ML19263E796
| ML19263E796 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 06/14/1979 |
| From: | Short T OMAHA PUBLIC POWER DISTRICT |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7906250288 | |
| Download: ML19263E796 (22) | |
Text
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Omaha Public Power District 1623 HARN2Y e O M AM A. NESRASMA se102 e TELEPHONE 536 4000 AREA CODE 402 June 14,1979 Director of Nuclear Reactor Regulation ATTN:
Mr. Robert W. Reid, Chief Operating Reactors Branch No, h U. S. Nuclear Regulatory Co =ission Washington, D. C.
20555
Reference:
Docket No. 50-285 Gentlemen:
The Omaha Public Power District received a telecopy from a member of your staff on April 26, 1978, requesting information in regard to radiation doses delivered through the Fort Calhoun Station equipment hatch during normal operation and after a LOCA.
In 1978, the District performed radiation surveys outside of the containment vessel in order to evaluate the Conmission's concerns. Detailed responses to the questions fored by the telecopy are attached.
Sincerely, f
', a ;%
., > T. E.I Short
[AssistantGeneralManager TES/KJM/BJH:Jcm Attach.
cc: LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue Suite 1100 Washington, D. C.
20036 2215 031 l
l M
79062502 Q g
0-1:
Demonstrate that the non-installation of the removable concrete block shield in the containment equipment hatch will not result in post-LOCA exposures in excess of the 10 CFR Part 100 guidelines at the exclusinn area boundary considering the cloud dose contribution, the gamma shine /skyshine dose contribution, and any neutron dose contribution from a Regulatory Guide 1.4 source tery Because of the complexity of the geometry, the post-LOCA gamma ray shine /skyshine dose contribution should be detemined using detailed transport calculations.
Response
An accident analysis evaluating the shielding effectiveness of the equipment hatch without additional shielding during an MHA was perfomed by the District in 1976. The following conservative assumptions were used for dose evaluations at the exclusion area boundary:
(1) The reactor was operating at full power prior to MHA.
(2) It was assumed tnat a receptor was standing at the exclusion area boundary and was exposed to radiation streaming through the equip-ment hatch opening.
(3) No credit was taken for the steel door in front of the equipment hatch.
(4) Credit was taken for a concrete wall (two feet thick and built of ordinary concrete with density of 2.35 gm/cc) in front of the equipment hatch opening.
(5) The assumptions related to the release of radioactive material from the fuel and containment were derived from Regulatory Guide 1.4.
2215 032
(6) The dose rates specified in 10 r,FR Part 100 were used as acceptance criteria.
The analysis demonstrated that the cumulative dose, due to radia-tion from the containment, at the exclusion boundary represents approxi-mately 4% of the doses specified in 10 CFR Part 100. We believe that the parameters utilized for the accident analysis are still valid and are very conservative.
It should be' pointed out that the meth0dology used for the above referenced accident analysis was derived from the following references:
1.
Reactor Shielding for Nuclear Engineers, N. M. Schaeffer (Editor),
TID-25951.
2.
Calculation of Distance Facto;s for Lower and Test Reactor Sites, DiNunno, et al, TID-14844.
Therefore, in light of the following factors, we believe that the requirements for any additional shielding in front of the equipment hatch are not warranted:
(a) conservative assumptions used for the above referenced accident analysis; (b) the resultant doses are a small fraction of 10 CFR Part 100; and (c) dose equivalent rates (see response to Q-2) during normal operation are ALARA.
2215 033
Q-2:
Demouatrate that, within the exclusion area, the dose rates due to neu-trons streaming through the equipment hatch opening during normal oper-ation are as low as is reasonably achievable (ALARA). As this is an operating facility, provide neutron dose equivalent rate measurement during normal operation in the unshielded radiation zone outside con-tainment as "best evidence" to support your ALARA conclusions. P ro -
vide the details of the monitoring program used to make these measure-ments including instrumentation used, sensitivity, and monitor location.
Response
First of all, it is necessary to reiterate the District's policy regarding the radiation designation of the equipment ha;ch area. The area immedi-ately outside the equipment hatch door is enclosed as part of the auxiliar:.-
building. This enclosed area is locked when not in use and is rarely occu-pied. This enclosed area is designated a radiation area with dose rates, due to neutrons and gammas, ranging from 4 to 11 mrem /hr at the sur-face of the equipment hatch door and dose rates, due to neutron and gam-mas, less than 1 mrem /hr in the truck dock area. These are the dose rates when the reactor is at 100% power. When the reactor is shut down.
this area can be designated as Zone 1 Uncontrolled due to the low radiatien.
Radiation Survey Program A radiation survey of the equipment hatch area was conducted in August 1978 in order to determine the dose rates due to neutrons streaming through the equipment Eatch opening during normal operation. Further details of the survey program are delineated as follows:
2215 034
~~
The plan view of the area in question is shown in Figure 1.
