ML20092J022
| ML20092J022 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 06/21/1984 |
| From: | Tucker H DUKE POWER CO. |
| To: | Adensam E, Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8406260331 | |
| Download: ML20092J022 (20) | |
Text
.
r DUKE POWER GOMPANY P.O. EKMC OG180 CHARLOTTE, N.C. 28242
. HAL II. TUCKER menon
.J"Z."". "."" -
June 21, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Ms. E. G. Adensam, Chief Licensing Branch No. 4
.Re: Catawba Nuclear Station Docket Nos. 50-413 and 50-414 i
Dear Mr. Denton:
As a result of preoperational testing of the Annulus Ventilation System at '
l Catawba, it was determined that reactor building in-leakage was higher than previously assumed. This higher in-leakage was attributed primarily to the use of foam fire stops in electrical penetrations as opposed to the multi-cable transits previously used at McGuire.
Appropriate FSAR sections have been revised to reflect the observed higher in-leakage. As shown on Table 15.0.12-1, these changes resulted in higher off-site doses for certain accidents. All doses remain well below 10 CFR 100 limits.
~
These revised pages will be included in Revision 11 to the FSAR.
l t
Very truly yours, ed
.C v v
Hal B. Tucker ROS/php Attachment cc: Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resident Inspector Catawba Nuclear Station Mr. Robert Guild, Esq.
[
Attorney-at-Law P. O. Box 12097
(
Charleston, South Carolina 29412
- /
8406260331 840621 PDR ADOCK 05000413 l
A PDR
r
-c.
Mr. Harold R._Denton, Director June 21, 1984 Page 2 cc: Palmetto Alliance.
21351 Devine Street i
Columbia, South Carolina 29205 Mr. Jesse L. Riley i
Carolina Environmental Study Group i
854 Henley Place Charlotte, North Carolina 28207 j
i L
I I
l I
f P
t i
CNS vided in Table 6.2.3-3.
The structural outline of the Reactor Building is iur-nished in Figure 3.8.1-1.
Additional plans and sections are shown in Figures 1.2.2-8 through 1.2.2-16.
6.2.3.3 Design Evaluation The results of an analysis of the functional capability of the Annulus Venti-lation System to depressurize and maintain a uniform negative pressure of >0.5 in. H O in the annulus following a design basis accident are provided in Table 2
6.2.3-2.
The pressure, temperature, and mass of annulus air is calculated by the Fortran IV program CANVENT for the entire accident transient, including the steady state conditions prior to the initiating event.
The containment is divided into three regions, where standard equations of heat transfer are applied.
No heat or mass transfer between regions is assumed except in the annulus.
The steady state, pre-accident temperatures are determined by an interactive process until successively calculated temperatures differ by less than some small predetermined amount.
The post-DBA transient conditions are calculated using the finite differences tech-nique.
The following assumptions are made for simplification and/or conservatism:
1)
The containment is divided into three regions. The temperature of each
[
region is uniform within that region.
2)
There are no temperature gradients in the vertical or circumferential dir-ections.
Thus, the model is one dimensional with heat transfer occurring only in the radial directions.
3)
All physical properties (e.g., heat capacity, thermal conductivity, emis-sivity, and density) are independent of temperature, except the density of air in the annulus.
4)
The air in the annulus behaves as an ideal gas and is uniformly mixed.
5)
The air in the annulus has a transmissivity of unity. Therefore, energy is transfered to and from the air only by natural convection.
6)
Radiative heat transfer occurs between the concrete reactor building and the steel containment building.
The surfaces are treated as gray bodies with a parallel, flat plate geometry, 7)
The equation for the heat transfer coefficient for the upper containment f to annulus air, treating the dome as a horizontal plate, is h =.22 (At)1 3 Similarly, treating the ice condenser and lower containment sections as verticalp3lates, the heat transfer coefficient to annulus air is h =
.19 (At)1 8)
For the transfer of heat from the containment air to the containment shell, a heat transfer coefficient that increases linearly in time from 8 8tu/hr -
6.2-47 Rev. 11
~
ft2 _ oF to some maximum value is assumed, followed by exponential decay at a rate of.025 sec 1 to some long-term value.
The steady-state calculations are based on the natural convection heat transfer coefficients previously mentioned.
