i0.00 i2.00 i4.0C
- TIME (SEC)-
l 1
-M
" O 7
A. ,. T w
,i ).
.V O
---===- -5 B
. a_ O O
m C "
'4 ~
M g
.j M g
en
.A - O
~
O (. ~
~k o m- N h a W r
.__f .-
1 o u_ . cc- --
~
O-m O - cd U UJ O_ O_ CD D- - - - - - - - , - W .. w ' _v H- w Lt --
--===._-.------
g O O 00'0d- 00'Od- 00'00I- 00'0PI- 00'08T-c0TM (81) 0V01 80013 8-22 e
..x. S4
~
C
~
O' Q
~
F' I G'. -
!8:. 7 F L O'O R LOAD +
i ~ L
. SUPPORT J4
- , p l
8 - I a -
. m. -
O '-' - l 4
.. w I
- l
[ 8 h ; L a Ng I
,d r l l W b.
I I f I f ll
- =
! g.E - l a2 . i, I ) .O - '
<C =
i O$ j' -Jg i cr$_ 4 O.i . o - J j' LL.$ , 2'.00 0.00 4'. 0 0 - 6'.00 8'.00'. i0.00 i2.00 I '4 . 0 ( ' ; I. TIME (SEC) 1
4 T
References:
' [1] . American,. Concrete-Institute, Code Requirements for
. Nuclear. Safety-Related Concrete Structures-(ACI-349-76)'.
- [ 2] - U.S. Nuclear; Regulatory Commission, NUREG-0800, July,
' 1981.-
~~
-[3} "Swanson[ Analysis-Systems, Inc., ANSYS-User Manual,.
, : February 1982.
6 O A e
~.
e d O ~ e I O 8-24
9.0 ENVIRONMENTAL-EVALUATION f3.y/ . 9.1 Summary
-Installation of High Density Spent Fuel Storage Racks at
- Oyster _ Creek Nuclear Station (OCNS) will increase the licensed storage capacity of the spent fuel from~1800 to a maximum of 2600 assemblies. Radiological' consequences of expanding the capacity have been evaluated with the objective of determining if there is I significant additional on site or off site radiological impact l -relative to that-previously reviewed and evaluated l. In -
addition, radiological. impact to operating personnel has been l evaluated to. ensure that exposures remain As Low As Is Reasonably l Achievable (ALARA).
-The. decay heat loading and the radiological burden to the spent fuel pool water are determined almost entirely by refueling operations. The frequency of refueling operations and the
, (N conduct-of refueling are-independent of the increased capacity of the storage pool, except that the increased capacity will reduce ! fuel movement and allow continued' normal operation. Since the
' fuel assemblies which will utilize- the bulk of the storage
. capacity (and will ultimately fill all incremental capacity above that of the existing design) are aged, their contribution to either- the peak ' decay-heat load or the - increased radiological impact,:in terms of increased-dose, is negligibly small. A study performed by the NRC 2 supports this conclusion. Consequently,
;the-increase in the storage capacity of the spent fuel pool will neither significantly - alter - the operating characteristics of the current pool nor result in a measurable change in impact on the environment. c-i I
9-1
F 9.2 Characteristics of Stored Fuel
.The currently authorized storage capacity of the oyster
. Creek. spent fuel-. pool is 1800 assemblies;'when fully. loaded,-the pool would contain . the 560 assemblies of a full core discharge
- and 1240: spent fuel assemblies with cooling times. ranging from 16' months.-t'o about 13 years.
Af ter _ the currently planned expansion, an additional 800 fuel assemblies can be stored in the expanded capacity racks with
- cooling times greater than about 10 years, based upon a fuel cycle. duration.of about 16' months.
Reduced fuel burnup or' increased cycle length would result
. in a . lower fission-product inventory of 1.onger storage (decay) periods respectively. Thus, the assumed storage pool composition should result in a conservative estimate of any additional thermal or_ radiological impact due to the- expanded storage
. . capacity.
, JI Because of radi'oactive decay, the heat generation rate and
- the~ intensity of_ gamma radiation-from the spent fuel assemblics
; decreases substantially with cdoling time. . After a cooling ti.me
. of about 4 years,3 the. decay heat generation rate is less than 2%
of . the rate at - 7 days--the nominal time at which depleted fuel assemblies are transferred to:the spent fue'l pool. The intensity of-gamma--radiation'is verylnearly proportional.to the decay heat
'and decreases with cooling time in a similar manner, i
The bulk' of heat loading is due to freshly discharged fuel; noreover, aged fuel - contributes very little to the total
. heat load. Therefore, it is not expected that this expansion will significantly increase the thermal dissipation to the environment. Since the intensity of gamma radiation follows the O
9-2 O w *+,.---,.-,4 -,4- .#
+ -c
-decline in'. decay heat generation' rate, it is:similarly concluded thatLthere would be' no significant increase" in . gamma radiation O due to the expanded. storage.
U
-It is important to-note that the aged. fuel in the expanded storage- ~ capacity will not- contain significant amounts of
. radioactive iodine or short-lived gaseous fission products, since these would have- decayed during the refueling period. The
- Krypton-85 which might L.escap's from defective fuel assemblies has been shown - to ' d'o so qu'ickly 2 (i.e. - within a short time after discharge from the core) . .Further, the residual Krypton-85 will be contained'within the fuel. pellet matrix and hence any leakage would occur . at very low rates 2 Cesium 134/137 2, is strongly
~
bound lwithin the fuel pellet matrix and its dissolution - rate in
. water is extremely small. Any cesium dissolved in the pool water
-isleasily. controllable in the' clean up system (demineralizer-ion exchanger ~ resin . bed) 2 Thus the planned storage expansion will not"significantly. increase the release of. gaseous radionuclides.
9.3 ,'Related Industry-Experience Experience with storing spent fuel underwater has ' been '
' substantial 2,3,5, These references show. that the pool water l- -activity, normally low, during ~ refueling periods experiences a small increase which decays rapidly with time. Typical
~ concentrations" of radionuclides in spent fuel pool water range from 10-4pCi/ml, with 'the . higher value associated -with refueling operations. References 2 and 4. also state that- the increase- in- pool water activity during refueling can be attributed'to:
L L a. dislodging (sloughing off) of. corrosion products on the h fuel assembly during transfer and handling operations t l O
~
i 9 9-3
- . . - . . , . . . ~ -
6
~ b. the- possible short-term exposure of fuel pellets to pool water via'a' cladding defect,+and-6
. c.' mixing -of = the spent fuel pool water ' with the ihigher
- activity _ reactor- coolant. Upon cessation. of the refueling ~ operations -the fuel pool water. and- the reactor coolant " system . would be isolated from each-other, 'thereby terminating transport of corrosion
. products from the Reactor. Coo.Lant System. Thus, deposition of- crud' is a function of refueling operations.and is not impacted by the expanded storage.
~
Furthermore,.it has beenJshowns that release of fission ,
- products- from failed fuel decreases rapidly after shutdown- to essentially. negligible levels. The fuel pellets -are made of _. inert Uo2 that have very. low solubility.in water and the propensity for corrosion of the cladding (Zircaloy J2) _ at spent fuel pool water-
- temperatures is virtually. n '.12 ,* . -Thus the .only mechanism available -for~ the release of the gaseous fission products is diffusion through the UO2 Pellet.
