ML20132F880
| ML20132F880 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 09/23/1985 |
| From: | Andrews R OMAHA PUBLIC POWER DISTRICT |
| To: | Butcher E Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-3.D.3.4, TASK-TM LIC-85-420, TAC-59783, NUDOCS 8510010426 | |
| Download: ML20132F880 (24) | |
Text
e Omaha Public Power District 1623 Harney Omaha, Nebraska 68102 402/536 4000 September 23, 1985 LIC-85-420 Mr. E. J. Butcher, Acting Chief Operating Reactors Branch #3 Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
References:
1.
Docket No. 50-285 2.
Letter OPPD (W. C. Jones) to NRC (D. G. Eisenhut) dated January 26, 1981.
3.
K. G. Murphy and K. M. Campe, " Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criterion 19," 13th AEC Air Cleaning Conference, August, 1974.
4.
San Onofre 2 and 3, FSAR, Appendix 15B, Dose Models Used to Evaluate the Environmental Consequences of Accidents.
Dear Mr. Butcher:
NUREG-0737, Item III.D.3.4 Control Room Habitability In accordance with NUREG-0737, Item III.D.3.4, Omaha Public Power District submitted a study which confirmed that 1) control room operating personnel would be adequately protected in the event of accidental releases of radio-nuclides or toxic gases, and 2) the plant could be safely operated or shut down in the event of design basis accidents.
The results of the study indicated that operator radiation exposure for a postulated LOCA were less than the limits established by the Standard Review Plan (SRP) 6.4 for beta skin and gamma whole body doses. However, the thyroid dose exceeded the SRP 6.4 limits. As a result, OPPD took corrective actions by installing an iodine monitor in the Control Room in order to monitor iodine exposure during a LOCA.
In addition, a shield wall was installed in the Control Room corridor to reduce the direct shine radiation from containment l
spray piping.
i The corrective actions and the study, (Reference 2) were reviewed and found to l
be acceptable by the NRC.
However, during the latter part of 1984, the NRC per-formed similar calculations using the Campe and Murphy methodology and the assumptions provided with the Reference 2 study.
Because the NRC results re-flected a higher whole body gamma dose, questions were raised concerning the X/Q dispersion values. Subsequently, OPPD reviewed the calculations support-ing the study and concluded that the initial analysis was acceptable and the 8510010426Bhh85 PDR ADOCK O PDR f
8 l
F tmpioumen g in a a opportunity assv4
LIC-85-420 Page Two results were conservative due to the diffused source model and virtual distance assumptions used. The NRC reviewers agreed that the calculations were conserva-tive. However, with no experimental data available to prove the virtual distance theory, the calculations were considered to be unacceptable.
In order to resolve the control room habitability issue, and to further reduce post accidedt radiation doses to the control room operators, OPPD will extend the control room air intake (VA-65) from the existing location (3.9 meters from the containment wall) to a distance of at least 10 meters from the containment wall. The new calculated doses using a 10 meter distance and NRC accepted meth-odology are lower than the limits identified in SRP 6.4.
A summary of the control room habitability calculated doses are included in Attachment A.
Justification for continued operation (including input data for the computer calculations, assumptions and the resulting doses), until shutdown for the scheduled refueling outage are included in Attachment B.
These doses represent the control room air intake in its present location of 3.9 meters from the containment wall. Attachment C represents the input data for the computer calculation, assumptions and resulting doses for the control room air intake relocated to 10 meters from the containment wall.
Checking and design verifica-tion of these calculations is currently being performed and will be finalized prior to plant restart.
The additional calculations performed and the proposed relocation of the VA-65 ductwork provide substantial justificat_ ion to complete resolution of the control room habitability issue. The relocation of the intake will be completed prior to restart after the Fall 1985 refueling outage.
Sincerely, R. L. Andrews Division Manager Nuclear Production RLA/AB/rs Attachments cc: LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N.W.
Washington, D. C.
20036 Mr. E. G. Tourigny, NRC Project Manager Mr. L. A. Yandell, NRC Senior Resident Inspector
ATTACHMENT A Summary of Control Room Habitability Calculated Doses Additional calculations have been performed for the existing configuration (3.9 meters duct distance) in order to justify continued operation until shutdown for the 1985 refueling outage. These calculations were done us-ing the Campe and Murphy methodology, (Reference 3) and plant specific assumptions. The results of the calculations and assumptions used are included in Attachment B.
The following doses were calculated:
Dose (thyroid)
Dose (Gamma Whole Body) 2.7 rem <
5 rem (1)
=
Dose (Beta Skin)
To resolve the Control Room habitability issue, the control room filtered air intake (VA-65) will be extended from its existing location (3.9 meters from the containment wall) to-at least 10 meters from the containment wall.
These calculations were done using the Campe and Murphy methodology, Refer-ence 3.
The results of the calculations and assumptions used are included in Attachment C.