9)
Circulation of refrigerating air in the ice condenser air ducts ceases at the initiation of the accident. Therefore, before the accident, heat is transferred to the refrigerating air by forced convection; whereas, after the accident the mechanism is natural convection.
The use of a forced convection heat transfer coefficient is eliminated by assuming the ice condenser walls are at the same temperature as the refrigerating air.
- 10) The annulus ventilation fan comes on instantaneously at full speed at a time determined by signal response times and fan characteristics.
Partial i
flow before this time is not considered.
At a later time, the fan flow capacity is decreased 15% due to an increased differential pressure across the annulus filter train.
- 11) A portion of the annulus ventilation fan flow is exhausted to the atmos-phere and the remainder is returned to the annulus.
The full fan flow is exhausted until the annulus pressure is reduced to -1.0 inches w.g.
From that point, the amount of air exhausted is that amount required to maintain the annulus pressure at -1.0 inches w.g.
- 12) Leakage of air across the concrete reactor building will be a maximum at l
-1.0 inches w.g.
This leakage is conservatively assumed to exist whenever the annulus is at a negative pressure.
No credit is taken for out leakage when the annulus pressure is greater than zero.
- 13) Thermal contact resistances are neglected.
- 14) For each of the three regions, heat transfer areas are lumped into one of three categories based on the inside radius of the containment shell, the midpoint of the annulus, and the midpoint of the reactor building.
This is assumed in order to avoid the continuous variation of area with radius associated with cylindrical geometry.
- 15) Outside temperatures remain unchanged during the course of the accident.
For steady-state calculations, the surface of the reactor building is at the outside temperature.
For the post-accident transient, the Reactor Building is considered an adiabatic wall.
- 16) The expansion of the containment shell, due to the pressure and temperature increase within, is calculated assuming each region is freestanding and independent of any other region.
6.2.3.4 Tests and Inspections Preoperational and periodic tests are described in Chapter 14 and the Technical Specifications respectively.
6.2-48 Rev. 11
-w
-.9
.7 7
- -., - + -
,,m,w-.,,
s- -.
Table 6.2.?-2 (Page 1)
Annulus Conditions vs. Time Following Design Basis Accident l
TIME ANNULUS ANNULUS PURGE FLOW RECIRCULATION LOWER CONTAINMENT 2 (SEC)
TEMP PRESSURE RATE FLOW TEMPERATURE
( R)
(IN. WATER)
(CFM)
(CFM)
(*R) j 1.
'498.94 0.550 0.
O.
690.33 2.
498.95 0.570 0.
O.
690.33 3.
498.96 0.598 0.
O.
690.33 4.
498.97 0.632 0.
O.
690.33 l
S.
498.98 0.671 0.
O.
690.33 6.
498.99 0.716 0.
O.
690.33 7.
499.00 0.764 0.
O.
690.33 8.
499.01 0.816 0.
O.
690.33 9.
499.02 0.871 0.
O.
690.33 10.
499.03 0.929 0.
O.
690.33 11.
499.05 0.988 0.
O.
690.33 12.
499.06 1.049 0.
O.
690.33 13.
499.08 1.111 0.
O.
690.33 14.
499.09 1.174-0.
O.
690.33 15.
499.11 1.239 0.
O.
690.33 16.
499.13 1.304 0.
O.
690.33 17.
499.16 1.369 0.
O.
690.33 18.
499.18 1.436 0.
O.
690.33 19.
499.21 1.503 0.
O.
690.33 20.
499.23 1.571 0.
O.
690.33 21.
499.27 1.639 0.
O.
690.33 22.
499.30 1.708 0.
O.
690.33 23.
499.33 1.650 9000.
O.
690.33 24.
499.37 1.592 9000.
O.
690.33 25.
499.41 1.534 9000.
O.
690.33 4
26.
499.45 1.475 9000.
O.
690.33 27, 499.50 1.416 9000.
O.
690.33 28.
499.54.
1.357 9000.
O.
690.33 i'
30.
499.65 1.239 9000.
O.
690.33 29.
499.59 1.298 9000.
O.
690.33 h
31.
499.70 1.181 9000.
O.
