ML20031D585
| ML20031D585 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 09/24/1981 |
| From: | Sinderman R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20031D553 | List: |
| References | |
| ISSUANCES-OLA, NUDOCS 8110130487 | |
| Download: ML20031D585 (19) | |
Text
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TESTIMONY OF ROGER W. SINDERMAN CONCERNING RADIATION DOSES TO THE PUBLIC FROM FUEL STORED NEAR THE SOUTH WALL OF THE SPENT FUEL POOL CHRISTA-MARIA CONTENTION 2 AND O'NEILL CONTENTION IIA I.
Introduction The purpose of this testimony is to respond to Christa-Maria contention 2 and O'Neill Contention IIA by showing that spent nuclear fuel stored near the tapered south wall of the spent fuel pool will not result in radiation dose rates off-site in excess of regulatory limits.
Consumers Power Company has performed calculations describing radiation doses to the nearest off-site location as a result of radio-active shine from spent fuel stored adjacent to the tapered 1/
south wall.
Two cases were considered as follows:
1.
The existing fuel channel rack remains in place as shown in Exhibit 1 and the location of fresh spent fuel is limited to that portion of the tapered wall occupied by the new type E Rack shown in the same Figure.
Same as above, but only fuel having decayed 2.
one year is placed in new-rack E.
II.
Discussion Calculations for each case described were made taking into consideration the thickness of the tapered wall, These calculations are set forth in Sinderman Exhibit 2.
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the nearest distance to the site property boundary in line
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with the direction of the radiation exposure, the shielding afforded by the containment building, and the shiciding of the air between the containment building and the site boundary.
Radiation attenuation caused by forest growth as well as that produced by the auxiliary building structure uns neglected to afford a degree of conservatism in the results obtained.
As has been previously indicated in Mr. Axtell's testimony concerning Christa-Maria Contention 2 and O'Neill Contention II-A, Consumers Power Company has chosen to keep the existing fuel channel rack (Rack C of Sinderman Exhibit 1) at its present location.
This ensures that spent fuel cannot dgs be stored near the thinnest section of the south wall of the spent fuel storage pool.
Further, only fuel having decayed for at least one year after removal from the reactor will be stored in the outer three rows of the new Type E storage rack (those rows most closely adjacent to the fuel pool wall).
Therefore, the results for Case Number 2 represent the actual conditions.
Case 1 is, however, presented to show that radiation dose rates at the site boundary would not be significant even if new opent fuel were to be stored in the "E" rack next to the tapered portion of the south wall. qv-
The dose rates immediately outside the south wall of the pool have been calculated by NUS, as described in the testimony of William Bell.
From this one can calculata the radiation dose rate at the site boundary, as follows.
The spent fuel in the "E" rack was represented by a disk source with an area equal to the area formed by the vertical plane of the array of spent fuel in the "E" rack.
The The radius of this disk was calculated to be 3.45 feet.
disk source term was calculated as that necessary to produce the dose rates calculated by NUS for the outside <C the south (that is, through an average 5.25 feet of concrete shielding).
wall Accounting for attenuation by the containment structure and
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intervening air, the appropriate formula to be used for cal-culating radiation dose at the site boundary is:
(pc c+Pa )
t d
In r +d B B e-R
=
ac YR f*
2 g
2 2 In(w +r )
2 y
4 is the dose rate in millirem per hour at the Where:
Ryg site boundary w is the pool wall thickness ii: feet for each case r is the disk radius in feet for each case
- V(~\\,
O wis the dose rate in millirem per hour at the R
outside of the pool wall in each case d is the distance to the site boundary (2900 feet)
B and B are the gamma build-up factors for air c
and containment respectively are the mass absorption co-efficients for air pa and pc and containment steel respectively is the thickness of the containment building and tc steel (3/4 of an inch)
See Hine and Brownell, RADIATION DOSIMETRY, Academic Press (1961), Chapter 16.