The following doses were calculated:
Dose (thyroid)
= 10.97 rem < 30 rem (1) 2.83 rem <
5 rem (1)
Dose (Gamma Whole Body)
=
Dose (Beta Skin)
Based on the above doses the operator exposures are below SRP 6.4 limits.
(1) NUREG-0800, Standard Review Plan (Section 6.4)
(N8-C )
ATTACISENT B Justification for Continued Operation (3.9 meter location)
I.
Assumptions The following specific assumptions were used in the analysis.
1.
The reactor core equilibrium noble gas and todine inventories are based on l ong-tenn operation at 100% of the ultimate core power level of 1500 Mwt.
2.
One hundred percent of the core equilibrium radioactive noble gas l
inventory is immediately available for leakage from the containment.
3.
Twenty-five percent of the core equilibrium radioactive iodine inven-tory is immediately available for leakage from the containment.
4.
The release of fission products to the containment is assumed to occur instantaneously and transported instantaneously to the environ-ment and Control Room assuming no decay during transport time.
5.
Of the iodine fission product inventory released to the containment, 91% is in the form of elemental iodine, 5% is in the fonn of partic-ulate iodine, and 4% is in the fonn of organic iodine.
6.
Radioactive Decay - Credit for radioactive decay for fission product concentrations located within the containment is assumed throughout the course of the accident to the Control Room.
7.
Containment Iodine Removal System - Credit for the removal of iodine from the containment buil di ng atmosphere is assumed during the course of the accident resulting from filtration in the iodine removal system.
The system consists of -four air handling units; two havi ng filtering capaci ty a nd the other two havi ng no filtering capaci ty.
-In the calculation of the radiological consequences an adsorption efficiency of 0.9 (90%) and operation of only one of the two containment filtering units is assumed.
Credit for elemental,
. particulate, and organic iodine removal is taken for the course of the accident.
8.
The following removal constants for the containment iodine removal systems are assumed in the analysis: (Ref. USAR Section 14.15.4)
Elemental Iodine 5.14 hr-1 Organic Iodine 5.14 hr-1 Particulate Iodine 5.14 hr-1 1
-(N8-C) 9.
The contaiment is assmed to leak at 0.05 vol.%/d for the duration of the accident.
This is conservative since under actual conditions the contaiment pressure decreases with time.
By reducing the leak rate as a function of containment pressure, the operator exposures will be less than calculated.
The test results fra 1983 integrated leak rate testing indicate 0.047%/ day at 95% confidence level.
10.
The radioactive leakage from contaiment is assmed to be unifonn fra the entire surface area of contaiment building.
Therefo re, the probability of leakage is the sane per unit surface area fran contai nment walls or from containment dome.
Three major leakage patNays have been identified as the following:
(a) The leakage from containment walls into the adjacent building, such as the spent fuel storage area and the renaining portions of Auxiliary Buil di ng, which will be eventually exhausted to the ventilation stack.
(b) The leakage from containment dome to the outside environment.
(c) The leakage fran contaiment walls to the outside enviroment.
Refer to Figure 1 for release pathways (a), (b), and (c).
The release patNay (a) and (b) will not increase the radionuclide concentrations at the fresh air intake since no downwash can occur at 1 m/sec wind velocity and due to 60 ft of elevation difference with the control. room intake.
However, in order to be conservative, pa tNay (b) from contaiment dome is assmed to be released as a ground level release at the same elevation as the control room i ntake.
Under accident conditions the release patNay (a), which is leakage into the Auxiliary Building, is most probable since the Electrical and Mechanical penetrations are located in the Auxiliary Buil ding and since any releases to the Auxiliary Building would be eventually exhausted to the ventilation stack.
The only exceptions are the releases from the containment wall into the switchgear room, cable spreading roon, lower electrical, capressor rom and rom 81 which exhaust to VA-41 at the same elevation as the Control Room intake.
However since VA-41 is downwind fran the Control Rom intake it will not contribute to the radionuclide concentrations at the intake.
Therefo re, assni ng release fran patNays (b) and (c) is extrenely conservative.
2
(N8-C)
Additionally, the release at all the elevations from pathway (c) is assumed to occur at the same elevation as the fresh air intake (ground level release and ground level receptor).
This is also an extremely conservative assumption since the releases from elevations up to 60 feet above the. fresh air intake cannot be observed at the elevation of the fresh air intake due to the fact that no signifi-cant downwash can occur at 1 m/sec wind velocity at close distance of 12 ft to the containment where the intake is located.
Higher wind vel oci ties, e.g.,
10m/sec which may result in downwash were considered.
However, X/Q was also reduced by a factor of 10 since X/Q and wind velocity are inversely proportional.
Therefore, even with downwash, the operator exposures were lower than the results shown in this report.
- 11. The X/Q dispersion model is assuned to be a diffused source model from Campe and Murphy (Ref. 3) with standard deviations in vertical crosswind (o z) and horizontal crosswind (o y) set equal to zero.
