ML19344A052
| ML19344A052 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 10/16/1978 |
| From: | Hiestand O DAIRYLAND POWER COOPERATIVE |
| To: | |
| References | |
| NUDOCS 7810270300 | |
| Download: ML19344A052 (39) | |
Text
EC PUBLIC DOCUMENT ROOM 10/16/78 IE qY b UNITED STATES OF AMERICA g
@ h NUCLEAR REGULATORY COMMISSION w
THIS DOCUMENT CONTAINS P0OR QUAUTY PAGES In the Matter of j
uocket No. 50-409
)
Amendment to DAIRYLAND POWER COOPERATIVE
)
Provisional Operating
)
License No. DPR-45 (La Crosse Boiling Water Reactor)
)
APPLICANT'S RESPONSE TO CREC'S SECOND, THIRD, AND FOURTH SET OF INTERROGATORIES Pursuant to 10 CFR f 2.740b(b), Dairyland Power Cooperative (Dairyland), the applicant for an amendment to Provisional Operating License No. DPR-45 in the above-captioned proceeding, hereby submits the following answers 1/
and objections in response to Intervenor Coulee Region l
Energy Coalition's (CREC) Second, Third, and Fourth Sets of Interrogatories to the Applicant and Requests for Production l
of Documents:
[
-1/
Dairyland is furnishing these responses in the hope of expediting this proceeding.
In doing so, Dairyland has purposefully limited its objections cnly to the most obvious cases and, unless otherwise indicated, Dairyland does not concede either (a) that the information sought by any of the subject interrogatories is relevant to any of the four CREC contentions identified in Appendix A to the Licensing Board's Prehearing Conference Orders (Sept. 5, 1978) which have been admitted as matters in controversy in this proceeding and to which the inquiry in this proceeding is limited, or (b) that this infor-mation is even reasonably calculated to lead to the discovery of admissible evidence.
Cf. 10 CFR S 2.740 (b) (1).
,/
75'l00 7G 300
-2 Second Set 2-1. and 2-2.
There is no evidence to date that significant degradation of fuel assemblies stored in water pools has occurred.
This is as expected since the cladding and structural materials selected for fuel assemblies are i
among the most corrosion resistant materials and are essen-tially compatible with the relatively benign water pool environment.
See the detailed study of spent nuclear fuel stored in water pools performed by Battelle (Reference 1).
This study concluded that fuel integrity could be maintained when fuel storage times were extended and fuel storage capacities. expanded.
The conclusion was based on examining the behavior of fuel currently stored in water pools and examining the potential degradation mechanisms under pool storage conditions.
1.
Zircaloy-clad fuel has been stored satisfactorily in pools up to 18 years; stainless-clad fuel has been stored up to 12 years.
2.
I.ow temperatures and favorable water chemistries are not likely to promote fuel cladding degrada-i tion.
3.
There are not obvious degradation mechamisms which operate on the cladding under pool storage conditions at rates which are likely to cause failures in the time frame of probable storage.
-3 Existing and future LACBWR fuel stored in the spent fuel pool uses Type 348 stainless steel as the cladding material.
This material was specifically' selected because of its excep-tional corrosion resistance and favorable nuclear characteristics.
It should be noted that in the future, Zircaloy clad fuel may also be used at LACBWR.
The specific areas of concern listed by the inter-venor in Interrogatory No. 2-1 are addressed seriatta below.
Corrosion Corrosion, as applied to nuclear reactors, is gen-erally meant to describe general surface chemical reactions with the water (or steam) environment and consequent degrada-tion of the material.
Stress corrosion, because of its po-tential significance to the integrity of nuclear reactor pressure boundaries is generally treated as a separate corrosion mechanism.
Stress corrosien is defined as the phenomenon of the cracking of metals caused by a combination of relatively high stress, a corrosive environment and the metallurigical condition of the material.
Accelerated Corrosion The chemical corrosion of the spent fuel assembly cladding and structural members may potentially be accelerated by the pool environment including the gamma radiation levels associated with the spent fuel.
Reference 1 reviewed the
-4 potential mechanisms for chemical corrosion and concluded that these mechanisms do not operate on the fuel assemblies at rates which are likely to cause failures during the probable storage time.
This conclusion is also supported by Reference 2 which states that corrosion is not expected to cause long term problems for storing fuel.
Stress Corrosion Cracking Stress corrosion cracking (SSC) can manifest itself as either transgranalar or intergranular crack propogation.
Transgranular propagation will occur at stress levels well below the yield value in the presence of a halogen con-taminant or will occur at high stress levels without a con-taminant.
Transgranular SSC is not a potentially significant degradation mechanism at LACBWR for the following reasons:
1.
The halide ions in the pool water are maintained at a very low concentration by circulating water through an ion exchange media so that the water is deionized and of a high purity.
The chlorides in the LACBWR pool water are routinely checked and maintained below 0.02 ppm.
2.
The residual stress levels in stored fuel assemblies are low (less than 5 to 10% of the room temperature yield strength) for the following reasons:
-5 a.
reactor exposure tends to relax high stresses from fabrication.
b.
pellet-clad interactions are minimized when the fuel cools down to pool temperatures.
residual gas pressures are low particularly c.
in nonpressurized fuel rods such as LACBWR.
Intergranular propagation requires the presence of high stress levels in the material (at or above the yield stress).
While intergranular SSC can be generated by high stress levels, crack propagation is accelerated if the material is sensitized and/or if a corrosion inducing agent is present (either halide ions or highly oxygenated water).
Inter-granular SSC is not a potentially significant degradation mechanism at LACBWR for the following reasons:
1.
The residual stress levels in stored fuel assemblies are typically well below the yield point.
2.
The stainless steel selected for the fuel cladding is Type 348 which is a stablized grade of stain-less steel developed to eliminate carbon precipi-tation and, consequently, intergranular corrosion (Reference 3).
3.
The halide ion concentration is maintained at a very low level.
t o
4
-6 Reference 1 supports the above conclusions by stating that there is nothing in current fuel storage ex-perience which suggests that stress corrosion cracking is an operative degradation mechanism for either stainless steel or Zircaloy.
Zircaloy is considered to be particularly Lanuned to stress corrosion cracking in an aqueous media.
Reference 2 states that considerable experience has been gained at the Savannah River Plan for 10 years storing packaged fuel assemblies in Type 304L stainless steel.
By maintaining the chloride concentration of the pool water below 5 ppm (at least 250 times higher than LACBWR's pool water), stress corrosion cracking has been avoided and other corrosive effects have been found to be minimal.
Type 304L is formulated to have a low carbon content, thereby minimizing the potential for carbide precipitation and, in turn, inter-granular SSC.
Its behavior supports the selection of Type 348 stainless as a suitable material to prevent intergranulas SSC.
Alteration in mechanical properties Mechanical properties of either cladding material are essentially unchanged by the pool environment.
The principal pool environment which has potential to affect mechanical properties is gamma irradiation.
The additional l
gamma radiation to which spent fuel cladding is subjected during fuel storage is only a small fraction of the gamma irradiation that the material has during its residence in
-7 the reactor (less than 10%).
An increase in gamma exposure of this magnitude will have no effect on the mechanical prop-erties of the stored fuel cladding.
Hydrogen absorption and precipitation Hydrogen absorption and precipitation is applicable only to Zircaloy since austenetic stainless steel is essentially not affected.
