ML20027A380
| ML20027A380 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/01/1978 |
| From: | Newton R WISCONSIN ELECTRIC POWER CO. |
| To: | WISCONSIN, STATE OF |
| References | |
| NUDOCS 7811240119 | |
| Download: ML20027A380 (39) | |
Text
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NRC PUBLIC DOCUMT%0d 4., # Q#
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g UNITED STATES OF AMERICA D'
4 NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD l
v In the Matter of ) Docket Nos. 50-266
-3 WISCONSIN ELECTRIC POWER l) -
COMPANY h Amendment to License Nos.
I DPR-24 and OPR-27
, (Point Beach Nuclear Plant, (Increase Spent Fuel
)I Units 1 and 2) Storage Capacity)
Z APPLICANT'S ANSWERS TO INTERROGATORIES -
PROPOUNDED BY THE STATE OF WISCONSIN .
ON OCTOBER 2, 1978 r- ,
b .
Interrogatory 1 Please state the type of airborne radioactive emissions expected from'theispent fuel pool. How will these emissions increase in quality and quantity-as a result of the increased fuel expansion? What model is utilized in making the quantative calculation as to the effect of the interim expansion?
\
RESPONSE
The potential for airborne radioactive effluents from stored fuel is discussed in Section 8 of Attachment A to the Application. As discussed therein, the only two isotopes which have any potential for increasing with increased storage are Kr-85 and H-3. While the inventories of these two nuclides will increase by a factor of approximately 3 if the expanded pool is filled to capacity, associated releases are expected to remain negligible. It is therefore not possible to accurately quantify the small increases, if* any, in the releases of these nuclides.
The following are the results of conservative bounding analyses. Oose calculations are censistent with NRC Regulatory Guide 1.109 and meteorological calculations are consistent with NRC Regulatory Guide 1.111. 7g1134Q % h 1-1
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(a) Less than 1% of H-3 in the storage pool originates directly from the spent ,
fuel. As explained in Section 8 of Attachment A to the Application, most i
of the H-3 in the spent fuel pool originates from other plant operations !
unrelated to the number of assemblies stored in the pool. The drunning i area vent exhausts ventilation air from the spent fuel pool area, the waste packaging area, and a portion of the auxiliary building; for this analysis, j all tritium released through the drumming area vent is assumed to originate [
from the spent fuel pool. With these grossly conservative assumptions and c
, based on current releases through the drumming area vent, an increase of
{
about 4 Curies is calculated. This release would result in a maximum dose l l
of 0.00028 mrem / year to an individual living near the site boundary. The l l
actual increase will be less, probably substantially less. '
(b) For Kr-85, the analysis is also ridiculously conservative. All the gases observed through the drunning area vent are assumed to originate from the spent fuel. Based on some observed data, about 80% of Drumming Area Vent releases consist of Xe-133. For this analysis, the remaining 20% is con-servatively assuming to be all Kr-85. With these assumptions, an increase 4
of about 150 Ci is calculated. This release would result in a maximum dose of 0.000031 mrem / year to an indiv'idual living near the site boundary. Again, this is a bounding analysis, not an estimate. Actual releases, if any, are expected to be substantially less.
As a practical matter, there is ess'entially no release of radioactivity from spent fuel assemblies after the first few months when the temperature has been reduced to the stage where there is no longer a substantial temperature differential between the fuel rods' and the pool water to drive nuclides out of the rods. Since all spent fuel assemblies would reside in the pool during this period, regardless of the ultimate capacity of the pool, i.e., whether or not new racks are installed, 1-2
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i the increased number of fuel assemblies ultimately residing in the pool would have an insignificant effect on the releases of airborne effluents from the pool.
Thus, the incremental number of curies calculated above would not be expected to actually be released.
Interrogatory 2 Do you plan to increase the air monitoring capability inside of the pool contain-ment structure? If not, why not? If your answer is yes, please describe the contemplated increase.
( RESPONSE:
The present air monitoring capabilities are appropriate for both present and .
planned storage. An increase in the capabilities is neither planned nor needed.
See response to Interrogatory 1.
Interrogatory 3 What is the calculated radioactive dose rate to a person standing next to the spent fuel storage pool after the expansion? What model is utilized in calculating this dose?
RESPONSE
The maximum dose received by a person ' standing at the edge of the spent fuel pool is virtually the same as the doses stated for the surface of the pool in Section 7.4 of Attachment A of the Application. The doses were ~ calculated by a point kernel technique used in a modified QAD computer program. QAD is the generic designation for a series of point-kernel computer programs designated for calculating the effects
, of gama rays that originate in a volume-distributed source. The QAD approach has been accepted nationally and is used with modifications appropriate to local computer hardware and specific applications.
1-3, 2-1. 3-1
. '. , . i
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Interrogatory 4 What is the probability that the radioactive releases from the Point Beach Nuclear
' Power Plant will combine with those from the Kewaunee Nuclear Power Plant? What meteorological conditions would have to exist for the radioactive plumes from these two plants to come in contact with each other and intermingle? Please set forth
- the model which you base your estimate upon.
RESPONSE
It is meteorologica11y impossible for the effluent plumes from both plants to intermingle between the plants, since the wind cannot be blowing north and !
t south at the same tilhe. In the event of a south wind, it is possible for some l i c6mbination of effluents to occur at some point north of both plants. At such !
i a point, the contribution from the southernmost plant would be negligible because i of the diffusion achieved over the rather substantial distance involved. A similar observation can be made for the converse case of a north wind and some point south of both plants. However, the more conservative case occurs at some point between the two plants, not from simultaneous contributions of both plants but from the alternate contributions from either plant dependent upon the meteoro- r logical frequency involved. To demonstrate that the cumulative effects of increased storage of spent fuel are negligible, we have considered a very con- ,
servative bounding case: two Point Beach sites imediately adjacent to each j other such that the south boundary of " Point Beach North" coincides with the l
north boundary of " Point Beach South". Doses at the coincident boundary are calculated applying known meteorology for north and south sectors. Assuming the
, relea:es for H-3 and Kr-85 as given in the response to Interrogatory 1, the doses are 0.000038 mrem per year from Kr-85 and 0.00035 mrem per year from H-3.
j This represents increases of 0.000007 and 0.00007 mrem per year,' respectively, as compared with the single plant doses presented in the response to Interrogatory 1.