The results of the calculations for each case are
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shown below:
TABLE I RADIATION DOSES AT THE SITE BOUNDARY RESULTING FROM SPENT FUEL STORED ADJACENT TO THE SOUTH FUEL POOL WALL ANNUAL DOSE AT REGULATORY LIMITS CASE SITE BOUNDARY 10 CFR 20 10 CFR 50 APPENDIX I 1
0.022 mr/yr 500 mr/yr 10 mr/yr 2
1.6 E-04 mr/yr 500 mr/yr 10 mr/yr The regulatory limit of 10 millirem /yr established by 13 CFR Part 50 Appendix I is a limit on plant effluents.
applies to doses due to releases of radioactive materials, It and not to doses due to radioactive shine such as those
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'j calculated in this testimony.
There are no releases of
's radioactive mater!
'.s to the environment through the south Therefore wall of the Big Rock Point Plant spent fuel pcol.
the Appendix I limit referred to in Christa-Maria contention 2 and O'Neill Contention IIA is not applicable, although it is shown in Table for illustrative purposes only.
III.
Conclusions In both cases, the dose at the site boundary due to shine through the south wall of the spent fuel pool is within the regulatory limits of 10 CFR 20 and the design objectives of 10 CPR 50, Appendix I.
In keeping with the constraints of maintaining radiation exposures te both the workers and the general public as low as reasonably achievable
( ALARA), however, Consumers Power Company will insure that (only) spent fuel having decayed for at least one year will be stored in the outer three rows of new rack E.
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SKBanbury, DAtc August 2h, 1981 PQygg[
Susacci DOSE RATE AT THE SOUTINEST SECTOR SITE 50UNDARY FR0t4 THE SPEUT FUEL POOL-BIG ROCK POINT PLANT INTCRNAL Conacsromocuct CC R!Earusich, P-2h-110
'SSB-81-13 The following calculations were performed by the writer at your request. A technical review by R;darusich note some conservatisms in the calculation; shielding provided by the control root vall and the containment vall thickness being greater than the 3/h" thickness used due to the angle of the rays.
A detailed description of the calculation methods used and resulting data is attached. Final results listed below:
Dose rate at the site boundary from fuel stored with one year decay is 1 58E-Oh mr/yr.
Dose rate at the site boundary from newly irradiated stored fuel is 2.17E-02 mr/yr.
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S 4
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4 (CIIRkSTA-MARIA CONTENTION 2 and O'NEILIi CONTENTION IIA)
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Paga 2 of 6 l
Dose to person at site boundary from spent fuel area.
l Reference RAE memo 25-79 (attached), the dose rate at El 600'6" is 38 mR/hr.
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Scaled Dvg C-3 to determine distance frca spent fuel area to site boundary in southeast sector.
l 2680' nearest property boundary t
. Approximately 2900' SW Sector distance line per technical specifications First assuming the source is a point source determine dose rate at site boundary I
without shielding.
From RAE calculation used h.1' (greatest distance therefore most conservative) as the distance from source to dose point which obtained the dose rate of 38 mR/hr outside the spent fuel pool area.
h.1 (38 mR/hr)(2h)(365) hrs = mR/yr 9 boundary 2900 yr= 0.67 mR/yr @ boundary without shielding i
Second determine dose rate using the shielding provided by the contaiment shell and air.
For determining mass attenuation coefficients and buildup vill assume 1 MeV ray.
Containment shell is 3/h" iron cm = 1,91 3
p = T.86 g/cm n
O y/, = 0.0599 cm'/g V
px = 0.89h Assume ux = 1 for buildup 3 = 1.87 em = 88.393 2900' of air
/
34,800 inches of air p = 0.001293 5 c" y/,= 0.0636 cm /g 9 x = T.26 l
To find buildup of air vill use average of buildups of aluminun and water B
= 17 1 17 1 + 13.8 15 5 buildup for air l-
=
v r
aluminum = 13.8 l
-UX I
- I Be I = (0.67)(1.87)(15 5)e (0.89h+T.26)
I = 5 58E-03 mR/yr at cite boundary Nov ve vill recalculate the dose using a disk source:
Radiation Dosimetry Hine & Brownell Page 763
References:
BRP Drawing M-101' BRP Drawing'M-103 Section B
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Paga 3 cf 6 Scaled dravi.,8 to find dimensions of fuel. pool.