Four cross sectional views of the in-containment equipment hatch door, out-side containment equipment hatch door, inside auxiliary building wall and, outside auxiliary building wall are shown in Figures 2, 3, 4, and 5, respectively. TLD (Thermoluminescence Dosimeter) test badges were placed on each of the above defined walls (cross sectional views) from August 3 through August 10, 1978 while the plant was operating at 100% of power level. Five TLD badges were placed on the in-contain-ment equipment hatch door at specific grid points as shown in Figure 2.
Twenty-one TLD badges were placed on the outside containment equip-ment hatch door as shown in Figure 3.
Seventy-two TLD badges were placed on the inside auxiliary building wall as shown by Figure 4.
Fifty-four TLD badges were placed on the outside auxiliary building wall as shown in Figure 5.
Results of the measurement data regarding the neutron dose eqaivalent rates are presented in Table 1 for each of the four groups. As it is obvious from the results that the dose equivalent rates, primarily from neutrons streaming through the equipment hatch opening, in the equipment hatch area are small; i. e., in the range of 0.2 mrem /hr to 0.4 mrem /hr as compared to the neutron dose equivalent rates in the containment. The confidence level in these dose rates can be enhanced by comparing with monthly TLD exposure results, taken during the months of September 1978 and October 1978, which indicated that the neutron dose equivalent rates were between 0.2 mrem /hr and 0.6 mrem /hr. The radiation survey pro-gram also indicated that the dose equivalent rates from neutrons and gammas, 2215 035
in the equipment hatch area, were between 0.3 mrem /hr and 0.8 mrem /
hr.
As 1) the dose rates in the equipment hatch area are small, 2) the area is designated a radiation area, and 3) the District's operating philosophy conforms with Regulatory Guide 8.10; the dose rates in the equipment hatch area, during normal operation are considered to be A LARA.
Type of Instrumentation The instrumentation (TLD badges) used for determining the neutron dose equivalent rates in the equipment hatch area is of standard geometry fur-nished by the Marshaw Chemical Company. Figure 6 shows the neutron TLD badge, which centains one G-7 card and one N-6 card. The model G-7 cards contain two TLD-700 chips (99.99% Lithium-7) and the Modal N-6 cards contain two TLD-600 chips (95. 6% Lithium-6). B oth type s o f cards utilize a teflon window to hold the TLD chips and both have the American Standard Information Interchange.
2215 036
EQUIPMENT HATCH AREA (PLAN VIEW)
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4 EL.1013 Roll-Up Door To i
Loading Dock
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- FIGURE I -
2215 037
Eauteusar HATCH OOOR
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DIA. = I4, 2215 038
-FIGURE 2-
EQUIPMENT HA TCH Dooe Outside Containment-Looking North W2
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DIA.=14' GRID = 3'x 3' on f 2215 039
- FIGURE 3 -
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AUXILIARY OUILDWG WAU.
Inside wall-Looking South E=
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g Reference Point GRID = 3'x3' 2215 040
-FIGURE 4 -
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AUXILIARY BUILOING WALL Outside wall - Looking North W=
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- EL.1034'- 0" GRIO = 3'x 3'
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Reference Point - Top Edge of Ramp 2215 041
-FIGURE 5 -
TABLE 1.
NEUTRON DOSE EQUIVALENT RATES Below are the readings of the test badges that were placed at the Fort Galhoun Plant from August 3 through August 10, 1978.
Placement of the badges are as follows:
Group 1 - In containment equipment hatch door.
Group 2 - Outside containment equipment hatch door.
Group 3 - Inside auxiliary building wall.
Group 4 - Outside auxiliary building wall.
Neutron Dose Rates Group 1 Badge (mrem /hr) 1 17.7 2
17.2 3
16.5 4
14.4 5
- 5. 9 Group 2 1
- 6. 6 2
- 6. 7 3
- 6. 7 4
4.1 5
- 5. 2 2215 042 6.