4-1
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While these doses are already negligible, it is. important to note that the releases assumed are grossly conservative, the distance between one plant and the other's site boundary is about 3.5 miles for the Point Beach-Kewaunee l
- situation, and spent fuel storage at Kewaunee is less than at Point Beach.
l Hence, the actual cumulative effects will be even less. !
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-. Interrogatory 5 What would be the maximum water temperature reached within the spent fuel pool were cores from both nuclear facilities at Point Beach offloaded into the pool?
In calculating this temperature, please assume that the offload of both cores would result in the spent fuel pool being full. Please state the model utilized in calculating this temperature, as well as any assumptions relied upon. How would an increase in the water temperature in the spent fuel pool affect the quantity and quality of radioactive emissions from the pool? Were a temperature increase in the pool to result in boiling and a loss of coolant, how would the quality and quantity of radioactive emissions from the pool be affected? Please state the model utilized and assumptions relied upon in making this determination.
What would be the radiation dose received by a person standing next to the spent fuel pool during such a rise in tenperature?
RESPONSE
Under the postulated conditions,1260 storage positions would be filled with 242 spaces available for unloading the two cores. The two core unloads in this-situation is not realistic but is evaluated for purposes of this interrogatory.
In order to evaluate the pool water temperature, it is first necessary to establish a time sequence of events and calculate the total heat load in the pool. 'The computer program identified in Section 4.2 of Attachment A to the Application has been utilized to develop the heat loads for this situation.
Figure 1 attached is a plot of the decay heat for a full core unload as a ,
function of time after reactor shutdown.
The heat load that would be in the spent fuel pool is established as follows, assuming that 13 days are required to unload a core and that the cores are unloaded sequentially with one day pause between unloadings:
1 l
5-1
.. a. decay heat from the second core unload 13 days after 11.5x106 BTUs/hr ;
shutdown of reactor (from Figure 1)
- b. remaining decay heat from the first core unload 27 days 8.3x106 BTUs/hr after shutdown of reactor (13 days for unloading the ;
4 first reactor,1 day pause,13 days for unloading second reactor; from Figure 1) >
- c. residual decay heat from 1260 in-storage assemblies 9.38x106 BTUs/hr [
(use 1280 assembly line from Table 4-1 of Attachment A to this Application - no further decay accounted for) !
Total decay heat load in pool 29.18x106 BTUs/hr l Normally, only one of the two cooling trains is used to maintain the temperature !
of the pool water at 120*F or less. Using both trains, the cooling system has a f design capability to maintain tne pool temperature at 120*F with a heat load of 28.2xl'06 .BTUs/hr. The above calculated number exceeds the design capability of '
the cooling system by 0.98x106 BTus/hr, or less than 3.5%. However, there is
, over 5% more heat transfer surface area in each heat exchanger (per the heat ;
exchanger technical manual data sheet) than is used to calculate the design t
capability. Thus, the cooling system could accomodate the above postulated ,
heat load and still maintain the pool temperature around 120*F. '
Since the pool water temperature is not expected to exceed the normal temperature (
of 120*F, there would be no affect on the quantity and quality of radioactive emissions from the pool for the situation assumed in this interrogatory.
If one were to arbitrarily assume the pool water reached the boiling point, the '
tritium in the amount of pool water boiled away would be released. If 1% of the
,( pool volume were to, be lost by the boiling, for example, the release would be p-a \ 5-2
~
about 0.4 Curies. There would be no effect on the fuel itself, since it is designated to withstand reactor temperatures in excess of 600*F, far greater than the temperature of water boiling in an open pool. There would be no significant increase in the release of other nuclides. By comparing this '
release of tritium with the releases and doses calculated in the response to l Interrogatory 1, it is concluded that the maximum dose to an individual living at the site boundary would be insignificant. Administrative pro edures and
_, ordinary good health practice would preclude the possibility of a worker i continuing to stand at the edge of the pool wb le it boiled; hence, any significant dose to workers is unlikely.
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Interrogatory 6 ,
What precautions have you taken to prevent the blockage of the coolant inflow and outflow pipes in the spent fuel pools? What would be the effect on the taoperature of the pool of a blockage of either the inflow or outflow pipe?
Would such a blockage cause an increase in radiaoctive emissions from the pool, due either to the inabi'ity to filter the pool's water during blockage or due to increased temperatures during blockage? If not, why not? If your answer is yes, please state the expected increase. Also state the model utilized and the assumptions relied upon in making your determination.
RESPONSE
The spent fuel pool coolant suction (outflow) pipe is located in the northwest
( corner of the north half of the spent fuel pool. The suction pipe enters the pool water vertically, from above, and is terminated three feet below the normal water surface elevation.
The spent fuel pool coolant discharge (inflow) piping is located in the south-west corner of the south half of the spent fuel pool. The piping enters the pool water vertically, from above, and is terminated ten feet below the normal r
water surface elevation.
The pool suction and discharge piping are designed so that blockage will not occur.
t Any items that would fall into the spent fuel pool would either float on the water '
surface or sink to the bottom of the s' pent fuel pool. As items floating on the l water surface are three feet above the suction pipe opening, it is inconceivable that anything could get sucked down and then up into the pool suction pipe. With respect to the discharge piping, the flow is from the pipe into the pool, and therefore the discharge piping would not become blocked.' ,
The pool suction and, discharge piping consksts of nominal ten-inch diameter (10.020"insidediameter) piping. Because the suction pipe feeds two cooling trains, it is reduced to nominal eight-inch diameter (7.981" inside diameter) 6-1
piping prior to reaching each pump. If somehow an eight-inch suction pipe was blocked, the cooling could simply be transferred to the other cooling train. l If somehow the ten-inch suction pipe was blocked, the cooling system would be shut down until the pipe was cleared. There is a set of bolted flanges in the ten-inch piping (originally installed for pressure testing purposes) and bolted flanges are used to connect the piping to the pumps. These flanges could be disconnected and the lines cleaned out if necessary.
If the cooling system were turned off, the pool water temperature would increase
( ,
during the time period for cleaning out the blocked line. Figure 2 shows the time required to heat up the spent fuel pool as a function of heat load.
