ML20040D681
| ML20040D681 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 01/27/1982 |
| From: | Rausch T COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8202020149 | |
| Download: ML20040D681 (49) | |
Text
2 Commonwealth ECson On3 First Nat4onal Plaza. Chicago,. Illinoys p Address Reply to: Post Othee Box 767 CNcago Ilknois 60690 S
RECSVED f?
P:.S Mr. Harold R. Denton, Director lig82a.b !
l Office of Nuclear Reactor Regulation Q unn %
q, kansatsiYn U.S. Nuclear Regulatory Commission UK A
Washington, DC 20555 q,
Subject:
Quad Cities Station Units 1 and 3 Transmittal of Supplemental 7 to Revision 1 o f the Licensing Report on High Density Fuel Racks NRC Docket Nos. 50-254 and 50-265 Reference (a):
T. A.
Ippolito letter to J. S. Abel dated May 19, 1981.
Dear Mr. Denton:
Enclosed is Supplement 7 to Revision 1 of the report prepared by Joseph Oat Corporation for Commonwealth Edison entitled
" Licensing Report on High Density Fuel Racks for Quad Cities Units 1 and 2."
This supplement provides resposes to a portion of the questions provided in Reference (a).
These responses are for questions numbered as follows:
12.3.1 12.3.2 In addition, Supplement 7 provides minor corrections to the remainder o f the report.
The following corrected sheets are attached:
Pages 11, 111, iv, 1-1, table 1.1, pages 2-1, 3-2, 3-4, 3-5, 4-2, 4-7, 4-16, fig. 4.1, pages 5-5, 5-15, 6-7, 6-27, 7-1, 7-2, 8-7, 8-12, 9-3, 10-1, 11-6, and all o f Section 12.
Please address any questions you may have concerning this matter to this of fice.
One (1) signed original and thirty-nine (39) Copies of this transmittal are provided for your use.
Very truly yours, j
fhel Thomas J. Rausch Nuclear Licensing Administrator 1m Enclosure k
cc:
Region III Inspector - Quad Cities h bI 8202020149 820127 PDR ADOCK 05000254 P
3337N PDR
f T
=
i Page
(}
4.4.4 Abnormal Positioning of Fuel Assembly 4-14 outside Storage Rack 4.4.5 Missing Absorber Plate.......................
4-14 4.4.6 Dropped Fuel Assembly Accident...............
4-14 4.4.7 Fuel Rack Lateral Movement..................
4-15 4-15 4.5 Summary 4-18 References 5.
THERMAL HYDRAULIC CONSIDERATIONS.......................
5-1 5.1 Heat Generation Calculations........................
5-1 5.2 Analysis of Pool Thermal Hydraulics.................
5-1 5-3 5.3 Results 5-4 References
=
4 6-1 6.
STRUCTURAL ANALYSIS 6-1 6.1 Analysis outline 6.2 Fuel Rack -
Fuel Assembly Model....................
6-3
(}
6.2.1 Assumptions................................
6-3 6.2.2 Model Description 6-4 6.2.3 Fluid Coupling............................
6-5 6.2.4 Damping......................................
6-6 6-6 6.2.5 Impact 6.2.6 Assembly of Dynamic Model....................
6-6 6-10 6.3 Stress Analysis 6.3.1 Stiffness Characteristics......
6-10 6.3.2 Combined Stresses and Corner Displacements...
6-11 6.4 Time Integration of the Equations of Motion........
6-12 6.5 Structural Acceptance Criteria 6-15 6.6 Accidents Associated with Rack Integrity.
..... 6-19 6-20 6.7 Results 6-27 References 7
O il 1
)
1 1
1
=
s Page 7-1 7.
ACCIDENT ANALYSES *
. *..*.... = * *.
7.1 Introduction..............
7-1 7-1 7.2 Results.
7-L 7.2.1 Fuel Pool 7-2 7
7.2.2 Cask Drop.
7.2.3 Reactor Building..............................
7-2 7-2 7.2.4 Chimney.
7-3 7.2.5 Refueling Accidents -
7.2.6 Radwaste Leaks and Spills.....................
7-5 7.2.7 Turbine Missiles..............................
7-5 7-5 References 8.
RADIOCHEMICAL CONSEQUENCES................................
8-1 8.1 Objectives and Assumptions.........
8-1 8.2 Operating Experience and Nature of Stored Fuel......
8-2 8.3 Consequences of Failed Fuel.........................
8-3 8.3.1 Methods of Analysis......
8-5 8.3.2 Fission-Product Radionuclide Concentrations..
8-6 T
8.3.3 Gaseous Releases from Failed Fuel............
8-7 8.4 Exposure for the Installation of New Racks..........
8-8 8-9 8.5 Conclusions 8-10 References 9.
POOL STRUCTURAL CALCULATIONS.............................
9-1 9-1 9.1 Introduction 7
9.2 Dynamic Analysis of Pool Floor......................
9-2 9.3 Qualification of Quad Cities Pool Floor..............
9-3 9-7 9.4 Summary-9-7 References..
10.
INSERVICE SURVEILLANCE PROGRAM FOR BORAFLEX NEUTRON ABSORBING MATERIAL.*.***...*........*.*....*.**..**..*.
10-1
................................... 10-1 10.1 Program Intent 10.2 Description of Specimens...........................
10-1 10-1 10.3 Test 10-2 10.4 Specimen Evaluation..................................
iii
1 1
s Page O
11-1 11.
COST / BENEFIT ASSESSMENT.................*.*.****.*********
11-1 11.1 Specific Needs for Spent Fuel Storage................
11.2 Cost of Modification................................. 11-2 11.3 Alternatives to Spent Fuel Storage Expansion........
11-2
- 11. 4 Re sou r ce Commi tme nt s................................
11-4
- 11. 5 Env i r onmental Ef f ects...............................
11-4 11-6 References...........................................
12-1 12.
RESPONSE TO NRC QUESTIONS..................*.*.*****
12.1 Questions of May 15, 1981.............................
12-2 7
12.2 Questions of May 18, 1981............................
12-5 12.3 Questions of May 19, 1981............................
12-11 12.4 Questions of June 16, 1981............................
12-18 O
O iv J
i t
1.
INTRODUCTION The purpose of this report is to provide descriptive information and performance and safety analyses on the installation and use of high-density spent fuel storage racks at Quad Cities station Units 1 and 2.
The corresponding request to change the Quad Cities Technical Specifications to allow the use of high-density spent fuel storage racks was submitted to the NRC via letter dated March 26, 1981.
Quad Cities nuclear power station consists of two generating units (Unit 1 and Unit 2), each with a General Electric BWR-3 reactor.
The station is owned by Commonwealth Edison Company (75%) and Iowa-Illinois Gas and Electric Company (25%), and is operated by Common-wealth Edison Company.
The two utilities share the electrical output in proportion to the ownership.
At the present time, the Quad Cities units have the following capacity in their spent fuel pools:
Storage racks for 2280 fuel assemblies.
e e
Storage racks for 354 control rods.
Storage allocation for 40 channels.
e 7
Table 1.1 shows the previous and projected fuel discharge schedule for Quad Cities Units 1 and 2.
After each operating cycle, approximately 150 to 200 fuel assemblies are transferred from the reactor to the spent fuel storage pool.
Considering the present com-bined, spent fuel storage capacity of 2280 fuel assemblies, Table 1.1 indicates that following the fall 1981 refueling outage, insufficient fuel storage capacity will exist to receive a full core discharge of 724 fuel assemblies.
Furthermore, following the 1983 refueling out-
- age, insufficient fuel storage capacity exists for a subsequent refueling discharge in excess of 158 fuel assemblies.
A limited number of spent fuel racks of the presently approved design are avail-able for installation in the Quad Cities Units 1 and 2 spent fuel storage pools, which could expand the storage capacity to 2920 fuel assemblies.
As shown in Table 1.1, with additional spent fuel storage 1-1
O O
O
- Table 1.1 Ouad citles station, Units 1 and 2 Fuel Assembly Discharges Total Discharged Discharged Assemblies Assemblies In Pool Rema ining Storage Capaci ty*
Year Unit 1 Unit 2 Followinq Refuel g l7 With Additional High-Density M ating Licensed Racks Racks 1974 64 144 208 2072 1975 0
4 212 2068 1976 156 164 532 1748 1977 184 0
716 1564 1978 0
180 896 1384 1979 192 180 1268 1012 1980 224 0
1492 788 1981 0
224 1716 564 1204 1982 224 0
1940 340 980 5744 1983 0
192 2132 148 788 5552 1984 184 184 2500 0
420 5184 1985 192 0
2692 228 4992 1986 0
204 2896 24 4788 1987 200 200 3296 0
4388 1988 200 0
3496 4188 1989 0
200 3696 3988 1990 200 200 4096 3588 1 31 200 0
4296 3388 1992 0
200 4496 3188 y
1993 200 200 4896 2788 1994 200 0
5096 2588 1995 0
200 5296 2388 1996 200 200 5696 1988 1997 200 0
5896 1788 1998 0
200 6096 1588 1999 200 200 6496 1188 2000 200 0
6696 988 2001 0
200 6896 788 2002 200 200 7296 388 2003
-200 0
7496 188 2004 0
200 7696 0
2005 200 200 8096 Ethe number ol~fociUons a~iTIK6le afEei~the completion of that year's scheduled ref ueling outage.