It :has been shown that at. low water temperatures (<l50*F) the' diffusion coefficient is extremely small 7. Therefore, the small : increase 'in activity of the spent ,
, fuel' pool water is .due to either crud transport, fission _ products -release,- or cross flow from the
. reactor coolant- system and is only a function of re. fueling operations. . It is reasonable to assume that the' increased ' capacity of the spent fuel pool will
-' reduce fuel handling operations, since fuel assemblies
~ will- Lnot - have - to be ' consolidated or shipped for an z extended period, thereby reducing the probability of
. increased pool water activity due to crud dislodging.
Thus,._- the expansion of fuel pool storage capacity will O 9-4 P twe+9 -- ,+r+-? g* +-o,g.- -,q-M-g ' re y ,m- *gy.-gp,. e w s h--g..mvm-g.-p--em-,y -%,Ae*-we--e.--- *.wm~ ,,.m,y-grg.cso.,qwa=-+em*gwei--n-wha.' e& ww+-me- e ve w w -rewemew,-m w w='
not:causo a'significant increase:in dose either on site or off. site. l G The corrosion properties of irradiated Zircalloy-2 cladding have' been reviewed ~1n References 2 and 5, and the conclusion. is- drawn that the corrosion 'of
. the cladding in spent - fuel pool' water is negligibly l sma],1. : The' minor increment'al heating of pool water,
~ due ' to' the expansion ~ of storage capacity, is far too
. small to materially af fect the corrosion properties of-
- 2ircall'oy-2 cladding. -
-9.4: Oyster Creek Operating Experience 4
Measurements have been made of the . principal radionuclide concentrations: in .the . Oyster Creek fuel storage pool with '980
-spent fuel assemblies in - storage.. Table 9-1 summarizes these
- o -measurements.
A-V
' Table 9-1 Observed Radionuclide' Concentrations in Spent: Fuel Storage Pool Water'
' Measured (uci/cc)
,. . Nuclide^ Date: May 31, 1983
-6 Mn-54 5.59 x 10
-5 Co-60 - 8.56 x 10
' 5 Cs-137 5.4 x 10 Th'ese observed radionuclide concentrations are generally
. comparable ' -to- industry experience in other spent fuel storage O
O . 9-5
-3 i
sk---t,y $ grws.,---c-e,- - - vs seg,r, ..y., ,p=m +.g,ww.we.- e-,me.-,-w ,9, -,e- e --e ,w-.,-w. ..w9.-.e,-e, y,a y*--*wez --my -,,,-e,w -,1,'e,,,,s.- -e mm..
pools.. Expanding the storage capacity of the Oyster Creek storage pool.is not expected to significantly alter the general g -magnitude 'of- radionuclide' concentrations, 'since the contribution from the aged fuel'will be.very low or negigible in comparison to that from recently- discharged fuel or from primary system carry-over during refueling. 9.5 Spent Fuel Pool Cooling and Cleanup System (FPCC)
'It has been shown-previously (Section 5 and Oyster Creek FDSAR Amendment '78) that the cooling system at. Oyster Creek is adequate to handle the expected . heat loads and maintain the temperature peaks within acceptable limits. It has been shown earlier in thi's section that the smal'1 increase in heat load due to the storage capacity expansion will neither significantly increase the thermal dissipation to the environment ' nor increase ,
the propensity for corrosion of the cladding.
/ It has also been 'shown that the crud deposition in the k) spent fuel pool water occurs during refueling outages and that the-planned expansion will.not increase crud deposition. The fuel pool clean-up system (filter and demineralizer) is designed to
- maintain fuel pool' water clarity and is operated and maintained in accordance with Oyster Creek plant procedures. The clean-up
! system takes a surface skim frem the fuel pool and cleans it through-a process of filtration and ' demineralization to prevent I crud build-up on the fuel pool walls at the water-to-air interface. .When necessary high press'ure water spray is also ! utilized, i l, The spent fuel pool water is sampled and analyzed j- periodically to confira proper operation of the pool clean-up j system. The- frequency of filter and resin replacement is L ' determined primarily by requirements for water clarity rather j than the loading of fission products radionuclides. i A) L , 9-6
The -- fuel; poolH demineralizer contains 150 cubic feet of
' ~
; ,. y mixed;bedibead type' resin. The cation resin is a~strongly acidic,
-(f highly cross linked,- sulfonated styrene-divinyl benzene-icopolymer. The anion resin is a ' strongly basic, -quaternary
- i. ammonium-poly (styrne _ divinyl benzene) resin.
The-quantity-ratio of cation to -anion T resin ' is . between 2 to 1 ' and .1 to 1. The demineralizer is instrumented with a conductivity - monitor which
. sounds.'an alarm in the control room on high' conductivity.
The' : fuel -. pool. filter consists . of plastic filter elements with ca precoat 'of filter material. The filter material is a mixture of ; fiber and' powdered resin. The fiber is cellulose -or plastic. The powdered-cation resin is a-sulfonic acid resin and
~'the' anion 'is'a'. strong basic resin. The ratios of fiber to cation
.to anion are adjusted- to optimize fuel pool water-treatment; requ.irements. . The filter is monitored for pressure drop and alarms =to control room on a high differential pressure.
The-present---annual quantity-.of solid.radwaste generated by the~ Spent Fuel Pool Purification System 'is about 230 ft3 The SFP mo ?ification is not' expected to result in a significantly higher que.:.*.ity _of solid radwaste. Table 9-2-summarizes the sampling frequency for the various parameters associated with .the spent fuel pool clean-up system cand lists pertinent limiting specifications. Neither impact on
-the-existing. procedures nor a.significant. increase in activity of l
the clean-up system f11ters-or resins is anticipated.
!9.6 Fuel Pool Radiation Levels n The measured radiation dose at the 1-foot level above the
^
Oys'ter Creek pool ranges from 7 to ,100 mr/hr (May 26, 1983).- Higher:. radiation dose rates above the pool #ce expected during i refueling ' operations, decreasing soon after completion of refueling.to the 7'mr/hr range. l . i I ' 9 E 9-2 ROUTINE SYSTEM ANALYSIS SHEET NO. 10 ANALYSIS FREQUENCY LIMITS SYSTEM PARAMETER REG. NORMAL LIMIT SET BY REQUIRED ACTUAL LIMIT OPERATING RANGE Spent Fuel Pool pH NA Monthly 5.5-8.0 5.6 - 6.8 Plart Proc. Conductivity NA Monthly 5.0 umho/cm 0.5 - 1.6 Plant Proc. Chloride ion NA Monthly 100 ppb <20 - 65 ppb Plant Proc. Silica NA Monthly 1.0 ppm <20 - 600 ppb Plant Proc. Sus. Solids UA Monthly 1.0 ppm <20 - 500 ppb Plant Proc. Gross Beta NA Monthly 1.0E-lpCi/ml 2.0E-4-5.0E-3 pCi/ml Plant Proc. Gross Alpha NA Monthly None 1.6E-8-2.0E-6 pCi/ml NA 02 ~ Tritium 'NA Monthly None 7.0E-3-6.0E-2 pCi/mi NA Gamma Isotopic NA Monthly None IE-4 -lE-2 pCi/mi NA Make up System Carbon filter Chlorine Res. NA Monthly <0.'1 <.1 ppm Plant Proc. ! Make up Sy, stem Cation Demin. pil NA Monthly 3.0 - 6.0 3.5 - 4.5 Plant Proc. i Make up Anion Demin. pil NA Monthly 7.0 - 10.0 8.0 - 9.7 Plant Proc. 1 Conductivity NA Contin- 100 umhos/cm <10 umhos/cm Plant Proc.