There would be no increase in radioactive emissions from the pool unless the water temperature were to reach the boiling point. Refer to the answer to Interrogatory 5 for the consequences of boiling.
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Since filtering is done intermittently, rather than continuously, the inability to filter during the postulated blockage would not effect snissions from the pool.
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Interrogatory 7 1 ,
If the coolant in the spent fuel pool were to boil away, what would be the radia-tion dose calculated to a person standing at the pool's edge? Please state the model utilized and the assumptions relied upon in making this calculation.
RESPONSE
u The radiation dose under the conditions specified in this interrogatory has not J
been calculated. The large loss of pool water and the uncovering of spent fuel
[ is an unacceptable condition which is precluded by the design features of the 3
spent fuel pool and plant. The design is such that water will always be covering shnt fuel assemblies stored in the pool and this water acts as the medium for
! removal of decay heat from the fuel assemblies and as a radiation shield. This requirement for coverage of fuel with water exists for the present pool configura-j tion and it will be required after the proposed rerack as well.
4 For boiling to occur in the spent fuel pool, both spent fuel pool cooling trains j would have to be inoperable for a period of time. The simultaneous failure of both cooling trains is not considered credible. The cooling system has been
.j seismically designed so that earthquake forces will not mechanically affect the system. The cooling system and components are all located substantially away from any high energy piping. Thus, the postulated failure of these piping systems would not affect the cooling system. The pump motors of the two cooling trains are powered from separate motor control centers. Thus, the failure of one motor control center would only affect one of the cooling trains. In addition, these motor control centers can be' individually powered by the two in-plant i
emergency diesel generators should electrical power from all off-site sources be lost. .
i 7-1
Interrogatory 8 What precautions have been taken by you to prevent the possibility of a simul- !
taneous loss of both storage pool coolant pumps? Were both pumps to fail, how long would it take for the coolant in the spent fuel pool to boil? Were both ,
pumps to fail, what emergency measures would you take to prevent coolant boiling i from occurring?
d
RESPONSE
In the response to Interrogatory 7, the precautions taken in the design of the cooling system to prevent the simultaneous loss of both cooling trains are described As stated in that response, the simultaneous failure of both cooling ;
trains is not considered credible. If one were to arbitrarily assume the failure of both pumps, the heatup times have been calculated and are presented in Figure 2 of the rasponse to Interrogatory 6. Figure 2 shows the time available to ;
restore at least one train to operation before certain temperatures are reached, ,
depending upon the heat load generated by the spent fuel in the pool at the time of postulated simultaneous pump failures. Obviously, the plant Maintenance Department (mechanical or electrical) would be required to restore the operating capability of at least one train as soon as possible commensurate with the situation. Spare pump parts are maintained on-site and replacement of these parts can be accomplished in a short t,ime period.
e e
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/
8-1
. - l,
- l Interrogatory 9 How would a loss of coolant in the pool, due to either a fracture of the pool liner !
or boiling away of the coolant, affect the integrity of the spent fuel storage l racks, due to increased thermal stresses? please assune in your calculation that the racks are filled with fresh spent fuel directly from the core. Please state the model utilized and other assumptions relied upon in making these calculations.
RESPONSE
A large loss of pool water and the uncovering of fuel is precluded by the design features of the pool and the plant. See response to Interrogatory 7.
r- In event of a leak, the pool water inventory can easily be maintained by adding
(.
water equal to the rate of leakage until the liner is repaired. Adding water to the spent fuel pool can be accomplished by many means. When the new cooling system was installed, an emergency cooling water makeup connection was included in the seismically designed service water supply piping. This connection was installed simply to provide a source of water for the spent fuel pool if required in an emergency. Water from this source can be added to the pool at a rate of 250 gpm for an indefinite time period. Some of the other sources for makeup water and their delivery capacities are as follows: reactor makeup water - 200 gpm for 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />, refueling water storage tank - 100 gpm for 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br />, water treatment plant - 85 gpm indefinitely, fire water system - 1,000 gpm indefinitely.
W O
e 9-1
Interrogatory 10 ,
What is the probability that a fuel assembly dropped during loading would crack or otherwise damage the pool liner?
RESPONSE
While it is possible, but extremely unlikely, that a dropped fuel assembly could damage the pool liner, no calculation of the probability of such an occurrence has been made. It should be noted, however, that the probability of liner damage will be substantially decreased by the reracking program because the area of the pool liner potentially exposed to a dropped fuel assembly will be decreased. :
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10-1
Interrogatory 11 l
In the event of damage to the pool and/or pool liner while the spent fuel pool is filled to capacity, how would repairs be made? Would repair necessitate removing the stored fuel assemblies from the pool? If so, where would these fuel assemblies be kept during repair?
1 RESP 0NSE:
Should a leak occur to the liner at some time in the future, the spent fuel will ;
be stored in the pool in storage locations as remote from the leak as is possible.
One empty rack module, with 110 storage locations or less, could be removed to [
provide access to the area of the leak for repair.
For the remote case when the pool is completely filled, two options exist; ship off-site to another pool for temporary storage 110 fuel assemblies, or place a rack in the cask handling area for temporary storage of fuel assemblies. Both options would allow fuel to be removed from the rack in the area of the leak and the rack to be removed to provide access for repairing the leak.
\
The repair procedure for repairing a leak in the pool liner would depend on the 1
location and' severity of the leak. A leak above the minimum water level over the top of the fuel could be repaired by dewatering to the level of the leak and weld repairing in the dry condition. If a leak were identified below the minimum water level, it could be repaired by welding using a diver. Diving work in fuel pools has been performed at Point Beach and other sites in the past. Undemater welding has also been, performed on stainless steel fuel pool liners similar to -
that of Point Beach.
If it was desired to avoid undemater diving work, a leak located on the bottom of the pool or below the minimum water level could be repaired by working inside an
- i 1 l 11-1 6
s evacuated chamber, such as a large diameter pipe caisson. The caisson would be Jacked against the liner, with a gasket on its leading edge, and the water pumped out. It might be necessary to remove a fuel rack to get at the damaged area.