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Hei6ht of pool 31' j
j Height of fuel 6'
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.=1%
l ol'f, N
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Equation for disk source where b = o h +
where: q = mil 11 curies y " h in N
2 f = gamma constant a = radius of disk
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b = perpendicular distance of dose point from axis of disk h = vertical distance of dose point from plane of disk In both cases it is assumed that There are two individual cases to calculate.
(3 the channel rack is stored near the thinnest portion of the spent fuel pool vall i
V and that fuel is stored next to the channel rack which has a concrete vall thicknes greater than h.5 feet.
Case 1 Fuel stored with 1 year decay Determine radius of disk - height of fuel is 6' vidth of spent fuel rack showing throu6h
>h.5' concrete thickness is 6.25' 6' X 6.26' = 37.5 sq ft 2 An equivalent circle ur E=r 11 9 = r 3.45' = r W
Dose rate = 2.5 mR/hr from NLJS Safety Analysis for BRP spent fuel case Graph of dose rate as a function of concrete thickness h = 5 25' or 160.0cm from RAE calculation is greatest distance between source and dose point and will therefore be r.ost conservative a = 3.h5' or 105 2cm radius of disk Solve for qr 2
gr = (Hv)(n )'
2+n" 2
O V
g h h
Paga 4 of 6 2
(2.5)(105 2 )
- 160 + 105.2 160 gr = 7 70E+0h Solve for dose rate at the site boundary - approximately 2900 feet or 8.839E+0 hem utilizing the shielding effects of the containment shell and air.
Determine buildup factors and pX's assume 0 5 MeVE containn.ent shell is 3/h" iron em = 1.91 3
p = 7.86 g/cm 2
8/p= 0.08ho em fg, pX= 1.3 Biron = 2.3 2900' of air em = 88,393 3
34,800" of air p = 0.001293 g/cm
"/p= 0.0870 cm /cm VX = 9.9 to find buildup for air vill use average of buildup 3 for aluminum and water O
38.9 + 77.6, 58.3 buildup for air
> _ i,,m. 33.9 2
Uvater
= 77.6 h2n2'(Be-UX}
RY a gr,2n 2
g a
h i
In 88,3932 + 105.2 (2.3)(58.3) e (1.3+9.9)
Ry=770$0h 105 2 88,393 By = 1.807E-08 mR/hr X 8760 hrs /yr
= 158E-Oh mR/yr @ site boundary
}
Case 2 Newly irradiated fuel stored Radius of disk is the came as in case 1 of 3.45' Done rate = ratio dose rates of fuel decayed one year of 38 mR/hr i.;
thinnest portion of wall and 2 5 mR/hr at thicker portion of vall and Dose newly irradiated fuel at thinnest portion of wall at 2300 mR/hr rate obtained from NUS Safety Analysis for Spent Fuel Pool.
_38_, 2300 25 X
Dose Rat = 151.2 mR/hr
(-
Paga 5 of 6 l
}
I.j Where: h = 5 25' or 160.0 cm j
Need to colve for gr a = 3.h5' or 105 2 cm
- )
fC' J
3 qP = (R )(a ) _
j In 'h2
,2' f
h l
,151.2)(105 2 )
l
(
]602 + 105 2
~
, 7 7, 160 qr = h.66E+06 i
Solve for dose rate at site boundary - approximately 2900 ft or 8.839E+ hem shielding of containment shield and air Determine buildup factors and pX's assume 1 ! EVE em = 1.91 3
Containment shell 3/h" fron p = 7.86 g/cm
"/p = 0.0599 uX = 0. 899 iron = 1.87 Assume px = 1 for buildup
- ,r-t em = 88,393 3
2900' of air p = 0.001293 g/cm 34,800 inches of air
"!p = 0.0636 cm fc 2
0X = 7.26 To find buildup for air vill use average of buildups for aluminum and vater 13*0 +
" 15*5
~
Daluminum = 13.8 2
water
= 17.1 2
2' n = EL in (Be ~ UX) b.d6E 4 88,3932 + 105 22 ' (1.87)(15 5)e (0.899+7 26)
T 2
2 F
3,
2 105 2 L
88,393 Y
= h.9h7E-06 mR/hr X 8760 hrs /yr R
Y
= h.33E-02 mR/yr 9 site boundary R
.....m._..
