ML19345E481
| ML19345E481 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 08/29/1977 |
| From: | Bixel D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| NUDOCS 8102040656 | |
| Download: ML19345E481 (4) | |
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- t Aitgast 29, 1977 Mr James G Keppler US Nuclear Regulatory Commission T99 Roosevelt Road Glen Ellyn, IL 6013T DOCKED 50-155 - LICENSE DPR BIG ROCK POINT PLANZ - 0FF-SITE DOSE ASSESSMENT By letter dated May 11, 1977 Consumers Power Company committed to forwarding the results of the study determining the off-site dose assessment for the postulated maximum credible accident resulting in 10% core meltdown. The purpose of this letter, therefore, is to provide these results.
-Site boundary dose rates and integrated doses have been calculated for direct shine resulting to each overland sector at Big Rock Point. Doses represent the Maximum Credible Accident (MCA) 10% core meltdown as described in the FHSR, Section 13 9 Integrated doses are presented in Table I.
A typical dose rate e.nd integrated dose curve (SSW sector) are shown in Figure I.
Calculations have not been audited since these results are an outgrowth of in-plant dose calculations which will not be completed until the end of the year as stipulated in our May 11, 1977 letter.
Previously calculated doses, as presented in FHSR, Section 13.9, were not see-tored.
FHSR Figure 13.6 indicates a 2-hour and 2h-hour integrate d dose of ap-
. proximately 12 millirad and 55 millirad, respectively, from direct shine at one-half (1/2) mile distance. The present calculations, as prestnted in Table I, indicate a range of site boundary doses from 14 to 220 mrad at two hours, and i
47 to 630 crad at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The major reason for the higher doses in the present cgiculations is the added conservative assumption that 10% of the airborne iodines are adsorbed on internal containment structures rather than returned totally to t&( liquid phase upon actuation of containment sprays. Thus, for direct shine cEhiderations, the adsorbed iodines contribute in the same manner as airborne iodines. All other assu=ptions (core release model, air and iron shielding for individual gamma energies, no credit for shielding by forest biomass, etc) are the same as set forth in FHSR, Section 13, for the 10% core melt MCA.
Doses as a function of direction were derived by subdividing the containment
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free volume into T1 individual volume segments. Major plant structures which
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present greater than two feet of concrete shielding (steam drum and control room
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shield wall) vere assuned to provide complete shielding for activity within the volume segments "behind" these shields relative to the site boundary of 'nterest.
It'should be noted, however, that actEil"c'alculations of control room shielding
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are{being performed for the in-plant portion of this study. Only the steel dome, air-from boundary to volume segments, and the two r.Qor concrete shields (as ap-
-prBjriate) were credited for shielding.
No credit was taken for any other inter-vening structures.
Direct shine doses presented in Table I will be added to plume and ground deposition doses' of the FESR, Section 13 9, to determine total dose for each sector, should the specified sector be downwind following a postulated LOCA.
Calculated in-plant radiation levels will allow confirmation that the lowest dose areas are designsted for post-accident assembly,' will identify expected area monitor response as a function of core inventory released to air, and will define times and locations which are acceptable for. entry to perform required emergency operations.
Changes to the Site Emergency Plan to allow inclusion of State of Michigan emer-gency classifications and protective action guides consistent with these classifi-caticns have been developed and are currently under review.
The calculated boundary doses and in-plant radiation levels vill be incorporated into implementation procedural aids following the completion of in-plant dose calculations and audit. It is expected that any required revisions of procedures, including aids, will be implemented by March 1978. Should any changes occur in values presented in Table I or Figure I, revised values will be submitted.
David A Bixel (Signed)
David A Bixel Nuclear Licensing Administrator i
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TABLE I
-f Direct Shine Dose by Sector
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Boundary Dist Doce @ 2 Fours Dose @ 24 Hours Dose 8 30 D Direction' (M)
(Rad)
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92G 0.14 0.k3 0.L8
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0.21 0.63 0.70 SE 890 0.091 0.2B 0 31 SSE 9h0 0.067:
0.21 0.23 s
1090 0.021 0.071 0.080 SSW-nho 0.01h 0.OkT 0.053 SW 990 0.OkT 0.15 0.165 ~
WSW 820 0.22 0.63 0 70 r
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