The existing LACBWR fuel, new fuel racks and pool liner are, therefore, not affected by this problem.
In general, Zircaloy is not affected in the pool environment unless galvanic-induced hydriding occurs.
Reference 1 states that the latter possibility is not likely for the following reasons:
1.
Relatively high purity water, 2.
Absence of direct Zircaloy-aluminum contacts, 3.
Oxide films on the irradiated Zircaloy fuel rods, and 4.
Normal pool temperatures are well below the range where rapid hydriding has been observed.
References l
1.
A. B. Johnson, Jr., Behavior of Spent Nuclear Fuel In Water Pool Storage, BNWL-2256, Battelle Pacific Northwest Labs, September, 1977.
2.
Draft Generic Environmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel, NUREG-0404, Volume 2, Appendix H, March, 1978.
3.
Properties of 200, 300 and 400 Series Stainless Steel by Republic Steel.
l
-8 2-3. The proposed monitoring system will involve monitoring of the radioactivity (
, and
) in the pool water.
If fuel is moved for other reasons, selected fuel assemblies may also be inspected visually and the results compared with previous visual inspections.
2-4. See answer to Question 2-3.
The industries' ex-perience with the storage of irradiated fuel assemblies in water for long periods of time has indicated that significant degradation does not occur while in storage, therefore moni-toring each individual fuel assembly while in storage does not appear to be warranted.
2-5. Based on extensive industry experience and DPC's own experience with failed fuel, DPC does not believe that the storage of failed fuel in the spent fuel pool will re-quire special monitoring.
2-6. Dairyland does not believe that fuel rods in any fuel assembly will deteriorate to the point that a problem l
will arise.
However, assemblies can always be re-moved, put in a special container for restorage if necessary, and returned to a storage position even if all 440 storage positions are filled.
2-7. (a)
Failed fuel can be stored in either tier.
(b)
There are no particular advantages or dis-advantages in storing failed fuel in either the upper tier or the lower tier.
l
-9 (c)
Failed fuel does not contribute to " crud" buildup.
(d)
There should be no significant effect.
(e)
See response to CREC Interrogatory No. 1-12 and 1-18.
2-8. Dairyland does not anticipate the need to move fuel assemblies from the lower to the upper level.
2-9. (a)
It is not anticipated that the presence of failed fuel will itself require any special treatment, sur-veillance, or handling or disposal difficulties or cause any exesss load on cooling filtering systems.
(b)
Actual operating experience at LACBWR.
2-10.Because of the very low solubility of UO2 in water, and since experience in the industry over many years had in-dicated essentially no additional degradation of defective fuel stored in water, it is unlikely that the defective fuel will need to be encapsulated for storage.
2-ll. (a)
The thickness of the crud on the fuel rods when discharged from the reactor is nominally less than 1.5 mils and remains essentially the same during storage.
The tendency of crud to increase corrosion rate is negligible.
See DPC's response to Interrogatory No. 2-1.
(b)
The basis for this conclusion is the fact that crud (similar to that on LACBWR fuel) is very common on ir-radiated fuel rods discharged from reactors, and the industry
-10 has not experienced any significant degradation in this fuel (including defective fuel) during extended storage in water pools.
See answers to Interrogatories 2-1 through 2-5.
(c)
Failed fuel will not cause more crud.
The crud is primarily oxides of copper, nickel, and iron.
The very low solubility of fuel material assures that it will not be transported to fuel surfaces by the coolant in
~
quanitities sufficient to contribute significantly to the crud layer.
(d)
Not applicable.
(e)
Not applicable.
(f)
Not applicable.
2-12. (a)
The Applicant does not anticipate that there will be any significant additional " problems" associated with handling spent fuel in conjunction with this license amend-The proposed rack design and the Applicant's proposed ment.
rack installation plan (see response to CREC Interrogatory No.1-27) are designed to reduce the number of fuel movements necessary to effect fuel transfers.
These move-ments are generally expected to involve routine fuel handling procedures.
(b)
No additional " complications" are expected due to the presence of failed fuel in the pool.
(c)
Actual operating experience involving the handling and moving of failed fuel elements.
i
-11 2-13. (a) and (b)
The Applicant has not canvassed other reactor operators and is unaware of any other such operators who have or have applied for a two-tier design for their spent fuel pools.
2-14. See the Applicant's response to Interrogatory No. 2-13.
2-15. The existing fuel handling and reloading operations are conducted with the spent fuel water level at the 700-foot level; therefore, no changes in the existing procedures are anticipated.
2-16. (b)
The transfer gate seal is a one-piece, 3/8 inch thick x 3-3/8 inch wide 30-40 durometer rubber gasket which extends along both sides and the bottom of the canal gate.
The gasket is glued and bo.lted to the side of the gate opposite the fuel pool and seals between the canal gate and the canal gate housing.
The gasket is compressed by bolts which, when tightened, press the gate against the housing.
Additional gasket compression results from hydrostatic pressure as the fuel pool water level is raised above the bottom of the canal gate.
(b)
Crud in the pool water will not prevent the transfer gate seal from functioning properly.
The potential for crud deposition on the canal gate housing is negligible becausc the gasket sealing surface is vertical.
When the canal gate is installed, the sealing gasket is compressed
-12 and prevents foreign matter from entering the seal area.
The thickness and compressibility of the sealing gasket enables it to seal around minor imperfections, providing further assurance that the gate will seal properly.
(c)
See response to Interrogatory No. 3-7.
2-17. (a)
See submittals to NRC by DPC letters LAC-5776, dated July 11, 1978, and LAC-5477, dated September 25, 1978.
As shown in these analyses, the reactivity of the storage array decreases with increasing temperature, decreasing water density, and compaction of the storage array.
(b).
If the very low probability cask drop accident should occur, essentially all particulate material or debris produced would be retained in the pool and the pool water would continue to provide shielding for this material.
The only significant radioactivity released to the containment atmosphere would be radioactive gases from fuel rods breached during the accident.
The resulting radiation levels would not prevent personnel from entering the Containment Building I
to perform repair work.
(c)
The suction for the pool cooling system is above the fuel storage racks and the flow rate is relatively low, therefore, significant broken pieces of fuel racks or l
fuel assemblies resulting from a cask drop will r.ot be carried into the system.
l 1
l
-13 (d)
Since the probability of a cask drop accident is considered to be extremely low, and since it has been shown that potential effects outside containment in the event of such an occurrence are not expected to be significant, (Ref. LAC-3187, dated June 13, 1975, and LAC-5341, dated June 7, 1978) the actual conditions after the event would dictate when and how the cleanup would be accomplished.
(e)
The specific man-rem dosage resulting from maintenace of the pool cooling system after a cask drop accident is dependent on the specific conditions and require-ments at the time.
However, as indicated above, pool cooling can be maintained without unacceptable man-rem exposure.
2-18. DPC has not stated that the principal effect of dropping a heavy object into the pocl will be to squeeze water from the rack.
This is one of the effects expected and is of interest when analyzing the effect on the reactivity of the system.
2-20. See DPC's response to CREC Interrogatory 1-22.
2-21. The total estimated cost of the rack change is
$800,000.00 including design, fabrication, and installation costs.
See Applicant's response to CREC Interrogatory No. 1-25 for estimated future operating and maintenance costs.
l
-14 2-22. See LAC-5477, dated September 25, 1978, and LAC-
)
l 3187, dated June 13, 1975.
