ML19296B975
| ML19296B975 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 01/30/1980 |
| From: | Jordan W, Sheldon K BIER, MILLS, CHRISTA-MARIA, ET AL, SHELDON, HARMON & WEISS |
| To: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| NUDOCS 8002220489 | |
| Download: ML19296B975 (32) | |
Text
.*
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
)
In the Matter of
)
)
CONSUMERS POWER COMPANY
)
Docket No. 50-155
)
(Big Rock Point Nuclear
)
Power Station)
)
)
INTERROGATORIES TO CONSUMERS POWER COMPANY (SET I)
Pursuant to 10 CPR S2.740, Intervenors Christa-Maria, JoAnn Biers, and James Mills request that these interrogatories be answered fully, in writing, and under oath by any employees or representatives of Consumers Power Company (" Licensee") who have personal knowledge of the facts or issues in question.
The answer to each interrogatory should contain the name and identification of each person supplying or contributing to the answer, whether or not he or she has verified the answer.
The answer should also explain the role of each individual in preparing the answer.
For the purpose of these interrogatories, (a) the term " spent fuel rack" refers to each com-pleted assembly that is or will be installed in the spent fuel pool to hold spent fuel.
An example appears at page 3-3 of the Spent Fuel Rack - Description and Safety Analysis prepared for this application and dated April 1979 (re-ferred to as the Description and Safety Ana-lysis).
The term includes both existing and proposed spent fuel racks unless otherwise specified.
8002220 W m
(b) the term " component" means one of the parts with which something is manufactured or con-structed.
For example, the components of a spent fuel rack include, but may not be limited to the spent fuel cells or cans, the top and bottom grids, the leveling legs, and the lead-in guides (See pp. 3-3 and 4-4 of the Description and Safety Analysis), as well as the rivets, nuts, bolts, plates, insulators, weld wires, and other parts of the spent fuel rack.
Similarly the pool liner constitutes a component of the spent fuel pool, as do any filters, demineralizers, water circulators, and other parts of the spent fuel pool system.
In addition, the term refers to all raw materials from which the component was made.
Each question is to be answered in four parts as follows:
A.
Provide the direct answer to the question.
B.
Identify and provide all documents and studies, and the particular parts thereof, relied upon by Licensee as the basis for the answer.
C.
Identify and provide all documents and studies, and the par'ticular parts thereof, examined but not relied upon by the Licensee, which pertain to the subject matter in question.
D.
Explain whether the Licensee or any independent contractor is presently engaged in further re-search or work that may bear on the issues covered in the interrogatory.
If so, please identify the research or work and the people responsible for it.
There are four sets of questions set out below.
In each set, the number of the relevant Contention precedes the number of the question. For example, the first question under Contention 3 is numbered 3-1.
Please use these numbers in providing your answers.
_3_
CONTENTION 2 Licensee states that increasing the spent fuel pool storage capacity will cause an increase in dose rates out-side a portion of the south wall of the spent fuel pool.
(Description and Safety Analysis, p.
7-1).
2-1 What is the basis for Licensee's statement concerning radiation doses outside the south wall?
Please include all technical calculations and other studies.
2-2 Describe and define the area referred to as "outside a portion of the south wall?"
2-3 What are the present dose rates outside the south wall?
2-4 What are the increased dose rates that Licensee expects outside the south wall?
a.
Where are these dose rates expected to occur?
b.
What studies have been done to determine whether increased dose rates will occur at other points outside the south wall?
2-5 What is the maximum distance from the south wall at which radiation from the spent fuel pool can be detected given the existing pool configuration?
a.
What would be the maximum distance from the south wall at which radiation from the spent fuel pool
~
would be detectable after the capacity of the spent fuel pool has been expanded, assuming the same conditions that gave rise to Licensee's prediction of increased dose rates outside the south wall?
b.
Answer Question 2-4(a) assuming the expanded ca-pacity spent fuel pool to be filled to capacity, including a recently offloaded full core.
2-6 What types of radiation would be involved in the in-creased doses outside the south wall?
Specify the radionuclides that would be released and the amounts of each.
2-7 What distinguishes the south wall from the other walls?
a.
Explain the difference between the dose rate pre-dicted outside the south wall adjacent to the new type "E" rack and the dose rate predicted outside the north wall adjmcent to the larger new type "F"
rack.
2-8 What are the present dose rates under the spent fuel pool floor and outside the north, west, and east walls?