- uous Chloride ion NA Monthly 50 ppb <20 - 40 ppb Plant Proc.
Silica NA Monthly 100 ppb <20 - 500 ppb Plant Proc.
p, . , - i
'Because of ? radioactive - decay, the total . contribution of allithelagedifuelto the dose rate at'the pool surface by direct 1
s )-[ cradiation will~ bela very . small . ( <4%) increase over.that from the more-recently discharged . . fuel. .Since 'the pool ' water affords
. adequate :shieldingnand no significant increase in radionuclide
~
it is concluded concentrations . in :the pool' water ' is expected, 3: ithat: the . . occupational dose rate above the surfac_e of the pool
- from direct radiation. will be essentially the ~ same- as that for
~
itheicurrently_ authorized storage pool. . c: The measurements were made' above the cen.ter area of the
-storage. pool:and at-the pool edge.
Higher readings which range
- from :25- to 100 mrem /hr are obtained at the . pool edge. Normally f :such'readingsfindicate-the possibility of crud. depositions on the L ' pool walls. However, as pointed out in Section'9.4, operatin'g
. experience.Lwith the --pool? clean-up s y .mem - has confirmed the absence. of. any _ significant crud buildup. - These higher measured -
! radiation '. doses at the pool edges can be attributed to other 4 ' sources. 'R'adihtion dose rate measurements were subsequently made
< of .- equipment' -temporarily stored inside the pool. These
- measurements show high dose rates of up to 20 Krem/hr from
. equipment such as_ _ control- rod ' blades, ; buckets and other waste
- material ~ which are submerged close to the walls. These latter
.' measurements'-explain the higher readings recorded above' the ' pool Lat1.the. pool edge.
Radiological- assessment around the perimeter of the spent. Jfuel shield--wall has been performed at 75'~and 95' elevations on
~
+- 1, May-21, 1983. The radiation level ~is less than 2mr/hr around the _ perimeter--at 95' elevation and along the outside of the North and
-. East- walls at175' elevation.
Along the west wall at the -= 75 ' e elevation the radiation ~1evel is higher and ranges from 25.to 60 mr/hrt the major source of this radiat. ion is the spent ~ fuel pool
~
I heatiexchangers. _ Measurements around the . perimeter of reactor j ' shield 1 wall also indicate radiation levels of less than 2mr/hr. Al ' Thei minimum : water :-gap between rack and ~ the pool wall for the exi~ sting' rack! and for the new racks is given in the following: v
'9-9
. Minimum Water: Gap?be' tween Rack and Pool Walls North South East West Existing : Rack's- 6.1/4" 9.1/4" 11 55/64" 5 39/64"
.New Racks.
-3 7/8" 1 1/2" 4 13/16" 2 1/2"
.The pool; wall consists of 6 feet thick of concrete shield. -
Hence
~
~ ~ the -~ decrease' in the . water gap of between 2 3/8" to 7. 3/4" will
; yield . insignificant ' increase in the radiation level around the shield-walle .
Measurements - made in the fuel building area failed to detect- ~any.. -Kr above the . minimum detectable concentration of 8
.2.5x10- DCi/cc.
-In view of the above, it is. concluded that the additional
. storage capacity of. the-' expanded spent . fuel pool wi,ll not
- measurably alter. the current - -level of radiological impact or
~ significantly alter the radiation dose to personnel occupying the
); fuel' pool' area. .
9.7' Radiation Protection Plan The~GPU Nuclear (GPUN) Radiation Protection Plan is based-on the regulations of the Nuclear Regulatory Commission (NRC) as
. contained in . Title 10 ~of the. Code,of Federal Regulations, Parts 1 9 ,- 20, .50,- and 71, and appropriate Regulatory Guides,
; specifically1 8. 8, Rev. 3 (1978), 8.10 Rev. 1-R (1975), 8.13, Rev. .1 (1975).- and 8.15 (1976)~. Specific details as to how the
'GPUN Radiation Protection- Plan is implemented shall be
. promulgated in the plant specific Radiological Controls Procedure Manual _ (RCPM) and shall include those applicable procedures
' addressed in Regulatory Guide 1.33 Rev. 2 (1978), Appendix A, paragraph 7, and paragraph 8 (aa), (bb). This GPUN Radiation Protection' Plan is - the ~ first .part of the RCPM. The total
' radiation exposure to personnel from 'all operations in the fuel pool area was approximately 21.750 man-rem from May, 198.2 to May,
~
p 1983. 'The operations performed were as follows: 9-10
- List of . Operations :in Fuel Pool ' Area' from ' May, 1982 to May,fl983
~
;A-)/
~
-1. . ' Removal - of spent' fuel racks, safety curtains and Control
-Rod Blade R'acks.
- :2 . : l Removal of equipment support stand.
-3. Underwater cutting of LPRM's.
- 4. Transfer. of liners- from pool into casksJ (some were.
transferred outlof water). 5.. Transfer control ' rod blades to . temporary _ storage, i6.- Removal o5 crusher shearer stand.
~1 . 'R'emoval of. jet pump system..
8.- -Relamping' underwater' lights. 19 . Removal and packaging of rad-waste.
- 10. Pack load and ship NAC-1-Nibbler Stand.
- 11. ; Package and ship Control Rod Blade Racks.
.12. Underwater TV maintenance.
- 13. -Removal and replacement of Fuel Poo7 Gates.
- 14. Installation-of New Fuel Racks.
n
- 15. Removal of tools from pool.
v-LArea1 radiation ' monitors provided . continuous surveillance-of radiation levels.. . The low cumulative radiation dose due to the' operations listed above' is indicative of the expected experience during the proposed expansion. 9.8- Re-racking Operation The-existing spentifuel racks willLbe removed, and the new racks - will be installed
- in' a : manner - which will minimize the-environmental impact and maintain. occupational exposure to levels As: Low As~Are--Reasonably Achievable (ALARA).
- e n
m s 9-11'- -
s . TheLfive' major' steps in this reracking' operation are: (i) ' Removal of spent fuel from existing racks.
'(ii)J Removal'of existing racks-1 (iii)iInstallation of new racks (iv)' Decontamination and-disposal of existing racks.that' are: removed..
; .(v) Loading:the new racks'with spent fuel. assemblies.
.In accordance with .'the basic- guide- lines of Regulatory Guide ~ 8.8 and other guidance _ as f given in the GPUN Radiation
' Protectioni . Plan, . for maintaining occupational , exposures ALA'RA, th'e fundamentals of time, distance and shielding will be employed at~every step.
Nine' ^hundred and eighty_ -(980) spent fuel assemblies will
- be.. stored. in the pool before- .and during~ the re-racking
~
- operation :using -- existing -fue1 handling equipment, fuel will be moved in.'such'a'mannerLthat the number of--handling operations and V :
di' stance of movement are_ optimized. .No. divers will be used, and-
~
therefore', shielding for - underwater human -participation in . fuel-
- movements 1is not a concern.-
- ,- In order that the time spent in the. radiation field of.all pc .onnel~.is such as to maintain ALARA exposure levels, each step
- l. will' be planned.- Procedures including organization -and
- responsibilities, , administrative procedures and detailed work
~
procedures wi11 ybe developed _ for each ' step to _ provide the most efficient accomplishment of each effort.. Each step will have j> f multiple' . signatures -prior to proceeding to the next to assure a
; . smooth re-racking operation. Radiation surveys will be taken at l .. each. location where work activities must be performed.