This is extremely conservative, since there is signi ficant turbu-lence within the building wake boundary.
- 12. All the activity released during H2 purge operation is assumed to be initiated 4 days after LOCA up to 30 days after LOCA.
This is conservative since the Fort Calhoun Updated Safety Analysis Report indicates initiation of H2 purge 5 days af ter LOCA up to 22 days af ter LOC A.
The contribution from H2 purge operation is negligible with the Auxiliary Building release flow rate of 72,500 CFM due to high exit velocity and no significant downwash from patNay (a).
As.
indicated above, downwash can occur at 10 m/sec wind velocity.
How-ever since X/Q decreases by a factor of 10 at 10 m/sec, and the loca-tion of the intake is 12 ft from containment wall and 60 ft below the ventilation exhaust, the contribution from H2 purge will be neg-ligible.
- 13. The adjustment factors stated in Table 1 of Campe and Murphy (Ref.
- 3) for Control Room occupancy have been modified as shown below:
TABLE 1 l
FACTORS USED TO CALCULATE EFFECTIVE RELATIVE CONCENTRATIONS FOR SELECTED TIME INTERVALS Adjustment Factors 0 - 8 hrs 8 - 24 hrs 1-4 days 4 - 30 days Occupancy 1
0.25 0.50 0.40 Wind Speed 1
0.67 0.50 0.33 Wind Direction 1
0.88 0.75 0.50 Overall Reduction 1 0.147 0.1875 0.C66 The occupancy factors for Control Roan operators is based on work schedule shown on Table 2.
3 L
(N8-C)
TABLE 2 OPERATOR WORK SCHEDULE 0 - 8 hrs
- 8 - 24 hrs 1 - 4 days 4 - 30 days Hours in Control 8 hrs 4 hrs 12 hrs / day Room Occupancy 1.0 (From 0.25 0.5 0.4 (From Factor Table 1)
Tab'le 1)
It is assmed that an operator spent a total of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per day during the fi rst 4 days of the accident in the Control Room.
- 14. A semi-infinite cloud equation for beta dose calculation was used from Campe and Murphy.
This equation is conservative since the Control Room volume is finite.
Beta exposure can be further reduced if credit is taken for attenuation of beta through the suspended ceiling at 9' above the floor level.
II.
Input Parameters for Computer Calculations LOCA II computer code from Combustion Engineering is used to calculate the operator Gamma, Beta and Iodine radiation exposures.
The equations l
used in the code are shown and discussed in SAN ON0FRE 2 & 3 FSAR Appen-dix 15B (Reference 4).
l Input Parameters for LOCA II are as Follows:
(1) Iodine Removal Constant l
5.14 hr-1 from Fort Calhoun USAR (Section 14.15.4)
(2) X/Q Atmospheric Dispersion Factor The equation from Campe and Murphy (Ref. 3) for diffused source is used:
l l
X/Q = [U(n cy az + a )]-1 XPE 4
(N8-C)
Where; X/Q = Relative Concentration Dispersion Factor ( sec/m3) ay,oz.=
Standard deviation of the gas concentration in horizontal crosswind and vertical crosswind respectively (m)
U = Wind velocity = 1 m/sec
-a = cross sectional area of contaimnent building = 1340 m2 from Fort Calhoun USAR 3
67.37 K
3
=
Ts/d)l 4
.(3.9/36)l 4 s=
Distance between containment and centrol room intake = 3.9 meter d=
Dianeter of containment = 36 meter X/Q = [1 m/sec (n-(0)-(0) +
1340
)]-1 67.37 + 2 X/Q = 5.18 x 10-2 sec/m3 (0-8 hr)
Multiplying the X/Q for 0-8 hours by reduction factors from Table 1 the X/Q for other time intervals is obtained.