2-23. (a)
See Applicant's response to Interrogatories No. 3-38, 3-39 and 1-23.
(b)
See Applicant's responae to subpart (a).
(c)
See Applicant's response to subpart (a).
(d)
Heat from the pool is ultimately dissipated in the Mississippi River.
The heat added to the Component Cooling System is removed by the Low Pressure Service Water System using a pair of heat exchangers.
The Low Pressure Service Water System discharges to the Circulating Water System.
(e)
- Yes, (f)
Continuously by an online monitor yielding an integrated dose from gross beta-gamma (nonisotopic).
Monthly samples are also drawn for confirmatory gross beta gamma counting.
(g)
Typically,1 x 10-7 uCi/ml or less.
2-24.to 2-26.
For the same reasons stated in its October 5, 1978 Response to CREC's First Set of Interrogatories, l
Dairyland objects to these interrogatories on the grounds of 1
relevance and materiality in that the information sought through these interrogatories concerns issues which go beyond the scope of the CREC contentions which were admitted as matters in controversy in this proceeding.
l
1 O
-15 l
Third Set 3-1. None.
3-2. See Applicant's response to CREC Interrogatory No. 1-18.
3-3. No.
3-4. See Applicant's response to CREC Interrogatory No. 1-18.
3-5. LACBWR design and operating capabilities and ex-perience.
3-6. Current data (see answer to Interrogatory No. 3-20) does not indicate any trend developing that relates FESW inventory to the frequencies of resin sluicing or filter changing.
The 3-month interval was chosen as a possible worst case. based upon conservative assumptions.
Actual operating experience would suggest a much longer interval between changes.
(See Applicant's response to OREC Inter-rogatory No. 1-18).
3-7. To Applicant's knowledge, it has nowhere stated that the fuel transfer gate will be able to maintain its integrity indefinitely.
However, the transfer gate is re-moved during each refueling outage at which time it is avail-able for inspection.
See " Environmental Impact Evaluation of Spent Fuel Pool Rack Modification," submitted to the NRC by DPC letter Reference No. LAC-5341.
3-8. It is not necessary from the standpoint of safety or radiological protection.
~
-16 3-9. Applicant does not anticipate that there will be any significant additional risk associated with raising the water level in the pool as planned.
The pool is designed to handle a water depth of 40 feet.
3-10. There are presently no known or anticipated con-ditions that would " dictate" that spent fuel recently dis-charged from the reactor be stored in the lower tier of the two-tier storage racks.
The statement was made to indicate that the recently discharged fuel could be stored in the lower i
level if desired.
3-11. The top of the new spent fuel racks will be at the 678' level.
Flow from the fuel pool cooling system enters the bottom of the pool, flows upward thrcugh the stored fuel, and is drawn off at the 679-foot level to return to the cooling system.
Since the spent fuel will still be located within chis flow path, the proposed increase in water level will not prevent the cooling system from properly cooling the spent fuel.
3-12. The major components of the activity on the pre-l sently installed racks consisc of activation and corrosion l
products.
The primary nuclides are cesium and cobalt isotopes.
137 3-13. The primary source contributors are cesium Cs and Cs) and cobalt (60Co) isotopes.
These are not con-sidered short-lived isotopes.
~
-17 3-14. There are no significant amounts of neutron radia-tion present in the fuel pool area, consequently, there will be no significant neutron activation of the storage racks.
3-15. See Applicant's response to Interrogatory Nos. 2-13 and 2-14.
3-16 and 3-17.
Yes, it is conservatively estimated to be
~
approximately 5 x 10 uCi/cc.
3-19, 3-20, 3-22, 3-23, and 3-24.
The FESW is sampled on approximately a monthly basis.
The following analyses are usually performed on the samples:
1.
pH determination 2.
conductivity measurements 3.
turbidity measurements 4.
chloride in concentration S.
gross B-y activity concentration l
6.
gross a activity concentration (since June 1977) l Results of analyses since January 1976 are summarized on the attached sheets while results of previous analyses are maintained at LACBWR's record storage facilities.
t
+
-18
'h FES3 G20SS ACTIVITY azSULTS_
?,
Grcss 4 7 (aci/ml)
Gross a (uci/:ml) i
[
?
Data Icn Exenancer Filter Ion Exenancer tilter u
In l
Out In In Out In 9-26-78 6.69E-3 4.14E-3 6.98E-3 7.53E-7
'1'58E-7
'i. 0 4E-7
' 9-12-78 5.73E-3 1.80E-3 4.96E '5.00E-7
<2.89E-8'.
' NAL
, 29-78 3.73E-3 8.60E-4 3.99E-3 8.01E-8
<4.54R-8 1.16E-7~
'8-22-78 3.452-3 5.94E-4 3.56E-3 2.73E-7
'4.83E-8 1.29E-7 8-15 3.68E-3 3.68E-4 3.96E-3 1.64E-7
<9.12E-8 1.46E-7' 2.25E-7-[I 7-25-78 3.67E-3 1.65E-4 4.25E 1.05E-7 (6.01E-8 2.5'4E-4lF NA-1.34E-7 5.35E-8 2.27E-7
' 11 2.64E-3 5-27.-78' 3-34E-3 1.318-4 3.22E-3
-1.12E-7 S.60E-s
- 8.39E-8
.1 6-13-78 3.33E-3].4.18E-4 2.87E-3 2.08E-7
'5.78E-8 6.93E-8:
5-23-78~
2.66E-3 4.79E-5 2.54E-3
'2.64E-7
<9.80E-8. " 1.85E-7l 5-9-78; 2.'47E 3.97E-5
' 2.56E-3 1.99E-7 4.19E-8 2.20E-7' s
.. 26-78 NS NS' l 3.17E-3 NS.
NS-2.55E 2.74E-3 f 3.76E 4.31E-R 3.56E-7' 4-25-78 2.58E-3 6.14E-5 4-18-78 NS NS l:
NS l 2.14E-7
<2.92E-8 2.82E-7 l
NS 1.99E-7 59.41E-8 2.098-7 i
4-11-78 NS NS a
NS l
NS 2.llE-7,l<9.49E-R l 1.05E-7 6
l 4-4-78 NS h
3-28-78 3,32E-3 2.46E-5 3.26E-3
<l.88E-7 !<l.88E <l.SSE-7 l g
3 2
3-23-78 3.09E-3
- 2. 70E -5 2.64E 3 3.15E-7
<8.12E-8:
2..'3E-7 f
3-6-78 2.95E-3:
2.33E-5
'3.06E-3 2.58E-7'
<7.32E 3.70E 4.33E-7l f
2-28-78 3.12E-3 3.54E-5 2.86E-3' 5.23E-7 1.71E-8 N
2.07E-S l 2.44s-3,
2.50E-7 1.78E-8 3.12E-7 f
2-21-78" 2.07E-3 2.18B-3l1.19E-5! 2.13E-3 l 3.58E-7 5.36E 1.523-7
-[
2-15-78 i.