7 Q 9-12 J
.,,,ww,ww~,y-,,v~,, w. ,- ,,,% y.,,,-,,,.. ..:,
The ~ general location of . the existing racks is ' given.. in
' Fig. 9.1. The module layout for the new racks has'been presented inE Fig. 2.1. LIn the first stage rows 4~and-5' of the existing - , racks.will be removed. Each rack assembly will be moved in a
- designated path. The racks will be lifted off the floor, decontaminated:by-hydrolaser and rinsed with demineralized water, and then removed ' f rom the pool. Prec,autions will be taken to prevent movement of fuel racks over fuel racks containing nuclear.
fuel.' .
' ~
iThe decontaminated ; old racks will finally.be shipped for burial, ^
.or sold.for scrap.
Our preliminary installation sequence calls for installing new . racks G , .H , J an'd ' K .. af ter removal of rows 4 and 5 of the existing' racks'. These new racks will be handled and installed
'using J, remote. handling devices and no divers < will be involved.
-Once'these new racks have been installed fuel-assemblies will-~be L loaded into it. At.the' appropriate time, the remaining fuel will
. .- be relocatedi to f acilitate . removal of rows 2 -: and
~
- 3. In the
.second- . stage, the- .new - racks: B, D, F Lwill be installed.
! Eventually,Erow ~1 will be ' removed and remaining new racks A, C
'and.E will be' installed.-
l - H . l l t l.- l~ 9-13
k (~ L). 3 d- o
= I I DE/ IN SIDE (REF. ) . =
rs v
~ ~_ -I -
_ _ Mr- y
'K-F s /
- L e
N ' ' K , CASK,7 ( - L AY D O W'N
'! A k 'A i, '
- I
- /
,k s
i ' u
, a
- ,- , w
. /
N 9 c L-- .. m I N w a ROW I ROW 2 ROW 3 ROW 4 'RO W 5 O
' .s N
, , i, i e FIG. 9 I, EXISTING RACK LOCATION S
-Theoverall re-racking operation (tentatively scheduled to 7 '-";st' art in; June, 1984'and-completed.by February, 1985) is expected to . require '- five ._ personnel .
The ' five major steps . listed at'the
'beginningl of th'is subsection, are estimated to result in
.approximately- 25 man-rem total; exposure, fairly equally distributed between the steps.
e
'9 . 9 ' ~~ Conclusions
- Based upon the industry _ experience and evaluations discussed in previous s'ec tions , the. following conclusions are made.
o' Minor increases in radiological burden to .the - pool water, if any, can be adequately handled' by the fuel pool cleanup system (filter and-demineralizer), thereby maintaining the radionuclide concentration in the. water
, at'an acceptably. low level'.
~
o No appreciable increase in solid radioactive wastes (i.e. . filter media and demineralizer resin) is anticipated, o --No increase 'n i release of radioactive gases is
- expected, .since any - long-lived inert radioactive gas 1 potentially_ available for release (i.e., Kr-85) wilf
~have leaked from' the fuel either in the reactor core j_ 'during operation or during the first few- months of residence in the pool. Furthermore, Ref. 2 (pp. 4-16) l: has shown airborne activity to be considerably- lower than. that allowable.;by Table 1 of 10CFR Part 20, Appendix B. Therefore, the planned expansion will not l
1-significantly. increase - the release of radioactivo
. gases.
f a e 1 , [ 9-15
4
~ o ;PersonnelJexposures resulting f rom crud' buildup along I
1 1the walls of -' the : Fuel' Pooli have been - insignificant in past experience.- The ' reason ., for this condition is . that' there . is no ~ accumulation of. radioactive contaminants
. calong the ' walls - of- the Fuel Pool' at the water to air
- interface.- - Accumulation- has ~:been prevented : due to an L - ef fective l surf ace skim design in ~the Fuel ' Pool - Water -
' Cleanup Systems-plus utilization'of.high pressure water -
~
. . spray' as . required .-
(Consequently, there ~is no. reason _ to ' expect crud . buildup' in ' the future: with increased pool capacity.) 9
. o: .'The existing radiation protection monitoring systems
.and . program' are adequate .to. detect. and warn of any unexpected: abnormal' increases -in radiation level.
.- This provides.. sufficient assdrance that personnel
. exposures can be i maintained as low as is reasonably t' , achievable.-
O . o 'No increase . in- corrosion' . of Zircaloy cladding is I
" expected,'.and there.is-sufficient evidence of-long-term fuel integrity to accommodate increased storage capacity.-
f ! L o The . totallexposure to. personnel occupying the fuel pool ;
~
l- ' area for all operations from May, 1982 to May, 1983 was
'21;75 man-rem. -No significant increase in personnel exposure is expected as a result of the increased storage. capacity.
F i ; I. l _. i L 9-16 ( L m g .nn--, v-, u -we -,,~-,re.,e. e,,- ,.,r-e ,,v,.m-,,.,,.---,-y_,..,w,w,,,.,*w,.,,e,,,,,,,.,--n
o - Expanding .the storage - capacity of the spent fuel pool
~
i
~
will-not significantly increase the;on site or off. site-
- radiolo'gical _ impact above that of the currently authorized storage ' capacity, nor is any significant
' increase- , in environmental radiological or non-radiological impact anticipated.
6 9 F 5 4 0 . O l' D . 9 . On 1* 9 9y+- -4 9 r- ---ta e' hP - 9"-6"~P*"e--* W W meP-r="'hW-"-*'$= -w yw94 '*We F--e-'--m'-N'Oe "wv-'~'9r- dW--"'wW"-wW-N'" =*9'#w--t*m v=--w--M*""#'--FwP--w'NMh8'e-Mw&
.._,.. . . ,~ . - .. .
4
'O
.w.
Q. . REFERENCES-TO'SECTION/9 _ y b j :- c
' "FDSAR, Amendment No.178",' Oyster Creek NuclearfStation,.
~
, 1
- 1. .
,1 Docket No..50-219.
~
! '2. ~NUREG'0575, " Handling and: Storage of Spent Light Water Power Reactor: Fuel, Vol. 1, Executive Summary-and Text,
.USNRC August 1979..
- 03. NUREG,0800,.USNRC Standard Review Plan ---Branch Technical ~
' Position;ASB9-2', Rev..'2, July.1981.
~
- 4. ,'
A'.B. Johnson, Jr., " Behavior of Spent Nuclear F:lel in- _ WaterzPool1 Storage,:'BNWL-2256,LSeptember 1977. 5.- J.R. Wee's,'"Corrosionlof. k Materials in Spent Fuel Storage
= Pools", BNL-NUREG-2021, July 1977.
- 6. . : J .M.-.. Wright , " Expected-Airrand Water Activities'in the Fuel Storage Canal"'.WAPD-PWR-CP 1723, (with addendum) undated.
~7. . cANSE5.4 Proposed Standard,." Method for' Calculating the Fractional Release of Volatile Fission Products from Oxide 1 Fuel",- American isuclear Society, issued: for. review 1981. -
~8.~ GPU Nuclear. Corporation Radiation Protection Plan,
- Document No. 10'00-PLN-4010 .01.