690.33 Rev. 11
Table 6.2.3-2 (Page 2)
Annulus Conditions vs. Time Following Design Basis Accident 1
} TIME ANNULUS ANNULUS PURGE FLOW RECIRCULATION LOWER CONTAINMENT 1 1
(SEC)
TEMP PRESSURE RATE FLOW TEMPERATURE
(*R)
(IN. WATER)
(CFM)
(CFM)
(*R) 32.
499.75 1.124 9000.
O.
690.33 33.
499.81 1.067 9000.
O.
690.33 34.
499.87 1.010 9000.
O.
690.33 35.
499.93 0.954 9000.
O.
690.33 36.
500.00 1.898 9000.
O.
690.33 j
37.
500.06 0.843 9000.
O.
690.33 38.
500.13 0.789 9000.
O.
690.33 39.
500.19 0.734 9000.
O.
690.33 J
40.
500.26 0.681 9000.
O.
690.33 41.
500.33 0.627 9000.
O.
690.33 42.
500.41 0.574 9000.
O.
690.33 43.
500.48 0.522
- 9000, r.
690.33 44.
500.55 0.470 9000.
O.
690.33 1
45.
500.63 0.418 9000.
O.
690.33 46.
500.71 0.371 9000.
O.
690.33 47.
500.79 0.325 9000.
O.
690.33 48.
500.86 0.279 9000.
O.
691.16 49.
500.94 0.234 9000.
O.
691.97 50.
501.03 0.189 9000.
O.
692.76 l
51.
501.11 0.145 9000.
O.
693.54 52.
501.19 0.101 9000.
O.
694.30 53.
501.27 0.058 9000.
O.
695.05 54.
501.36 0.015 9000.
O.
695.79 55.
501.44
-0.014 9000.
O.
695.48 56.
501.52
-0.037 9000.
O.
695.19 j
57.
501.60
-0.060 9000.
O.
694.89 58.
501.68
-0.084 9000.
O.
694.61 a
59.
501.75
-0.107 9000.
O.
694.33 i
60.
501.83
-0.130 9000.
O.
694.05 61.
501.91
-0.153 9000.
O.
694.30 i
e 62.
501.99
-0.176 9000.
O.
694.54 Rev. 11
(
m-,
._.m.
i i
I Table 6.2.3-2 (Page 3) i Annulus Conditions vs. Time Following Design Basis Accident l TIME
' ANNULUS ANNULUS PURGE FLOW.
RECIRCULATION LOWER CONTAINMENT 1 (SEC)
TEMP PRESSURE.
RATE FLOW TEMPERATURE
(*R)
(IN. WATER)
(CFM)
(CFM)
(*R) 63.
502.08
-0.198-9000.
O.
694.78 I
64.
502.16'
-0.220 9000.
O.
695.02 i
65.
502.24
-0.243 9000.
O.
695.25 66.
502.32
-0.264 9000.
O.
695.48 67.
502.41
-0.291 9000.
O.
695.70 68.
502.49
-0.317 9000.
O.
695.92 69.
502.58
-0.343 9000.
O.
696.14 l
70.
502.66-
-0,369 9000.
O.
696.36 1
71.
502.75
-0.394 9000.
O.
696.57 72.
502.83
-0.420 9000.
O.
696.78 i
73.
502.92
-0.445 9000.
O.
696.73 l
74.
503.01
-0.470 9000.
O.
696.41
}
75.
503.09
-0.495 9000.
O.
696.10 76.
503.18
-0.520 9000.
O.
695.79 l
77.
503.27
-0.545 9000.
O.
695.48 78.
503.36
-0.570 ~
9000.
O.
695.18 j
79.
503.45
-0.595 9000.
O.
694.89 l
80.
503.53
-0.620 9000.
O.
694.59 i
81.
503.62
-0.645 9000.
O.
694.30 82.
503.71
-0.669-9000.
O.
694.02 i.
83.
503.80
-0.694 9000.
O.
693.74
)[
84.
503.89
-0.718 9000.
O.
693.46 j
85.
503.98
-0.743 9000.
O.
693.18 4
86.
504.07
-0.767 9000.
O.
692.91 i
87.
504.16
-0.792 9000.
O.
692.64 1
88.
504.25
-0.816 9000.
O.
692.37 89.
504.34
.-0.841 9000.