2.
GENERAL ARRANGEMENT d
The high-density spent fuel racks consist of individual cells with a 6-inch-square cross section, each of which accommodates a single BWR fuel assembly.
The cell walls consist of a neutron absorber sandwiched between sheets of stainless steel.
The cells are arranged in modules of varying numbers of cells with a 6.22-inch center-to-center spacing.
The high-density racks are engineered to achieve the dual objec-tive of maximum protection against structural loadings (such as ground motion) and the maximization of available storage locations.
In general, a greater width to height aspect ratio provides greater margin against rigid body tipping.
Hence, the modules are made as wide as possible within the constraints of transportation and site-handling capabilities.
The high-density spent fuel racks will be installed in the Unit 1 and Unit 2 spent fuel pools, each of which is 7
33 feet wide by 41 feet long.
J O
The oued Citiee Ueit 1 goo 1 w111 cone in 19 hieh-density fue1 racks in 7 different module sizes.
The module types are labelled A through G in Figure 2.1, which also shows their relative placement.
There will be a total of 3714 storage locations in the Quad Cities Unit 1 pool.
The Quad Cities Unit 2 pool will contain 20 high-density fuel
{
racks in 6 different module sizes.
The module types are labelled A through F in Figure 2.2, which also shows their relative placement.
There will be a total of 3970 storage locations in the Quad Cities Unit 2 pool.
Table 2.1 gives the detailed module data (e.g., weight, quantity, l
I and number of storage locations).
The spent fuel rack modules are not anchored to the pool floor or connected to the pool walls.
The minimum gap between any two spent i
fuel rack modules will be 3.0 inches at all locations. The minimum gap between the fuel pool wall and spent fuel rack modules will be i
2-1
O O
O 41 *-0" k
Pool North WaN 4g-x s s 3, h
s n
s s
a Platform Channel Rack Mh Contaminated Equipment 3-}
Fuel Prep. Machines 22" J
(
G Pipe Cut-Storage Pit 54" Ref. 4 g
B3 Off 4'0" 2%~ k-182 Cells (16 x 16)
'88 c'"'
114 x i3) i00%
S..
Hef. T-Oi 02 I"*l 135 Cells 135 Cells 88%" x 82" (14314) 88%" Sq.
(2 Places)
(15 x 9)
(15 x 9) 94%" x 56%~
_94%" x 56%"
l l A1 A4 C2 B2 El 210 Cells 210 Cells 182 Caus 196 CeHs 224 Cells (15 x 14)
(15 x 14)
(14 x 13)
(14 x 14)
(16 x 14) 94%" x 88%"
94%" x 88%"
88%" x 82" 88%" Sq.
100%" x 88%"
s 3" Typ.
i I
33' 0" g
GI A2 AS C3 B3 210 CeNs 210 Cells 182 CeRs 196 CeHs 192 Cells (15 x 14)
(15 x 14)
(14 x 13)
(14 x 14)
(16 x 12) 100%" x 75%"
94%" x 88%"
94%" x 88%"
88%" x 82" 88%" Sq.
10,, --*
+
i i l
1 a
l ls h
Cask A3 A6 C4 B4 Area 210 Cells 210 Cens 182 CeHs 196 Cous 9' 0" Sq' (15 x 14)
(15 x 14)
(14 x 13)
(14 x 14) 9'4%"
94%" x 88%"
94%" x 88%"
88%" x 82" 88%" Sq.
' l 9"
l i
i l
y o
s s
s s
s s
s s
,s s
s s
s ps s s s s s s
+
3" Typ.
- - 5%" Ref.
Gate 30" 9
Figure 2.1 Section of Rack Modules Arrangement Above Base Plate Quad Cities Unit 1 (3714 Cells)
O O
O 41'0" Pool North WaN Gate 9"
e 3" Typ.
3' 0" e
+
k t.
A7 A10 C5 B5 E2
}
210 Cells 210 Cens 182 Cells 196 Cells 224 Cells (15 x 14)
(15 x 14)
(14 x 13)
(14 x 14)
(16 x 14) 94%" x 88%"
94%" x 88%"
88%" x 82" 88%" Sq.
100%" x 88%"
e.
i I
i 1
AB A11 C6 B6 E3 210 Cells 210 Cells 182 Caus 196 Cous 224 Cells b 9%"
10" - *
(15 x 14)
(15 x 14)
(14 x 13)
(14 x 14)
(16 x 14) 94%" x 88%"
94%" x 88%"
88%"x 82" 88%" Sq.
100%" x 88%"
e i
3"Typ.
l
}
33'0"
{
A9 A12 C7 B7 E4 210 Cells 210Cens 182 Cens 196 CeNs 224 Cens (15 x 14)
(15 x 14)
(14 x 13)
(14 x 14)
(16 x 14) 94%" x 88%"
94%" x 88%"
88%~ z 82" 88%" Sq.
100%" x 88%"
e e
i i
D3 D4 C8 88 F2 135 Ceus 135 Cens 182 Cens 196 CeNs 256 Cens (15 x 9)
(15 x 9)
(14 x 13)
(14 x 14)
(16 x 16) 2%" I
_ 94%" x 56%"
94%" x 56%"
88%" x 82" 88%" Sq.
100%" Sq.
b,
Contaminated Equipment 3" Ref.
C Storage Pit 54"
$- { -
~} 22" Platform Channel Rack Fuel i
Prep Machines y_
o Pipe Cut-Off 4'.0" M 41"
+
12%" Ref.
Below Water Level m
(2 Placn) 2 g
Figure 2.2 Section of Rack Modules Arrangement Above Base Plate Quad Cities Unit 2 (3970 Cells)
=
The " ell" and " tee" elements are constructed similarly using (v. ')
angular sub-element "B," and flat sub-element "C".
Having fabricated the required quantities of the "crucifcrm," " tees," and " ells," the assembly is performed in a specially designed fixture which serves the function of maintaining dimensional accuracy while welding all the contiguous spokes of all elements using fillet welds.
Figure 3.4 shows the fillet welds in a 4 x 4 array.
In this manner, the cells are produced which are bonded to each other along their long edges, thus, in effect, forming an " egg-crate".
The bottom ends of the cell walls are welded to the base plate.
Machined sleeve elements are positioned in the holes drilled in the base plate concentric with the cell center lines, and attached to the base plate through circular fillet welds (Figure 3. 5).
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 bearing surface are minimized.
The central hole in the sleeve provides the coolant flow path for heat transport from the fuel assembly cladding.
Lateral holes in the cell walls (Figure
- 3. 5) provide the additional flow path in the unlikely event that the main coolant flow path is clogged.
The rack assembly is typically supported on four plate-type sup-ports.
The supports elevate the module base plate 6.5 inches above the pool floor level, thus creating the water plenum for coolant flow.
Figure 3.6 shows vertical and horizontal cross sections of a typical support leg.
The box-shaped support structure is treated as a " linear type" support for stress analysis purposes.
The welds are sized to produce large margins of safety consistent with the rest of the sup-port region.
Lateral holes in the support foot plates provide flow 7
paths for coolant flow to the storage locations (see Figure 3.6).
O 3-2
applicable portions of Revision 15 of topical report CE-1-A, O
co o=~eetta taiso= co=9e=v o=e11tv
^=="re=ce vroere= cor ""c1eer Generating Stations.
Revision 15 of this report, dated January 2,
1981, was approved by the NRC in February 1981.
V.
Other References (a)
NRC Regulatory Guides, Division 1, Regulatory Guides 1.13, 1.29, 1.71, 1.85, 1.92, and 1.124 (Revisions effective as of April 1980).
(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)
NRC Standard Review Plan, Sections 3.8.3 and 3.8.4.
(d)
NRC Standard Review Plan, Section 9.1.2 (as applicable to spent fuel racks).
(e)
"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.
3.3 INSTALLATION AND LEVELING 7
The new spent fuel storage racks will arrive at the site by
- truck, packaged on their
- sides, and secured to shipping rigs.
Unloading of the packaged racks will be conducted by station personnel.
The racks will be brought through the reactor building receiving bay equipment air lock and lifted to the spent fuel pool operating floor elevation using the overhead crane.
Racks will then be secured to the upending cradle shown on Figure 3.7 and uprighted vertically onto their legs and moved to their temporary storage locations on the operating floor.
This operation will also be performed using the overhead crane.
Procedures and specifications will be used to control all operations required to remove existing and install new spent fuel racks.
A sequencing system will be employed for relocation of spent O
ree1 itat# the eoote-t#itte117.
evere1 existi=e recx
~111 ae 3-4
emptied of spent fuel and removed from the pool, thereby creating the required space for the first new racks to be installed.
Relocation of fuel to the new racks will then allow additional existing racks to be removed.
No old racks or new racks will be lifted over stored fuel or near enough to fuel so that any postulated lifting rig failure would result in any fuel damage.
A diver will assist in leveling the new racks with shims during the installation project.