5.18 x 10-2 x 0.147 = 7.61 x 10-3 sec/m3 X/Q
=
(8-24 hr)
X/Q = 5.18 x 10-2 x 0.1875 = 9.71 x 10-3 sec/m3 (1-4 days)
X/Q = 5.18 x 10-2 x 0.066 = 3.41 x 10-3 sec/m3 (4-30 days)
(3) Containment Leak Rates The leak rates from containment are as follows:
Contairment
% Leak Rate I
Time Interval in 24 hr Period 0-8 hr 0.05%
l 8-24 hr 0.05%
1-4 days 0.05%
4-30 days 0.05%
L h
I I
5 w
(N8-C)
(a) Leakage Area Adjacent to Auxiliary Building Leakage pathway (a) as discussed in the assumptions:
The total surface area of containment enclosed in the Auxiliary Building will be calculated as follows:
as shown on Drawing 11405-A-8 (attached) the containment has been divided in four major segments; 130' Segment with Roof Elevation of 1057' 130* x n xD Arc length
= W 130* x n x 116' = 131.6 f t
= W Height of 130* segment = El.1057' - El.995' = 62 f t Area of 130' segment = 62' x 131.6' = 8159 f t2 112* Segment with Roof Elevation of 1045' 112' x x xD = 113.4 f t Arc length
= W Height of segment = El.1045' - El. 995' = 50 ft Area of 130' segment = 50' x 113.4' = 5670 f t2 36* Segment with Roof Elevation of 1083' 36' x x x D = 36.4 f t Arc length
=
366' Height of segment = El.1083' - El. 995' = 88 ft
+
Area of segment = 88' x 36.4'= 3203 f t2 6
(N8-C) 27' Segment with Roof Elevation of 1035'-8 27* x n x D = 27.3 f t Arc length
=
TW Height of the segment = E1.1035'.8 - El. 995' = 40.8 f t Area of the segment = 40.8' x 27.3' = 1114 f t2 Total surface area of containment wall enclosed in Auxil-iary Building = 130' segment area + 112* segment area + 36' segment area + 27' segment area.
= 8159 + 5670 + 3203 + 1114 = 18146 f t2 (b) Leakage Area From Containment Dome Leakage pathway (b) as discussed in the assumptions:
Contalment due ( circular ) = (116)2n = 10568 ft2 Area approximation T
(c) Leakage Area From Containment Walls to the Outside Environment Leakage patkay (c) as discussed in the asstaptions:
To calculate the surface area of the containment walls leaking to the outside enviroment; total surface area of containment wall is calculated and then subtraced by the area of contaiment enclosed in the Auxiliary Building:
Total surface area = (E1.1119'-E1. 995') n D = 45189 f t2 Surface area of containment wall leaking to outside enviroment = 45189 - 18146 = 27043 ft2 As stated in the assumptions, the leakage from ventilation stack patNay (a) will not contribute to the concentration of radionuclides at the control room intake since no downwash can occur due to the following reasons:
l l
l l
l l
7
(N8-C) i The release fra Auxiliary Building ventilation stack is assmed at 72500 CFM. Since no darnwash can occur if w > 5; (fra Reg. Guide l
ii-1.111) l Where u = wind velocity w = release velocity = (72500 CFM)/34.9 ft2 cross sectional area of stack
= 2077 ft/ min or 10.55 m/sec 10.55 m/sec > 5 thus no downwash can occur 1 m/sec In addi tion the ventilation stack exhaust is 60 feet higher in elevation than the Control Room intake and due to close proximity of the intake to the contaiment wall, even if downwash were to occur it would not affect the intake which is 12 ft away from containment.
Therefore the only section of contaiment leakage which may con-tribute to the radionuclide concentration at the Control Room intake is the release from walls of contaiment exposed to the outside environment (pathway c) and release from containment dome (pathway b).
As stated earlier the release fra the contaiment walls at different elevations is all assumed to be condensed and released at the same elevation as the Control Room i ntake.
The total leakage rate from pathway (c) and (b) are calculated as follows:
total leakage fra contaiment is assmed at 0.05% for the duration of the accident; 3
Contalment x Containment = 0.0005/ day x 1.05 x 106 ft3 = 5.25 x 102 f t / day Leak Rate Free Volume 3
2 f t / day + 24 hr/ day = 2.1875 x 101 3
ft /hr 5.25 x 10 The leakage for the containment walls and dome exposed to the envi roment assmi ng unifom release fra all surface areas is calculated as follows:
i
(N8-C)
Contalment Wall Area Exposed Leak Rate From Total to Outside & Containment Dome Area Containment Wall
= Containment x Total Containment Exposed to Outside Leak Rate Surface Area (Wall & Dome) ft /hr) x 27043 ft2 +10568 ft2 3
(2.1875 x 101
=
45189 f t2 + 10568 ft2 1.475 x 101 3
ft /hr
=
ft /hr = 1.405 x 10-5 hr-1 3
Fraction. Containment Leak Rate = 1.475 x 101 of Leak Rate Contaiment Free Volume 1.05 x 106 ft.
3 The above leak rate will be assumed for time intervals 0-8 hr, 8-24 hr, 1-4 day, and 4-30 days.
(4) Radionuclide Activities in Containment at Time Zero After LOCA The source term activies are based on release of 100% of core noble gases and 25% of core iodines to the contaiment atmosphere:
NUCLIDE ACTIVITY (Ci)
KR-85 0.33 x 106 KR-85m 0.11 x 108 KR-87 0.19 x 108 KR-88 0.28 x 108 XE-131m 0.29 x 106 XE-133 0.85 x 108 XE-135 0.15 x 108 XE-135m 0.17 x 108 XE-138 0.68 x 108 I-131 0.10 x 108 I-132 0.15 x 108 I-133 0.21 x 108 I-134 0.23 x 108 I-135 0.20 x 108 (5) Breathing Rates The breathing rates are based on values specified by Reg. Guide 1.4.