O.
692.11 i'
90.
504.44
-0.865 9000.
O.
691.85 l
91.
504.53
-0.889 9000.
O.
691.59 92.
504.62
-0.914 9000.
O.
691.34 j
1 93.
504.71
-0.938 9000.
O.
691.09 Rev. 11
}
Table 6.2.3-2 (Page 4)
Annulus Conditions vs. Time Following Design Basis Accident l
TIME ANNULUS ANNULUS PURGE FLOW RECIRCULATION LOWER CONTAINMENT 1 (SEC)
TEMP PRESSURE RATE FLOW TEMPERATURE i
(*R)
(IN. WATER)
(CFM)
(CFM-)
(*R) i 19 4.
504.80
-0.962 2000 0
690.84 95.
504.89
-0.987 9000 0
690.59 96.
504.99
-1.000 7436 1564 690.35 97.
505.08
-1.000 7285 1715 690.11 98.
505.17
-1.000' 7283 1717 689.87 99.
505.26
-1.000 7282 1718 689.63 100.
505.35
-1.000 7280 1720 689.40 150.
509.93
-1.000 7026 1974 679.96 200.
514.28
-1.000 6724 2276 680.35 250.
518.34
-1.000 6367 2633 682.52 300.
522.11
-1.000 6011 2989 684.00 350.
525.59
-1.000 5658 3342 682.79 j
400.
528.76
-1.000 5311 3689 682.09 450.
531.64
-1.000 4983 4017 681.96 i
500.
534.24
-1.000 4680 4320 681.85 550.
536.57
-1.000 4402 4598 681.75 600.
538.67
-1.000 4158 4842 681.65 650.
540.54
-1.000 3871 5129 674.10 700.
542.18
-1.000 3600 5400 667.11 4
750.
543.57
-1.000 3334 5666 662.85 800.
544.73
-1.000 3106 5894 661.04 850.
545.68
-1.000 2937 6063 659.35 900.
546.46
-1.000 2772 6228 657.75 950.
547.09
-1.000 2631 6369 656.23 1000.
547.60
-1.000 2521 6479 656.49 1100.
548.35
-1.000 2380 6620 657.26 1200.
548.88'
-1.000 2302 6698 657.95 1300.
549.29
-1.000 2259 6741 658.19 1400.
549.62
-1.000 2220 6780 657.82 1500.
549.90
-1.000 2193 6807 657.48 1600.
550.13
-1.000 2172 6828 657.16 t
Rev. 11 s
Table 6.2.3-2 (Page 5)
Annulus Conditions vs. Time Following Design Basis Accident l
TIME ANNULUS ANNULUS PURGE FLOW RECIRCULATION LOWER CONTAINMENT 1 (SEC)
TEMP PRESSURE RATE FLOW TEMPERATURE
( R)
(IN. WATER)
(CFM)
(CFM)
(*R) 1700.
550.32
-1.000 2156 6844 656.86 1800.
550.47
-1.000 2113 6887 652.85 1900.
550.52
-1.000 2049 6951 649.06 2000.
550.46
-1.000 1984 7016 645.46 2100.
550.30
-1.000 1951 7049 644.91 2200.
550.09
-1.000' 1945 7055 644.39 2300.
549.88
-1.000 1958 7042 643.89 2400.
549.69
-1.000 1972 7028 643.41 2500.
549.54
-1.000 1995 7005 642.95 2600.
549.43
-1.000 2024 6976 642.52 2700.
549.37
-1.000 2052 6948 642.09 i
2800.
549.36
-1.000 2078 6922 641.69 2900.
549.40
-1.000 2105 6895 642.40 3000.
549.50
-1.000 2096 6904 643.04 4
3100.
549.52
-1.000 1994 7006 641.98 3200.
549.38
-1.000 1948 7052 640.96 3300.
549.17
-1.000 1926 7074 639.97 f
3400.
548.92
-1.000 1920 7080 639.01 i
3500.
548.68
-1.000 1925 7075 638.65
{
3600.
548.45
-1.000 1937 7063 638.31 3700.
548.24
-1.000 1952 7048 637.97 3800.
548.06
-1.000 1965 7035 637.65 3900.
547.91
-1.000 1978 7022 637.33 4000.