This will necessitate maintaining separation between the diver and the spent fuel stored in nearby racks.
Initial washdown of the existing racks will be performed at the central decontamination area on the fuel handling floor.
Various methods are being investigated for disposal of the old racks including burial, shredding followed by burial, and decontamination.
The final decision as to disposal method will be based upon ALARA and cost considerations.
The new racks will be unpackaged at the temporary storage area, Od lifted and transported to the decontamination area using the lifting frame and rigging assembly shown on Figure 3.8.
Four sets of holes allow the frame to accommodate all seven new rack configurations.
The lifting rods and plugs shown on Figure 3.8 will extend and thread into the leg portion of each rack. This assembly will also be used to lower the racks into final pool positions.
In addition to the procedures which will be developed for rack handling, other areas which will be addressed are:
acceptance procedures; equipment and specifications for removal of existing rack supports where necessary; interim fuel pool liner repair guidelines; and controls for final disposal of existing racks.
O 3-5
O Top End Spacer C. * :: C.
Strip f
II fI a
n Cd
's l
g
=
i 6
U i
u a
Berafier 4,
Y l*g
=
g
,. /'
/
Fuel Bundle g
u.
. '/
is y
a
',,e
'E E
u
=
s 2
- s 1
3 J
i Sleeve (see detaa)
E j
s Bottom End i
Spacer Strip I
(4) %@ Flow Padis
'l i
i l
e s o
M) p[NT f
b Base Plate 45o u
t_%
- P t
l 3%"
Dia.
5%"
0.D.
Sleeve Detaa O-Figure 3.5 Typical Cell Elevation Revissn 1
(
The design basis fuel assembly is an 8 x 8 array of fuel rods (BWR type) containing U02 at a maximum nodal enrichment of 3.2% U-235 by we ig h t.
In the analysis, all axial nodes were assumed to be uniformly enriched to this value.
Fuel assemblies containing gadolinium burn-able poison or assemblies of other configurations or enrichments, 7 x 7 array, may also be safely accommodated in the spent e.g.,
fuel storage racks provided the maximum reactivity is less than or equal to the reactivity of the design basis fuel assembly.
To ensure the true reactivity will always be less than the calcu-lated reactivity, the following conservative assumptions were made:
Moderator is considered to be pure, demineralized, unborated o
water at a
temperature corresponding to the highest reactivity.
Lattice of storage racks is considered to be infinite in all o
directions; i.e., no credit is taken for axial or radial
(])
neutron leakage.
Neutron absorption in minor ctructural members is neglected; o
i.e., spacers and Inconel springc are replaced by water, o
Pure zirconium is considered to be used for cladding and flow channel; i.e., higher neutron absorption of alloying materials in Zircaloy is neglected.
Each fuel assembly is at maximum enrichment and reactivity.
o 4.2 Geometric and Calculational Models 4.2.1 Reference Fuel Assembly The reference design basis fuel assembly, illustrated in Figure 4.1, is an 8 x 8 array of fuel rods (type 8 x 8R) with two of the 7
central rods O
replaced by Zircaloy " water-rods."
The square Zircaloy channel surrounding the fuel has walls 0.080 incher thick and has a nominal outside dimension of 5.438 inches.
A maximum nodal enrichment 4-2
an infinite' array of storage cells.
The Boraflex absorber has a nomi-nal thickness of 0.070 inch and a nominal B-10 areal density of 2
0.01728 gram B-10 per cm,
{
l 4.3 Reference Subcriticality and Mechanical Tolerance Variations 4.3.1 Nominal Case (8 x 8R Fuel Assembly of 3.2 wt% U-235, Maximum Nodal Enrichment)
Under normal conditions, with nominal dimensions, the calculated k= is 0.9155 1 0.0036 (la with 140 generations).
For a one-sided tol-erance factor of 1.879, corresponding to 95% probability at a 95% con-fidence limit with 140 generations, the maximum deviation of k= is 0.0067.
4.3.2 '-
Maximum Enrichment Capability For the nominal nodal enrichment of 3.2 wt% U-235 and the selected Boraflex loading (0.01728 g B-10/cm2 nominal), there is a margin between the limiting reactivity (k= of 0.95) and the calculated reactivity including all uncertainties.
Additional calculations were performed to estinate the maximum uniform nodal enrichment which the racks could s'afely accommodate without exceeding the limiting reactivity.
Results of these calculations indicate a limiting nodal enrichment of 3.40 wt% U-235.
For this enrichment, the estimated k= is 0.9305 1 0.0036 (la).
With a uniform nodal enrichment of 3.40 7
wt% U-235, the maximum k= would not exceed 0.95, including all uncertainties, with 95% probability at a 95% confidence level (see Section 4.5.2 below).
4.3.3 Alternative Fuel Assemblies 4.3.3.1 Alternative Geometry and Enrichments The alternative 8 x 8 fuel assembly of 2.62 wt% U-235 maximum nodal enrichment will have an appreciably lower reactivity than the reference 3.2% enriched asseably, because of the lower enrichment.
For the-7 x 7 assembly at a conservatively assumed nodal enrichment of 2.8,wt% U-235, AMPX-KENO calculations with nominal dimensions yielded akeff of 0.890 1 0.005, which is substantially less than that of the 4-7
- - - - - - - - - - ~ - - -
d Table 4.4 Summary of Criticality Calculations O
- C a _s e,
k = or6k=
Comment Normal Conditions k=,
reference
- 0.9155 Section 4.3.1, includes gap correction Calculational bias
+0.0036 Section 4.2.3.2 Uncertainties'
'M Bias 10.0028 Section 4.2.3.2 Calculational 10.0067 Section 4.3.1
+
Mechanical
_0.0098 Section 4.3.10, i
h-Table 4.2
~
+0.0122 Statistical combination f
Total 0.9191 + 0.0122 Maximum km 0.931 h
Abnormal and Accident Conditions
' Decreased temperat"ure
+0.0007 Maximum water c
density
,m L.
'e Increased temperature or void negative l
l FO'51 element t
positioning negative i
Fuel
- channel bowing negative l
-f l _.
' Lost / missing absorber l
plate (1 of 30)
+0.0031 Conservative 7
Fuel handling accident negligible s
Lateral rack movement negligible E.
t
^
l 4-16 9
..--,5
-,9
- w
Fuel Stainless Steel 7 6 0.64" E Water Rod r5mese,esm=4meeseem if!@$OOoooOlillN ul oo00o oo l s
- /
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! 00000000 Ol h
10000000000;g
! Arna===mna==uwd s
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~
o,
,,e Unradiated Zirconium Fuel Channel h
Bulged Zirconium Fuel Channel Figure 4.1 Nominal Geometric Model of Quad Cities Spent Fuel Storage Rack Cell O
Revis on 1
2.
The heat exchangers are assumed to have maximum foul-ing.
Thus, the temperature effectiveness, S,
for the heat exchangers utilized in the analysis are the lowest postulated values: S= 0.52 for fuel pool coolers, 0.385 for RHR heat exchangers.
S is calculated from FSAR and heat exchanger technical data sheets.
3.
No heat loss is assumed to take place through the con-crete floor.
4.
No credit is taken for the improvement in the film coefficients of the heat exchangers as the operating temperature rises.
Thus, the film coefficient used in the computations are lower bounds.
5.
No credit is taken for evaporation of the pool water.
The basic energy conservation relationship for the pool heat exchanger
()
system yields:
C st, g
_ g2 -O t dT 1
3 (5.1.2) where C=
Thermal capacity of stored water in the pool.
t t=
Temperature of pool water at time, T O=
Heat generation rate due to stored fuel assemblies in y
the pool.
Q is a known function of time, T from the 1
7 preceding section.
Q=
Heat removed in the two fuel pool coolers.
2 0=
Heat removed in the RHR heat exchanger (Q =
if RHR is 3
3 not used).
O 5-5
the local water temperature also reaches its maximum.
Furthermore, no credit is taken for axial conduction of heat along the rod.
The highly conservative model thus constructed leads to simple algebraic equations which directly give the maximum local cladding temperature, tc*
5.2.2 Results Table 5.2.1 gives the maximum local cladding temperature, tc' "U the instant the pool bulk temperature has attained its maximum value.
It is quite possible, however, that the peak cladding temperature occurs at the instant of maximum value of q, i.e., at the instant when g
the fuel assembly is first placed in a storage location.
Table 5.2.2 gives the maximum local cladding temperature at t =0.
It is to be l7 noted that there are wide margins to local boiling in all cases.
The local boiling temperature near the top of the fuel cladding is 2400F.
Furthermore, the cladding temperature must be somewhat higher than the boiling temperature to initiate and sustain nucleate boiling.
The
()
above considerations indicate that a comfortable margin against the initiation of localized boiling exists in all cases.
O 5-15
~
~
20 nonlinear gap elements, and 18 nonlinear friction elements are b
v used.
A summary of spring-damper, gap, and friction elements with their connectivity and purpose is presented in Table 6.2.
If we restrict the simulation model to two dimensions (one hori-zontal motion plus vertical motion, for example) for the purposes of model clarification only, then a descriptive model of the simulated structure which includes all necessary spring, gap, and friction ele-ments is shown in Figure 6.3.