(6) The Control Room Free Volume 100,000 f t3 This value represents the actual dimensions of the Control Roan space.
USAR value of 45,100 represents free volume of the Control Roan below the suspended ceilings.
9 to k
(N8-C)
(7) Control Room Infiltration Flow 3
1 1100 CFM x 60 min /hr = 66,000 ft /hr (8), Build Up Rate for Radionuclide Concentration Build Up Rate = 66000 ft3 /hr = 0.66 hr-1 100,000 f t3 (9) The Semi-Infinite Cloud Approximation for Gamma Dose and Beta Dose D
seni. = 0.25-(X/Q) I Qi Ei fra Campe and Murphy y
1 A geometry factor for a finite volume from Campe and Murphy is appli ed 1173
= 23.95 GF = 1173
=
T 38 0.338 3
(V)
(100,000 f t )
D
= 0.25-(X/Q) I Qi Ei y
T535 1
Dg Semi. = 0.23-(X/Q) g Qi Ei 1
III. OPERATOR EXPOSURES (1) The integrated iodine thyroid dose to the Control Rom operators for the duration of LOCA from airborne activity in Control room is DI = 12.81 Rem The gamma whole body and beta skin integrated radiation dose to the Control Room operators for duration of LOCA from airborne activity in Control Rom are:
2.1 Rem < 5 Rem (SRP 6.4 limit)
D
=
y Dg = 25.0 Rem < 30 Rem (SRP 6.4 limit)
NOTE:
The actual Beta dose calculated in the air is 60 rem, however, the ratio of Seta skin factors fra Reg. Guide lieta air 1.109 were applied.
Taking credit for Beta attenua-tion through the outer dead layer of skin is stated in Reg. Guide 1.109.
Refer to attached output sheet from LOCA II B Program. for conversion of beta air dose to beta skin dose.
10
- P
(N8-C)
(2) The operator gamma whole body exposure from radioactive cloud over the Control Room roof for duration of the LOCA is:
Whole Body < 0.1 REM (3) The operator gamma whole body exposures from iodine filter in Room 81 and other sources i.e. containnent spray piping and containnent building for duration of LOCA is:
Whole Body < 0.5 REM Total Operator Exposures From all Radioactive Sources:
2.7 REM D
=
y D
= 25.6 REM g
DI = 12.81 REM All the exposures are below SRP.6.4 limits.
IV.
CONCLUSIONS The operator integrated dose for whole body gamma, beta skin, and iodine thyroid calculated for duration of LOCA are less than the limits of Standard Review Plan Section 6.4.
The dose calculations are extranely conservative as stated in various sections of this report.
Major conservatisms which will further reduce the above doses are:
(1) The containment leak rate tests at Fort Calhoun indicate leak rate of 0.047%/ day.
Reduction in leak rate as a function of post LOCA containment pressure can further reduce the operator exposures.
(2) The actual y and 0 z values are larger than zero asstaned in this report which can decrease the X/Q value which will in turn reduce operator exposures.
(3) The releases fran pathway (c) or the exposed walls of containnent are assumed to be ground level releases at the same elevation of the Control Roan -i ntake.
Taking credi t for release at di fferent elevations would further reduce the calculated dose by a factor of two or more.
Therefore, the calculated doses are believed to be a factor of two or more higher than actual if a more realistic set of assumptions were used.
11
- \\
FIGURE 1 1m/SEC WIND TURBINE $M#
BUILDING
_rA.*H lNoWN81 W
[ CONTRdi_ H00M FILTEPED INTA<E D
ROOF
\\
L. 1057' l g
CC'lTAINMENT EL. 1119'
/.UXILIARY CijILOING VEtJTILATION STACK SPENT FUEL STORAGE AREA ROOF EL. 1045' ROOF EL. 1083' PLAN VIEW im/SEC WIrlD AUXILIARY BUILDING VENTILATION 3
STACK tT V
1m/3EC WIND
\\1e
/
^
/
o ASSUMEO GROUND LEVEL PELEASE AUXILIARY 31 21~ "
LINE BUILDING I
SPENT TURBINE FUEL AUXILIARY BUILDING BUILUING l
STORAGE
@~
/
@ THE LEAKAGE TO ADJACENT AUXILIARY OUILDING AND EVENTUAL RELEACE FROM THE VENTILATION STACK.
j
@ THE LEAKAGE FRCM CCNTAIN!^2NT DOME HAS ALSO BEEN ASSUMED TO BE RELEASED AT ASSUMED GROUNO LEVEL RELEASE LINE SHOWN ACOVE.
)
@ THE LEAKAGE FRCM EXPOEED PORTIONS OF CCNTAINMENT WALL.
THE PATHWAY ' c" RELEASES AT DIFFERENT ELEVATIONS HAVE ALL DEEN CONOENSEO AND RELEASED AT THE AGSUMED GROUND LEVEL LINE SHOWN ABOVE.