2-9 What dose rates are predicted to occur under the spent fuel pool floor and outside the north, west, and east walls after fuel storage expansion has been completed?
a.
Where are these dose rates predicted'to occur?
b.
What is the basis for these predictions?
c.
What studies have been done to determine what dose rates will occur under the pool floor and outside the north, west, and east walls?
2-10 What is the maximum distance below the floor of the spent fuel pool, from the north, east, and west walls, and above the spent fuel pool at which radiation from the pool can be detected given the existing pool configuration?
a.
Answer this question assuming that the capacity of the pool has been expanded as requested and assuming the same conditions that gave rise to Licensce's pre-diction of increased dose rates outside the south wall.
b.
Answer Question 2-10(a) assuming the expanding capa-city spent fuel pool to be filled to capacity, includ-ing a recently off-loaded full core.
2-11 Describe the physical characteristics of the area, both above and below ground, outside the spent fuel pool.
a.
How close is the nearest gr.oundwater to the spent fuel pool?
b.
Has groundwater ever been tested for contamination by the spent fuel pool?
If so, what were the results?
c.
Is the area outside the spent fuel pool where increased doses are predicted classified as a " restricted area" or an " unrestricted area" under 10 CFR 20. 3 (a) (14),
(17)?
2-12 Explain why the radiation dose rates discussed in response to Questions 2-3, 2-4, and 2-7 will not exceed the permis-sible levels established by 10 CFR 20.101 and 20.105.
2-13 Explain all Licensee proposals to reduce these increased dose rates, including, but not limited to, the use of additional shielding and the implementation of administra-tive controls.
Provide the justification and technical calculations for the reduction in dose rates from each of these proposals, the estimated cost of each proposal, and any present commitment by the Licensee to any of the proposals.
2-14 Explain in detail why the Licensee anticipates that expansion of the spent fuel pool capacity by over 100%
will cause "no increase in dose rates over those previously experienced due to radio-nuclides...at the edge of the pool." (Design and Safety Analysis, p.
7-1). What are these dose rates and precisely where and how are they measured?
2-15 Explain in detail why the Licensee anticipates that expansion of the 9 pent fuel pool capacity will not affect the frequency of replacement of the domineralizer resin beds. (Letter of December 19, 1979, p. 5).
a.
Licensee states that " spent fuel pool water is typically cycled through the radwaste system domineralizer on an annual basis" (Letter of December 19, 1979, p.
5).
Explain under what circumstances more frequent filtering would be necessary.
IIas more frequent filtering ever been required or performed?
Explain.
b.
Approximately what percentage of radioactive saturation of the domineralizer resin beds is attributable to the spent fuel pool water?
c.
What exactly is the "little noticeable effect" that annual cycling of spent fuel pool water has had on the frequency of replacement of the resin beds?
2-17 Explain and provide the technical calculations for the conclusion that the crushing of a recently off-loaded full core and of all the stored fuel in the remainder of the 441 available storage cans would produce doses at the site boundary of less than 8% and 3% of the thyroid and whole body dose limits specified in 10 CFR 100.
Provide a copy of Table 8-1 of Licensee's July 1, 1974, letter to the AEC concerning this conclusion.
CONTENTION 3 3-1 Is type 304 austenitic stainless steel used in any of the existing spent fuel racks or elsewhere in the spent fuel pool?
Which racks and where?
3-2 Answer Question 3-1 with respect to the proposed addi-tional racks.
3-3 Provide the following information with respect to the proposed additional spent fuel racks and their compo-nents. Please provide a separate set of answers for each spent fuel rack.
If answers are the same for different racks, they need not be repeated.
a.
Deecribe the rack, identifying each component.
b.
Who will manufacture the rack?
c.
From whom will Licensee purchase the rack?
d.
Who will manufacture each component of the rack?
e.
What is the composition (metalic, chemical, physi-cal) of each component?
f.
Provide a detailed description of the manufacture of each component from the raw materials to the completed product.
Identify and explain every test or quality assurance check that is performed on the component during the manufacturing process
_9_
or after manufacture is completed, but prior to receipt by Licensee.
g.
Identify and explain all terts, research, and quality assurance checks and all test, research and quality assurance reports and other documents relied upon, referred to, or otherwise available to Licensee concerning the performance or use of the spent fuel rack components (and the materials from which they are made) in general and in the spent fuel pool environment in particular.