s Paga 6 of 6 In coing these two cases there was no consideration made for the decay of the fuel l
over the year time period.
This first year, the newly irradiated fuel vill drop significantly in dose rate in an exponential fashion and then level off.
(Almost in half).
To'be conservative, we vill divide the dose rate obtained for the two cases by two.
Dose rate for newly irradiated fuel = k.33E-02 mR/yr 1.58E-04 mR/yr Dose rate for fuel decayed 1 year
=
h.33E-02 + 1 58E-04/2 = 2.17E-02 mH/yr average dose rate for newly irradiated fuel The average dose rate for 1 yr old fuel vill be conservatively assumed to remain constant at 1 58E-Oh mR/yr O
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CHRISTA-MARIA CONTENTION 2 l
AND l
O'NEILL CONTENTION IIA l
A.
THE CONTENTIONS Christa-Maria Contention 2:
I The increase in fuel stored in the Big Rock pool will I
result in an increase in the amount of radiation re-leased to the environment at the south wall of the storage pool where there is less shielding, according i
to the licensee's Description and Safety Analysis.
This increment in the level of radiation released to the environment enhances the risks to the health and safety of the public in the vicinity of the plant.
O'Neill Contention IIA:
The routine releases of radioactivity during the in-I stallation of new racks, the loading of those racks, and storage of fuel in the racks will exceed the ex-posure of workers, as will the releases of radioactivity through the south wall of the pool exceed the limits imposed by Appendix I to CFR Part 50 on exposure to the general public.
B.
MATERIAL FACTS AS TO WHICH THERE IS NO GENUINE ISSUE TO BE HEARD.
1.
The south wall of the spent fuel pool tapers from a maximum thickness of 5 ft. 9 inches to a minimum thickness of 3-1/2 feet.
(Affidavit of Charles Axtell at p.4).
2.
If freshly discharged spent fuel were to be stored l
adjacent to the 3-1/2 foot thick portion of t5'e south wall of the spent fuel pool, the dose on the outside of the spent fuel pool wall would be 2300 millirem /hr.
(Affidavit of William Bell at p.4).
O, L'
. /"N U
3.
If spent fuel which has a decay time of one year or more were to be stored adjacent to the 3-1/2 foot thick portion of the south wall of the spent fuel pool, the dose on the outside of the spent fuel pool wall would be 37.3 millirem /hr.
(Affi-d..vit of William Bell at p.4).
4.
The dose rate due to radioactive shine from stored spent fuel at all other locations outside the spent fuel pool walls and floor will be less than 2.0 millirem /hr., *ollowing the proposed installa-tion of new storage racks.
(Affidavit of William Bell, Ex. 1).
5.
Following the proposed rack replacement, the exist-ing channel rack will not be moved from its present location at the western (thinnest) portion of the south wall of the spent fuel pool.
(Affidavit of Charles Axtell, at p.8 and Attachment C, Affidavit of Roger Sinderman Concerning Radiation Doses To The Public From Fuel Stored Near The South Wall of The Spent Fuel Pool at p.2).
6.
It is impossible to store spent fuel in the channel rack.
(Affidavit of Charles Axtell at pp.7-8).
7.
Therefore, the NUS calculations of 2300 mrem /hr.
and 37.3 mre:a/hr. are an overestimate of actual dose rates which could occur at the outside of the i_
_ : ($)
south wall of the spent fuel pool due to radio-l active shine from spent fuel stored in the Big l
i Rock Point pool following the proposed license amendments.