The beam springs Ks, KB at each level, l7-which represent a rack segment treated as a structural beam,4 are located in Table 6.2 as linear springs 2, 3, 6, 7, 10, 11, 14, and 15.
The extensional spring Kg, which simulates the lowest elastic motion of the rack in extension relative to the rack base, is given by linear spring 37 in Table 6.2.
The remaining spring-dampers either have zero coefficienes (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 KI, 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 Kg are modelled by elements 9,10 in Table 6.2 for the 2-D case. Note that the local elasticity of the concrete floor is included in K To simulate sliding potential, friction elements 2 7
s.
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 rotational friction springs KR, reflect the rota-tional elasticity of the foundation.
The nonlinearity of these springs (friction 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.
l i
! O l
6-7
REFERENCES TO SECTION 6 1.
" Seismic Design Classification,"
Rev. 3, 1978.
2.
" Friction Coef ficients 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 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 Earthquake and Vehicle Engineering" by S.
Levy and J.
P.
D.
Wilkinson, McGraw Hill, 1976.
5.
General Electric specification 22A5866, Rev.
1, Appendix II,
" Fuel Assembly Structural Characteristics."
6.
R.
J.
Fritz, "The Ef fect of Liguids 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.
8.
J. T. Oden, " Mechanics of Elastic Structures," McGraw-Hill, N.Y.,
1967.
9.
R.
M.
- Rivello,
" Theory and Analysis of Flight Structures,"
l l
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, Section 3.8.4, Rev. 1, 1981.
7 15.
U.S. Nuclear Regulatory Commission, Standard Review Plan, NUREG-0800, Section 3.8.5, Rev. 1, 1981.
16.
U.S.
Nuclear Regulatory Commission, Regulatory Guide 1.124, l
l
" Design Limits and Loading Combinations for Class 1 Linear-Type l
Component Supports, November 1976.
l 6-27 l
7.-
ACCIDENT ANALYSIS OO 7.1 Introduction The Quad Cities Station Safety Analysis Report, and other docu-ments, have presented results of analyses of several types of accidents which could potentially affect the spent fuel storage pools.
Installation of the proposed high density racks will enable Licensees to store increased amounts of spent fuel in the Quad Cities spent fuel pools.
Accordingly, accidents involving the spent fuel pools have been reevaluated to ensure that the proposed spent fuel pool modification does not change the present degree of assurance of public health and safety.
The following accidents have been considered:
o Fuel Pool - Earthquake Loading Loss of water o
Cask Drop
(}
o Reactor Building - Earthquake Loading Tornado Loading & Missiles o
Chimney - Wind Loading o
Refueling Accidents - Dropped Fuel Dropped Gates Dropped Channel Measuring Device o
Radwaste Leaks and Spills o
Turbine Missiles 7.2 Results of Accident Reevaluation 7
7.2.1 Fuel Pool The effects of earthquake loadings on the fuel racks and spent fuel pool floor are discussed in Sections 6.0 and 9.0 respectively of this report.
The loss of cooling water in the spent fuel pool is discussed in report Section 5.1.2.
O 7-1 l
7.2.2 Cask Drop O
On May 31, 1973, Commonwealth Edison submitted Dresden Station Special Report No. 28 to the NRC.
The report stated that it was applicable for Quad Cities Station also.
Addendums 1 and 2 were submitted on July 2 and August 10, 1973.
This report, " Analysis and Procedures for Handling General Electric 1F-300 Spent Fuol Shipping Cask", contained " Cask Drop Analyses".
The report was accepted by the NRC letter of September 10, -1973, from D.
J.
Skovholt to J. S. Abel.
The report concluded that the fuel pools l7 could withstand a drop of the 1F-300 cask.
Subsequent to Special Report No. 28, modifications-were made to 3
the Reactor Building crane which preclude postulated drops of a 100-ton-spent fuel shipping cask.
These modifications are des-cribed in Quad-Cities Special Report No.16 transmitted by letter from Commonwealth Edison Company to the NRC dated November 8,
1974.
Supplementary information was transmitted to the NRC by
()
letters dated June 10 and December 8,.1975 and February 9, March 2,
and March 29, 1976.
The NRC approved the modifications 'and associated changes in the Technical Specifications in the letter of January 27, 1977 to Commonwealth Edison Company.
Therefore, the reracked spent fuel pool will not be subject to a cask drop accident analysis.
l 7.2.3 Reactor Building The ability of the reactor building to resist earthquakes and l
tornadoes has not been af fected by the spent fuel pool reracking, except as described in Section 9.0 of this report.
Therefore, except for the described differences, the information presently contained in the FSAR is still valid.
7.2.4 Chimney The accident involving the chimney is described in Quad Cities FSAR Section 12.2.1.2.
The scenario described therein is that the top 250 feet of the chimney break off during a tornado and 7-2
8.2.2.4 Airborne Radionuclides Because of radioactive decay, Kr-85 will be the only significant contributor to any potential increase in airborne radionuclide concentrations above that currently authorized.
For the current Quad Cities storage pools, Kr-85 has not been detected at the reactor building vent (i.e.,
any Kr-85 present is less than the minimum detectable concentration of 6 x 10-6 4C/cc to 9 x 10-6 F C/cc).
As discussed in Section 8.3.3 below, no significant increase in Kr-85 concentration from the aged fuel is expected.
Consequently, expanding the spent fuel storage capacity will not impose any significant radiological burden from airborne radionuclides.
8.3 Consequences of Failed Fuel Escape of fission-products from failed fuel stored in the spent fuel pool will contribute to the radionuclide concentrations. in the pool water.
However, calculations described below indicate that the O
reato"uc tae co"ceatreet "- ero= eettea ruet ere c a=taered1r 1e==
than the concentrations of corrosion-product radionuclides and, there-fore, the aged fuel in the expanded storage pool will not contribute significantly to the onsite or offs'te radiological impact.
The decay heat generated in spent fuel rapidly decreases (by radioactive decay) following removal from the reactor and, in the aged fuel, will be very small ( < 5% of that in freshly-removed fuel).
Fuel temperatures and internal gas pressures will correspondingly decrease 1
with time.
Johnson also cites evidence to confirm that UO2 is inert to the relatively cool water of spent fuel storage pools.
Therefore, l7 the release rate of fission-products from any defective rods among the aged fuel is expected to be negligibly small.
Release of fission-products from failed fuel probably results from water leaching or diffusion of material plated out or absorbed in n
v 8-7
increasing the radiological impact to the reactor building atmosphere as a result of expanding the capacity of the spent fuel storage pool.
(Short-lived noble-gas radionuclides and other volatile fission-1 oroducts, such as iodine, are not present in the aged fuel.) Johnson concludes that the radioactive fission gases will have been largely expelled from defective fuel rods during reactor operation and, therefore, are not available for release during fuel storage.
This is expected, since the noble gases are chemically inert and there are no plate-out or hold-up mechanisms in the fuel-clad gap of the fuel element.
Measurements above the Quad Cities storage pools failed to detect any Kr-85 above the minimum detection level.
The small amount of chemically inert Kr-85 that might be absorbed on the surface of a fuel assembly and released slowly during storage, is believed to be insignificant, particularly in the aged fuel.
Since UO2 is chemically inert to cool water, diffusion of Kr-85 entrapped within the UO2 fuel matrix would be the remaining source for Kr-85 release.
Based on the method outlined in the proposed ANS 5.4 2
()
standard on fission gas release, the diffusion coefficient in the aged fuel at spent fuel pool temperatures will be negligibly small.
Consequently, diffusion release of Kr-85 from aged fuel will be 7
negligible in accord with Johnson's findings.l It is concluded that the incremental radiological impact from the l
release of Kr-85 with the expanded-capacity spent fuel storage pool will be negligibly small.
8.4 Exposure for the Installation of New Racks The existing spent fuel racks will be removed, and the new racks will be installed in a manner which will maintain occupational expo-sure to levels as low as reasonably achievable (ALARA).
The following l
8-12
To derive an effective static uniform pressure
- load, for
()
subsequent strength analysis, max is compared with the exact solution for a statically loaded plate.
Using values appropriate for the Quad The effective pressure Cities pool floor, we may determine qs/Ws.
associated with the maximum dynamic deflection S max is then obtained from the equation-( [sI 6 max 9e
- The following effective static loads are obtained from the floor slab dynamic results:
ge (SSE) = 7.25 KIPS /sq.ft.
qe (OBE) = 4.224 KIPS /sq.ft.
9.3 Qualification of Quad Cities Pool Floor
()
Table 9.1 summarizes the loadings used in the qualification of 7
the pool floor.
l t
l 9-3
10.
INSERVICE " SURVEILLANCE PROGRAM FOR BORAFLEX NEUTRON ABSORBING I
MATERIAL 10.1 Program Intent A sampling plan 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 program is intended to be conducted in a manner which allows access to' representative absorber material samples without disrupting the integrity of the 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.
I 10.2 Description of Specimens The absorber material, henceforth referred to as " poison," used O
in the surveillance program must be representative of the material used within the storage system.
It must be of the same composition, produced by the same method, and certified to the same criteria as the 4
production lot poison.