The requested documents include, but are not limited to, documents showing the history of the manufac-ture, performance, or use of the spent fuel rack components; documents provided by the manufact-urer, vendor, a' consultant, or other sources; documents recommending or otherwise judging the components.
h.
Identify and explain any problems that have been identified in the use of the spent fuel rack components, both in general and in the spent fuel pool environment in particular.
Explain why the problems do not threaten the safe operation of the spent fuel pool.
i.
Identify and explain all changes that may occur in the spent fuel rack components during the time
~
they are in the spent fuel pool.
Changes include, but are not limited to chemical or physical reac-tions.
Explain the impact of these changes on the integrity of the spent fuel rack and the safe operation of the spent fuel pool.
j.
Answer questions 3-3 (e), (f), (g) and (h) with respect to the spent fuel rack.
k.
Who will install the spent fuel rack?
1.
Provide a detailed description of the installation of the spent fuel rack.
Identify and explain every test or quality assurance check that is performed by either the installer, the Licensee, or anyone else, to assure correct and safe instal-lation of the r'ack.
m.
Answer questions 3-3 (f) and (g) with respect to installation of the spent fuel rack.
n.
What is the design life of this spent fuel rack?
Provide the basis for your answer and explain how the conclusion was reached.
Include all calcula-tions, computer programs or models, experiential data or information, assumptions, and the justifi-cation for all assumptions.
o.
Explain the physical, chemical, radioactive, and other stresses to which the spent fuel rack or its components will be subjected in the spent fuel pool environment.
p.
Explain the effects of these stresses on the spent fuel rack or its components.
q.
Explain the synergistic effects of combined stresses, including scismic stress, on the spent fuel rack or its components.
r.
Explain why Licensee decided to use this spent fuel rack, as opposed to other possible racks,
.taking into account at least the design of the rack, the components from which the rack will be made, and the experience using this type of rack in other spent fuel pools.
3-4 Please answer all parts of Question 3-3 with respect to the existing spent fuel racks, substituting the words
" relocate" and " relocation" for " install" and "installa-tion."
Where the questions use the future tense, please answer with respect to past events and experience as well as with respect to the future.
3-5 Please answer all parts of Question 3-3 with respect to the spent fuel pool liner.
Where the questions use the future tense, please answer with respect to past events and experience as well as with respect to the future.
3-6 Please identify all other components of the spent fuel pool that will be subjected to the spent fuel pool environment and answer all parts of Question 3-3 with respect to those components.
Where the questions use the future tense, please answer with respect to past events and experience as well as with respect to the future.
3-7 Dr. John Weeks, an NRC consultant, recently informed the NRC that a number of stress corrosion cracks have been discovered in Type 304 austenitic stainless steel pipe containing boric acid in spent fuel storage pools.
Letter to Herbert Conrad, NRC Division of Systems Safety, October 30, 1979).
Weeks stated that PWR spent fuel pools frequently contain several thousand ppm of boric acid, which is used with demineralizer resins to remove radioactive halides from the pool water.
(BNL-NUREG 23021, pp. 3-4).
Please provide the following information concerning the use of boric acid, the function that boric acid performs in other spent fuel pools, and the chemistry of the spent fuel pool water:
a.
Has boric acid or boron ever been used in connec-tion with the spent fuel storage pool at Big Rock?
b.
If boric acid or boron has been used in the spent fuel pool, please explain how and when it has been used and provide a complete description of its function.
c.
Even if not used for any specific purpose, has boric acid or boron ever been present in the pool water?
If so, in what amounts, when, why, and if not intentional, how was it discovered?
d.
Will boric acid or boron be used in connection with the spent fuel pool after the proposed addi-tional spent fuel racks have been installed?
How and for what purpose?
e.
Will boron or boric acid be present in the spent fuel pool water in any amount after the proposed additional racks have been installed?
f.
If neither boron nor boric acid will be used with the domineralizer resins to filter radioactive halides from the pool w'ater, please explain in detail how that function will be performed.
g.
If neitner boron nor boric acid will be used in any way in connection with the spent fuel pool, please explain how the functions that it serves in other spent fuel pools will be performed at Big Rock.
h.
If the answers to Question 3-7(d) and (e) are "no,"
what tests or studies has the Licensee parformed or referred to for assurance that corrosion cracking and other problems found in spent fuel pools where boron or boric acid are present will not occur at Big Rock?
i.
What is the expected normal water chemistry in the expanded capacity spent fuel pool?
j.