- 5. 6 7
5.3 8
5.1 9
- 3. 9 10 0.1 11 4.2
Neutron Dose Badge Rate (mrem /hr) 12
- 2. 6 13
- 3. 0 14 3.9 15
- 2. 3 16 4.1 17
- 3. 9 18
- 3. 6 19
- 2. 6 20
- 2. 6 21 3.6 Group 3
.1
- 0. 2 2
0.2 3
- 0. 2 4
- 0. 2 5
- 0. 2 10 0.1 11
- 0. 2 12 0.2 2215 043
'3 o2 14 0.1 19
- 0. 2 20 0.1 21
- 0. 2 22
- 0. 2
Neutron Dose Badge Rate (mrem /hr) 23
- 0. 2 24
- 0. 2 25
- 0. 2 26
- 0. 2 27
- 0. 2 28
- 0. 2 29
- 0. 4 30 0.2 31
- 0. 3 32
- 0. 3 33 0.2 34
- 0. 2 35
- 0. 2 36
- 0. 2 37 0.2 3E 0.3 39
- 0. 2 40
- 0. 0 4'
2215 044 42
- 0. 3 43
- 0. 3 44
- 0. 2 45
- 0. 0 46
- 0. 2
Badge Neutron Dose Rate (mrem /hr) 47 0.3 48
- 0. 2 49 0.2 50 0.3 51
- 0. 3 52
- 0. 2 53
- 0. 2 54
- 0. 2 55 0.2 56 0.3 57
- 1. 3 58 0.3 59 0.3 60 0.2 61
- 0. 2 62
- 0. 2 63
- 0. 2 64 0.2 2215 045 65
- 0. 3 66
- 0. 2 67
- 0. 2 68 0.2 69
- 0. 2 70
- 0. 2
Neutron Dose Badge Rate (mrem /hr) 71
- 0. 2 72
- 0. 2 Group 4 1
0 2
0 3
0 4
0 5
0 6
0 7
0 8
0 9
0 10 0
11 0
12 0
13 0
14 0
15 0
16 0
17 0
2215 046 18 0
19 0
20 0
21 0
22 0
Neutron Dose Badge Rate (mrem /hr) 23 0
24 0
25 0
26 0
27 0
28 0
29 0
30 0
31 0
32 0
33 0
34 0
35 0
36 0
37 0
38 0
39 0
40 0
41 0
42 0
4 2215 047 44 0
45 0
46 0
Neutron Dose Badge Rate (mrem'/hr) 47 0
48 0
49 0
50 0
51 0
52 0
53 0
54 0
2215 048 f
W.
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TLD-600 Figur e 6.
Neutron TLD Badge 2215 049 W
Code (ASCII) identification matrix.
Chip No.1 (G-7) measures gamma and beta radiation with energies greater 2
than that which will just penetrate the teflon window of 13. 5 mg/cm. Chip No.2 (G-7) is covered front and back by aluminum shields 0.25 inches thick (0.064 cm,170 mg/cm2), and as a consequence is shielded from all beta radiation up to about 0.6 Mev. Thus, Chip No. 2 (G-7) measures only gamma radiation and bata particles with energies greater than about 0. 6 Mev. The reading of Chip No. 2 (G-7) is taken to be a measure of the gamma radiation dose, and the reading of Chip No.1 (G-7) minus the reading of Chip No. 2 (G-7) is taken to be a measure of the beta radiation dose.
Chip No.1 of the N-6 card receives neutrons, both slow and fast, and also gamma radiation. Chip No. 2 (N-6) receives no slow neutrons because all slow neutrons arestopped by the 0.015-inch (O. 038 cm) cadmium shields which cover Chip No. 2 (N-6) front and back. Thus, Chip No. 2 (N-6) receives
- only fast neutrons and gamma radiation.
Slow Neutron Sensitivity Determination
- 1) For the purpose of the neutron sensitivity determination, a TLD badge 6
with TLD-600 (enriched in Li ) on one side and TLD-700 (enriched in Li7) on the other side is placed in the lucite jig and irradiated for a predeter-mined time. TLD-600 has high sensitivity for thermal neutrons (6 = 980b) whereas TLD-700 is insensitive to neutrons (6 = 0. 033b) and is used here as a Y-ray monitor.
- 2) The irradiation of a similar neutron badge is done in the same manner and for the same time, except that both sides are covered with cadmium.
2215 050
's
- 3) The badges are read out with a -calibrated reader in the usual manner.
4} The neutron response R (in units of nanocoulombs) of a badge is obtained 6 - R ; Where:
as R= R 7
R6b, R6c are the bare and covered TLD-600 responses, resp.
are the bare and covered TLD-700 R7b, R7e re spons es, resp.
- 5) The slow neutron response R, is given by the difference between the bare response Rb and the cadmium-covered response, R, as c
Rb = R6b - R7b Rc=R6c 7c
~
R, = Rb-R e
- 6) The sensitivity Sn of TLD-600 is determined by, S
n Dn Where:
D is the slow neutron dose in mrem.
n D is found from the TLD irradiation n
time T and the slow neutron flux F s
according to :
9 2
n/c'm / rem)
Dn = F,T/(10 109 n/cm2 per rem is the fluence-dose conversion factor for neutrons of thermal energy up to about 10 kev.
Fast Neutron Sensitivity Determination In order to determine the TLD sensitivity for other than slow neutron energies, TLDs must be given known neutron doses at several discrete, or average neutron energies. To find the fast neutron sensitivity response,
~
2215 051
~
neutron sources, are chosen to cover a wide range of neutron energies in order to define the sensitivlty response as a function of neutron energy.
Such a response curve can then be used to obtain TLD neutron sensitiv-ity at any desired neutron energy within the jurisdiction of the curve.
The entire neutron sensitivity response Sn (E) can be modeled as:
Sn = aEb g TOT (Li6).
It is assumed that S follows the total neutron cross section of Li-6, n
FTOT, as a function of neutron energy E, and is proportional to it accord-b ing to the weak energy dependence of aE, where a and b are constants which were determined empirically. This technique depends on determin-ing an effective neutron energy for a multi-energy neutron sperctrum. The sensitivity response function above is fitted using both absorbed doses (mrad) and equivalent doses (mrem) in determining Sn.
2215 052
_