L i
2-6-78{' 2.00E-3 1.60E-5 t 2,10E-3 i <$. 20E-8. l 3.lOE-7 3.00E-7
'I m
i
{n 1-31-78; 2.54E-3 4.5 9E-5 ! 2.25E-3 4 3.99E-7
<7.llE-8 3.112-7 i L
NS = Not Sampled h
NA
- No Data Available
{
~
~
_ -... _ _.. -. _W K
-19 k
FEsw CROSS ACTTVITY RESULTS-3 Gross :i-v (uC1/ml) n Grcas 4: (uCi/sl) g Date Ion Exchanger Filtcr Ion Exchanger Filter y
In out In in I
out In 2.83E-5l3.12E-3 6.71E-7 6.013-7) 6.27E-7 1-24-78 3.10E-3 1-16-78 3.61E-3 2.33E-5 1.092-4 5.47E-7
<5.47E-8k1.06E-7 1-9-78 3.96E-3 3.28E-S 3.77E-3 8.39E-7 1.71E-8 4.88E-7 1
J 1.3-73 3.11E-3 9.30E-5 3.32E-3 8.56E-7 1.68E-7 9.04E-7 H
- j. -
k a
12-27-77 3.53E-3 2.51E-5 3.27E-3 1.24E-6 1.122-8 6.81E-7 x
12-19-77 3.915-3 2.59E-5 4.27E-3 5.80E-7 3.013-8 6.93E-7 12-12-77 4.78E-3
<9.42E-5 4.74E-3
<5.90E-7
<5.90E-7
<5.90E-7 2.73E-3l4.40E-5!5.86E-3 1.60E-6
<5.52E-8 2.87E=6 12-6-77 11 29-77;
,.,4.23E-3
.. 3. 5.2E--4
,,6.
9 6 E - 3
, 2.,83E.6, 2.91E-7,,,,NA,
,,.[
M io7Erf ""G8h4"35D3E-3'-
"I.~ 5' 5'E-7 75.97sT1 i.41E2T;l :
- i 3
8 1 T--2 T-77~
11-15 1. 42E-2 3.. 9 7 E.5.
l.47E-2 1.14E-6
<4.2SE.8 1.05E.6.L..
I ll-7-77/ '
9.482-3 2.77E-5 S.35E-3
- 6. 92E 1. 8SE-8 l 6.07E-7 b
2.19E-5f2.99E-3" i
4.36E-7
<2.91E-S 2.91E-7 j
10-31-77) 2.83E-3 2.74E-5l3.09E-3l7.53E-7 10-26-77 2.77E-3
<l.40E-2 1.14E-6 i
4 f
3.81E-3;1 2.05E-5 l 4.08E-3' l.21E-6 l < 3. 7 92-7 1.21E-6 10-20-77 f10-10-77l 1.02E-4 l 7.70E-3 f 1.91E 1.79E-7f5.60E-6 3.42E-3 10 77 3.69E-3 1.-23E-4 3.95E-3 6.68E-7
<l.21E-7 7.882-7 ;
L
=
i i
9-26-77 3.73E-3
<4.10E-5 3.67E-3 1.16E-6
< 5.81 E-7' l.31E-6 5
4 3
1.5SE-6 l < 2.21E-7 2.13E-6 j
i 9-20-77 3.72E-3 1.17E-4 4.50E-3 l
2 i
9-12-77 4.04E-3 1.27E-4 l 5.70E-3 1.36E-6 2.04E-7 1.36E-6 5
1.
1.88E-7 f 1.69E-6 h
9-8-77 4.39E-3 9.62E-51 4.59E-3 1.06E-6 t
4.72E-3l 1.29E-4 4.92E-3l 1.25E-6 l 2.63E-7 [ 1.98E-6 l j
9-2--77 8-26-77f5.33E-3 1.36E-6 f <l.78E-7 f 1.42E-6 NA j
NA F.
E NS = Not Sanpled s
T NA = No Data Available
!?
e 5
l
'l
-20 6
i FESW GROSS ACTIVITY RESULT 5 k
Gross s fr (uci/ml)
Grcas a (aci/nl) _
Data Ion Excnanger Filter l
Icn Excnancer Filter j
In cut In i
in cut in p
I l
8-15-77 5.95E-3 3.27E-5 5.53E-3 1.92E-6 3.95E-7 1.86E-6 8-8-77' 7.88E-3 1.02E-4 8.70E-3 1.80E-6
<.1.68E-7.
3.31E-6.
[
8-2-77' l.50E-3
<9.83E-5 NS 4.28E-6
<2.31E-7.-
NS
~
7-28-77 9.74E 1.275-4 NS NA-2.15E-7 N S.-
7-22-77 2.60E-2 2.5SE-4 NS 3.62E-6 2.13E-7 NS
/
7-11-77 1.91E-2 6.59E-4
,N S l 1.29E-6 1.99E-7 NS I
Ns 6.52E-6 1.53E-7' N5 F
7-6-77 L 24E-2 l < 4.60E-7 a
1.10E }
- 5. 732-6"l
,[
3.32E-7
< l.56F7 NS N
6-27-77
~
5-25-77 4.862-2 2.54E-5 NS
--~*'
-[
d-26-77 1.23E-3" <3.21E-6
\\
[
3-29-77 2~.11E-3~
1.15E-3
\\
2-22-77 3.39E-3
1.47E-5
~
'/
6.14E-5I
\\
1-25-77 1.53E-3~
s
_l 12-28-76 2.50E-3' - 3. 40E-5 h
[
t 2.'2dE-51
[
11-23-76.
2.22E-3
~ <1.67E5f f
j l10-26-76 2
2.22r-3 9-28-76 4.31E-3
<1.47E-5l
[\\
,[
\\...
8-24-76l 3.83E-3
<l.92E-5
[
-\\
6-22-76 4.70E-3 1.56E-5 5-25-76 3.42E-3 3.OSE-3l
/
\\
2.78E-3 1.64E-5l
/
\\
f4-30-76
[
2-24-76
- 2. 90E-- 3
- 1. 64E-5 '
3.04E-3f1.G4E-5[
I k
1-27-76 f
4 NS = Not sampled
~
NA = No Dates Available t
FESW C H ICAL ANALYSES
-21. ~
l Data l Location pH conductivity Cl ppm Turbidity eP 9-26-78 FI 5.19 1.60-
<0.02 0.11 9-26-78 11 5.25 1.70 0.15 k
9-26-78 IO' 5.14' l.70 0.11'
)
9-12-78
.FI 5.36-1.57
' ' fo~.'13
~'
' l'. 67 :
0.15 3
9-12-78 II 5.40 9-12-78 IO 5.28 1.74 0.06 i
8-29-78l PI 5.53 1.70 O.12 l
l 0.13
~
"- 29-78 II 5.48 1.70 i
O.14' f
8-29-78 10 5.36 2.00-s I
l 0.37 8-22-78 FI 5.45-1.74 i
8-22-78l
}
O'.31 IT 5.45 1.76 l
]
s--22-78 Io 5.33 2.00 0.40 S
j[.
8-15-78 FI 5.48 1.62.
[
0.12-t I
8-15-78 l II 5.47-1.74 0.09-l}
}
IO 5.36!
1;95 l
l 0.11 i
j 8-15-78f i
l 1.70 0.46 i
7-25-78 FI 5.76 t
t 7-25-78l II 5.73 l
1.60 l
{
0.46
{
f f7-25-78f I
0.45 Io S.60 1.77
~7-11-78 FI 5.50 1.72 0.43
)
i j
g u
i 0.45 5
7-11-78 II 5.45 1.68 i
1 11-78 IO C48 1.96 0.42 I
i f
0.41
[.