What are the maximum expected ranges of the water' chemistry in the spent fuel pool?
O CONTENTION 7 7-1 Please explain whether there is or will be any airborne radiation in the atmosphere above the spent fuel pool (but within the containment) as a result of spent fuel pool operation, including fuel movement and pool mainte-nance activities, and including all credible accidents.
What is the source of the airborne radiation?
What type of radiation is it?
7-2 How are airborne radiation levels in the atmosphere above the spent fuel pool (but within the containment) monitored?
Please provide a complete description of the monitoring that has been done to date, including complete records of the airborne radiation levels.
7-3 Does Licensee project any increase in the above air-borne radiation levels as a result of increased fuel storage? If so, under what conditions?
If not, please justify your answer.
7-4 If Licensee projects any increase in the above airborne radiation levels, please explain the projection and provide all technical calculations and other supporting i
material.
7-5 How has the above airborne radiation been removed from the containment atmosphere to date?
How does Licensee intend to remove it after the spent fuel pool storage capacity has been expanded?
7-6 Please explain all of the possible pathways that the above airborne radiation could follow, including any ways that the radiation could be released to the atmos-phere outside the containment.
7-7 Assuming the spent fuel pool to be filled to capacity, including a recent addition of a full core, what would be the level of airborne radiation in the atmosphere above the spent fuel pool (but within the containment) ?
7-8 Provide all projections of increased dose rates in the atmosphere outside the containment as a result of the increased storage of spent fuel.
Justify your answer.
7-9 What is the role of the containment itself in prevent-ing the escape of airborne radiation from the spent fuel pool into the outside atmosphere?
7-ic Have any questions been raised concerning the ability of this containment or similar containments to perform that function?
Please explain.
7-11 In a letter dated October 19, 1979, the Licensee responded to staff questions issued September 4, 1979 Question #2 concerned leak collection and detection f roin the spent fuel pool.
The Licensee explained that
"[p]eriodic inspections" were made at collection lines which terminated at an "open basin" and drained into the reactor building sump and the liquid radwaste system.
a.
How often are such " periodic inspections" made and by whom?
b.
Does the open collection basin hold any leakage that may occur, or does this basin simply drain into the reactor building sump?
c.
Is it possible to observe directly the liquid level of the open collection basin?
d.
The Licensee stated that "[m]oisture observed during periodic inspection is believed to be due to condensation,rather than pool leakage."
What is the basis of this belief?
Is there a chemical distinction between such condensation and spent fuel pool water?
e.
Have there been any evaporation losses from the open basin?
If so, what has happened to the resulting airborne radiation?
9 e
CONTENTION 9 9-1 Does Licensee have an emergency plan for the evacuation and protection of plant workers and the surrounding pcpulation in the event of an accident at the Big Rock Plant? Dlease provide a copy.
9-2 What is the maximum release of radioactivity assumed as the basis for the emergency plan?
Justify this assump-tion.
9-3 What is the worst case accident or condition assumed as the basis for the emergency plan?
Justify this assump-tion.
9-4 What other assumptions were made in deciding the extent of evacuation or protection provided for in the emergency plan?
Justify these assumptions.
9-5 Was the existence of the spent fuel pool taken into ac-count in developing the emergency plan?
Explain in detail.
9-6 To what extent is the spent fuel pool assumed to contri-bute to the maximum release of radioactivity assumed for emergency planning?
9-7 Are there any conditions under which a spent fuel pool accident alone is assumed to justify activation of the emergency plan to some degree?
Explain in detail.
9-8 What state and local government officials were consulted in developing the emergency plan?
Please describe all such consultations.
9-9 What state and local officials were consulted or contacted concerning the emergency plan from the time that it was developed until March 28, 1979?
Please describe all such contacts.
9-10 Please answer Questions 9-8 and 9-9 with respect to private organizations or individuals, including par-ticularly major employers, hospitals, churches, agricul-tural or farming organizations, and major retailers.
9-11 Please answer Questions 9-8 and 9-9 with respect to federal officials, including particularly representatives of the Departments of Defense, Agriculture, and the In-terior.
9-12 What assurances does Licensee have that the emergency plan will be workable under optimal weather conditions?
9-13 What assurances does Licensee have that the emergency plan will be workable under the worst weather conditions normally experienced each winter?
Under the worst a
credible weather conditions?