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O CLIOE GAMMA RAVS BETQ RAVS (MEV/015)
/,pe.A g.g. Ou/j" N 7.
MR-85 0.0021 0.228 kR-85M 0.1560 0.354 MR-07 0.8560 1.014 MR-88 2.0000 0.307 AE-131M 0.0220 0.135 AE-133 0.0455 0.126 XE-135 0.2480 0.310 xE-135M 0.4400 0.090 XE-138 0.9320 0.565 1-131 0.3750 0.209 I-132 2.2900 0.421 1-133 0.6360 0.403 1-134 2.5000 0.558 I-135 1.4570 0.475 51TE BOUNDARY WHOLE 800Y 005ES DUE 10 I
$ fosp, ppC6 fo/09-2/
Of gg gy m g uatt NUCLIOE GAMMA RAY 5 BETA RAYS M
(/7'? tE4 -
ffpg/,pe) / b ## ~
TIMES 1 TIME *4 TIME =1 TIME =4
(.Bria ) x A k,p=. ScrA Do.54 H.bcap L&M df SAW
^
KR-85 3.47E-05 3.37E-04 8.30E-02 8.06E-01 X V /- 3) m e o.5'y
/,44 MR-85M 4.16E-02 4.49E-02 2.08E+00 2.25E+00
~7 U ( - 3)
/. W 6 (- 5)
~-
. 3,g/
KR-87 1.23E-01 1.23E-01 3.21E+00 3.22E+00 X f 7 4 (- 4).- "
'/ 97 C-31 i
d KR-88 1.00E+00 1.04E+00 3.39E+00 3.51E+00
/*
- 3 C-2 7
- z., 7 p (, < 3
=
y g
XE-133 1.86E-01 7.37E-01 1.13E+01 4.49E+01
- 7. 4 3 (-J) xE-131M 3.17E-04 1.83E-03 4.28E-02 2.48E-01 m
- 'US o g (- 4 3
~
,1' g, g
/f.48
- f. o g e. 3 3 '"
XE-135 1.33E-01 1.65E-01 3.65E+00 4.55E+00y /. W-3 )
xE-135M 4.32E-03 4.32E-03 1.95E-02 1.95E-07 z. V6 (- 3)'"
xE-138 3.01E-02 3.01E-02 4.02E-01 4.02E,
I-131 5.72E-05 5.72E-05 7.02E-04 7.03E-04 I-132 3.37E-04 3.37E-04 1.36E-03 1.36E-03 N 1-133 1.88E-04 1.88E-04 2.62E-03 2.62E-03 1-134 3.33E-04 3.33E-04 1.63E-03 1.63E-03 1-135 3.58E-04 J.5AE-04_
2.57E-03 2.57E-03..
NUCLIOE TOTAL OF GAMMA A b BET er#/
2./.S ':::". 26 NN[M IN M/A gg TIME = 1 TIME a 4 KR-85 8.31E-02 8.07E-01 HR-85M 2.12E+00 2.29E+00 kR-87 3.33E+00 3.34E+00 KR-88 4.39E+00 4.54E+00 XE-131M 4.31E-02 2.50E-01 XE-133 1.15E+01 4.57E+01 XE-135 3.79E+00 4.72E+00 "4
XE-138 4.32E-01 4.32L-01 gg
__ g./
Q p [4 fhp)
XE-135M 2.38E-02 2.38E-02 1-131 7.59E-04 7.60E-04
~
P g g g g, J y,4 g g 1-132 1.70E-03 1.70E-03 I-133 2.81E-03 2.81E-03
- ?
1-134 1.97E-03 1.97E-03 s
1-135 2.93E-03 2.93E-03 TOTAL 2.57E+01 6.21E+01 RA05 ML.S Alat
SUMMARY
- MESSAGE NUMBER - COUNT 208 63
(N8-C)
ATTACISENT C Calculations for Relocation of the Fresh Air Intake to 10 meters I.
Assumptions l
The following assumptions were used in calculating the operator exposures by relocating the filtered air (VA-65) intake to 10 meters away from con-tainment wall:
1.
The reactor core equilibrium noble gas and iodine inventories are l
based on long-te nn operation at 100% of the ultimate core power level of 1500 Nt.
2.
One hundred percent of the core equilibrium radioactive noble gas inventory is immediately available for leakage from the containment.
3.
Twenty-five percent of the core equilibrium radioactive iodine inven-tory is immediately available for leakage from the containment.
4.
The release-of fission products to the containment is assumed to occur instantaneously and transported instantaneously to the environ-ment and Control Rom assuming no decay during transport time.
5.
Of the iodine fission product inventory released to the containment, 91% is in the fonn of elemental iodine, 5% is in the fonn of partic-ulate iodine, and 4% is in the fonn of organic iodine.