6-27-78 FI 5.45 1.90 6-27-78 II 5.40 1.91 0 ~. 4 0 g
6-27-78 IO S.30 l
1.92 l
l 0.40 h
f PI = Filter Inlet II - Icn Exchanger Inlet i;
10 = lon Exchanger Outlet j
5
_v
-22 2s FESW CIEMICAL MIALYSES j
i 3
Date Location pH Conductivity' C1" ppm Turbidity
'j l
6-13-78 FI 5.35 2.10
<0.02
- 0. 33 j
6-13-78 II 5.34 2.10 0.32 1
6-13 -7 8 Io 5.25 2.10
~0.32
{
5-23-78 FI 5.42 1,76 O.35 5-23-78 II 5.42 l
l'822 0;35 5-23-78 IO 5.27 2.' 10 ;
0.36 5-9-78 FI 5.46 1.35
- 0.61 5-9-78 II 5.45 1.61 0'.54 5-9-78 Io 5.31 1.84
' 0e54 i
4-25-78 FI 5.85 1.90 j
0.36 l
,IO. 33 4-25-78 II 5.33 1,.60 4-25-78 IO 5.28 2.00-
- 0.31 l
l 2.04 0.36 3-28-78 FI 5.70 1
3-20-78 II 5.50-1.39 0'.38 3-28-78 10 5.50 l
1.63 l
0.35 l
3-6-78 PI 5.65 1.50 c.49 f
0.48 3-6-78 II 5.55 1.50 i
l 3-6-78 10 5.50-1.64 0.48 2-28-78 FT 5.50 2.04-l 0.55
(
2-28-78 II 5.65 l
1.40' 0.59
"~
r 2-28-78 10 5.' 6 5.
1.36 0.55 i
2-21-78 FI 5.70 l.15
{
0.68 2-21-78 TI 5.75 1.12 0.70 j
2-21-78j Io 5.75 1.30 0.67 i
i FI. = Filter Inlet II - Ion Exchanger Inlet I0 = Ion Exchanger Outlet
c v
sf O
b
-3
{
FES'.i CEICA : N4AI,fSES n
I Date Lccaticai pd Conductivit*f Cl ppm l Turbidit*.'
J
~4 2-15-78 FI 5.48 1,39
<0.02 0.90 n
t II 5.48 1.39 0.78 0
2-15-78l i
j f
I, 2-15-78 10 5.41 1.50 0.75 i
l 0.46 l
2-6-78 FI 5.70 1.37 l
r j
2-6-78 II 5.75 1,42
.0.46 2-6-78 IO 5.70 1.55 0.46 l
l 1-31-7 8 l 0.61 FI 5.60 1.18 l
0.63 1-31-78 I II 5.60 1.20 i
i l
1-31-78 !
IO-5.60 l
1.36 l
I 0.59 I
1-24-7 8 l FI 5.50 l
1.55 0.62 3
l 0.62 i
g l
1-24-78 II
[
5.50 l.47 j
t o
i 1-24-78},
1.80 l
1 0.60 i
IO 5.40 i
YI 5.80 1.36 I
i L-16-78' l
0.38 l
l I
0.38 1-16-78l 11 i.
5.80.
- 1. 3 6..
f i
3 1
1-16-7 8 l 10 5.65 1.48 0.40 t.
f i
i l-9-73; F1
(
5.00 l
1.32 0.34 l'
f I
g 1- 0-78 !
II 4.90 l
1.42 j
0.38 i
i 0.31 l
1-9-7ti-10
}
4.90 l
- 1..:0 l
l
- 0. 12
}
1-3-7 8 l FI l
5.80 1.43 5.80 l
1.33 l
'l O.36 1-3-7.94 II i
i i
i i
1-3-78i 10 5'.60
(
1.50 l
j 0.25 c
}
l l
I f
l I
I e
i e
i I
l
\\'
i i
FI - Filter Inict 11 = Ion Excnanger Inlet IO a Ion Exch.mger Cutict
O t
s i
-24 PESW cre;jfcAL xyJYSES t
conductivity l Cl ppm l Turbidity Location!
pu i
Data t
FI 5.75 1.35
<0.02 I,
O.so 12-27-77l 0.43 l
l 5.30 l
1.35 f12-27-77l a
TT 0.38 l 12-27-77 l 10 5.70 1.67 i
l 0.42 t
!! 12-19-77 Fr 5.42 j
1.51
=
l 0.42 l
l 1.57 5.45 12-10-77 TI f
0.42 l
5.34 1.79
=
12-19-77 Io FI 5.55 l
1.63 l
l 0.36
,12-12-77l l
0.36 12-12-77l II 5.55 l
1.68 l
l f
' Shh f
j 0.34 l12-12-77!
2.00
=
IC FI 5.70 1.36 l
1 0.56
!12-6-7.7l
=
t l
0.52 12-6-77l TT 5.65 i
1.60 f
l 0.61 l
8 12-6-77i 10 5.55 l
i 1.90
=
4 l
5.02
(
1.52 0.65 i~
l
=
11-29-77l FI 0.58 l
,11-29-77l II 5.04 l
1.52 a
5 0.53 IO 4.87 l
11-20-77l 1.60 j
a t
I i
i O.44 l
l 11-21-77.
YI 5.70 l
1.17 i,
il 11-21-77!
II 5.72 l
1.09 l
l 0.46 l
=
f 1
0.41 l
l 5.65 l
1.20 l11-21-77!
To l
i 1.73 0.39 I"11-15-77'l FI 5.72 l
=
i 0.40 11-15-77 l 11 5.77 l
1.71 0.38 l-l l
i
/ 11-15-77 IO
]
S.60 1.94 11-7-77l l
0,.24 FT 5.74 1.08
=
[
c.43 I
II 5.80
. 12 l
11-7-77!
=
f 0.37 I
11-7-77f 1.30
=
IO 5.75 FI a Filter Inlet II = [cn Exchanger Inlet IO = Icn Exchanger Outlet
-.o
]
f.
g E>
M
-25 i
FESW CIIO!ICAL IdiALYSES_
3 5:
Location l p11
, Conductivity Cl ppm Turbidity y
Date
+,
10-31-77 FI l
5.80 1.12
,i <0.02 0.de k
i I
I10-31-77 0.44
~
II 3.85 1.22 l-0 i
1 10-31-77 IO 5.70 1.33 0.46 i
4
- I 5
i 10-26-77 FI 5.80 1.11 O.48 i.,
i 10-26-77 II 5.85 1.26 I.
0.48
.i j10-26-77 10 5.70 1.44 l
0.52 i
I 0.48
- 10-20-77 FI 5.91 J
1.01
~i l
l10-20-77 II I
5.80-l 1.03 0.50 l
0.56 10-20-77 TO 5.70 1.24 i
10-10-77 FI 5.75
[
0.90 i
t 0.45 0.50 l
II 5.80 0.95 l
10-10-77 l l
l
'10-10-77l IO 5.70 O.48 1.09 i
FI 5.70 1.34
}
0.32
!,10- 6-77 f.
l 0.30 i
f i
II 5.95 1.30 l
f10-6-77 i
10- 6-77 l 10 5.65 i
1.44 l
0.30 l
9-26-77j FT 5.73 1.21 i
0.42 l
II 5.7G 1.21 0-26-77l 0.45
=
i 9-26-77 i Io 5.61 1.38 l
O.40 l
l 0.~ 3 2 0.98 9-20-77 FI 5.85 II l
5.80 l
c.99 l
l 0.36 l
9-20-77l l 9-20-77 10 j.