9-14 Please explain all actions that Licensee has taken or consultations, conversations, or meetings that Licensee has had concerning emergency planning for the Big Rock plant since March 28, 1979.
9-15 Has Licensee examined the question of whether increas-ing the amount of spent fuel stored at the Big Rock plant will have any impact on emergency planning require-ments?
9-16 What is the total amount of radioactivity from each radioactive source that would be stored in the existing spent fuel pool when filled to capacity?
9-17 Please answer Question 9-16 with respect to the expanded capacity spent fuel pool when filled to capacity.
9-18 Is it possible for a loss of coolant accident in the spent fuel pool to lead to a fuel melt reaction in the pool?
Please explain in detail.
9-19 Describe the three worst spent fuel pool related accidents that the Licensee acasiders credible.
a.
What would be the environmental and radiological consequences of these accidents with 193 fuel assemblies in the spent fuel pool?
b.
What would be the environmental and radiological consequences of these accidents with 441 fuel assemblies in the spent fuel pool?
9-20 Please describe all actions that the Licensee would take if each of the accidents described in response to Question 9-19 were to occur.
9-21 What actions would Licensee take in each case discussed in response to Question 9-20 if the accident occurred while the containment was contaminated such that it could not be entered?
What if the containment became contaminated as a result of a reactor accident imnediately following the spent fuel pool accident?
9-22 What is the basis for the conclusion that the new 9 inch center to center spacing of storage cans yields a km of 0.886 for the proposed storage system under nominal conditions at 68 F.
(Description and Safety Analysis p. 4-1) Provide the technical calculations and related studies.
9-23 Answer Question 9-22 with respect to the conclusion that the worst case situation would result in km less than 0.950 with a confidence level of 95%.
a.
Explain the consequences if km were to exceed 0.950.
9-24 Justify assumptions "a" "g,"
which are referred to as being the bases for the referrence design calculation.
(Description and Safety Analysis, p. 4-2).
Explain how each " nominal condition" is controlled, the extent of actual variation in each condition, and the extent to which any such variation increases the final km.
In addition, please provide the following information with respect to each of the assumptions:
a.
Was the calculation for the reference design based on the use of all three fuels?
What was the percentage of each fuel?
For what type of fuel and for what period of time does the Licensee have contracts?
b.
What are the consequences and probability of non-uniform distribution of the enriched U234?
Provide the technical basis for using 3.80 w/o U235 as
" nominal."
c.
How often, to what extent, and why do water tempera-ture deviations from 68 F occur in the storage pool?
(i) Explain whether the pool water contains any chemicals that assist in filtration.
(ii)
Table 1 of BNL-NUREG 23021 provides the storage pool environments for 27 nuclear reactors. The lowest pool temperature listed is 105 F.
How does and how will the Licensee maintain the pool temp at 68 F particularly when the pool is filled to capacity?
d.
To what extent are burnable poison and cobalt targets present in fuel assemblies?
e.
Why are infinite rather than actuel dimensions assumed for the fuel rack cells?
f.
Where are control rods and fuel channels stored?
g.
To what extent do rack structural materials, other than storage can stainless steel, absorb neutrons.
9-25 Licenseo presents a " Worst Case Analysis of Tolerances and Calculational Uncertainties" in its Description and Safety Analysis, pp. 4-3 to 4-7.
Under nominal condi-tions the new storage rack design is calculated to yield a neutron multiplication factor (kco) of 0.8862 (p. 4-6).
Provide the technical justification for each parameter and the resultant 6 km for " worst tolerances" as listed on p.
4-6.
a.
Rack Can Pitch.
(i) Define in detail " Rack Can Pitch." (ii) Explain in detail why " Rack Can Pitch" was evaluated with a margin of error of i O.12 inches.
(iii) Does a greater margin of error for
" Rack Can Pitch" produce a greater /1 km ?
b.
Rack Can Size.
(i). Define " Rack Can Size."
(ii)
Explain in detail why " Rack Can Size" was evaluated with a margin of error of i 0.03 in.
(iii)
Justify the final 6 k Does a greater margin of m.
error for " Rack Can Size" produce a greater [i ke (iv) Justify the final 6,k co.
c.
Minimum Water Gap.
(i) Explain whether " Minimum Water Gap" refers to the space between the fuel assembly and the inner walls of the SS-304 storage cans.(ii)
What exactly is the size of the " Minimum Water Gap" used to calculate this 6 km ?