547.78
-1.000 2075 6925 637.02 1
4100.
548.40
-1.000 3001 5999 640.23 J
4200.
550.26
-1.000 3368 5632 643.35 4300.
552.57
-1.000 3440 5560 646.40 I
4400.
554.90
-1.000 3300 5700 649.39 4500.
557.06
-1.000 3180 5820 652.30 4600.
558.97
-1.000 3052 5948 655.15 4700.
560.65
-1.000 2935 6065 657.94 Rev. 11 l
Table 6.2.3-2 (Page 6)
Annulus Conditions vs. Time Following Design Basis Accident TIME ANNULUS ANNULUS PURGE FLOW RECIRCULATION LOWER CONTAINMENT 1 (SEC)
TEMP PRESSURE RATE FLOW TEMPERATURE
( R)
(IN. WATER)
(CFM)
(CFM)
(*R) 4800.
562.11
-1.000 2841 6159 660.67.
4900.
563.41
-1.000 2778 6222 663.35 5000.
564.61
-1.000 2734 6266 665.31 5500.
569.05
-1.000 2479 6521 665.67 6000.
571.17
-1.000 2266 6734 666.01 6500.
572.04
-1.000 2216 6784 666.17 7000.
572.57
-1.000 2190 6810 665.20 7500.
572.91
-1.000 2180 6820 664.30 8000.
573.21
-1.000 2178 6822 663.46 8500.
573.49
-1.000 2179 6821 662.67 9000.
573.77
-1.000 2180 6820 661.93 9500.
574.06
-1.000 2181 6819 661.22 10000.
574.34
-1.000 2182 6818 660.56 11000.
574.89
-1.000 2183 6817 659.32 12000.
575.43
-1.000 2184 6816 658.18 13000.
575.95
-1.000 2185 6815 657.14 14000.
576.44
-1.000 2185 6815 656.17 15000.
576.92
-1.000 2186 6814 655.28 16000.
577.37
-1.000 2186 6814 654.44 17000.
577.80
-1.000 2187 6813 653.65 18000.
578.20
-1.000 2187 6813 652.90 19000.
578.58
-1.000 2187 6813 652.20 20000.
578.94
-1.000 2187 6813 651.53 30000.
581.45
-1.000 2188 6812 646.25 40000.
582.58
-1.000 2187 6813 642.51 50000.
582.93
-1.000 2186 6814 639.60 60000.
582.88
-1.000 2184 6816 637.23 70000.
582.61
-1.000 2183 6817 635.22 80000.
582.25
-1.000 2181 6819 633.48 90000.
581.84
-1.000 2179 6821 631.95 h
- 100000, 581.42
-1.000 2178 6822 630.57 Rev. 11
j Table 6.2.3-2 (Page 7)
Annulus Conditions vs. Time Following Design Basis Accident TIME ANNULUS ANNULUS PURGE FLOW RECIRCULATION ' LOWER CONTAINMENT 1 (SEC)
TEMP PRESSURE RATE FLOW TEMPERATURE
(*R)
(IN. WATER)
(CFM)
(CFM)
(*R)
- 110000, 581.66
' 1.000 2183 6817 630.57 5
120000.
581.99.
-1.000 2184 6816 630.57 130000.
582.36
-1.000 2185 6815 630.57 l
140000.
582.72
-1.000 2187 6813 630.57 150000.
583.07
-1.000 2188 6812 630.57 160000.
583.42
-1.000 2189 6811 630.57 7'
170000.
583.77
-1.000.
2191 6809 630.57 l
180000.
584,13
-1.000 2192 6808 630.57 190000.
584.47
-1.000 2193 6807 630.57 i
200000.
584.81
-1.000 2194 6804 630.57
{
250000.
586.41
'1.000 2200 6800 630.57
)
300000.
587.85
-1.000 2205 6795 630.57 i
j I
I 2The lower containment temperature shown is not a calculated value but is input from straight line approximations.
See Figure 6.2.1-6 for the calculated lower compartment temperatures for a LOCA.
i I
?
I i
t I
j Rev. 11 r
.g,
..e
-r~...-
.,,-,,_v.
-,,,y,,
,7, -, _, _, _-. - -
._.,,_,m.,,.
m....
._.v_,-.__..