Se72 l
1,17 i.
l 0.34 t.
FI l
S.75 l
9-12-77 (;
0.99 i
l 0.31
}
f l
5.72 I
1.03 l
0.36 i
9-12-77 TT j_"
0.36 l
l 9-12-77 l 1.25 5.60 To FI = Filter inlet I! = Tcn Exchanger Inlet TO s Ion Exchangest outlet
ri 4k'
-26 FESW CHEMICAL ANT *LYsrs uv' s
Da te Location pH l Conductivity Cl pp:2 l ?nrbidity
~
I
^:
9-8-77 F1 5.76 1
1.05
<0.02 0.34 f
9-8-77 II 5.78 1.06 i
0.33 j
i i:
O.33 9-9-77 Io 5.60 1.19 i
L 9-2-77 FI 5.76 1.01 0.49 0.50 9-2-77 II 5.74 1.04 9-2-77 IO S.55 1.18 l
0.43 i
i 1
8-26-77 fit 5.80 1.12 l
0.45 l
1.12 0.35 3-26-77 II 5.90 f
I i
i 8-26-77 {
IO l
5.70 i
1.24 t
0.35 I
i 1
8-15-77 (
FI 5.80 l
1.14
)
0.41 l 8-15-77 !
II 5.80 l
1.14 l
0.45 l
0.45 8-15-77l IO 5.70 1.33 FI 5.70 1.12 0.35
{ 8-8-77 l j
8-8-77 !
II 5.70 1.12 l
0.33 i
i i
i i
8-8-77 IO 5.60 1.30 j
0.32
_i 3-2-77f II 3.63 i
1.13 0.42 1
8-2-77 l IO 3.49 l
1.39 5
0.39 i-0.38 j 7-28-77 !
II 5.80 1.05 6
j 0.21 1.52
{
7-28-77 IO 5.55 1, 7-22-77 II 5.90 l
1.20 0.36 h
0.36 7-22-77!
10 5.50 1.24 l
NA 1
7-11-77 II J
5.74 0.99 IO 5.59 j
1.20 NA l
7-11-77l l
l II 5.92 c.94 f
7-6-77l 0.21 FI = Filter Inlet II = Ion Exchanger Inlet IO = Ton Exchanger Outlet
FEET CITE 4TCAL MALYSES
_27 t
j i
i Date j 'ocatica i pli l Conductivity Cl ppm j Turbidirf i
l 7-6-77l TO 6.12 l
1.12
<0.02 O.39 3
l l
0.24
}
5-25-77l 11 5.82 l
0.S6 I
f l
f 5.87 k'
l l
l 0.73 5-25-77l 0.35 10 1
'l 4-26-77 II 6.02 l
0.41 j
0.41 l
r 4-26-77 l IO 6.08 0.34 j
0.24 f
5.90 l
0.68 l
3-29-77l 0.12 II l
6.00 0.30
[
l 0.09
! 3-29-77 !
10 I
l 2-22-77 i II 5.90 0.70 0.12 i 2-22-77 l 10 S.95 0.40 l
i 0.06 i
i 1-25-77 TI I,
6.04 l
0.43 l
i 0.04
_,i j
I 0.04 1
j 1-25-77 '
IO g
6.10 0.39 l
i I
li e
l 0.08 I
i 12-28-76'l 11 5.95 i
0.56 1
0.03 i
I il2-28-76 IO 5.90 0.55 t
f l
5.30 l
1.10 j
11-23-76 i II 0.03 10 l
5.70 1
1.24 O.02 l.11-23-76 l
! 10-26-76 !
II 5.30 l.01 I
0.02 i
I j
0.02 i
S.70 l
1.09
- 10-26-76,
10 l
i 0.05 l
f l
l
}
9-2H-7 6 i II 5.85
~
1.06
--w l.
l 0.07 j 9-28-76l 10 5.70 l
1.24 l
II 5.90 0.95 h
f3-24-76
'O.05 f
l l
l 0.05
. l8-24-76l 1.13 10 5.80 l
I 5-22-76 I II I
7.10 I
1.02 l
l 0.04 t
?
I j
l 0.04 1
2 l
G.40 l
1.24 t-6-22-76 i To l
FI = Filter inlet II = Icn Exchanger Inlet 10 = Icn Exchanger Outlet
te s
FESW CIL.:UCAL ANALYSES
-28 f
i 9 _ _ - _ _ _,
f Dato T.ccati.cn !
pII Ccnductivity Cl FF:s i Turbidity
}
1 5
II t
6.32 0.94 l
<0.02 0.04 i,
5-25-76l.
6 0 17 d,
l
{
5-25-76 l IO 7.23 0.77 l
II l
6.65 i
0.63
!.'4-30-76l n.05-1
.['
l l
0 202~
l 4-30-76 l 0.39 6.62 IO l
2-24-76 j IT 6.00 l
l 1.10 i
0.-03 i
i i
0,04 5.80 t
1.20
'i 2-24-76 j 10 I
l 5.73 l
1.13 l
0.04
!l-27-76 II l
'O.04' IO 5.70 l
1.33 l
~
1-27-7!i I
}
}
l I'
i I
i.
}
l 1_ _
i j
l
.l I
t i
i I
i l
I i
l' 1
l I
i
}
i i
l t
1
+
)
1 i
l.
1 k
I j
I 4
1 i
i l
-)
i
(
i e
I l
F l
t I
t l
i i-i i
j j
l I
t
}
I l
\\
t j
i I-i i
I f.
4 1
+
i i
I l
[
S 1
i l
.j -
l I
i i
I j
i l
r FI = Filter In:ct II = Icn Exchanger Inlet 10 = Icn Exchanger Outlet N
-29 3-21. The gross gamma activity concentration (1 x 10-3 uCi/ml) of the FESW as reported in the June 7, 1978, submittal was obtained by summing the activity of each gamma component as listed in Section 3.1 of that sub-mittal.
The average gross activity for 1978 (using values to the filter inlet) is 3/16 x 10-3 uC1/ml.
3-25. The latest composition of radioisotopes by weight in the pool coolant is summarized below:
Isotope Coolant Inventory - Grams Mn-54 5.5E-7 Co-58 8.9E-8 Co-60 4.3E-5 Sb-122 1.2E-8 Cs-134 3.5E-5 Cs-137 1.8E-3 Co-144 1.3E-5 3-26. The " crud" is primarily composed of the oxides of iron, copper, and nickel with smaller amounts of magnesium and manganese oxides.
3-27. The activity and quantity of " crud" in the pool has not been precisely ' determined, but experience has shown l
that it does not cause significant problems in the operation of the storage pool.
3-28. The " crud" that is on the fuel rods when they are discharged from the reactor will generally remain on the fuel rods unless removed.
l l
l t
{
l l
-30 3-29. The thickness of the crud on the fuel rods when discharged from the reactor is nominally less than 1.5 mils on those rods on which crud has formed and remains essentially the same during storage.
3-30. Crud formation is not increased due to the presence of failed fuel.
3-31. For a short time period after refueling some xenon isotopes are produced due to the radioactive decay of their precursors.
However, as noted in the Applicant's response to Interrogatory No. 1-20, the quantities are negligible.