(iii)
What is the effect on the gs km if, for any reason, all water is drained from the storage can and rack?
(iv)
Under what conditions would the maximum f km be calculated for the minimum water k mfor the minimum gap?
(v)
Justify the final a water gap?
d.
Can Stainless Steel Composition.
(i) How do variations in the composition of austenitic type 304 stainless steel affect the g3 km ?
Explain the basis for determbling the probability of any such variation.
(ii) What extent of variation and what probability of variation in the composition of the type 304 stainless steel were assumed in computing the b km for "Can Stainless Steel Composition?"
Justify those assumptions.
e.
Can Wall Thickness.
(i) Explain in detail why the Can Wall Thickness was evaluated with a margin of error of + 0.01 inches (p. 4-8).
(ii) Explain the effect on the 3 kmfor can wall thickness.
f.
Eccentric Loading of Fuel.
(i)
Please describe in detail the process of unloading fuel from the reac-tor and moving it to the pool and the process of moving fuel within the pool.
(ii)
What would prevent a fuel assembly from dropping between any two storage racks to the bottom of the pool?
(iii) What guarantee is provided to ensure that administrative rules concerning the handling of casks are followed by the nuclear technicians?
What degree of human error is factored into this gs km?
(iv)
Provide a detailed technical justifi-cation for the computation of this g ka as 0.0158.
9-26 Explain in detail why the algebraic sum of the "wcrst tolerancer, gs km" overestimates the effects of com-bining such tolerances.
Explain and justify the calcu-lation and use of the root mean square of all the tolerances.
9-27 Explain and justify the use of the figure 3.96 for the maximum enrichment value of U-235 (Description and Safety Analysis, p.
4-8).
9-28 Explain why the "as-fabricated limiting design fuel bundle is expected to have a non-uniform enrichment distribution" (p.4-1, emphasis added),
t 9-29 Provide a detailed explanation of "[t] he expected maximum average steam void near the top of the fuel assembly" during a pump failure which causes surface boiling of pool water (p.
4-9, emphasis supplied).
06 Provide the technical calculations which yield 20.6 for this figure.
9-30 Explain how the risk of a critical reaction occurring in a storage rack with 12 inch spacing (k m = 0.80) compares with the risk of a critical reaction occurring in a storage rack with nine inch spacing (k 0.886).
=
m 9-31 Explain how the plant " emergency action level guide-lines" will be affected by the addition of 248 spent fuel assemblies.
9-32 Explain how a reactor core melt would affect the spent fuel in the storage pool.
9-33 Describe in detail the " failed fuel fractions" noted on page 6 of the November 19, 1979 letter from Licensee to the NRC.
a.
What is the effect of failed fuel assemblies on the spent fuel pool.
b.
Will these failed fuel assemblies be removed with the failed storage rack?
If so, how and where will the failed fuel assemblies be disposed of?
9-34 Provide a diagram showing the location of the sight glass located on the south wall of the fuel pool for observation of the surge tank.
a.
Where is the sight glass relative to the new storage rack "E?"
9-35 In responding to a question from the NRC, Licensee has revised its estimates of total "one-time" worker radia-tion exposure resulting from the removal of a failed fuel rack and installation of the new racks.
(Letter of October 19, 1979, p.
10-11).
Please provide the following information relevant to that discussion:
a.
A copy of revised Subsection 3.2.3 of the Environ-mental Impact Evaluation.
b.
A copy of " Table 9-1."
c.
An explanation and technical basis for each of the figures in Table 9-1.
d.
An explanation of the apparent discrepancy between the 23.3 man-Rems total exposure figure shown on page 10 of Licensee's letter and the 18.2 man-Rems total exposure figure shown on the following page.
9-36 Please provide the following information with respect to the failed fuel rack:
a.
How will the failed rack be decontaminated?
b.
Where will the failed rack be stored, and how will it be protected?
c.
To what extent will the failed rack remain radioactive after it has been " decontaminated."
9-37 Provide a copy of the cask drop analysis submitted to the AEC on July 1, 1974.
What weight casks are now used during loading operations?
What weight casks will be used in the future?
9-38 In the Licensee's October 1, 1979, reply to NRC ques-tions dated August 17, 1979, Licensee states that it has corrected a "significant overconservatism in the previous prediction of long-term spent fuel heat load."
p.
2.).
a.
What was the overconservative prediction ?
b.
Why was it overconservative?
c.
What assumptions and calculations produced the overconservative prediction?
d.