.......,.,m
Table 6.2.3-3 (Page 1)
Dual Containment Characteristics I.
Secondary Containment Design Information A.
Free Volume, ft3 484,090 B.
Pressure, inches of water, gauge 1.
Normal Operation 0.0 l
2.
Postaccident 5.-0.5 C.
Leak Rate at Postaccident Pressure (weight %/ day) 0.2 D.
. Exhaust Fans See Figure 9.4.9-1 E.
Filters See Figure 9.4.9-1 II. Transient Analysis A.
Initial Conditions 1.
Pressure, inches of water, gauge 0.0 2.
Temperature, F
81.1 i
3.
Outside Air Temperature, *F 95 4.
Thickness of Secondary Containment, in Wall 36 Dome 27 5.
Thickness of Primary Containment, in Wall 0.75 Dome 0.688 8.
Thermal Characteristics 1.
-Primary Containment Wall a.
Coefficient of Thermal Expansion, 1/"F 8.4E-06 b.
Modulus of Elasticity, psi 2.9E+07 c.
Thermal Conductivity, Btu /hr-ft-F 25 d.
Specific Heat, Btu /lb *F 0.113 2.
Secondary Containment Wall a.
Thermal Conductivity, Btu /hr-ft-F 0.92 b.
Specific Heat, Btu /lb *F 0.21 Rev. 11
CN,5 radioisotopes following a LOCA by filtering and recirculating a large. lumi of annulus air relative to the volume discharged for negative pressure maintenance; and (3) provide long-term fission product removal capacity by decay and filtration.
This system is provided with two independent, 100 percent capacity ventilation filter systems complete with fans, filters, dampers, ductwork, supports and con-trol systems for each unit.
This meets the single failure criteria.
Switchover between redundant trains is accomplished manually by the operator.
Electrical and control component separation is maintained between trains.
All essential system components, including fans, filter trains, de:ipers, duct-work, and supports are designed to withstand the Safe Shutdown Earthquake.
Essential electrical components required for ventilation of the annulus during accident conditions are connected to emergency Class 1E standby power.
p 9.4.9.2
System Description
The Annulus Ventilation System is shown on Figure 9.4.9-1 and consists of redundant ventilation subsystems for each unit.
Each ventilation subsystem consists of a filter train, fan, dampers, associated ductwork, supports and s )
control systems.
The Annulus Ventilation System filter trains are described j in Section 12.3.3.
/-
The Annulus Ventilation System functions to discharge sufficient air'from the annulus to effect a negative pressure w'th respect to the containment and the atmosphere 60 seconds following a LOCA.
Subsequent to attaining a negative pressure, additional air is discharged as necessary to maintain the pressure l
at or below -0.5 inches water gauge.
In order to mix the inleakage in as large a volume as possible, a large flow of air is displaced from the upper level of the annulus and passed through the filter train before bei.ng returned to the annulus at a low level.
Both the suction and return air flow is accom-plished using ring-type distribution headers in the annulus.
The Annulus Ventilation System is activated by the safety injection signal (Ss).
Upon receipt of this signal the recirculation dampers and discharge dampers are aligned to exhaust 9000 cfm to the unit vent until the annulus negative pressure: (
is >0.5 inches water gauge.
The recirculation dampers and discharge dampers then modulate to exhaust air as required to maintain the annulus negative pressure at -0.5 inches water gauge.
Computer code CANVENT has been developed by Duke Power Company to analyze the thermal effects of a loss-of-coolant accident (LOCA) in a Westinghouse " ice condenser" containment.
CANVENT is capable of evaluating the following factors:
(a) Steady state (pre-LOCA) radial temperature distributions corresponding to fixed outside Reactor Building and inside containment temperatures.