The sample coupon must be of similar thickness as the poison used within.the storage system.
Figure 10.1 shows a 7
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 in a position and with tolerances similar to that designed into the storage system.
The jacket would be closed by tack welding in such a manner as to retain its form throughout the use period yet allow rapid and. easy opening 4
without contributing mechanical damage to the poison specimen J
{
contained within.
The jacket will permit wetting and venting of the specimen similar to the actual rack environment.
i 10.3 Test The test conditions represent the vented conditions of the cruci-form elements.
The samples will be located adjacent to the fuel racks i
I 10-1 4
.,. _... ~
_,..m.
l Table 11.1 Quad Cities Station: Projection for Loss of Full Core Discharge Capability (FCDC) and Reload Discharge Capability (RDC) l7 Currently Available Spent Fuel Racks 2280 Capacity
=
1556-Capacity with FCDC*
=
2080 Capacity with RDC
=
9/81 Lose FCDC 3/83 Lose RDC Currently "On Site" Spent Fuel Racks **
2440 i
Capacity
=
1716 7
Capacity with FCDC
=
2240 Capacity with RDC
=
12/82 Lose FCDC 3/84 Lose RDC Currently Licensed Spent Fuel Racks
(
Capacity 2920
=
2196
}
Capacity with FCDC
=
2720 Capacity with RDC
=
j 3/84 Lose FCDC 7
{
5/86 Lose RDC j
High-Density Spent Fuel Racks i
7684 Capacity
=
6960 L
Capacity with FCDC
=
7484 I
Capacity with RDC
=
I Lose FCDC 3/02 7
3/04 Lose RDC
- Full core capacity = 724
- Eight (8) racks (160 spaces) on site, not in pool, 7
but available for repair and use O
11-6
12.
RESPONSES TO NRC QUESTIONS
/3 Given below are NRC questions concerning the Licensing Report on High-Density Spent Fuel Racks for Quad Cities Units 1 and 2.
They are listed by date of transmittal. Also given below are responses to those questions.
7 12.1 Questions from T. A. Ippolito to J. S. Abel transmitted on May 15, 1981 12.1.1 Question:
As a result of replacing the fuel pool racks, there is an appreciable increase in the applied load to the fuel pool concrete floor. Indicate the method and the code used in the analysis of the concrete fuel pool slab.
The method end codes used in the analysis of the concrete fuel pool Response: slab are contained in Supplement 5 to Revision 1 of Licensing Report Section 9.0, Pool Structural Calculations, submitted to the NRC on November 2, 1981.
12.1.2 Question:
Provide the floor response spectra or the time history used in the analysis of the spent fuel racks and state the source of this information.
Response: Section 6.7 of Supplement 4 to Rev. 1 of the Licensing Report submitted on 10/19/81 gives the source of the time history data.
[~')
Figures 6.9 and 6.10 of Section 6.7 depict horizontal and vertical
\\/
pool floor accelerations used in the racks analyses.
12.1.3 Question:
Indicate the damping value used in the analysis of spent fuel racks and state whether this value conforms with Regulatory Guide 1.61.
Response: Paragraph 6.2.4 of Supplement 4 to Rev. 1 of the Licensing Report submitted on 10/19/81 states that 1% damping was used in the analysis of the spent fuel racks.
This value is consistent with that used in the FSAR and conservative with that permitted by Regulatory Guide 1.61.
12.1.4 Question:
Indicate whether material, fabrication, installation, and quality control conform with the ASME code, Subsection NF.
7 Response: Yes, material, fabrication, inspection and quality control conform with ASME code, Subsection NF.
OV 12-1
12.1.5 Question:
Indicate if there is any possibility that the shipping cask may drop onto the fuel pool liner or on to the fuel pool racks and what design considerations are given to the fuel pool liner and racks.
Response: Section 10.1.2 of the Quad-Cities FSAR describes the fuel pool structure and leak detection system.
In regard to cask drop this section references the Dresden-2/3 FSAR ~(Dockets 50-237/50-249)
Amendment 16/17, Section 11, Fuel Pool Damage Protection.
In response to NRC question 2.9.3.11, Section 10 of Amendment 11 of the Quad-Cities FSAR describes the fuel pool liner design and additional details of the leakage detection system.
Dresden Special Report No. 28 transmitted to the NRC from Commonwealth Edison by letter dated May 31, 1973, provides a structural analysis which concludes that a dropped cask will not penetrate the bottom of the pool. This report also applies to Quad-Cities. Addenda Nos. 1
& 2 transmitted to the NRC by letters dated July 2,1973 and August 10, 1973 provide additional information.
Modifications have been made to the Reactor Building crane handling system which preclude postulated drops of a 100-ton-spent fuel shipping cask.
These modifications are described in Quad-Cities Special Report No. 16 transmitted by letter from Commonwealth Edison Company to the NRC dated November 8, 1974.
Supplementary information was transmitted to the NRC by letters dated June 10 and December 8, 1975 and February 9, March 2, and March 29, 1976.. The NRC approved the modifications and associated changes in the Technical Specifications in the letter of January 27, 1977 to Commonwealth Edison Company.
12.1.6 Question:
Provide the names of the codes and standards used in the fuel pool liner design.
Response: The liner was designed in 1968 to ASME Section VIII, Subsection B, Part UW and ASME Section IX.
Weld Procedure Qualifications were made in accordance with Section VIII, Q10 thru Q19.
All exposed plate, shapes, and hardware was purchased in 1968 to ASTM A167, Type 304.
The floor is %" plate while the walls are 3/16" plate.
All plate was purchased hot rolled, annealed, and pickled followed by cold rolling and polishing.
O 12-2 l
l
N
\\_,/
12.1.7 Question:
With regard to the fuel assembly drop on the top of the rack, provide the following:
Detailed description of the method used to satisfy the accept-a.
ance criteria for dropped fuel accident I and II.
b.
Comparison between drops in the tilted position, straight drop and on the corner of the rack.
c.
Indicate whether other modes of failure of the racks exist beside crushing.
The method used in analyzing the type I and II dropped fuel Response: a.
accidents was the classical one of a rigid body impacting the end of a plate or solid rod. The fuel assembly was assumed to be rigid and the drag effects of the water mass were conservatively ignored.
Thus, the impact velocity could be determined by elementary dynamics.
The impact of the rigid body fuel assembly on the fuel cell lead-in edge and base plate was used to determine the maximum magnitude of stress induced in these members.
b.
A straight drop hitting the top of the rack or the base plate of the rack produces the highest local stress levels in the r-(m,g
/
rack.
The results of the fuel assembly drop analyses were previously given in the response to NRC question 12.2.3.h.
Crushing and widespread plastic deformation were the modes of l7 c.
failure examined.
Plastic deformation could cause the geometric dimensions and tolerances used in the criticality analysis to be violated, but did not occur for the cases analyzed.
12.1.8 Question:
Indicate in detail the methodology used to demonstrate the leaktight integrity of the fuel pool liner when subjected to either the postulated fuel assembly drop or the cask drop over the spent fuel pool liner. The heavy drop should be analyzed for the tilted position and straight drop.
Response: The methodology and results of a fuel assembly drop within a fuel storage cell are described in Section 6.6 of the Licensing Report and further clarified in the responses to NRC questions 12.1.7 and 12.2.3.h.
The consequences of dropping a fuel assembly outside of a fuel storage cell are described in the FSAR.
The probability of such an accident will be reduced considerably when the new high s_-
12-3 i
l
density racks are installed as much more of the floor liner will be O
protected by the racks.
The cask drop over the spent fuel pool liner is addressed in Section 7.0 of the Licensing Report and further discussed in the response to NRC question 12.1.5.
The results of a cask drop accident are not affected by the modification.
12.1.9 Question:
Indicate whether the proposed fuel storage pool modifications conform with the staff position on " Fuel Pool Storage and Handling Application", dated April, 1978, including revisions dated January, 1979.
If any deviations exist, identify and justify these deviations.
Response Yes, the guidance is followed, with the exception of the Technical Specification for maximum enrichment.
This is because of the variety of enrichments in the fuel and the existence. of the l7 subcriticality specification of k,gg less than or equal to 0.95.
12.1.10 Question:
The seismic analysis as presented in the submittal is not clear. Indicate in detail how all the seismic models and parameters (Figure 6.1, 6.3, 6.4, 6.5, 6.6, 6.7 and 6.8, the friction forces and floor response spectra) fit together to predict the seismic stresses.
Indicate the interrelationship among the models.
Response: See Revision 1 to Chapter 6,
Seismic Analyses Description, submitted to the NRC by letter from T. J. Rausch to H. R. Denton on June 24, 1981.
12.1.11 Question:
Because different type modules were used in the proposed modification with different sizes and weights, indicate which type was used in the seismic and sliding analysis.
Indicate also how other types were qualified for the postulated loadings.
Response: Section 6.7 of Supplement 4 to Rev. 1 of the Licensing Report submitted on 10/19/81 indicates rack types, sizes, and weights used in the seismic and sliding analyses.
O 12-4
(f 12.2 Questions from T. A.