(Affidavit of William Bell at p.3; 1
l Affidavit of Charles Axtell at pp. 7-9).
l l
8.
Licensee will store only spent fuel with a decay l
l time of one year or more in the outer three rows l
l of the proposed 9 x 9 fuel rack near the south wall of the spent fuel pool.
(Affidavit of Charles l
Axtoll at pp. 8-9; Affidavit of Roger Sinderman 1
concerning Radiation Doses to the Public From Fuel Stored Near the South Wall of the Spent Puel Pcol at p.2).
9.
Therefore, the dose rates outside the south wall of the spent fuel pool due to radioactive shine from 1 year old stored spent fuel will be approxi-mately 2 mrem /hr.
(Affidavit of Charles Axtell at p.8).
10.
A dose rate outside the south wall of 2 mrem /hr.
due to radioactive shine from stored spent fuel is small in comparison with the existing dose rate of about 30-40 mrem /hr. due to the filter sock i
tank.
I l
11.
The area outside the tapered portion of the south wall of the spent fuel pool is a radiologically
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l
~
__ O controlled area which is infrequently entered by plant workmen.
(Affidavit of Charles Axtell at p.8).
12.
An incremental dose rate of approximately 2 mrem /hr in the infrequently entered, radiologically con-trolled area outside the south wall of the spent fuel pool would not present a radiation hazard to plant workers or result in occupational exposure exceeding 10 CFR Part 20 limits.
(Affidavit of Charles Axtell at pp. 9-11, 17-20).
4 13.
The annual dose rate at the site boundary due to radioactive shine of one year old spent fuel through the south wall of the spent fuci pool 4
would be.00016 millirem /hr.
(Affidavit of Roger Sinderman Concerning Radiation Doses to the Public from Fuel Stored Near the South Wall of the Spent Fuel Pool at p.4 and Exhibit 2).
14.
This dose rate was conservatively calculated since the attenuation due to forest growth and the auxiliary building structure were neglected.
(Affidavit of Roger Sinderman Concerning Radiation Doses to the Public from Fuel Stored Near the South Wall of the Spent Fuel Pool at p.2).
15.
A dose rate to the public at the site bour.dary of-00016 millirem /yr. is within the legal limit
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l O
imposed by 10 CFR Part 20 and within 10 CFR Part 50 Appendix I, and therefore radioactive shine through the south wall of the spent fuel pool would not present a risk to the public health and safety.
(Affidavit of Roger Sinderman Concerning Radiation Doses to the Public from Fuel Stored Near the South Wall of the Spent Fuel Pool, at pp.4-5).
16.
The total radiation exposure to plant workers during the proposed rack replacement operation will be about 18.2 man-rem.
(Affidavit of Charles Axtell at pp. 11-16).
17.
After installation of the new storage racks, the storage of additional spent fuel will not increase dose rates in the spent fuel pool area.
(Affi-davit of Charles Axtoll at p. 19; NRC Safety Eval-untion at pp. 3-13 and 3-14).
18.
Refueling operations, which will not be affected by the installation of new storage racks, are the principle cause of occupational exposure to workers from the spent fuel pool since little worker activity occurs over or near the spent fuel pool during routine operations.
(Affidavit of Charles Axtell at p.19).
d L
_ O 19.
Occupational Exposure will not increase due to storage of additional spent fuel in the Big Rock Point spent fuel pool.
(Affidavit of Charles Axtell at p.19; NRC Safety Evaluation Report at pp. 3-13 and 3-14).
20.
The radiation protection procedures to be used at Big Rock Point Plant for the proposed rack replace-ment operation and thereafter are in compliance with all applicable federal regulations.
(Affi-davit of Charles Axtell at p.19).
21.
Individual occupational doses during the rack re-placement operation and thereafter will be main-tained below 10 CFR Part 20 limits by these radiation protection procedures.
(Affidavit of Charles Axtell at pp. 17-19).
22.