6.
Radioactive Decay - Credit for radioactive decay for fission product concentrations located within the containment is assumed throughout the course of the accident to the Control Rom.
7.
Containment Iodine Removal System - Credit. for the removal of iodine frm the containment buil di ng atmosphere is assumed during the course of the accident resulting fr a filtration in the iodine removal sys tem.
The system consists of four air handling units; two havi ng filtering capaci ty and the other two havi ng no filtering capaci ty.
In the calculation of the radiological consequences an adsorption efficiency of 0.9 (90%) and operation of only one of the two containment filtering units is assumed.
Credi t for elemental,
particulate, and organic iodine removal is taken for the course of the accident.
8.
The following removal constants for the containment iodine removal systems are assumed in the analysis: (Ref. USAR Section 14.15.4)
Elemental Iodine 5.14 hr-1 Organic Iodine 5.14 hr-1 Particulate Iodine 5.14 hr-1 l
1
(N8-C) 9.
The containnent is assuned to leak at 0.1 vol.%/d for the duration of the accident per Fort Calhoun Technical Specification. _This is conservative since under actual conditions the containnent pressure decreases with time.
By reducing the leak rate as a function of containnent pressure, the operator exposures will be less than cal culated.
The test results from 1983 integrated leak rate testing indicate 0.047%/ day at 95% confidence level.
- 10. All the leakage from containment walls is assumed to be a ground level release and ground level receptor at the sane elevation as the control room filtered intake.
- 11. The X/Q dispersion model is assuned to be a diffused' source model from Campe and Murphy (Ref. 3) with standard deviations in vertical cros swi nd ( z) and horizontal cross wind (oy) set ecual to zero.
This is extremely conservative, since there is signi"icant turbu-lence within the buil ding wake boundary.
The wi nd vel oci ty was assumed at 1 m/sec for calculations X/Q.
- 12. The contribution from H2 purge operation is negligible from ventilation stack, with the Auxiliary Building release flow rate of 72,500 CFM due to high exit velocity and no downwash.
Downwash can occur at 10 m/sec wind velocity, however since X/Q decreases by a.
factor of 10 at 10 m/sec, and the location of the intake is 12 ft from containment wall and 60 f t below the ventilation stack exhaust, the contribution from H2 purge will be negligible.
- 13. The control room filtered intake flow rate was-assumed at 1200 CFM.
- 14. The adjustment factors stated in Table 1 of Campe and Murphy (Ref.
- 3) for Control Room occupancy have been used to adjust X/Q values.
- 15. A semi-infinite cloud equation for beta dose calculation was used from Campe and Murphy.
This equation is conservative since the Control Room volume is finite.
Beta exposure can be further reduced if credit is 'taken for attenuation of beta through the suspended ceiling at 9' above the floor level.
II.
Calculations LOCA~ II canputer code from Combustion Engineering is used to calculate the operator, ' Gamma, Beta and Iodine radiation exposures.
The equations used in the code are shown and discussed in the SAN ON0FRE 2 & 3 FSAR Appendix 15B (Ref. 4).
Input Parameters for LOCA II are as Follows:
(1) Iodine Removal Constant 5.14 hr-1 from Fort Calhoun USAR (Section 14.15.4) 2
(N8-C)
(2) X/Q Atmospheric Dispersion Factor The equation from Campe and Murphy (Ref. 3) for diffused source is used:
X/Q = [U(n 3y 0z + a )]-1 XPE Where; X/Q = Relative Concentration Dispersion Factor (sec/m3)
Standard deviation of the gas concentration in o y, az
=
horizontal crosswind and vertical crosswind respectively (m)
U = Wird velocity = 1 m/sec a = cross sectional area of contatrinent building = 1340 m2 fra Fort Calhoun USAR 3
18.03 K
3
=
Ti7d)l 4 (10/36)l 4 S=
Distance between containment and control room intake = 10 meter d=
Diameter of contalment = 36 meter X/Q = [1 m/sec (n -(0)-(0) +
1340
)]-1 18.03 + 2 X/Q = 1.49 x 10-2 sec/m3 (0-8 hr)
Multiplyi ng the X/Q for 0-8 hours by reduction factors from Table 1 of Campe and Murphy, will provide the X/Q's for other time intervals; 1.49 x 10-2 x 0.59 = 8.8 x 10-3 sec/m3
'X/Q
=
(8-24 hr)
X/Q = 1.49 x 10-2 x 0.23 = 3.4 x 10-3 sec/m3 (1-4 days)
X/Q = 1.49 x 10-2 x 0.066 = 9.9 x 10-3 sec/m3 (4-30 days) 3
(N8-C)
(3) Containment Leak Rates The leak rates from containment are as follows:
Containment
% Leak Rate Time Interval in 24 hr Period 0-8 hr 0.1%
8-24 hr 0.1%
1 0.1%
1-4 days 4-30 days 0.1%
NOTE:
0.1% is convervativ.e since actual integrated leak rate test at Fort Calhoun indicated 0.047% at 95% confidence level during 1983 testing.