Single fuel racks can be removed without removing adjacent racks.
q N e i*
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G 11-2
1 4
- l Interrogatory 12 , ,
In your response to Question A-12, you cite long-tenn radiation studies documented in BISCO Report 1047-1. Under what conditions, were these studies conducted; by
" conditions", I am referring to the gama flux to which the boraflex I plates were subjected, the time period in which they were subjected to gansna flux, and the medium (water, air, etc.) in which the experiments took place.
RESPONSE
A sumary of the results of previous testing of the Boraflex poison material is contained in an eleven page Wisconsin Electric, Nuclear Projects Office memorandum of June 26, 1978; see copy attached hereto. The estimated gansna radiation exposures I
c6ntained within this memorandum were preliminary numbers;.the correct numbers are as presented in the October 10, 1978 response to NRC question C-2.
Additional testing is planned to comence at the University of Michigan on or about October 16,1978. Samples in three different environments will be exposed to various levels of gama radiation. The environments will be air, deionized water, and deionized water with 2000 ppm boron in the form of boric acid (each
, sample ~will be in its own container). Control samples will also be maintained in corresponding environments so that the relative effects can be determined.
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June 26, 1978
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.3., Mr. T. R. Wilson / File 4.9.5 . ;.d
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....;;.(.:{..;U.5:-(
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. POISON MATERIAL FOR THE 7; . .NN' *~ " J' l PGNP SPtnT fuct siu h6E ACKS - .- . ~.c. u4..- -
l
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.. .- .q . . - . ~r- . . .-- - .
. Because of the recent problems experienced with the Connecticut' Yankee a' ;
plant spent fuel storage racks (off-gassing of the poison material with attendant j t 5.. bulging of the encapsulating steel and subsequent stuck fuel assemblies). this :
memorandum is written to compare pertinent poison material parameters and to .
' summarize the known poison material testing results for the Point Beach high density '
. spent fuel racks. ,
The poison material to be used'in the PBNP spent fuel racks is called !
.'-- Boraflex and is a silicone rubber, boron carbide comoound with a minimum B4C loading of 34.8% by weight. The material is fabricated by Brand Industrial Services, Inc.
(BISCO) who has prepared a report (No.1047-1) that presents the results of the testing already conducted on the Boraflex material. One copy of this report is in the Nuclear Projects Office; Attachment A hereto is based upon the BISCO report. :
In addition to the testing that has already been concluded, bachter !
Associates has advised that additional tests are being performed. Because of the l
~ off-gassing situation, both at Connecticut Yankee and as noted in the test reports, !
the fabrication process for the poison material has been changed to include oven- ,.
drying of the baron carbide material and oven-curing of the formulated Boraflex !
material. All of the testing is, or will be, repeated with the oven-cured material. '
Also the high. temperature soak tests in borated water (see Attachment A, item 4) are continued and have accumulated about 280 days of testing to date.
(- The problem that occurred at Connecticut Yankee (stuck fuel assemblies)'
l sshould not develop at Point Beach because of a basic design difference: the poison l
' material in the Point Beach racks will be contained within a tight-fitting stainless ;
steel " bucket" open to the water where the Connecticut Yankee poison material was i completely enclosed. Thus, generated gas will be able to escape rather than bulge :
the poison material container. Table 1 sumarizes sace of the differences between j
. the Connecticut Yankee and Point Beach poison materials and storage racks. ,
.i To further evaluate the acceptability of the Boraflex poison material, i a the follwing para eters are presented. The radiation dosages are based upon the l following cases; " fresh" - where every six months a re:ently discharged spent '
fu21 assecbly is placed in the same position with the previously installed assembly being relocated and "three equal" - wn.re a spent fuel assemoly is stored for about 13 years in the same position ar i then replaced with a recently discharged i spent assembly.
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, . lir, T, it. tillson - Page Two l
. June 25, 1978 -
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l ~=v-*v - a. Conservatively estimated : =+ ^* -1$ x 10lE t ~ " fresh"' ~""~'
^-P M '
gama radiation exposure, M.~ ' 6 x 1011'
, ' ' "three equal
- rads
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a -'
[. '
.1.. .:.-
- b. llater temperature around poison
170 1. . expected '
.$ - (Calc.123510, Pg. 31), *F _.. less than 240 worst case ;
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- c. Poison material temperature , 178 . e expected i (Calc.128S13, Pg. 5), *F 248 worst case ,
- n r. ....... -
- d. North pool exit water temperature 150*F. . ! 5mrst case !
(with cooling - Calc. 123510, . i
".g Pg.31) ~
.. e. Approx. surface area of a poison . , 2502.6 .e in2 ..
3
~s .,
s slab (0.1in.x8.5in.x145.5in.) .- .
- f. Estimated total poison material 7,087,363 in2 surface area in pool {
(2832 of "e") !
Attachment A test results show that the material releases gas due to both !
irradiation and exposure to high-temperature borated water. Because tte testing ->
W was perforced individually, the conoined effects are not known. Also, during the tests the Doraflex was not covered with stainless steel sheets and thus this effect on the gas release rato is not known, and the spent fuel pool will be tt a much lower temperature and have a lower baron content and their effects are not known.
However if it is assumed that the data of Attachment A, item 4, is '
applicable- (5 in3,oTgas/in2 of surface area with 35% of the gas released in the 1 first 25 days following installation), the following would result: -
. , r
.y .. South Pool -
4 -
.. . .- t N .' a. 803 storage positions.with about 1500 poison slabs
'~'
.; ; . b. 2500 in2 x 1500 slabs x 5.0 in3 of cas -
= 18,750 x 103 in3 of gas slah in' of slab area .
,l c. 18,750 x 103 in3 of gas x 1 f3 = 10.85 x 103 f3 of gas -
1728 ind -
l
<. d. 10.85 x 103 f3 of gas x 1 dav x 1 hr = 0.301 cfm
~
25 days 24 hrs 60 min.
A gas release rate of 0.3 cfm (one 8-inch cube a minute) is not very signi- !
ficant with respect to gas volume. While the !!arth pool is to be reracked first, '
the florth pool will contain less poison slabs and therefore the gas re". ease rate
. would be smaller for the !! orth pool. .
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12-3 .