J B 6 4 4 l '.,' , e 9-18 [d) y
' O O
. 41 0 . INSERVICE SURVEILLANCE ~ PROGRAM FOR BORAFLEX NEUTRON ~ ABSORBING MATERIAL r
10kliProgram Intent: A sampling program - to verify the integrity of the neutron absorber material employed in the ~high-density fuel racks in the long-term. environment is described-in this'section. -
~ '
The. surveillance program is - designed for the spent fuel pool sinc ~e the Boraflex used;in these racks will experience long term radiation. ,
- The : program- is intended ~ to be conducted in a manner which allows. access to the: : representative- absorber material samples without disrupting the ' integrity of the ~ entire fuel. storage
[ system.~ The program :is tailored to evaluate- the material in normal'
~use . mode, .and to forecast future changes using the data base developed.-
- O ;10.2-Description of Specimens:
. .The abso'rber material, : henceforth referred to as " poison",
used in the surveillance program .murst be representative. of the
' material; used within the storage 1 system. It must be of the . same composition, produced by the same. method, and certified to the same criteria- as the production' lot poison. The sample coupon must be of.similar thicknes's as the poison used 'within the storage system l ~ and!not ~ le'ss than : 5 - 3/4" x 3" inches on a side. Figure 10.1 shows l : a [ typical coupon; Each poison ~- specimen must be encased in 'a
; stainless! steel jacket of an identical alloy to that used in~the
' storage.. system,. formed;so.as~to encase the poison material and fix
! it' ini a position and with toleranceis ' similar to that design used for . the' ' storage system.-- The jacket has to be closed by tack l - L i I. pO
~
f
^
10-1 1
-.._u__., . - - - . . _ . - . _ _._..._. _. . _ - _ . - . _ . . . _ _ . . _,_._.,,_,_.m_..,_ _ . _ . . _ . , , _ .
rwelding : in . such a - manner as ~.to retain its form throughout the test
' period iand stillL allow rapid. and easy. opening without causing mechanical- : damage to - the poison specimen. contained within. The
~ jacket should permit wetting and venting of the specimen similar to theractual rack environment.
- 10. 3 . 'Te s t :
. The test-conditions represent thi ventid conditions of the box elements. The samples are toibe located adjacent to the_ fuel racks and' suspended from the. spent fuel pool wall. Eighteen test samples
~
j are . to be f abricated in accordance ,with Figure 10.1 and installed tin the pool when the; racks are installed. , The procedure. for fabrication .and testing of samples is as given below: a..The samples should- be cut to- size and- weighed carefully in milligrams. r J/] . b'. The length , width, and the average thickness of each specimon is to-be. measured and recorded.
- c. The. sainples should be fabricated in accordance with l Figure 10.1 and-installed in the pool.
d.. Two samples :should be removed at each time interval according to the schedule shown'in Table 10.1. I' 10.4 Specimen Evaluation:
-After the removal of the jacketed poison specimen ' f rom the
~ fuel. pool ~ at a; designated time, a careful ' evaluation of that
~
1 specimen ~should-be made to determine its actual condition as well. as ,its apparent durability for continued function. Separation of p -the' poison fromL-the stainless steel specimen jacket must be , -performed carefully to avoid mechanical damage to the poison specimen. . Immediately .af ter the removal, the specimen and jacket section should visually be examined for any effects of L-10-2 6
environmental -' exposure.- Specific ~ attention' should be directed to
- the _ examination :of ' the Estainless steel. jacket for any evidence of t-physical._ degradation. Functional evaluation of the poison material can be accomplished by:the following measurements:
- a. A 1 neutron. radiograph of thefpoison specimen aids in
= the'dete.mination'of'the maintenance of uniformity of .
- the boron distribution.
- b. Neutron attenuation measurements will. allow evaluation'of the continued nuclear ef fectiveness of' the -= : poison. : Consideration must be given, in the analysis of the attenuation . ~ measurements ,' for the
~
level of accuracy of such measure ~ments as indicated' by the degree of - repeatability normally observed by
. the testing agency.
, , c. A measurement of the hardness of the-poison material will: establish the continuance of ' physical and
' ~
- structural- durability.- 'The hardness acceptability
' criterion requires that the -specimen hardness will not exceed'the hardne'ss listed in the qualifying test document-- for. laboratory test specimen irradiated to 1011 rad s ~. ~ The actualihardness measurement should
' be - made ' . af ter, the specimen has been withdrawn from n the pool and allowed to air dry for.not less than 48 L
hours - to allow for a meaningful correlation with the preirradiated sample,
- d. Measurement ~ of the length, the width, and the g average. thickness .and- comparison- with the g pre-exposure data will indicate dimensional stability within ~ the ivariation , range reported in the Boraflex laboratory' test. reports.
10-3
. v --y, . e,, ..e.- ,+w,,m4-,w,,.y---+.-w,v.s,w.-,+--e-w+.,w+-..#-----,e,,--m,-e.,~r.,..w, e. - w ,- --,,--- - w--%ww-ee-er.-.e.-=.-e- --=w-w-m. in
1 A detailed - procedure paraphrasing the- intent of this program will~ be
- prepared ~ - for. step-by-step execution. of - the test procedure and:
Q . interpretation 'of the - test' data. i O e W O l: l 4 [..' i l 10-4 y - - - - - - , -, - y 9r --- .-%,,,3 3,9-- ,-,,,,.,w,-,.--y,-w., v. ,w.g..,e ,-,..w..,.w.-mw.,,,,.-w.,. -....-re.--.-,m.,%,----- - - - _ . - _ _ - _ - _ _ _ - - - - -
TABLE'10.1 _ . Time Schedule-for. Removing Coupors , Date Installed
' INITIAL FINAL WEIGHT PIT
' WEIGHT- WEIGHT . CHANGE PENETRATION SCHEDULE (mg/Cm2 -Yr) (mg/Cm2 -Yr) (mg/Cm 2 -Yr) mil /Yr .
l- .
~2 90 day 3
1 4 . 180 day
'5 6 1 Year i
0 _7 '
'\ 8 5 Year -
9 10 10 Year . 11 . 12 '- 15. Year: l 13 l
- 14. 20 Year
- 15-16 30 Year l=
17' , 18 , 40 Year 10-5 I. T t T' e k- 4 -- -
'-w-w- ur ' r v pTNTP -wwY'-9'-**N--T*Wtr*---sw*w
TOP (vD = 0 e m .
$ ._._(=
a. E o 2 . / A a ' g- TACK WELD / - l _ N ,-
' I 7,;; g .O d.T H K ,X ,-
$4f' /i 7sW.SST. ,- l
? 304 L STRIP e -
n /% / /
/
Nl (4 SIDES ). (
/b' Ik
/ /
l l l l gp#
$ l
/
?k{ g NEUTRON ,-
l [. j%'/ ABSORBER ,- hN . j%j - i l hbf . I, '
$r +
- /r 2 r_;
3 l s%s
/
- .063 THK SST. 304 L
, x . (4 FIG.10.1 - TEST COUPON 1 10-6
I
- 11. COST / BENEFIT ASSESSMENT l
-A cost / benefit assessment has been. prepared in accordance with the requirements of Section V,'Part 1.1 .The purpose of the assessment is to demonstrate that the -installation of high-density spent fuel storage racks is the most advantageous
- means 'of' handling spent fuel, considering the needs of our customers fo.r a dependable source of electric power.