Ippolito to J. S. Abel transmitted on May 18, 1981 12.2.1 Question:
When Section - 5.1, Heat Generation Calculations, is provided, include the following information:
a.
Indicate the minimum elapted time between shutdown and when the dis-charged fuel is in the spent pool for all anticipated fuel discharge cycles.
Reponse: See Section 5 of Supplement 2 to Revision 1 of the Licensing Reports submitted to the NRC by letter from T. J.
Rausch to H. R.
Denton dated August 10, 1981.
b.
For Units 1 and 2 spent fuel pools, indicate the number of fuel assemb-lies and their respective decay times of all fuel that will be in the pools when recacking occurs.
Response: See Revision 1 of Licensing Report submitted to the NRC by letter from T.
J.
Rausch to H. R. Denton on June 24, 1981.
c.
It is noted in the FSAR that portions of the RER system may be used to augment the spent fuel pool cooling system by inserting spool pieces in the spent fuel pool cooling lines shown in Figure 10.2.1.
In this E
regard, indicate the length of time required to install these spool pieces and describe the capability of the RHR system to remove the heat from the spent fuel pool over a range of pool temperatures and with and without the spent fuel pool cooling system in operation.
Response: The time required to install the spool pisces is discussed in the response to question 12.2.2.
The capability of the RHR system to remove heat from the spent fuel pools is discussed in Section 5 of Supplement 2 to Revision 1 of the licensing report, submitted to the NRC by letter from T. J.
Rausch to H.
R.
Denton dated August 10, 1981.
d.
For Units 1 and 2 indicate the length, width and depth of the spent fuel pools and the minimum volume of water in each when all storage racks are filled with fuel assemblies.
Response s As shown in Section 2 of the licensing report, the length and width of each pool are 41 feet and 33 feet respectively.
The depth of water in each pool is 39 feet. As stated in Section 5 of Supplement 2 to Revision 1 of the licensing report, submitted to the NRC by letter from T. J. Rausch to H. R. Denton dated August 10, 1981, the water inventories in the Quad-Cities Unit 1 and Unit 2 spent fuel pools are 44887 and 44471 cubic feet respectively when all racks are in place in the pools and every storage location is occupied.
O 12-5
Figure 2.1 and 2.2 of the March 26, 1981 submittal shows that the down-V e.
comer region, i.e., space between the racks and walls of the pool, is quite small. Further, the vertical dimension of the water plenum formed by the base plate of storage racks and the pool bottom is 6-1/2 inches.
Assuming the maximum heat load is adversely located in the storage racks demonstrate that sufficient circulation will occur to preclude nucleate boiling.
Response: See Section 5 of Supplement 2 to Revision 1 of the Licensing Report, submitted to the NRC by letter ~2om T. J. Rausch to H.
R.
Denton dated August 10, 1981.
12.2.2 Question:
Assuming the reactor is operating at power when it becomes necessary to utilize the RER system to cool the spent fuel pool, describe and discuss the steps that must be taken and the elapsed time before the RER system can be placed in the fuel pool cooling mode of operation.
Response: Using the Residual Heat Removal (RHR) System for fuel pool cooling will render one of the two loops (two pumps and one heat exchanger) unavailable for use in any of the safety functions (LPCI or containment cooling). Quad Cities Technical Specifications allow LPCI and one loop of containment cooling to be inoperable during reactor operation as long as 1) the other loop of containment O
cooling is available, both core spray systems are operable, and V
both diesel generators are operable, and 2) the loop used for fuel pool cooling is returned to normal within seven days, or the reactor shall be shut down.
Once it has been determined that supplemental fuel pool cooling using RHR is necessary, the RHR/LPCI Mode Outage Report Surveillance would be performed, and crews would be dispatched to install the two spool pieces which join the fuel pool cooling system to RHR. When this has been accomplished, the valving operations may begin. This involves the closing of several motor-operated valves, racking out the breaker on another motor-operated valve, and the opening of two manual valves near the fuel pool cooling heat exchangers.
Next, the RHR Shutdown Cooling Mode suction header must be filled and vented and the RER system vented. Finally, the RHR service water system is started and an RHR pump is started to t
l commence fuel pool cooling.
The total elapsed time would be l
approximately three hours if two maintenance crews were available l
(one for each spool piece) or four hours if a single crew installed both spool pieces. At times when no maintenance crew is on site, an additional one to two hours would be required to assemble the f
necessary personnel.
i l
t O
12-6 l
12.2.3 Question:
O For both Units 1 and 2 spent fuel pool reracking operations, provide the following additional information:
Assuming a load drop, describe and discuss, with the aid of drawings, the a.
travel paths of the new and existing storaga racks with respect to plant equipment that may be needed to attain a cold safe shutdown or to raitigate the consequences of an accident.
Response: Diagrams will be prepared before moving racks based upon results of NUREG-0612 studies, b.
Provide the weights of the racks. Describe and demonstrate the adequacy of the lifting rig attachment points, on the new and old racks, to withstand the maximum forces that will be experienced during the load handling operations.
Response: The weight of the racks is contained in the Revision 1 Licensing Report submitted to the NRC on June 24, 1981 by letter from T. J.
Rausch to H. R. Denton.
Lifting rig requirements are not yet defined and will be submitted later.
c.
With the aid of a drawing, describe the lifting rigs that will be employed in handling the racks and demonstrate their adequacy.
Response: The lifting rigs that will be used to handle the fuel racks are described in Supplement 5 to Revision 1 of Licensing Report Section b
3.3 submitted to the NRC on November 2, 1981. Figures 3-7 and 3-8 show these rigs. Both lifting devices will be analyzed to assure their adequacy to safely handle the fuel racks.
d.
Assuming stored spent fuel is in the pool when the storage racks are being removed or installed, demonstrate that the stored spent fuel is not within the area of influence of dropped racks should one or more of legs of the lifting rig fails.
Response: An installation sequence has been developed whereby there will be no old or new racks containing stored fuel immediately adjacent to the location where a rack is being lifted. Therefore, a racn can be dropped vertically or assume a hanging angle and still not land on stored fuel.
FSAR Figures 12.1.1 and 12.1.2 shows a trantfer canal joining Unit 1 pool e.
with Unit 2 pool. Assuming a significant num'>st of loads are transferred between the two pools, describe the merits of providing additional protection in the form of a cover over those storage racks directly under this frequently travelled path.
12-7
Response: The assumption that a significant number of loads will be trans-
\\
ferred between.the pools is incorrect. Both pools are nearly full which precludes significant. transfers of fuel.
With regard to adding a cover, this cover would only add another heavy object consideration in additon to thermal cooling concerns.
f.
For both Units 1 and 2, with the aid of drawings, sequentially describe the movement of the stored spent fuel assemblies and storage racks in order to reduce the possiblity of fuel damage in the event of a load drop during the reracking operations.
Response All work will be planned in advance and detailed procedures de-veloped to reduce the possibility of load drops and resultant fuel damage.
the limited space between tS e storage racks and the pool g.
Considering t
walls, describe the travel paths and laydown area for various pool gates.
Demonstrate that the consequences of a dropped gate are acceptable or that one can reasonably assume that dropping of the gates is very unlikely.
Response Pool gates will have to be moved over stored fuel in the new racks.
Although a gate has never been dropped at Quad Cities, an analysis of such a drop of the heaviest gate from the highest potential elevation above the racks has been performed.
The ':nethod and results are given in Section 7.2.5 of the Licensing Report and show that no permanent deformation of the rack cells would occur.
l h.
Using Figure 3.7, describe and discuss the ability of the high density storage racks to protect the stored spent fuel assemblies from damage following a load drop.
Response: Two fuel assembly drop conditions are described in Section 6.6.
~
Accident I, where the fuel assembly is postulated to drop and impact the base plate, the maximum deformation of the plate is approximately 0.5".
It is proved that the base plate is not pierced. The analysis is based on a very conservative model which ignores the fluid drag of water in the cells, and does not account for material strain hardening.
Accident Condition II postulates that the fuel assembly drops on top of the rack and impacts at its weakest location. Maximum local stress in the region of impact is 22900 psi which is below the material yield point.
i.
In regard to the potential for damage to stored spent fuel resulting from l
light load drops (i.e., one fuel assembly and its associated handling tool when dropped from its maximum carrying height), it was assumed that
!O 12-8
e all lesser loads that are handled above stored spent fuel would cause
()
less damage if dropped. Verify that this assumption was correct, e.g.,
indicate that all lesser loads when dropped from their maximum elevation would impart less kinetic energy upon impact with the tops of the fuel assemblies and or storage racks.
Response Very few loads of any magnitude are permitted to be handled over the spent fuel pool.
The only items transported over spent fuel are other fuel assemblies, pool canal gates,- and a fuel channel measuring device.
The effects of dropping these non-fuel objects onto the spent fuel storage racks were analyzed and the results reported in Section 7.2.5 of the Licensing Report.
12.2.4 Question:
Since Figure 2.2 shows that essentially all available space in Unit 2 pool
)
will be occupied by storage racks, therefore, all Unit 2 stored spent fuel must be moved to Unit 1 pool via the transfer canal before it can be loaded into the shielded shipping cask. Describe and discuss what measures will be taken to reduce the possibility of fuel assembly damage resulting from the additional fuel handling operations.