4.17 x 10-5.hr-1
.001/ day Fraction of
=
=
Leak Rate for 24 hrs / day Containnent (4) Radionuclide Activities in Containment at Time Zero After LOCA The source tenn activies are based on release of 100% of core noble gases and 25% of core iodines to the containnent atmos-phere:
NUCLIDE ACTIVITY (Ci)
KR-85 0.33 x 106 KR-85m 0.11 x 108 KR-87 0.19 x 108 KR-88 0.28 x 108 XE-131m 0.29 x 106 XE-133 0.85 x 108 XE-135 0.15 x 108 XE-135m 0.17 x 108 XE-138 0.68 x 108 I-131 0.10 x 108 I-132 0.15 x 108 I-133 0.21 x 108 I-134 0.23 x 108 I-135 0.20 x 108 (5) 8reathing Rates The breathing rates are based on values spect fied by Reg. Guide 1.4.
1 4
I
(N8-C)
(6) The Control Room Free Volume 100,000 ft3 (7) Control Room Infiltration Flow 3
.1200 CFM x 60 min /hr = 72,000 ft /hr (8) Build Up Rate for Radionuclide Concentration Build Up Rate = 72,000 ft 3 /hr = 0.72 hr-1 100,000 f t3 (9) The Semi-Infinite Cloud Approximation for Gansna Dose and Beta Dose D
semi, = 0.25-(X/Q) j Qi Ei fran 'Campe and Murphy 7
A geometry factor for. a finite volume from Campe and Murphy is appli ed 1173
= 23.95 GF = 1173
=
- T338 0.338 3
-(V)
(100,000 ft )
-D
= 0.25-(X/Q)-
Qi Ei 7
2T.T5 D Semi, = 0.23-(X/Q)-
Qi Ei g
OPERATOR EXPOSURES
.(1) The integrated iodine thyroid dose to the Control Roan operators for the duration of LOCA from airborne activity in Control room is DI = 10.97 Rem < 30 ran (SRP 6.4 limit)
The ganma whole body and beta skin integrated radiation dose to the Control Roan operators for duration of LOCA~ fran airborne activity in Control Room are:
2.23 Rem < 5 Rem (SRP.6.4 limit)
D
=
Y D
= 26.99 Rem < 30 Rem (SRP.6.4 limit)
B
~
NOTE:
The actual Beta dose calculated in the air is 67.3 ren, how-ever, the ratio of Beta skin factors fran Reg. Guide 1.109 Beta air were applied.
Taking credit for Beta attenuation through the outer dead layer of skin is stated in Reg. Guide 1.109.
i i
5
(N8-C)
(2) The operator gamma whole body exposure from radioactive cloud over the Control Room roof for duration of the LOCA is:
Whole Body < 0.1 REM (3) The operator gamma whole body exposures from iodine filter in Room 81 and other sources i.e. containnent spray piping and contairment building for duration of LOCA is:
Whole Body < 0.5 REM Total Operator Exposures From all Radioactive Sources:
Dv = 2.83 REM DB = 26.99 REM DI = 10.97 REM All the exposures are below SRP.6.4 limits.
III. CONCLUSIONS The operator integrated dose for whole body gamma, beta skin, and iodine thyroid cal culated for duration of LOCA are. less than the limits of Standard Review Plan Section 6.4.
The dose calculations are extracely conservative as stated in. vari ou s sections of this report.
Major conservatisms which will further reduce the above doses-are:
(1) The containment leak rate tests at Fort Calhoun indicate leak rate of 0.047%/ day, however by using the conservative value of 0.1%/ day the operator doses calculated above are increased by more than a factor of 2.
Reduction in ' leak rate as a function of post LOCA containment pressure can further reduce the operator exposures.
(2) The actual oy and az values are larger than zero assumed in this report which can decrease the X/Q value which will in turn reduce operator exposures.
(3)
It. is extrenely conservative and unreali stic to assume that all activi ty from containment walls will be released at the same elevation as the fresh air intake through a solid ' concrete wall (ground level release and receptor).
Therefo re, the calculated doses are believed to be a factor of four or more higher than actual if a more realistic set of assumptions were used as it has been demonstrated in Attachment B for justification for contirF ued operation.
There is a strong need to establish standards which will allow realistic release pathways for radionuclides ' from containment walls during a LOCA in order to obtain a more realistic operator exposures.
6 l
(N8-C)
IV. -
CORRECTIVE ACTIONS In order to '. increase the margin of safety for operator exposures during LOCA, control room filtered intake (VA-65) will be relocated to at least 10 meters away fran containment wall.
In addition, an iodine monitor was.
installed in Control Room for continuous monitoring of post LOCA iodine in the Control Roan.
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