(b) Radial temperature distributions in the steel containment and crdcreie Reactor Building during post-LOCA transient.
o 9.4-21 Rev. 11 i
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TAB!.E 15.0.12-1 (Page 1)
[_
.0FFSITE DOSES (Rem)
.3 FSAR Exclusion Area Boundary
"' Low Population Zone Acci. dent Section Whole Body Thyroid Whole Body,'
Thyroid Main Steam Line Break 15.1.5 Case 1 (No iodine spike) 8.6E-2 7.6 4.4E-3 2.6E-1 Case 2 (Pre-spike) 7.4E-3 2.8 5.3E-4 9.6E-2 Case 3 (Coincident spike) 7.4E-3 2.4 5.3E-4 8.5E-2 Loss of Power 15.2.6 Case 1 (No iodine spike) 4.5E-3 7.0E-2 5.9E-4 6.5E-3 Case 2 (Pre-spike) 4.5E-3 7.3E-2 5.9E-4 7.6E-3 Case 3 (Coincident ypike) 4.5E-3 7.2E-2 5.9E-4 8.2E-3 RodEjectionAccidedt 15.4.8 Side Release 5,1E-2 4.8 1.1E-2.-
2.1 l
Primary,'y Side Release Secondar 3.3E-2 1.2 1.1E-3 3.8E-2 Instrument Line Break 15.6.2 Case 1 (No iodine spike) 1.6E-1 3.2E-1 5.1E-3 1.0E-2 Case 2 (Pre-spike) 1.8E-1 1.9E+1 6.0E-3 6.3E-1 Case 3 (Coincident spike) 1.8E-1 5.2 6.0E-3 1.7E-1
. Steam Generator Tube Rupture 15.6.3 Case 1 (No iodine spike) 6.4E-1 1.5 2.1E-2 8.8E-2 Case 2 (Pre-spike) 7.1E-1 4.4E+1 2.4E-2 1.5 Case 3 (Coincident spike) 7.0E-1 1.2E+1 2.3E-2 4.6E-1
~
Design Basis Accident 15.6.5 Case 1 (With ECCS leakage) 3.0 1.2E+2 7.6E-1 5.1E+1 L.
Case 2 (Without ECCS.. leakage) 3.0 1.0E+2 7.6E-1 4.6E+1 Waste Gas Decay. Tank /
15.7.1 Rupture
.5.0E-1 1.6E-2.
Rev. 11
t CNS The following conservative assumptions are used in the analysis of the releise of radioactivity to the environment in the event of a postulated rod ejection accident.
A summary of parameters used in the analysis is given in Table 15.4.8-2.
1.
Ten percent of the gap activity is released to the containment atmosphere.
2.
50 percent of the iodines and 100 percent of the noble gases in the melted fuel are released.
l 3.
50 percent of the iodine released are deposited in the sump.
4.
Annulus activity, which is exhausted prior to the time at which the annulus reaches a negative pressure of -0.25 in.w.g., is unfiltered.
5.
ECCS leakage occurs at twice the maximum operational leakage.
6.
ECCS leakage begins at the earliest possible time sump recirculation can begin.
7.
Bypass leakage is 7 percent.
8.
The effective annulus volume is 50 percent of the actual volume.
(
9.
The annulus filters become fouled at 900 seconds resulting in a 15 percent reduction in flow.
10.
Elemental iodine removal by the ice condenser begins at 600 seconds and continues for 3328.3 seconds with a removal efficiency of 30 percent.
11.
One of the containment air return fans is assumed to fail.
- 12. The containment leak rate is fifty percent of the Technical Specifications limit after 1 day.
13.
Iodine partition factor for ECCS leakage is 0.1 for the course of the ac-cident.
14.
No credit is taken for the auxiliary building filters for ECCS leakage.
15.
The redundant hydrogen recombiners fail; therefore, purges are required for hydrogen control.
(The following assumptions apply to the secondary side analysis).
16.
All the activity released is mixed instantaneously with the entire reactor coolant volume.
J 15.4-34 Rev. 11
CNS 17.
The primary to secondary leak rate is 1 gal / min.
18.
The iodine partition factor is 0.1.
19.
The steam release terminates in 120. seconds.
t
- 20. All noble gases which leak to the secondary side are released.
21.
The primary and secondary coolant concentrations are at the maximum al-lowed by technical specifications.
l Based on the foregoing model, the primary and secondary side releases may be calculated as well as the offsite doses.
The doses, given in Table 15.4.8-2, are well below the limits of 25 rem whole body and 300 rem thyroid established in 10CFR100.
I 15.4.8.4 Conclusions l-Even on a pessimistic basis, the analysis indicate that the described fuel and clad limits are not exceeded.