- --_ . _ _ . _ - _ _ _ _ _ _ _ _ _ _ _ _ . _ --_-- z-- - - - - - - - - ~~ ~
sir, i. n. Wilson - Page Three Jun2 26,1978
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t . The estimated gamma radiation to which the poison material would be -
.:1
~' . exposed during 40 years has been conservatively esticated at.1.5 x 1012 rads. It must be noted that tnis exposure is greater 'tnan the reported: testing exposure and
- . thus the anticipated Doraflex behavior is not known. The Cobalt 60 testing showed N-that at an exposure of 7 x 100 rads gama the material had become quite brittle.
Because of the high temperature (about 300*F), the results are in question but it
?- should be noted that the thermal aging tests (at 350*F) did not produce embrittle-ment to the extent that the gama radiation exposure showed and thus the gama radiation is concluded to be a cajor affect. -
Based upon the testing results (Attachment A) and tbo analy[is of the Point Beach racks, the following conclusions can be reached concerning the Boraflex poison material. .
- 1. The material will beceme embrittled due to the gaca radiation but because it is contained within a bucket (see fiRC. submittal,
(~ Attachcent B, Page 2-4) it will be retained in place.
, 2. The material is acceptable in a baron water environment with dimensions decreasing.
t 2 3. Off-gassing util occur probably for an extended period of time but not at a very large release rate. .
. .t
\.
ORIGINAL SIGNED BY
.- _- D. L DlLL -
/ldk ,
. .- D. L. Dill
( Attachment ,
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Copies to Messrs. Sol Burstein w/ attachment * '
' G. A. Reed w/ attachment .'
File 4.9.5 w/ attachment "
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COMPARIS0ft OF SPCflT FUCL RACR POIS0tl MATERIALS .
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[~ Parameter , Yankee -*
Point Beach I
. 4 n..n. . . . . , n .a.:.. .. . . . .. .. .... .. . .
- 1. Type of poison .. 8 4C plates with binder B4 C in a silicone rubber
~ ~ " ~
- 2. Manufacturer - .. . CarborundumCo.(2) BISCO -
. . . . . n.- .. - .
.; a., .
- 3. Completely encapsulated? Yes T ' .,1
. , . ~.' , .
. . .. c .
No .
[;. 'O'T
, 4. Previously tested? .Yes .' '~Yes .i. . , .
l.'. :. . .- '
a) irradiation 2 x 1011 rads ~8.5 x 1017. neutron by slectron beam 7 x 109 rads gamma b) in water .
Yes. . $ . .'.
~
Yes, with 3000 ppm baron I'.
. ' c) thermal cycling to 350*F~ f ' . .'7 ".' , '
Yes, at 240*F i(. ' d) off-gassing consti- H2 - 18% by vol. ~-
-40.9
. tuents and - ~
. . percentage 02- 3% by vol. 6 . .
~ ~
. . CO2 - 8% by vol. .
- ' 2- NZ - 69% by vol. ' 33.7 .
i -
f
,' . CH4 - 1% by vol . 19.5 5.. Apparent Min. Gap .
l' . ' ' .~ ~ '
between FA size and ,, .0,' f ' . , c. .-
storage position.
inch 0.190 .
4.480
- 6. Racks installed Summer 1977 Sumer 1979
'(1) All data from Licensee Event Report CTHNP1, 78-004/01 T 0, dated 5/12/78.
(2) Also was the supplier for the Kewaunee storage racks poison material.
Would not tell WPS the binder composition but since the development of the problem it is believed that the binder composition has been provided to NRC. WPS has cancelled purchase order since the development of the Connecticut Yankee. problem and is now working.with a company in Germany. ,
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f.."- SU M ARY OF BISCO REPORT 1047-1 -
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.The subject report (Revision .1 dated May 5,1978) is' entitled "Boraflex 1 "
Suitability Report" and is compiled and published by Brand Industrial Service Inc.
of Park Ridge, Illinois. The report is in a loose-leaf 3-ring binder, and is about one inch thick. The report includes various sunnaries, data sheets, and BISCO promotional literature. . The following excerpts are taken from various portions of the report. .. . , . , .,
- . ., ..... . . ..c _
, *f ,
- 1. Thermal Aging Tests , ' -icy .- :. . .. . , '.] -
These tests were performed in a controlled temperature oven at 177'c (about 350*F) and at 190*C (about 375'F). Tests of physical prop'erties were con-s ducted at various times during the thermal aging testing periods which were
( about 245 days and 210 days respectively. The results were as follows:
l *. . -. . . .:i 4 G Time, hrs. Durometer Tensile - Elongation G177 0190 ,; 9177 @l90 9177 0190 l .
7 days 9 RT 53 53 ;. 460 '460 ,.116 - 116 240 63 63 549 444 .78 84
~ ' ~
480 59 62 ,
404- 397 84 90 .
960. ' 64 . 62 . 3.~. 490 364 ~ ' 78 92 1920 -
62 63 404 353 80 83
' 63 275 2880 -
,. 60 -
x . 4080 .
62 65 271 263 64 70 .
. .. . v.
5160 62 64 267 , 247 ~ 68 70 5880 63 -
. . . 278 - 57 -
From the above data, at 177*C (350*F) the property changes seem to have stabilized after about 3000 hours0.0347 days <br />0.833 hours <br />0.00496 weeks <br />0.00114 months <br /> (about 125 days). The testing at 190*C does not appear to have been long enough to stabilize the properties.
- 2. Effects of Gamma Radiation from a Cobalt 60 Source The data is presented on page 4. The samples were not cooled during the testing and it is estimated that the temperature reached about 200*F. ,
The data shows that gamma irradiation makes the material brittic. '
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~: 3. Irradiation Testing of 50% G4C Boraficx " "- ' " " :
The samples were irradiated and tested at the t!niversity of Michigan. ' The ~~ ~ -
long term irradiation data is presented on page 5..
. . , . .s. .
Z.q
7 .
The irradiation of 8.53 x 1017 N/cm2 produced a weight lo'ss 5f about 6% .
with a decrease in all of the dimensions. .
..-- !. 1. .; # . r.. . ..
- 4. Prel. Report: Exposure of BISCO Boraflex I to High Temperature Borated Water !