The material presented in Section 9 shows that the , proposed pool storage .densification will have no significant impact. on the - . human environment. Similarly, NRC. precedent establishes ' - that alternatives and economic costs need not be discussed when' there is no significant environmental impact. U However,.the assessment presented herein is intended to satisfy NRC's need for.information. ll.I Specific Needs for Spent Fuel Storage
-G:
U Disposal of Oycter Creek spent nuclear fuel is scheduled to be - carried out by the Department - of Energy in approximately - 1998. As Oyster' Creek spent fuel may not be accorded a high priority under the DOE program, GPU is seeking to provide a spent
. fuel storage capacity to' support approximately twelve ad<11tional
~
' years - of nominal operation. . No other contractual arrangements
~
exist.for the interim storage.or-reprocessing of spent fuel from . Oyster Creek; therefore, increased storage density in the Oyster L Creek ' fuel' pool is the only viable option under consideration. Table 1.1, the fuel discharge schedule, indicates' that with the !. =high density spent fuel racks, loss of full core discharge capability (FCDC) will occur in 1992.
-In addition to spent fuel, storage is available in the i oyster Creek spent fuel storage pool for miselllaneous equipment
. (i.e. fuel channels).
11-1
, , , . . , , , - - - . .,--,-n,--, , , , ,,.-,,,rw--,,-.-, --,..n,..,,.-n,.--n,- ..-,.~,,,--,,----,-n-,...
. l'% .
(j.
~
11.2 Cost of Spent Fuel Storage. The design and manufacture of the spent fuel storage racks
.will be. undertaken by the organizations described in Section 1.
It . is . expected: that the -total project cost will be between
-6 and 7 million dollars.
11.3 Alternatives to Spent Fuel Storage
'GPU Nuclear (GPUN) has considered the various alternatives to the proposed onsite spent fuel storage. These alternatives are as follows:
- o S_hipment of fuel to a reprocessing or independent spent fuel storage / disposal facility No commercial spent fuel reprocessing facilities are presently operating in the United States. In addition, GPUN has not obtained commercial spent fuel storage commitments for fuel from- oyster Creek. The Department of Energy Away-From-Reactor Storage program has been terminated. Therefore, spent fuel acceptance j and disposal , by the Department of Energy is not an alternative to increased on-site pool storage capacity, i
o shipment of fuel to another reactor site l-Shipment of Oyster Creek fuel to another reactor site l I O i i 11-2 \ 6 v-._,. , . - . , , -,,..,c-- ..----,-c.,-- ,..-.m. v.--m ,,r, ---ww-w-r<w -er -wy r- W -------+-G--nwreer- ==-w='-- erg *---~vue--wevrw-*'V"F
9 4 Lcould; provide short term relief 'to the storage capacity p' problem.' However, transshipment of spent fuel merely 5 -serves..to transfer the problem to another site and does not result - in any additional- net long-term storage
. capacity. Accordingly,. GPUN does not consider the l transshipment of spent fuel to be an appropriate l
' alternative to high-density spent fuel storage at the site.
o: -Not operating the plant after the' current spent fuel-storage capacity is exhausted As indicated in NUREG-0575, " Final Environmental Impact Statement on Handling and Storage of Spent Light Water Power- Reactor. Fuel,"2 the replacement of nuclear power by coal-generating capacity would cause excess mortality to rise from 0.59-1.70 to 15-120 per year for
'0.8 GWY(e). - Based on these ' . f acts , not operating the plant or shutting down the plant after exhaustion of spent -fu'el discharge capacity are not viable alternatives to high density storage in the spent fuel pool. The . total ' expenditure of 6-7 million for the high
~
density racks ~is small compared to the estimated value of replacement power equivalent to the plant's energy output: S.'310,000 per day in 1983~and S 500,000 per day in 1990-1991.- i. The subject of the comparative economics associated with various- spent- fuel options is the subject of Chapter 6 of NUREG-0575.. Although the material. presented is generic, it is of value- in ~ comparing the. costs of the various options. Of ' the
- options. presented ~in Chapter 6 of NUREG-0575,.high-density spent Lfuel' storage at the - site is the most economic ~ option at $18 per
~ ~
11-3 6 i...E---
,,-,--,,.-m--,, .-,,,--y-,,-,..er ,,.-,,--,,-,w,,,,nw,.-,,-. ,r-,.= th--s*p---vew*=r-'1----"=--Wm -o * -
Kg U . -- The price off"Away From Reactor. (AFR)" fuel ' storage, if available, would be Sil5 per' KgU. This- corresponds to 0.5
]
v' mill /Kwh' from ~ a .1000 MWe power reactor for AFR storage. The marginal ~ cost of high density spent fuel racks per KgU for Oyster Creek is'$13.00. 211'.'4 Resource Commitments _ The' expansion of the Oyster Creek spent fuel. storage capacity.will require the_following primary resources:
~
O Stainless steel - 260,000 pounds o Boraflex ~ neutron absorber - 12,500 pounds - of which 6,250 pounds ~ic Boron Carbide'(BgC) pawder. The requirement for stainless s. teel represents a small fraction- of the total domestic production for 1983. Although the fraction ~ of domestic production -of BgC, required for the-
-fabrication, is somewhat higher than that for. stainless steel, it is. unlikely that . the commitment. of BgC to this project will affect ~ other . alternatives. Experience has shown. that the production - of . BgC is highly. variable. and depends on need, but
, could easily be- expanded to accommodate additional domestic
. n e e' d s .-
l t. 6 f
- i l~ .
I -[h l l-Il-4
REFERENCES TO SECTION 11 r
\ 1. B.K.' Grimes, "OT Position for Review and Acceptance of Spent
. Fuel Storages and Handling Applications," April 14, 1978.
- 2. ~NUREG-0575, " Handling and Storage of Spent Light Water Power Reactor Fue' 'Vol. 1-3, USNRC, August, 1979.
- 3. " Mineral' Facts and Problems," Bureau of Mines Bulletin 671, 1980. .
F O i. l t A 11-5
, b -
'O
. . .-_ . = . _ . _ - . _ ._ . _ _ . _ _ _ _ _ _ _ ___ _ _ . ._
- 12. DESIGN-CONTROL AND FABRICATION INTERFACE j 12.1 Introduction. -
In this chapter. an abstract of the design control from. Joseph Oat's Q.A. System is. presented in a flow' chart form. . This program has . been accepted by the ASME for' engineered fabrication of ASME Section~III, Class l', 2, 3 and MC components. The-program has been found to be acceptable to NRC ' audit teams, as well as to the
, special projects such as the Clinch River Breeder Reactor Plant and
; the U.S. Department of Defense.
12.2 Personnel The personnel-categories involved in the operations are:
- l. General Manager (G.M.)
~2.- Chief Engineer (C.E.-)
- 3. Project Engineer (P.E.) ,
4.. , Professional Engineer ,
- 5. Designated ~ Analysts (D.A.) ,
6.- Designated Draftsman'
- 7. Contract Administrator (C.A.)
- 8. Transmittal Clerk (T.C.)