It is concluded that there is no danger of sudden fuel dispersal into the coolant.
Since the peak pressure does not exceed that which would cause stresses to exceed the faulted condition stress limits, it is concluded that there is no danger of further consequential damage to the Reactor Coolant System.
The analyses have demonstrated that upper limit in fission pro-duct release as a result of a number of fuel rods entering DNB amounts to ten per-cent.
Parameters recommended for use in determining the radioactivity released to atmosphere for a rod ejection accident are given in Table 15.4.8-2.
The Re-actor Coolant System integrated ' reak flow to Containment following a rod u
ejection accident is shown in Figure 15.4.8-5.
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l 15.4-35 Rev. 11 f
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TABLE 15.4.8-2 (Page 3) i Parameters for Postulated Rod Ejection Accident Analysis Conservative Realistic I
i b.
Dose conversion assumptions Regulatory Guides same 1.4 and 1.109 r
c.
Doses (Rem)
Primary side-Exclusion area boundary Whole body 5.1E-02 i
Thyroid 4.8 Low population zone i
Whole body 1.1E-02 Thyroid 2.1 Secondary side Exclusion area boundary l
Whole body
- 3. 3E Thyroid '
1.2 Low population zone Whole body 1.1E-03 i
Thyroid 3.8E-02 e
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Rev. 11 3 -r n_
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CNS 3.
Annulus activity which is exhausted prior to the time at which the annulus
~
reaches a negative pressure of -0.25 in.w.g. is unfiltered.
4.
ECCS leakage begins at the earliest possible time sump recirculation can begin.
5.
ECCS leakage occurs at twice the maximum operational leakage.
6.
Bypass leakage is 7 percent.
7.
The effective annulus volume is 50 percent of the actual volume.
8.
The annulus filters become fouled at 900 seconds resulting in a 15 percent reduction in flow.
9.
Elemental iodine removal by the ice condenser begins at 600 seconds and continues for 3328.3 seconds with a removal efficiency of 30 percent.
- 10. One of the containment air return fans is assumed to fail.
- 11. The containment leak rate is fifty percent of the Technical Specification limit after 1 day.
12.
Iodine partition factor for ECCS leakage is 0.1 for the course of the accident.
13.
No credit is taken for the auxiliary building filters for ECCS leakage.
- 14. The redundant hyrdogen recombiners fail; therefore, purges are required for hydrogen control.
The resulting offsite doses presented in Table 15.6.5-10 are below the limits of 25 rem whole body and 300 rem thyroid established in 10CFR100.
15.6.5.4.2 Control Room Operator Dose The maximum postulated dose to a control room operator is determined based on the releases of a Design Basis Accident.
In addition to the parameters and as-sumptions listed in Section 15.6.5.4.1, the following apply:
1.
The control room pressurization rate is 4,000 cfm; the filtered recir-culation rate is 2,000 cfm.
2.
The unfiltered inleakage into the control room is 10 cfm.
3.
Other assumptions are listed in Table 15.6.5-11.
I 15.6-18 Rev. 11
TABLE 15.6.5-10 (Page 3)
' Parameters for Postulated Design Basis Accident Analysis Conservative Realistic c.
Doses (Rem)
Case 1 (With ECCS leakage)
Exclusion Area Boundary Whole Body 3.0 Thyroid 1.2E+02 Low Population Zone Whole Body 7.6E-01
. Thyroid 5.1E+01 Case E (Without ECCS leakage)
Exclusion Area Boundary Whole Body 3.0 Thyroid 1.0E+02 Low Population Zone Whole Body 7.6E-01 Thyroid 4.6E+01 e
Rev. 11
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TABLE 15.6.5-11 (Page 2)
Parameters for Postulated Design Basis Accident Control Room Analysis h
Conservative Realistic 3.'
Dispersion data a.
Control room intake X/Q (sec/m )
3 0-8 hrs 9.9E-04 8-24 hrs 7.2E-04 1-4 days 5.1E-04 4 + days 2.8E-04 4.
Dose data i
a.
Method of dose calculations Standard Review Plan 6.4 b.
Dose conversion assumptions Regulatory Guides 1.4, 1.109 c.
Doses (Rem) l-Whole body 4.6E-01 Thyroid 2.1E+01 Skin 9.0 h
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