Samples containing 50% B4C were immersed in 240*F boiated water (3000 ppm ;
boron) for over 4700 hours0.0544 days <br />1.306 hours <br />0.00777 weeks <br />0.00179 months <br /> (about 200 days). The water pH was adjusted with ,
sodium hydroxide to a range of 9.0 to 9.5. -l . !
The tests were interrupted at intervals of 40 days, 80 days, 150 days and ;
199 days' for measurement of the physical parameters. Some of the data is ;
presented on pages 6 and 7. . .
.[ " l>
.( -
The data shows that while the sample dimensions decreased by about 1%, the
.2
. sample mass increased initially by about 0.8% and then decreased but remained
, greater (by about 0.25%) than the original mass. The density (initially -
.about 114.6 lbs./ft3) increased by about 4.6% and gas was evolved. The gases were identified as hydrogen, methane, ethane, and carbon dioxide, but the :
ratio of each was not determined. The gas evolution decreased as a function '
t of time as shown in the following figure: -
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.5. The Effect of Combined Gamma and heutron Radia' tion on the' Hydrogen Content ',
of 81500 (15-11 ticutron Shiciding Haterial . + . , . .
, : - 4. .This material is intended to be used to attenuate high ene,rgy neutrons escaping 7- from the area between a reactor vessel and the primary shield wall of a pres- y
.1'9 surized water reactor. This is a silicone' resin material having a relatively high hydrogen content. ;' -
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.y.
. . The tests were conducted at the University of Michigan.. ' Cumulative irradia-.
Oi tion was greater than 2 x 1011 rads with the gamma component exceeding -
2 x 1010 rads and the integrated fast neutron dose was in excess of 10 81 N/cm2 (where e = 1 to 10 Mev). .
- , . .. _,.y. . . , .
- c !:,
~
. The following data was obtained: ,
! ,. i ; :'. j . '.
~
Control Sample Irradiated Sample c
^
s Specific Gravity - l'.156 -
1.219 >
b 1 Hydrogen (weight) 5.69 .
' - 5.63 -
Hydrogen density (g/cc) 0.0658 "
- 0.0686 Carbon, % 41.21 42.68
Silica, 5 of sample 61.31 55.50 -
. Oxygen, 1 -, ,
. .s . 25.76 ,
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= . .. O r, - .
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. . EFFECT OF RADIATION ON BISCO NSI'. .
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EXPOS.*ID TO A COBALT 60 SOURCE
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Do::e Tensile ' Elongation , Elastic
~ I. i.f.:.:n..a.r.ii.h.
. .. : (PSI)_ 7. Modulus -
. .a .
'.1. :
t
. , .. . ,e o .. .
. 510 .
68
. .. 750 :
~
16 516 . 55 938 .
. . in '
~ 60' ~550 -
40 ' 1375 -
. :< N :'-.
~ :..-
111 ,
504.. ..,....,.-t 38 -
1326 - -
~
164 -
. 553' -
23 2404 - -
{ . .713 896
~ 3.3 27,151 - -
t .
.1" x 1" Tensile Bar pulled @ 10 inches /rainute S .gelo: . .
Stress for (7.) -
Dn.4e .." -
207. Compression Dyna.raic Comp. See
']'7 fj., nt:q:,idtt. ,
17~. ' ,' ,
(PSI) at 207. Cor.m .
, . -r. . :-
N . . , . . .
O 19.' .
2.57.
34 -
206 -
0
' ^
GP. -
396 O 119 652 0
. 48'i ,
2756 (shattered) 100 i . ,
St.:..Ph : 1.125" Dir. x 1" thich button compressed 207. f_! 1 inch / min.
[
12-9 1
. . . p'
- g
~. - *
.. ,:} }
. . , ^*~
(. . .
. . BISCO Silicon -50% B 4C Somple ~'
p;. .. *.
+.
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g
... .. . Long Tctm Inadiation *
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- c. .
. . ; ,. . . . Test Sequence .
' .c
_,. i.; ;';.;/ 50% B C.
4 Sample e -
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.. ....v.. ..
. . .e,,: <.~ .
. . .. . -.a .: .; . . .,a. .i:,;1 .. ,...
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7.. 'y.y',
. l ". .;. r. .. . . " . ' . . '
s~.
, . : ..- . . n ., . . , . .. .
~...:. . .. ...
~ '
L. h.. .;'... .P. Pre-fr:adiation Dim-nsions (in) . ..T..*3. :; * ".J
- r. : -. . .":.' , ?. i.b,,;. . . .,
... ..... .. .t ...... . ,. .w
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. : :. . ...:.: ., ..~ . '.. .
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. ?. !' O.317 * .y .
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. :.? : ' .. '. '. . .
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~.
.:.:c. 1,g.:.......?w
.e .-:.. .>... ;..?: : . .
".~..4 M sn .!8.53 11x10 x 109 -
- y: -.
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... Gas' Gamma Evolution Dose m, (Rads) .
"$.... .i.:i. ' 7 . ..s
- 5.7 .
' * ' . '. 3 -
.(
. ...-. C linder Pressure BuiIdu *.. M. , . .~1. . : : &. . . .. .' .-,.e . S .
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4 ; . . . . . ...M. Rate PSI hr
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g ., . . . . . , . .. . . . .. . . . - .
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~@{.?' . . .3,{[h .5@50xygen (%) '.Q.$$jhiN>5.93. ::.,. ', '[;-.
. '* '5; -.
. ;:. .<; Nltrogen (%)
- l . .- 6,':.Q;. 33.72 ^ . ; ': -
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ost rro . t.ic ion imensicns in
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l Oc.igItaal ) based or Single Lbng Dimen=iona,1 Chr.tige
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_SAMD.f. t:2 40 d.a.y.n. 8 0 .d a..y_:. 150 dayu .' 1 9 --9 d ty y , .
': *- :1 -
0.00t -0.93?. -0.93t -0.930 . ,
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2- O.00 -0.93 - 0 . 9 3 ..~ -0.93-i
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9 4 0.00 -0.93 -0.93 .40.93 .-
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5 -
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- -0.93
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TABLE III -
- . ~ .
p , ,
~
i '.
Mar.n .'Itability of Tm nficy. I (S Change front original) .
ShMt's':5 4 0 dn.yn.. CO davn -:- ' 150 da.ys, 199 dayn, 1
~
. :0.822 .