The. flow of work is shown in the following flow charts. Flow Chart-il (pp. 12-3 and 12-4)-shows the job progress sequence from L ,
. its initiation.. ' Flow Chart #2 (p. 12-5)- gives the operation l- sequence following . customer feedback to the . initial document and the customer! generated documents.
l' All documents to be treated in course of a job are divided into'five types as.noted in the footnote of-Flow Chart #2 on page 12-5. 7 This~ operational flow chart gives the minimum number of steps required in the processing of'a contract. Additional personnel may be called upon for expert help by the ' Project Engineer wherever [ . L 12-1 L t
-- - ~ . , _ . _ . . - _ _ . _ . . . , _ . . . . _ _ _ . . _ . . _ _ _ . _ _ . _ . . _ _ _ . _ . - - _ . . . . _ . - . - -
deemed necessary. For example, the practical advice of the Shop superintendent in determining the- feasibility or economy of a design or the advice .of the Quality Control Manager regarding NDT and material testing requirements are frequent types of help sought i ,, ; by the Project Engineer. These are necessary steps for high quality design, although not essential for. meeting quality
-assurance requirements.
The procedure itself is self-explanatory as laid out in the two Flow Charts.- 12.3~ Flow Charts: The flow chart follows'from pages 12-3 to 12-5, and.in case of 4
- customer's' comment for any document, the recycling channel is same as shown on page 12-5.
If there is any deviation in any dimension 'or tolerance
. according.to the fabrication drawing, O.C. sends.a deviation notice to the Project Engineer. The Project Engineer evaluates and decides if it meets the customer.'s specification and drawings. If it does not meet the customer's specification, drawing, or both,
~
.the dev'iation notice is sent to the customer for evaluation. The final decision is based on customer's review.
s A I 1 h g
~
12-2
FLOW CHART *
*1
i r CUSTOf1ER FOltWARDS CONTRACT DOCUl1ENTS TO OAT p CONTRACT ADMINISTRATOR (C.A.) RECEIVES PURCIIASE OltDER AND CUSTOMER SPECIFICATIONS 6
~ l C.A.
, EXAMINES TIIE CONTRACTUAL TERf1S WITl! Tl!E !!ELP OF ,_
SALES PERSONNEL AND k GENERAL MANAGER l CUSTO:!. p REVISED CONTRACT DOCUMENT L aw- -- q, 3 C.A. l C
- A-1 ACCEPT AND ACKNOWLE; C ACCEPT WITIl EXCEPTI W lCUSTO?t PURCl!ASE ORDER g g d
C.A. p ASSIGN JOB NUM W H r lC.A. k PREPARE'JOD. FILE. FORWARD PREPARE DATA FOR SPECIFICATIONS AND ESTIMATE DOCUMENT SUBf1ITTAL FOR)1
- WORK MATERIAL TO PROJECT AND FORMARD TO ENGINEER TRANSMITTAL CLERK (T.C.:
U l C. A.
- CALL PROJECT REVIEW MEETING WITl! CIIIEF ENGINEER (C.E.) AND PROJECT ENGINEER (P.E.)
y l P.E. CUSTOtiER 2 REVIEW TECllNICAL DOCU?!ENT e ANSWER / INCORPORATE FAT!! FORWARD COtt!!ENTS TO CUSTO?!ER TECIINICAL DOCUMENT 12-5 COMMENTS 1
.O 12-3 v
u
. I!
l*
.- . . . . . - .~ _ - .- .
i .
, *.. ,L 3 .
' ' lP.E.
PREPARE SOFTWARE SilEET s FORWARD COPIES TO CllIEF
. ENGIi4EER (C' . E . ) AND C.A.
1 l-P.E.
,, PERFORM PRELIP11 NARY DESIGN
- lP.E.
- PREPARE DESIGN .)
REPORTS . l DRAFTSMAN
.DCEIGNATED DP.AFTS? TAN F PREPARES CONCEPTUAL .DRAffINGS ' l ENGINCER & REQ'JIRED DETAILED DRAWINGS, B".
REVIEhna in , U DESIGNATED. ENGINEER I l>nAFTb; tat' FORWARD CONCEPTilAL DEC
~
DESIGNATED CilECKER DRAFTS!!AN To isEsicNyrHD ANALYST ( CliECKS DRAWINGS AND Btt
, U PROFESSIONAL ENGINEElt lP.C. l U.A.
CERTIFIES T!!E REPORT REVIEWS DRAWINGS AND DM. PREPARE SPECIALIZED REPORTS TRAh5.31iT,w CaLH6 (T.C.) 4
'. ISSUE DRAWINGS & BM TO Q.C., p EllOP PURCIIASING AND ENGINEERING S IIEDULE PRE-FAB MEETINr; REVIEWER ANALYST REVIEWS Tiln REPORT
.g.
- S!!OP, O.C.,_PURCllASING, P.E.
)- HELD PRE-FABRICATION FIEETING TO , 'V . DISCUSS & CO!! MENT ON DWGS . & BM u gp.g, PROPUS5 J O;mL UNG1;iLLE GIVES~ DRAWING BACK TO DESIGNATED EXAMINE AND CERTIP'l DRAFTSMAN FOR CORRECTION Tile REPORT l DESIGNATED ANALYST DRAWINGS APPROVED BY DESIGNATED ANALYST (NOT Btt)
K t P.E.
~
APPROVAL AND REI. EASE BY PROJECT ENGINEER TO T.C. l TRANSi11TTAL. CLERK (T.C.) ISSUE TO CUSTOMER WITil DOCU!1ENT SUBMITTAL FORM COPIES TO C.A. AND P.E. t i CUSTOMER t RESPOND TO SUDMITTAL 1 2 12=4 9q c . ;- I;I
.i t -.
-(,t
( L
t
.l.-
FLOW CI: ART f2 DOCUMENTS FROM* CUSTOMER
,m 2
\
/ 'L]
'{
3 h CUSTOMER lc; EXCEPTIONS I LOG INTO IC.A. STATUS FIL; AS APPLICABLE { ! l i i REVIE!? DOCUttEN'I U HITl! G.M. l P.E. g REVIEW AND !!AKE PRELIMINARY I COf1MENTS. SET LATEST RESPONSE i DATE. l C. A. ACCEPT Tl!E DOCUt1ENT. 1 3 DISTRIBijTE AS 4 APPLICABLE V 4 l DRAFTSMAN l D. A. DE3IGNATED DRAnsnAL REVIEW THE REPORT WITH COMfiENTS
, 011ECKS Tl!E DRAWING PREPARE RESPO'ISE / ~N APPROVED l D.A. \s' REPARE CUSTOMER APPROVAL l C.E. RS 3 ,
APPROVES COM!iENTS y RESPONSE l . t. . ANSWERS OPPICIALLY d E TO CUSTOMER IF REQUIRED I
, l1 l 2 3 l 4l Y T.(
e C.A. FILES lP.E. IN ENGINEER-EXAMINE IMPACT ON EXISTING COPY OF RESPONSE ING FILE DESIGN DOCUMENTS. !!ODIFY TO C. A. FOP. JOB PILI OTilERS IF REQUIRED l1l2 3 l4l ~
.p T.C. -
3 l ~~ TRANSMIT TO CUSTOMER WITil DOCUMENT StinMITTAL FORM TP REOUIRED DOCUMa.NT l'aPLS: r~s h CUSTOMER PREPARED DRAliINGS h OAT PPEPArrD DRAWINGS CONTRACTUAL (NON-TECllN1 CAL)DOCU-( x_j j @ CUSTOMER PREPARED REPORTS OAT PREP % PED REPORTS MENTS Y p 12-5 b jl n
- 13. -QUALITY ASSURANCE PROGRAM V[ - 13.1 Introduction This chapter provides a general description of the Quality Assurance J Program that- is implemented to assure that the quality ,
objectives of-the contract specification are met.