Why and how did Licensee make the " overconservative prediction" errors that it is now correcting?
e.
Explain in detail and justify by technical calcula-tions the following revisions to Table 6-1 of the Description and Safety Analysis.
These revisions are shown in Appendix A to its letter of October 1, 1979:
i.
The reduction in normal refueling heat load.
ii.
The reduction in full core offload heat load.
iii.
The reduction in maximum pool temperature due to failure of a spent fuel cooling pump for normal refueling and full core offload.
iv.
The increase in time before the storage pool would boil after an electrical failure (for both normal refueling and full core offload conditions).
9-39 In its October 1, 1979, letter to the NRC, Licensee states that it has not considered the " double pump failure of the spent fuel cooling system considered in the Descrip-tion and Safety Analysis...in accordance with the NRC guidance on spent fuel pool modifications..."
(p.2.).
a.
Provide the justification for the Licensee's conclusion that' Big Rock Point is exempt from cur-rently applicable NRC guidance on pool modification.
b.
Is Big Rock incapable of complying with currently applicable NRC guidance on spent fuel modifications?
Please provide the basis for your answer.
c.
Provide the technical calculations for the maximum thermal load assuming a double cooling pump failure.
d.
Verify that this maximum thermal load has-been con-sidered in the analysis of the existing racks, liner, and concrete pool structure.
a t'- 4 0 Provide a copy of the Licensee's letter dated June 28, 1977, which discusses the health and safety consequences of the loss of integrity of a dropped or struck fuel assembly (October 1, 1979 p.
7).
9-41 Explain the consequences of worst case drop (24-ton spent fuel transfer cask) on the edge or corner of a rack should the safety slings fail.
9-42 The October 1, 1979, revision of Section 6 of the Design and Safety Analysis provides a revised assembly o
inlet temperature of 132 F for a single active failure condition following a full core offload 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown (Appendix a p.
6-4).
a.
Provide the technical calculations for this re-duced inlet temperature and explain why the ori-ginal value has reported as 153 F.
b.
Explain and provide the technical calculations for the conclusion that the bulk temperature will not exceed 132 F for a full core offload, c.
Explain and provide the technical calculations for concluding that under these conditions, "the maxi-mum fuel rod surface temperature will not exceed 198 F, providing a 39 margin to local boiling."
(Id).
o
~
d.
Explain and provide the technical calculations for the conclusion that the " margin to bulk boiling is 0
77 F in the assemblies and 58 F in the intercell region."
(Id).
c.
Explain in detail why questions 9-42(c) and (d) represent the limiting thermal condition in the pool.
Respectfully submitted, W
fn f~ek :~
f William S.
Jofdan, III N.
2
/.k '.
Ln I-
~
Karin P.
Sheldon SHELDON, HARf40N & WEISS 1725 "I"
- Street, N.W.,
Suite 506 Washington, D.C.
20006 Counsel for Christa-Maria, et al.
Dated:
January 30, 1980
e UNITED STATES OF AMERICA NUCLEAR REGULATORY COfD1ISSION BEFORE TIIE ATOhiIC SAFETY AND LICENSING BOARD
)
In the Matter of
)
)
CONSUf1ERS POWER COMPANY
)
Docket No. 50-155
)
(Big Rock Point Nuclear
)
Power Station)
)
)
CERTIFICATE OF SERVICE I hereby certify that copies of "Christa-Maria, Et Al, Interrogatories to Consumers Power Company (Set I),"
were mailed postage pre-paid this 31st day of January, 1980, to the fallowing:
Herbert Grossman, Esq.
Janice E. Moore, Esquire Atomic Safety and Licensing Counsel for NRC Staff Board Panel U.S.
Nuclear Regulatory U.S.
Nuclear Regulatory Commission Commission Washington, D.C.
20555 Washington, D.C.
20555 Dr. Oscar II. Paris Joseph Gallo, Esquire Atomic Safety and Licensing Isham, Lincoln & Beale Board Panel 1050 17th Street, N.W.
U.S.
Nuclear Regulatory Commist, ion Suite 701 Washington, D.C.
20555 Washington, D.C.
20036 Mr. Frederic J.
Shon John O'Neill, II Atomic Safety and Licensing Route 2, Box 44 Board Panel Maple City, Michigan 49664 U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Docketing and Service Section U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 W: W1 && dc y Wil1iam S.,J6rdan,/III a/
.