+0.834 +0.59% +D.47% -
2 '. 10.89 +0.59 +0.41 40.40 .
3 +0.01 +1.8G +0.52 40.35 N -
, 4 +1.03 +0.75 .+0.59 +0.3G
". '5 +0.79 +0 64 . +0.40 M . 27 -
G +0.07 -0.08 -0.43 -0.4d A. .rne,p 40.740 + 0.774 +0.3Ge +0.2*'
\
. . l
.. . \
L. .-
. 12-11
~ .
' p- . .
. . . . s 7
. . , ... TADC.E IV - -
( ,.,
, fr nnity Stubility of Dornfic:: I (h Clmngo Cco:a 0 igiuni.)
. . . . .~
.' St.itt>I.E? 40 dayu 80 dayn 150 day: 199 cL2yu .
1 ..,.+0.82%
+1.604 -
+1.87% f
+5.270 ..,s,
. 2 +0.42 +1.11 ~+1.88 +5.21
. 't ,
. 3 -
. -0.10 +2.06 +1.93 +4.G6 , .
, , .... e , . .
- - .- 4 +1.44 +1.23 +1.06 .. +4.19 5 -
+1.21 +1.59 +1.42 +4.10 *
. .G , +0.32 -0.09 ' +1.07
'! +4.30
.t . .
., Averago +0.69% +1.25% +4.G20
( +1.540 '
Gas evolution of the Doraflex I samples was continuou.Ely annitored as describcd in the test proceduren and '
, r61xtetod in the following tabics: . .
s.
s TADT.E V -
~. . .,
, Acculaulat.cd gas volume evolved (Cubic inches per Sq. In.
- of sancple arca) " ~
T1tm (days) - -
' ' ' TOTAL Evor.vr:0 Cns (in3/3n?]
N. .
40 .
1.48 50 : .. 1.73 C0 . . .
2.61 100 , 3.09 150
- 4.15 -
193
- 4.90 e .
g e
, 12-12
,s.
- - Interrogatory 13 Do you presently monitor the groundwater around the Point Beach Nuclear Power Plant for radioactivity? If your answer is no, do you plan to install ground-water monitoring equipment to monitor releases from the spent fuel storage pool after expansion? If not, why not?
RESPONSE
Groundwater beneath the site is sampled on a quarterly basis at the plant well, located just south of the switchyard. In addition, lakewater is sampled on a monthly basis at 5 points along the shoreline, the natural terminus of ground-water gradients in the area.
k.
Interrogatory 14 If your answer to the above interrogatory is yes, please state whether you plan to increasethe describing or changes change your present groundwater monitoring system in any way, contemplated.
If you do not contemplate changing your present groundwater monitoring system, state the reasons for this decision.
RESPONSE
Changes are neither contemplated nor needed. A significant leak would be detected by the presently available indicators: wellwater sampling, shoreline lakewater l
sampling, pool leak detection system, and indications of unusually high quantities of makeup water to the pool. Further, as explained in Section 7.4 of Attachment A to the Application, increased fuel storage is not expected to result in any !
significant increase in the radionuclide concentrations in the pool water.
~
i 13-1, 14-1
Interrogatory 15 How many fuel assemblies are presently stored at the NFS plant in West Valley, New York? What precautions are planned in order to insure that any fuel assemblies returned from NFS arrive safe and intact? What procedures are planned should a number of the fuel assembliss returned arrive in deteriorated condition?
RESPONSE
Wisconsin Electric currently has 114 fuel assemblies in storage at the NFS plant in West Valley, New York.
( Phecautions to insure that spent fuel returned from NFS is safe and intact are addressed in our previous response to NRC Question C-19 in our October 10,1978, ,
submittal.
Special procedures to specifically deal with fuel assemblies which are returned and arrive in a deteriorated condition are not required. The fuel assemblies are intact at NFS and are expected to show no signs of physical deterioration resulting from storage and shipment. This is borne out by the general experience of the nuclear industry in respect to the integrity of spent fuel while in water pool storage and during transport, including our recent experience of shipping six fuel assemblies from NFS to Battelle N'orthwest Laboratories (BNWL) in Hanford, Washington. In any event, even if a returned fuel assembly did arrive in a {
deteriorated condition, the potential releases and contamination to pool water l would be far less than the increased contamination to pool water that normally i occurs at each refueling. The spent fuel pool filter a'nd cleanup system would remove any contaminants from the pool water in a short time period.
l l 15-1
- 1. . . .- . . . . - -
i
- . l l
Interrogatory 16 What is the expected increase in occupational exposure due to the daily operation of the expanded spent fuel disposal pool? Please state the assumptions relied upon in making this calculation.
RESPONSE: !
As discussed in Section 7.4 of Attachment A to the Application, radiation dose ,
rates with the expanded storage capacity are not expected to be significantly different from those encountered for the present design. In fact, for one ;
particular case discussed in Section 7.4 the dose rate will actually go down.
Furthermore, no additional worker activity in the vicinity of the spent fuel pool is anticipated as a result of the increased storage capacity. Hence, there will be no increase in occupational exposure due to the daily operation of the expanded spent fuel pool.
Interrogatory 17 What is the expected increase in radiation exposure to the public due to operation of the expanded spent fuel storage pool? Please state the assumptions relied upon in making this calculation.
RESPONSE
See the response to Interrogatory 1.
l i'
i I
l 16-1,17I J l- ;
..s
{ .
Interrogatory 18 StatethUtechnicalbasisuponwhichyoubelievethatthespentfuelstoredin the pool will retain its integrity for the entire period of licensing.
RESPCtlSE:
The Point Beach fuel and cladding are designed for use in a borated water environ-ment in the operating reactor under conditions much more severe than that which 7 ,. will be experienced in the spent fuel pool. In the operating reactor the fuel is designed to be exposed to neutron irradiation, temperatures above 600*F, and
( pressure of 2250 psia without significant corrosion or loss of fuel rod cladding integrity.. For in-reactor corrosion rates at temperatures of 500*F, it would n take approximately 2,200 years to penetrate the Zircaloy-4 cladding. Since all
?
of the structural material of the fuel assembly have the same or better corrosion i
resistance characteristics than Zircaloy-4, the structural integrity of the assembly would also remain intact for as long as that of the fuel cladding.