Individual occupational exposure in the vicinity of the spent fuel pool during the proposed rack replacement operation and thereafter will be main-tained as low as reasonably achievable.
(Affi-davit of Charles Axcell at pp. 12-14, 17-19 and Attachment D; NRC Safety Evaluation Report at pp. 3-12 through 3-14).
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V C.
DISCUSSION The first part of Mr. Axtell's affidavit serves as an introduction to and description of the Big Rock Point spent fuel pool.
The south wall of the spent fuel pool at the level of the stored spent fuel tapers to a minimum thickness of 3-1/2 feet.
Christa-Maria Contention 2 and O'Neill Conten-tion IIA express concern about potential exposure to plant workers and to the public due to radioactive shine through the south wall.
The affidavits of William Bell and Charles Axtell demonstrate that, taking into account the location of the proposed storage racks and licensee's commitment to store only spent fuel which has decayed at least one year in the outer three rows of the spent fuel rack nearest the south wall, the dose on the outside of the south wall will be about 2 millirem /hr.
Moreover, this dose rate will occur in an infrequently visited area which is already radiologically controlled due to the presence of the filter sock tank (30-40 mrem /hr.).
Therefore, the incremental dose rate of 2 mrem /hr.
in this area due to radioactive shine through the south wall presents no hazard to plant personnel and is as low as rea-conably. achievable.
The affidavit of Roger Sinderman establishes that the dose rate to the public at the site boundary due to radioactive shire through the south wall, conservatively O
1 E
_ AV calculated, is.00016 mrem /nr., which of course is trivial and well within regulatory limits.
The affidavit of Charles Axtell explains the radiation procedures in effect at Big Rock Point Plant, and the specific measures which will be taken during and after the proposed rack replacement operation to ensure that exposure to plant workers is within regulatory limits and as low as reasonably achievable.
These affidavits conclusively establish that radia-i tion from spent fuel through the south wall of the spent fuel pool is not a problem.
With respect to potential occupa-tional exposure to plant workers (other than due to shine through.the south wall) due to the proposed rack replacement l
operation and thereafter, Mr. Axtell's affidavit and the NRC Staff's Safety Evaluation Report both describe the measures which will be taken to maintain exposures within regulatory limits and'as low as reasonably achievable.
Both PJ. A).tell and the NRC Staff conclude that following rack replacement storage of additional spent fuel will not result in addi-u tional occupational exposura.
Therefore, there is no genuine issue of material fact even with respect to those aspects of Christa-Maria Contention 2 and O'Neill Contention IIA which do not involve the south wall.
Summary disposition of both contentions is appropriate.
'O
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STATE OF MICHIGAN
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UNITED STATES OF AMERICA Y
NUCLEAR REGULATORY COMMISSION BEFORE THE ATCHIC SAFETY AND LICENSING BOARD In the Matter of
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CONSUMERS POWER COMPANY
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Docket No. 50-155
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(Spent Fuel Fool Erpansion)
(Big Rock Point Nuclear Plant
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AFFIDAVIT OF CHARLES E. AXTELL I, Charles E. Axtell, of lawful age, being first duly sworn, do state as follows:
I am employed by Consumers Power Company as the Plant Health Physicist at the Big Rock Point Plant. I have held this job for 13 years. In this job my responsibilities include monitoring and controlling personnel exposure, ALARA considerations, controlling off-site releases of radioactive materials, and plant water chemistry. My resume has been pre-viously submitted.
I am the author of the Testimony of Charles E. Axtell concerning Christa-Maria's Contention 2 and O'Neill Contention 11.2.
To the bebt of my knowledge and belief, the statements in this affidavit and in the above testimony and Figures and Attachments thereto are true and correct.
S' ? 0Yh Charles E. Aztell SUBSCRIBFD AND SWORN TO before me this Mday of October, 1981 RK* Y) 44 Notary Public
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PATRICIA E. KUJAWsKl Notery Public,ChartenisgM
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Eugene A, ffzieazic' "fcD't"f"^'*.05"'r **
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March 6,1983 k
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