Any gases escaping to the containment atmosphere are treated with all other gases in the off-gas stream.
They are passed through various filter banks and holdup tanks before being exhausted to the stack and released to the environment.
3-32. It is contemplated that the old racks will be stored at LACBWR for a maximum period of 2 weeks to 1 month following removal from the spent fuel storage pool.
3-33. No significant personnel exposures are anticipated.
Storage of the racks would be in areas not normally occupied by personnel and a total dose of 20 mrem or less is expected in the time period referenced in the answer to Interrogacory No. 3-32.
3-34. The four gallons per hour represents the minimal amounts referred to in this section.
This value is primarily due to the FESW system leak rate.
O
-31 3-35. Based on prior operating experience, it is estimated that system leakage could increase by approximately 2 gallons per hour.
3-36. The evaporation rate is not dependent upon hydro-static conditions per se.
3-37. The requested diagrams will be made available to intervenors or objections to the production of same will be made within the timeframe specified in the NRC Rules of Practice.
3-38. The Component Cooling System is a closed loop and does not have a " source" or " discharge" in the sense of once through cooling.
Any additional makeup water required is drawn from the demineralized water system.
3-39. The major components in the component cooling system are located in the Turbine Building.
3-40. The appropriate weight and function of objects that may be moved above the pool include:
(1)
Fuel Shipping Cask - 35 tons; Function: Receive spent fuel for shipping.
(2)
Fuel Element -- 385 pounds; Function:
Fule the reactor core.
l l
(3)
Em~ergency Core Cooling System Assembly -- 4,000 pounds; Function:
Provdes emergency core spray coolant to the core in the remote event of a LOCA.
~
-32 (4)
Core Spray Assembly Storage Rack -- 1,500 pounds; Function:
Serves as a support for the core spray assembly when the assembly is removed from the reactor for refueling.
(5)
Fuel Inspection Fixture -- 150 pounds; Function:
Serves as a support for fuel elements during inspection of elements.
(6)
Fuel Shipping Container - 175 pounds; Function:
Holds fuel element while monitoring fuel for leaks.
(7)
TV Camera and Rigging -- 200 younds; Function:
Services as a support frame and means to permit remote under-water examination of miscellaneous items.
(8)
Control Rod -- 275 pounds; Function:
Control reactivity in reactor core.
(9)
Shroud Can Plug -- 160 pounds; Function:
Services as orifice for unfueled reactor core locations.
(10) Source Holder and Source -- 90 pounds; Function:
Holds reactor core startup sources.
~
(11) Miscellaneous Equipment -- Up to approximately 75 pounds; Function:
Equipment and tools used to perform tasks or in.auections underwater.
-33 Fourth Set 4-1. At the present time Dairyland does not anticipate that there will be any need to expand the storage pool beyond the proposed amendment.
l 4-2. Dairyland Power Cooperative has no existing plans for expansion of the spent fuel storage pool beyond the l
440-element total capacity proposed at this time.
4-3. The intent of the expanded fuel storage pool capacity is to have the capability to store spent fuel elements at LACBWR until at least 1991.
4-4. Spent fuel still at LACBWR would be disposed of in accordance with the NRC policies in effect at the time of decommissioning and the mode of decommissioning selected.
4-5. Dairyland obj ects to this interrogatory on the grounds that the information requested is irrelevant to the issues raised in CREC's contentions.
See Dairyland's October 5, 1978 Response to CREC's First Set of Interrogatories.
4-6. The water in the spent fuel storage pool will be processed through the waste water treatment system and dis-posed of as low level liquid waste.
4-7. None, no need exists for this service.
4-8. Spent fuel rods have been stored in pools for up-wards of 18 years.
There is reason to believe that it can-not be stored for much longer periods.
See Johnson, A B.,
Jr.,
-34 Behavior of Spent Nuclear Fuel in Water Pool Storage, BNWL-2256, Battelle Pacific Northwest Laboratories, Richland, Washington, September, 1977).
4-9 and 4-10.
LACBWR's personnel training program meets the requirements of 10 CFR 19, Section 19.12 (Instructions to Workers) ANSI /ANS N 18.1-1971, and ANSI /ANS-3.1-1978.
These requirements are implemented by formal lectures, written procedures, and a special work permit (SWP) issuance program.
4-11. DPC does not belive that there are any significant additional hazards associated with storage in the spent fuel pool.
4-12. See DPC responses to CREC Interrogatory Nos. 2-1 and 2-2.
4-13. No.
No appreciable benefits would be derived from the use of distilled water in lieu of the demineralized water currently being used.
4-14. See Applicant's responses to Interrogatories 19, 20, 22, 23, and 24 of Set 3.
The background information requested by CREC for each of the persons answering these interrogatories was pre-viously provided by Applicant.
The affidavits required by 10 CFR $ 2.740b(b) are attached as Appendix A hereto.
Respectful y submitted,
'Tu
. S. Hiestand Attorney for Dairyland Power Corp.
i OF COUNSEL Kevin P. Gallen
~
UNITED STATES OF MERICA m
NUCIE.AR REGULATORT COMMISSION
. s
~
)
Docket No. 50-409 In the Matter 5f
)
Amendment to
~ O 8
DAIRYIJWD PCWER CocPERATIVE
. )-
Provisional Operating
)
License No. DPR-45 y
'(La Crosse Boiling Water: Raactor); }
' m g
- ' AETIDAVIT zCP SEYMOUR J. RArrui p
State of Wiscons'im County of Vernen:
Sey:ncur J. Raffety. beingJfirst.-duly sworn, on cath.says as
'c -
~*
follows-il.. - 'rhat: he :is 'employedJ by Dairyland Power Cooperative, 2515 radt hvenue South,= La' Crosse, Wisconsin, as. Reactor - Engineer..
3
.~
~ >
' 2. - That he is duly; authorized to answer the;Intarrogatories.
~
j numbered 8,.10, :ll, 17, :18. & : 22 propounded by the Coulee Region j
Energy Coalition.under date of ' September IS,19.78, on behalf of the 9
~
Applicant Dairyland Pcwer Cooperative. (Set 2).
.3.
That' the above-:nentioned and attached answers 'are true-and' ccrrect to' the best o. f his knowladge and belief.
f l $
// "
^?
~
' np l J%c[-
l S. J. Rhffany Subscribed and sworn toi before me this.
)3 day ~of October,~1978.
3
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UIIITED STATES OF AMERICA NUCL".AR REGlJLATORY COW.ISSICII In the Matter of
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Docket No. 50-409
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Ar.cnhont to DAIRYLAND POWER COOPEPATIVE
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Provisional Operatinq
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License No. DPR-45 I
(La Crosse Boiling Water Reactor)
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1 AFFIDAVIT OF NORMAN L. TiOEFERT State of Wisconsin:
County of Vornons i
Norman L. Ecefort being first-duly sworn, on cath says as j
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-l follows:
l.
That he is employed by Dairyland Power Cocperative, 2615 East Avenuo South, La Crosse, Wisconsin, as Mechanical Engineer.
i 2.
That he is duly authorized to answer the Intorrogatories numbered 9, 12-16, 19-21, & 23, propounded by the Coulee Region r
Energy Coalition under date of Septenher 18, 1976, on behalf of the Applicant Dairyland Pcwer Cooperativo. (Set 2).
3.