Response: It will not be necessary to move all Unit 2 fuel thru the Unit 1 pool when it becomes possible to ship fuel. The racks in the Unit 2 cask handling area will not be installed unless required. If they were installed, they could be removed to facilitate the use of a cask later. In addition, all fuel movements will be accomplished by approved procedures to reduce the possibility of fuel assembly k
damage.
12.2.5 Question:
For both Unit 1 and Unit 2 storage pools, starting with the total decay heat load that will exist in each pool following the reracking operations, provide the following information:
a plot of the pool's maximum anticipated total decay heat load resulting a.
l from normal discharges versus time until each pool has reached its l
storage capacity.
Response: Decay heat loads for several limiting cases are discussed in l
Section 5 of Supplement 2 to Revision 1 of the Licensing Report, i
submitted to the NRC by letter from T. J. Rausch to H. R. Denton dated August 10, 1981.
b.
Verify that all decay heat calculations have been made in accordance with ASB technical position 9-2.
12-9 l
i
e Response: All decay heat calculations have been made in accordance with Branch Technical Position APCSB 9-2 (now ASB 9-2).
a plot of the pool's water temperature versus time for each discharge c.
where the total decay heat exceeds the capacity of the spent fuel pool cooling system. Indicate what cooling systems are in operation and their respective capacities.
Response: See Section 5 of Supplement 2 to Revision 1 of the Licensing Report, submitted to the NRC by letter from T. J.
Rausch to H.
R.
Denton dated August 10, 1981.
d.
a plot of maximum decay heat load in each pool, assuming a full core-discharge at each of the normally scheduled refueling periods.
Response: See Section 5 of Supplement 2 to Revision 1 of the Licensing Report, submitted to the NRC by letter f rom T. J.
Rausch to H.
R.
Denton dated August 10, 1981.
a plot of the pool's water temperature versus time following each full e.
core discharge assumed in Item d above.
Indicate what cooling systems are in operation and their respective capacities.
Response: See Section 5 of Supplement 2 to Revision 1 of the Licensing Report, submitted to the NRC by letter from T. J. Rausch to H.
R. Denton dated August 10, 1981.
f.
Assuming the maximum heat load exists in Unit 1 and Unit 2 pools when all external cooling was lost, indicate the time interval before boiling occurs and the boil off rate.
Response: See Section 5 of Supplement 2 to Revision 1 of the Licensing keport, submitted to the NRC by letter from T.
J. Rausch to H.
R.
Denton dated August 10, 1981.
t i
9 Describe and discuss the sources of makeup water, the quantity avail-able, their respective makeup rates and the steps that must be carried out and the elapsed time before the makeup water will be availaole at the pools.
Response: There are 3 sources of makeup water available to the spent fuel pool. They are:
1.
Using the condensate transfer pumps, water from the condensate i
storage tanks can be transferred to the skimmer surge tanks. Three 17 pumps can be started in minutes.
Per FSAR Section 10.2-3, this system can deliver approximately 550 gpm of water which is cooler l7 l
than that normally found in the spent fuel pool.
l O
12-10
1 i
2.
Water from the condensate storage tanks can also be transferred O
to the spent fuel pool utilizing the RHR pumps.- This method will require the installation of a spool piece which will require about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to install.
(See response to Question 12.2.2).
The amount of water available from this source is conservatively J
estimated to be 1000 gpm due to all flow coming into the pool via '
one 6 inch header.
3.
River water can be delivered to the spent. fuel pool within 30 minutes by use of fire hoses and one or both fire pumps. Each pump 4
l can deliver 3,200 gpm.
12.2.6 Question:
Since the RHR system will be required to augment the spent fuel cooling system for some period of time following a discharge, describe and discuss how it j
will be verified that the decay heat load has decayed to a value within the capacity of the spent fuel pool cooling system and, therefore, allowing the RHR system to be safely returned to its safety function mode of operation.
Response: It has been CECO's experience that the RER is not required for
.l either a reload or full core discharge. If it was required, its use j7 would be phased out by throttling back the RHR and observing if the-pool temperature remains stable. If it is stable, the spool pieces would be removed and the RHR returned to its safety function.
12.3 Questions from T. A. Ippolito to J. S. Abel transmitted on May 19, 1981 12.3.1 Question:
Discuss in some detail, the procedure that will be used for (1) removal of the fuel rods from the present racks, (2) removal and disposal of the ~ racks themselves (i.e., crating them intact or cutting and drumming them), (3)
-l 7 installation of the new high density racks and (4) loading them - with ' the presently stored spent fuel rods.
In this discussion include, in a step by step fashion, the number of people involved in each step of the procedure including divers if necessary, the dose rate they will be exposed to, the time spent in this radiation field and the estimated man-rem required for each step of the operation.
Response: The procedures for reracking have not been finalized and minor changes may occur.
Reracking is expected to be done in several steps starting with Unit 1.
These steps, and the number of workers 7
required for each step are given in table 12.3.1-1.
Exposures are given for dose rates of 3 mrem / hour and 7 mrem / hour as this will be the general range of the radiation field that the workers will experience.
O 12-11 e
~ - - -
O Table 12.3.1-1 Estimated Personnel Exposure During Re-Racking of Quad Cities 1 & 2 Spent Fuel Pools Total Exposure Job With Dose Rates No.
Length of Step Workers (hrs) 3 mrem /hr 7 mrem /hr Unit 1 Pool Remove Empty Racks (3)
Rows, D,E,F,X,Y,Z 5
140 2.100 4.900 Clean Racks 2
32 0.192 0.448 Clean & Vacuum Pool 2
32 0.192 0.448 Cut & weld two circula-3+
tion pipes 1 diver (l) 6 0.390 0.462-Install 5 new high 2+2 7
O~
density racks divers (2) 14 0.308 0.420 Move fuel to high density racks (4) 2 96 0.576 1.344 Install 1 high density 2+2 rack divers 3
0.063 0.087 Remove Row C Racks 2
52 0.312 0.728 Clean Racks 2
16 0.096 0.224 Install 2 high density 2+2 racks divers 6
0.126 0.174 Move remaining fuel to high density rack 2
20 0.120 0.280 Remove racks in Rows A&B 5
102 1.530 3.570 Clean racks 2
24 0.144 0.336 Install remaining 11 Unit 1 high density Spent Fuel 2+2
()
Racks divers 70 0.913 1.473 12-12
O Table 12.3.1-1 (Continued)
Estimated Personnel Exposure During Re-Racking of Quad Cities 1 & 2 Spent Fuel Pools Total Exposure Job With Dose Rates No.
Length
- of Step Workers (hrs) 3 mrem /hr 7 mrem /hr Unit 2 Pool Move fuel from(5)
Rows A,B,C,D,X,Y to Unit 1 high density racks 2
460 2.760 6.440 Clean Unit 2 pool 2
32 0.192 0.448 Remove racks in Rows A,B,C,X 5
156 2.340 5.460 Clean Racks 2
32 0.192 0.448 Cut & weld 2 circula-3+1 tion pipes diver 6
0.390 0.462 Install 8 new high 2+2 density racks divers 22 0.490 0.666 Move fuel from Rows E,F,Z to high density racks in Unit 2 pool 2
46 0.276 0.644 Clean pool 2
32 0.192 0.448 Remove remaining existing racks 5
120 1.800 4.200 Clean racks 2
22 0.132 0.308 Install remaining 12 2+2 high density racks divers 75 0.988 1.588 Crate or shread(6) existing racks 3
120 1.08 2.52
( )
Totals Exposure (Man-Rem) 17.894 38.526 12-13
..J
ll Table 12.3.1-1 (Continued)
Estimated Personnel Exposure During Re-Racking
~'
of Quad Cities 1 & 2 Spent Fuel Pools Notes:
(1)
Recent Dresden experience indicates diver exposure for cutting'and welding one pipe is 0.168 man-rem
.7 (2)
Recent-Dresden experience indicates diver exposure for rack-installation is an average of 0.0224 man-rem per rack (3)
Rack removal is assumed to be 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per rack 4
~
(4)
Moving fuel from one rack to another in the same pool is estimated to be 5 assemblies per hour (5)
Moving fuel between pools is estimated to take place at 2 assemblies per hour (6)
Information on exposure from chemical decontamination is not yet available.
However, the exposure is expected to be less than that O
given for crating or shreading 12-14
--..~. ---..- -----
(
e
~
d
,a w
Rack designations "A" through "F" represent existing rows in each L
spent fuel pool. In addition, each pool has four extra rack X, Y, and 21 and Z2 See attached figures 12.3.1-1 and 12.3.1-2.
If Commonwealth Edison Co. does not receive approval to install
'r'acks in a timely f ashion, eight additional existing racks of the present type may have to be installed.
Starting with the Unit 1 pool, which has fewer stored spent fuel 3
assuablies, the empty racks in rows, D,
E, F,
and X,
Y, Z are
~
rettoved utilizing the building overhead crane, and given a watsr wash to remove loose contamination.