In the relatively mild spent fuel pool environment, any deleterious effects of the borated water on fuel and cladding are reduced to relative insignificance even if the spent fuel pool temperature should increase substantially.
e 4
The good perfomance of fuel stored underwater in pools is supported by extensive
[, , successful experience with storage of spent fuel in water pools ac discussed in the .Oraft Environmental Impact Statement prepared by the U. S. Department of Energy; 00E/EIS-0015-0, Storage of U. S. Spent Power ' Reactor Fuel, August 1978, wherein it states: "The technology of water-cooled basin storage is well- developed, and water basins have been successfully used for receiving and storing spent nuclear fuel since the beginning of the nuclear age, more than 18-1 l I,
'{
. l .
. g 30 years ago. Spent fuel has been stored without any significant incident or
' detriment to the surrounding environment or population. Further, the storage
'has been accomplished without any serious deterioration of the fuel cladding.(l)" .
'I( ,
P 3
\
N. -
[
l (1) A. B. Johnson, Behavica of Spent Nuclear Fuel in Water Pool Storage.
USERDA Report BNWL-2256, Battelle Pacific Northwest Laboratories, Richland, Washington (September 1977) 4'
18-2 s.
~ f. n i
Interrogatory 19 ,
Please state the average, median and maximum burnup of the spent fuel which will I be stored in the fuel pool. How does the burnup of the fuel affect your estimate ;
of long-term fuel integrity? Please be specific. Please state the name of all i technical studies and/or experiements with which you are familiar, whether ;
completed or ongoing, which assess the integrity over a forty-year period of spent fuel having a burnup as high as that of the spent fuel with the maximum l burnup expected to be placed within the Point Beach spent fuel pool. l I
RESPONSE: (
It is not possible to state the precise average, median and aaximum burnup of .
spent fuel which will be stored in the spent fuel pool, because future fuel ;
( assembly and fuel cycle designs have not yet been specifically developed.
Typically, discharged fuel has achieved burnups ranging from about 21,000 MWD /MTU. i
\
up to 40,000 MWD /MTU.. Currently, the region average discharge burnups are l l
targeted on a burnup of 33,000 MWD /MTU. with expected maximum and minimum burnups of about 37,000 MWD /MTU. and 28,000 MWD /MTU. respectively. Higher region average discharge burnups may be achieved in the future. Higher burnups would have ;
i little. If any, incremental impact on spent fuel integrity because the storage i
duty of the fuel in a spent fuel pool environment is so much less limiting than :
that experienced in an operating reactor. Fuel temperatures, dy r loads and thermally induced stresses on all portions of the fuel assemblies win be much !
lower. Thus, fuel assen61y burnup does not have a significant effect on long-i term fuel integrity under spent fuel storage conditions.. We are not aware of whether o= not studies of long-term fuel integrity have specifically included !
. spent fuel having burnups in excess of the maximum expected at Point Beach. I 19-1
- s. . l Interrogatory 20
. How will the integrity of the fuel rods in the spent fuel pool be monitored?
RESPONSE: !
The spent fuel pool water is, and will be, monitored on a regular basis by '
laboratory analysis of water samples.
- 4. 9C25D I Roger A. Newton Senior Nuclear Engineer Subscribed and sworn to before me this 1st day of November,1978 M s 4.
7
~
/ ,ete w..--
l Notary Public State of Wisconsin My comission expires I/-FA'O
([ ,
S t
s 20-1
,r ,'
- 4
..c. .
t .
November 1,1978 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of Docket Nos. 50-266 50-301 WISCONSIN ELECTRIC POWER COMPANY Amendment to License Nos.
-, (Point Beach Nuclear Plant, (Increase Spent Fuel Onits 1 and 2) Storage Capacity)
AFFIDAVIT OF SERVICE I hereby affirm that copies of " Applicant's Answers to Interrogatories Propounded by the State of Wisconsin on October 2,1978", were served upon those persons en the attached Service List by deposit in the United States mail, postage prepaid, this 1st day of November,1978.
Roger A. Newton 1 Senior Nuclear, Engineer Subscribed and sworn to before me this 1st day of November,1978.
,YY//
/ ~ . . n .-
Notary Public,, State of Wisconsin My commissio expires /- /s '..'!?
i
9
. +.',c. V j
l : -
l UNITED STATES OF' AMERICA I
! NUCLEAR REGULATORY COMMISSI0t; . l 4
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD f'
/
In the Matter of l} Docket Nos. 50-266 l l 50-301 t WISCONSIN ELECTRIC POWER l i COMPANY l Amendment to License Nos. ,
5 DPR-24 and DPR-27 I (Point Beach Nuclear Plant, (Increase Spent Fuel h
j -
Units 1 and 2) Storage Capacity) j f
f SERVICE LIST p a !
Marshall E. Miller, Esq. Bruce A. Berson, Esq.
Chairman Office of the Executive Legal Director Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comission U. S. Nuclear Regulatory Comission Washington, D. C. 20555
__ Washington, D. C. 20555 Ms. Mary Lou Jacobi Dr. Emeth A. Luebke Vice Chairperson Atomic Safety and Licensing Board Lakeshore Citizens for Safe Energy ;
U. S. Nuclear Regulatory Comission 932 N. 5th Street i Washington, D. C. 20555 Manitowoc. Wisconsin 54220 I l
Dr. Paul W. Purdom Patrick W. Walsh, Esq. !
245 Gulph Hills Road Assistant Attorney General l
. Radnor, Pennsylvania 19087 The State of Wisconsin !
Department of Justice ;
( Docketing and Service Section 114' East State Capitol i Office of the Secretary . Madison, Wisconsin 53702 ~
U. S. Nuclear Regulatory Comission -
Washington, D. C. 20555 Ms. Sandra Bast 1112 N. lith Street i Atomic Safety and Licensing Appeal Manitowoc, Wisconsin 54220 !
Board :
U. S. Nuclear Regulatory Comissicn Ms. Jame Schaefer i Washington, D. C. 20555 3741 Koehler Drive l Sheboygan, Wi,sconsin 53081 l i
. ~ . .
_p.- -w -e+,--- - ,