That the above-centioned and attached answers are truo and correct to the best of his knowledge and belief.
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i Subscribed and sworn to before me this i '3 day of October, 1973.
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.r UNITED STATES OF A'3 ERICA NUCLEAR REGULATORY CCMMISSION l
In the Matter of
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Dcchet No. 50-409 f
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Amen 6 ant to 4
DAIRYLAND POWER COOPERATIVE
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Provisional Operating
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Licenso No. DPR-45 (I.a Crosse Boiling Water Reactor)
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AFFIDAVIT OF SEYMOUR J. RMTETY State of Wisconsin:
County of Vernon:
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Seymcur J. Raffety being < first, duly sworn, on cath -says as L
follows:
j 1.
That he is employed by Dairyland Power Cooperative, l
2615 East Avenue South, La Crosse, Wisconsin, as Reactor Engineer.
2.
That he is duly-authorized to answar the Interrogatories j
numbered 10, 15, 28-30, propounded by the Couleo Region Energy l
Coalition under date of September 18, 1978, on behalf of the l
l Applicant Dairyland Power Cooperativo. (Set 3).
3.
That the above-mentioned and attached answers are true and correct to the best of his knowledga and belief.
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V S. J./ Raffety Subscribed and sworn to before me this f 'A day of October, 1978.
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UNITED STATES OF 7eE.RICA 3C
^7 UUr7 Mut PIGULATCRY CC.9.ISSION S
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Docket No. 50-409 d
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?.cendment to UA1RYLAND POWER COOPERATIVE
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Provisional Operating
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License No. DPR 15 E
(La Crosse Boiling Water: Reactor)
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AFFIDAVIT OP ROBERT J. PPMCE i
State of Wisconsin:
County of Vernen:
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Robert J. Prince being first duly sworn, on oath says as follows:
af That he is e= ployed by Dairyland Pcwer Cooperative, e
1.
2615 East Avenue South, La Crosse, Wisconsin, as Radiation fag Protection Engineer.
N That he is duly authorized to anawer the Interrogatories g
2.
F nu.tered 1-6,84, 12-14, 16, 17, 19-27, 31-34, propounded by the y
J 19, 1978, on y
Coulee Region Energy Coalition under date of September l
a behalf of the Applicant Dairyland Power cooperative.
(Set 3).
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That the above--mentioned and attached answers are true l
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and correct to the best of his kncwledge and belief.
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Subscrif;sd: and sworn to before me this r
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UNITF.D STAT 2S OF AMERICA NUCLZAR REGUIJLNRY COMMISSION In the Matter of
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Docket no. 50-400
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Acendment to DAIR1D ND POWER CCCPERATIV2
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Provisional Cpsrating
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License No. DPR-45 (La Crosse Boiling Water Reactor)
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AFFIDAVIT OF NORMAN L. HOE?ERT State of Wisconsin:
County of Vernon:
Norman L. Hoefert being first v.tly sworn, en oath says as follows:
1.
That he is employed by Dairyland Power Cooperative, 2615 East Avenue South, La Crosse, Wisconsin, as Mechanical Engineer.
2.
That he is ' duly authori=ed to answer the Interrcqatories numbered 7, 11, s 35-40,1 propounded by the coulee Region Encrgy Coalition under data of September 18, 1978, en behalf of the Applicant Dairyland Pcwer cooperative. (Set 3).
3.
That the above-mentioned and attached answers are trua and correct to the best of his knowledge and belief.
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N. Li/ Hoofart i
subscribed and sworn to before me this
/3 day cf October,19 N,
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UNITED STATES OF AMERICA UUCLEAR REGMTORY teriIS$IO!!
In the Matter of
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cocket no.30-409
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M9nrinent to DAIRYLAND PCWER COOPERATIVE
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Provisional Cpurating
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License No. DFB-45 (I,a Crosse Boiling Watcr Reactor)
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AFFIDAVIT OF RCBERT J. PRINCE s
State of Wisconsin:
County of Vernen:
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Robert J. Prince being first duly sworn, on cath says as follows:
1.
That he is employed by Dairyland Power Ccoperative, 2615 East Avenue South, La Crosse, Wisconsin, as Radiation 5
Protection Engineer.
2.
That he is duly authorized to answer the Intorrogatories nunbered 9,.10, 12 & 14, propounded by the Coulee Region Energy Coalition under date of September 18, 1978, en hohalf of the Applicant Dairyland Power Cooperative. (Set 4).
j 3.
That the above-mentioned and attached answers are true and correct to the best of his knowledge and belief.
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1 io U?iITED STATES OF AMERICA R
NUCIE.AR REGULATORY COMMISSION
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In the Matter of
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Docket No. 50-409 f
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Anendment to DAIRYLAND POWER COOPERATIVE
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Provisional Operating
)
License No. DPR-45 (La Crosso Boiling Water Reactor)
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AFFIDAVIT OF SEYMOUR J. RAFFETY 1
State of Wisconsin:
County of Vernont
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Seymour J. Raffety being first duly sworn, on oath says ca
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That he is employed by Dairyland Power Cooperative, 7
4 2515 East Avenue South, La Crosse, Wisconsin, as Reactor Engineer.
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2.
That he is duly authorized to answer the Interrogatories numbered 1-4, 6-8, 11 & 13, propounded by the coulee Region Energy
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Coalition under date of September 18, 1978,_on behalf of the 4
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Applicant Dairyland Power Cooperative. (Set 4).
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3.
That the above-rentioned and attached answers are 3
true and correct to the best of his knowledge and belief.
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t Subscribed..and sworn to before me this
/3 day of Cctieber, 1978.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In'the Matter of
)
Docket No. 50-409
)
Amendment to DAIRYLAND POWER COOPERATIVE
)
Provisional Operating
)
License No. DPR-45 (La Crosse Boiling Water Reactor
)
CERTIFICATE OF SERVICE Service has on this day been effected by personal delivery or first class mail on the following persons:
Charles Bechhoefer, Esq., Chrm.
Docketing & Service Section Atomic Safety and Licensing Office of the Secretary Board Panel U.S. Nuclear Regulatory U.S. Nuclear Regulatory Commission l
Commission Washington, D.C.
20555 Washington, D.C.
20555 Atomic Safety and Licensing Mr. Ralph S. Decker Board Panel l
Route 4 U.S. Nuclear Regulatory l
Box 190D Commission Cambridge, Maryland 21613 Washington, D.C.
20555 Atomic Safety and Licensing Dr. George C. Anderson Appeal Board Department of Oceanography U.S. Nuclear Regulatory l
University of Washington Commission Seattle, Washington 98195 Washington, D.C.
20555 l
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-2 Colleen Woodhead, Esquire Office of Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Richard J. Goddard, Esquire Office of Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Richard Shimshak Plant Superintendent Dairyland Power Cooperative La Crosse Boiling Water Reactor Genoa, Wisconsin 54632 Fritz Schubert, Esquire Staff Attorney Dairyland Power Cooperative 2615 East Avenue, South La Crosse, Wisconsin 54601 Coulee Region Energy Coalition P. O. Box 1583 La Crosse, Wisconsin 54601 David S. Simpson Rt. 3 Box 34 Durand, Wisconsin 54736 Ellen Sabelko 929 Cameron Eau Claire, Wisconsin 54701 b+u
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- 0. S. Hiestand Dated:
October 16, 1978
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