The pool is then cleaned and k
five new high density spent fuel storage racks (F1, El, G1, B1, and B2 - See Figure 2.1, Section 2.0) are installed.
The building h
overhead crane is used in installing new racks as well as divers at the bottom of the pool for floor preparation, setting of racks, and i
shimning to level the racks.
Spent fuel is then moved trom rows A, B, and C to fill new racks F1 and El and the pool again cleaned. New rack B3 is then installed.
Existing spent fuel racks from row C are removed and new racks, C1 and C2, are installed.
The remaining spent fuel assemblies are g
moved from row A to new rack B1, the remaining existing racks in rows A and B removed and the remaining new racks installed.
7 The first step in reracking the Unit 2 spent fuel pool is to move fuel assemblies from rows A, B, C, D, and racks X, and Y to Unit 1 high density racks B1, B2, G1, B3, B4, and Cl.
Racks in rows A, B, C, and rack X are removed and cleaned, and high density racks A7, A8, A9, A10, All, A12, A13, D3, and D4 are installed (see Figure 7
2.2) The spent fuel remaining in the existing racks in rows E, F, e
[
and Z is then moved to the newly installed high density racks A7 and 1
A8 and the existing racks in rows D, E, F, and Z are removed and 5
cleaned.
The rest of the high density spent fuel racks are then installed in the Unit 2 spent fuel pool.
By using a sequence of spent fuel and rack movement such as that presented above, movement of racks and fuel over stored spent fuel is avoided.
In addition, radiation exposure of workers performing the reracking is minimized.
Final disposal of the racks can be done by one of several methods.
At this time, three options are under consideration.
1.
Crate or box the racks into small pieces which would be packed into drums for burial.
With the escalating costs per unit volume for radioactive waste burial, this may not be an economically practical alternative.
2.
Shread the racks into small pieces which would be packed into drums for burial.
The resultant volume reduction would translate into redaced burial costs.
LO 12-15
e 3.
Chemically decontaminate the racks and dispose of the bulk of
[
the caterial as clean scrap.
The decontamination chemicals will be reduced in volume, and buried.
This is the method planned for use for the racks at the Dresden Station.
The process will be evaluated for applicability to the Quad Cities racks.
final decision as the method of disposal will depend on the The 7
results of the Dresden work and a complete analysis of expected man-hours, radiation (following ALARA guidelines), and costs.
> All of the foregoing operations will be performed utilizing formal station procedures.
These may be existing procedures or modifications thereof, or new procedures, as required.
12.3.2 Qi:estion:
Demonstrate that the method used for removal and disposal of the old racks will ptovide ALARA exposure.
Response
The procedure that is expected to be used to rerack the Quad Cities spent fuel pools is given in the response to Question 12.3.1.
As indicated in this response, the procedure is designed in part to minimize personnel exposure to radiation. This is done by keeping 7
rack removal and installation activities as far as practical from
)
stored spent fuel.
As indicated, the method of disposal of spent d
fuel rack has not been determined, but a major consideration in the decision is personnel exposure. Minimizing personnel exposure, as well as as economics and practicality of operations, will be considered in the final disposal decision.
12.3.3 Question:
j What radiation levels will be used to determine whether the racks to be 7
disposed are identified as clean or radioactive racks.
l7 2
Response: 1000 DPM per 100cm is considered clean.
12.3.4 Question:
Identify the important radionuclides and tgep pgespntCs,Sgncentragfons 3
(pci/cc) in the spent fuel pool water including Cs, Co, and Co.
I What is the external dose equivalent (DE) rate (mrem /hr) from these radionuclides? Consider these DE rates at the edge and center of the pool.
Response: See Section 8 of Supplement 2 to Revision 1 of the Licensing Report, submitted to the NRC by letter from T.
J.
Rausch to H.
R.
Denton dated August 10, 1981.
O 12-16
m..
..-s O
12.3.5 Question:
Provide an estimate of the increase in annual man-rem from more frequent changing of the domineralizer resin and filter cartridge.
Responses As discussed in Section 8 of Revision 1 of the Licensing Report, the proposed modification will have a negligible annual effect on the pool cleanup system; therefore, there is expected to be no increase in the annual frequency of changing of the filter domineralizer resin.
+
12.3.6 Question:
Discuss the build-up of crud (e.g., ' Co, Co) along the sides of the pool l7 and the removal methods that will be used to reduce radiation levels at the edge of the pool to ALARA.
Responses A buildup of crud as a result of this proposed modification would mean that the concentration of crud in the pool water has increased.
Because the cleanup system removes essentially all crud deposited in the pool water from one refueling long before the next refueling, j
a measurable buildup will not occur. -(See Section 8 of Revision 1 of the licensing submittal.)
In addition, operating experience to date indicates no significant buildup of crud along the sides of the pool.
f 12.3.7 Question:
l Provide an estimate of the total man-rem to be received by personnel oc-cupying the spent fuel pool area based on all operations in that area in cluding those resulting from 4, 5, and 6 above.
Describe the impact of the modification on these estimates.
Response: As discussed in revised Section 8 in Supplement 2 of Revision 1 of the Licensing Report, there is expected to be negligible to no increase in man-rem as a result of the modification.
Assuming a radiation dose of 4 mr/hr around and above the pool (see Section 8 of Supplement 2 to Revision 1 of the Licensing Report) and occupancy of 5000 man-hour during refueling and 4000 man-hour /yr at other times, the total exposures are 20 man-rem and 16 man-rem /yr respectively.
12.3.8 Question:
Identify the monitoring systems that will be used, and its location in the spent fuel pool-area, that would warn personnel whenever there is an inadvertent increase in radiation levels that could trigger the alarm set-point.
O 1247 l
l
Response: There are six monitoring systems with set-points of 5 mr/hr to 100 fl mr/hr presently monitoring the spent fuel pool area.
These are V
deemed adequate for personnel protection.
12.3.9 Question:
Describe the methods used to preclude spent fuel pool water from overflowing onto the spent fuel pool area floors.
Response: There are skimmers and a surge tank which will take up water displaced by the new racks.
12.3.10 Question:
Specify the present dose rate in occupied areas outside the spent fuel pool concrete shield wall and provide an estimate of the potential increase of this dose rate if the space between the spent fuel and inside concrete shield wall is reduced due to the modification.
Response: The present (S/26/81) dose rates everywhere outside the spent fuel pool shield walls are 2 mr/hr. As seen in Figures 2.1 and 2.2 of the licensing submittal, there are at least nine inches of water between the outside of the new spent fuel racks and the thick, concrete walls of the spent fuel pool.
This amount of water plus the concrete supplies sufficient attenuation that the dose rate outside the walls is negligible and changes in this dose rate due to increased spent fuel storage are not measurable. Also, there are no normally occupied spaces immediately adjacent to the concrete O
shield walls.
12.4 Questions from T. A. Ippolito to J. S. Abel transmitted on June 16, 1981 12.4.1 Question:
Describe the samples and instrument readings and the frequency of measurement that are performed to monitor the water purity and need for spent fuel pool cleanup system demineralizer resin and filter replacement. How will these be r
affected by the proposed action?
Response Water purity is monitored by a continuous conductivity meter installed on the inlet to the fuel pool demineralizers, and by periodic grab samples for laboratory analysis.
Once a week a representative grab sample is obtained from the fuel pool demineralizer inlet line.
The analyses performed are pH, chloride, silica, and turbidity.
The activity checks are gross beta and gross alpha counts.
C>
12-18
s)
Once a month a sample from the same location is obtained for a gamma isotopic analysis.
All major peaks are identified.
All identifiable isotopes are quantified, and an LLD is determined for Kr-85.
The criteria for a demineralizer backwash and precoat is a consistent excursion from the chemistry
- limits, or high differential pressure across the demineralizer. Each demineralizer has differential pressure instrumentation installed which will alarm in the Unit's control room and the radwaste control room if a preset value is exceeded.
The proposed change is not expected to alter the chemistry or radiochemistry of the spent fuel poolf consequently, the described measurements will not be changed.
12.4.2 Question:
State the chemical and radiochemical limits to be used in monitoring the spent fuel pool water and initiating correcting action.
Provide the basis for establishing these limits, giving consideration to conductivity, gross gamma and iodine activity, demineralizer and/or filter differential pressure, demineralizer decontamination factors, pH, and crud level.
Response The chemical and radiochemistry limits used in monitoring the spent fuel pool water are as follows:
Conductivity
< l.0 mho/cm pH 6.0 - 7.5 Chloride
<0.500 ppm 7
Silica
< l. 0 ppm Turbidity None Gross Beta
< lE-02 p Ci/ml Gross Alpha
< 1E-05 p Ci/ml If any of the above limits are exceeded the recommended action is to backwash and precoat the fuel pool demineralizer.
The basis for the water chemistry limits is the G.E. Water Quality document (22Al286, Rev. 0) that provides the water specifications for various plant systems. The limits are set to minimize corrosion and to maintain the water in a " crystal clear" condition.
The radiochemistry limits have been established based on operating experience as action levels below which personnel exposure in the vicinity of the spent fuel pools is minimized.
The demineralizers are backwashed if differential pressure exceeds 25 psid for protection of the filter elements.
O 12-19
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