, 13.2 General-The Quality Assurance Program used on this project is base'd upon the-. system described in Joseph Oat's Nuclear Quality Assurance Manual. This system is designed to provide a flexible, but a ,
' highly controlled system for the design, manufacture and testing of customized components in accordance .with various Codes, specifications, and regulatory requirements.'The Joseph Oat Nuclear Quality Assurance Program has been accepted by ASME and found to be adequate by NRC ' audit team.
The philosophy behind' Oat's Quality Assurance System is that
, p it shall provide for all controls necessary to fulfill the contract
- -requirements with sufficient simplicity to make' it functional on a i
day to day basis. As this system is applied to most of the j- -contracts which Joseph Oat obtains, implementation of it is almost l second nature to Oat'.s personnel. The system readily - adapts to. i different designs and component configurations, making possible the l construction ~ of many : varied forms of equipment. The highlights of this . system, as addressed -in the following paragraphs, provide an overview of the system and how it has been applied to the customer
-specifications and-regulations.
13.3 System Highlights: The design control- is organized to provide for careful review of-all contract requirements to extract.each individual design and i, quality criteria. These criteria .are - translated into design and quality-control documents customized to the' contract requirements and completely reviewed and approved by responsible personnel. O l
, 13-1
P The- system- for. control of purchased material entails l generating -detailed descriptions of each' individual item of Lmaterial- along .w ith specifications' for any special requirements-
- such . as . impact- testing, corrosion . . testing,- monitoring, or
~ witnessing' of chemical analysis, - provision' of overcheck specimens,
~
special treatments or conditioning.of material, source inspection, Land provision of documentation'of performance of any~of the above. 4 . Material receipt -inspection. ' includes a complete check of all material ~ and its '. documentation. ' 'Upon acceptance, each item of ' material is ~ individually listed on a control sheet issued once a week to assure that only' accepted material goes into fabrication. 6 The' fabrication control system provides that a shop traveler is prepared. for each . subassembly and assembly in each contract. The; traveler is generated specifically to provide step by step Linstructions :for fabrication, inspection, testing, cleaning, j- . packaging, etc. which address all standard and special requirements of the contract specifications. Special- attention is given to deployment of fabrication sequence and. inspection steps to preclude the possibility 'of' missing poison -sheets or incorrect sheets (incorrectfB10 loading). Due to the tendency. of_ contract specifications to require special examination' techniques or test procedures, all nondestructive examination procedures and test procedures are custom written to. apply to each given component within a contract. The system provides for qualification and written certification of' personnel performing quality related activities including nondestructive examination and fabrication inspection, _ welding, engineering, production'_ supervision and auditing.
.other, requirements of a solid quality control system are fully covered as'specified in the Quality Assurance Manual including m
O j-13-2 e
- . -.-,_ -. _ -,_., ..._ ,~.- .._.-..._. L . -...-..-__. _ _, _ ..-.. _ .. _ . _ ... ____...- ._.
document control, control of measuring and test equipment, control of nonconforming material and parts, corrective action auditing and 4other areas as specified. 13.4 Summary:
, . Joseph oat Corporation's Quality Assurance System provides the full measure of . quality assurance required by the contract. All special requirements of the specifications are covered including source inspection of material and witnessing of material-testing by the Engineer, furnishing..of material certifications and. test reports.within five days of shipment, and obtaining verifica. tion of qualification testing of poison' materials. Oat has a long history of providing excellent- quality- control over. a wide range of equipment types'such as the high density fuel racks.
8 L , t I 6 e O
, 13-3
. 14 . - PRODUCTION CONTROL
,p
~V 14.1
Introduction:
Production control at Joseph oat Corporation is based on the use ~ o f . a c r i t i c a l _ p a t h' diagram (CPD). A critical path diagram (CPD) is' developed for each component manufactured at' Joseph Oat
, Corporation.. The critical path diagram consists of a detailed
~
. - ~ breakdown ~'of- the operations _ required to fabricate each part, subassembly and total assembly required to' complete the finished product. The. . critical- path diagram is- arranged to show inter-relationship of all parts and sub-assemblies, including milestone dates for the completion of each operation.to assure that.
all parts and subassemblies are completed in time - to support the
. overall fabricat, ion schedule.
l'4.2 Procurement: - A bill of materials-- is generated for' every component to be
;. manufactured. The bill of material is reviewed against the CPD to l determine - the required delivery date for each item of material.
' This !information is given to the Purchasing Department to be used as the basis. for purchase _ delivery requirements. The Purchasing Department - has - a ' full-time Expeditor to continuously review _the scheduled delivery of all materials from suppliers. Problems are l reported ~ t'o- the Purchasing - Agent who is responsible for assuring on-time deliv'ery of all materials. Expediting- visits to the supplier -in question are performed by the; Purchasing Agent or Expeditor ~ whenever~ necessary. In addition, ' Production Control
-reviews the_-received materials on each component on a weekly
- basis. Any;unreceived item of material which is within 2 weeks of
~
- its critical' required _ date is reported to Purchasing and to the ,
General Manager. The General Manager institutes the corrective action which is necessary-to maintain the required delivery. I. I e !. 9 Ln Ly - - 14-1 i
-. - ;.. u u___..-....._._._~.-__-.
14.3 shop Floor Planning:
.. Daily' work assignments on the production : floor are generated
~
by ' the Plant Manager. All work assignments are pla'nned out in
-writing' a week in advance. Work assignments are based on completing the operations necessary 'to maintain the schedule required by the critical path diagrams. The work assignment sheet
, is . checked each week by Production Control to assure that all required work"is scheduled.
-14.4 operations control and Coordination's-The-_ critical path diagram for each compcnent is monitored continuously by the Production Control Department. Once a week, each. component's status is determined and recorded on the CPD. The
. diagram is then. reviewed to identify _ any operations which : are not on: or. ahead of schedule.. All such operations are reported to the General.' Manager, the Plant . Manager, and to top management.
Production Control meets with the . Plant Manager to determine the action necessary to bring the operation back on schedule.. The work L g schedule for .the ' following week is ~ revised as necessary to assure
# performance of the work required to support the delivery schedule.
i 14.5 Reporting: The complete status of _ each component -in the fabrication schedule is reported to management by Production Control-every week
}
in the form of updated critical path diagrams.- This information is used.by management for future work load planning, scheduling, and reporting status to the customer. O 14-2
_ __ _ _ .-_ ~ _ _ _ _ _ . - . _ _ .- _ _____ .
15.0 CONCLUSION
.g,
'We have determined that the proposed modification to the spent fuel pool . storage racks is acceptable because (1) the l structural ' design. -is. adequate, (2) the new storage . racks will preclude' criticality for the currently approved Oyster Creek fuel
. assemblies, (3) the spent fuel pool can.be adequately cooled, (4)
' the modification will be completed without damage to stored fuel .
assemblies sufficient to cause criticality and (5)-the quality of ~ the human environment :will1 not be significantly affected. We ' have, . therefore, - determined that (1). there is reasonable , assurance - that . the health and safety of the ' public will.not be endangered by. operation in - the proposed manner . and '(2) such ; activities will be conducted in compliance with the Commission's regulations .and the issuance of this amendment will not be
-inimical to the common defense and-security or to the health and ,
- safety of the public. ,
O V . d I J 1 8 6 L.) 15-1
'F---v-- ,r p e --wegeyy -pm - gy. <gw< a-64we-p9a-. .+ ggg- g awu-,,.eue .%.N.}}