ML20211Q356
| ML20211Q356 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 07/18/1986 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Andrews H GREELEY, CO |
| Shared Package | |
| ML20198J191 | List: |
| References | |
| NUDOCS 8607250024 | |
| Download: ML20211Q356 (5) | |
Text
l c
=o p
k UNITED STATES g
g NUCLEAR REGULATORY COMMISSION 5
j WASHINGTON, D. C. 20555 July 18, 1986 Docket No. 50-267 Mr. Harold L. Andrews Sheriff, Weld County Greeley Office Post Office Box 759 Greeley, Colorado 80632
Dear Sheriff Andrews:
I am replying to your letter dated July 2,1986 to Chairman Lando W. Zech, Jr., of the Nuclear Regulatory Comission, concerning the safety of the Fort St. Vrain Nuclear Generating Station in light of the recent accident at the Chernobyl facility in the Soviet Union.
You expressed interest in two catastrophic accident scenarios:
- 1) complete destruction of the containment building with total release of all fission products, and 2) a major internal malfunction with containment of the radio-activity. You also inquired about " entombment" of the plant, as is being done at Chernobyl.
The hazards relative to the operation of Fort St. Vrain were initially examined by the Division of Reactor Licensing of the U. S. Atomic Energy Comission in a S,afety Evaluation published on January 20, 1972.
In this Safety Evaluation ~, 'the staff anticipated the scenarios discussed in your F
letter.
In particular, in Section 4.4 of this report, the staff reviewed the consequences of the radioactivity released in a rapid depressurization l
of the reactor vessel, designated as Design Basis Accident No. 2.
Enclosure i
1 indicates the radiation doses that could result from this accident. These consequences, which are considered conservative, were found to be acceptable in accordance with the Commission's standards.
In the extremely unlikely event that this accident should occur, the utility would shut down the reactor and remove the remaining decay heat from the reactor core. However, it is anticipated that no further specific actions would be required to prevent additional releases of radioactivity or provide protection to the public.
In Section 4.5 of the Safety Evaluation, the staff also examined the consequences of a major internal malfunction, i.e., permanent loss of forced circulation cooling (designated as Design Basis Accident No. 1). The staff found that the reactor vessel could effectively contain the radioactivity released from the fuel particles with acceptable consequences for the public health and safety. Using conservative assumptions, the staff's evaluation indicated that the radiation doses would be so low as to not require evacuation of the local residents (see Enclosure 2).
8607250024 860718 ADOCKOSOOg7 DR
I l
July 18, 1986 Mr. Harold L. Andrews Since the massive reactor vessel serves as an effective barrier to contain radioactive materials, " entombment" of the reactor is not envisioned either as a necessary precaution, or as a remedial action in the event of an accident.
Other public officials in Colorado have inquired recently about the safety of Fort St. Vrain relative to the Chernobyl accident. Enclosure 3 is a copy of a letter dated May 29, 1986 to Governor Lamm which provides a discussion of the NRC staff's post-Chernobyl reassessment of Fort St. Vrain.
Please note that all docketed materials concerning the safety of Fort St.
Vrain Nuclear Generating Station are available to you at the Greeley Public Library, City Complex Building in Greeley.
Sincerely, o
arold R. Denton, Dir ctor Office of Nuclear Reactor Regulation
Enclosures:
As stated
~
L.
W:
July 18, 1986 Mr. Harold L. Andrews -
Since the massive reactor vessel serves as an effective barrier to contain radioactive materials, " entombment" of the reactor is not envisioned either as a necessary precaution, or as a remedial action in the event of an accident.
Other public officials in Colorado have inquired recently about the safety of Fort St. Vrain relative to the Chernobyl accident. Enclosure 3 is a copy of a letter dated May 29, 1986 to Governor Lamm which provides a discussion of the NRC staff's post-Chernobyl reassessment of Fort St. Vrain.
Please note that all docketed materials concerning the safety of Fort St.
Vrain Nuclear Generating Station are available to you at the Greeley Public Library, City Complex Building in Greeley.
Sincerely, Original signed by Richard Vollmer for Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosures:
As stated DISTRIBUTION:
, Docket File; NRC PDR' Local PDR HDenton/RVollmer ERossi DCrutchfield Glainas SSPD Reading OELD PPAS, Green Ticket No. 001929 Mossburg/ Toms FMiraglia/MSchaaf PNoonan CHinson KHeftner Olynch HBerkow JTaylor, IE RMartin, RIV R0 Williams, PSC B:SSPD DPW SPD DPWRL-B:
.SSPD n
chi KHeitner:ac Olynch HBerJpw 7/17/g6 07N7/86 07/G/86 07/lff/86 07/p/86 DPWM 4DIR f
D NRf:
/~/
FMir<
Aa er HDei. n )'
07/
86 7/g /86 01/ /86
.?
TABLE 4.2 ACCIDENT ASSUMPTIONS RAPID DEPRESSURIZATION ACCIDENT:
DBA-2 ITEM STAFF APPLICANT Coolant Fission Product Inventory Design Design Coolant Lost Unfiltered 100%
100%
i 3.47 x 10 'm /sec 3.47 x 10 m /sec
~
~
Inhalation Pate Meterological Model Volume Source Area Source Pasquill Condition "F"
"G" Wind Speed 0.3m/sec 1 m/sec Dilution Factor:
-4 3
-4 3
X/Q (Exclusion Radius) 8.4 x 10 sec/m 4 x 10 sec/m X/Q (LPZ) 4 x 10-5sec/m3 4 x lo-5 ec/m3 s
Deses - (2 Hr. at exclusion radius) due to Coolant-Borne Activity Thyroid 11 rem 5 rem Whole Body
(
Bone
< 0.2 rem
< 0.1 rem Plateout Inventories - Source Calculated Allowable Design I-131 ' equivalent 5000 C1/ loop 5500 Ci/ loop Sr-90 equivalent 140 C1/ loop 160 Ci/ loop Cr Fraction of Plateout Inventory Released Iodine 0.06 0.0057 Strontium 0.05 0.01 Doses (2 Er. at exclusion radius) due to removed Plateout Activity Thyroid 139 rem 12.4 rem Bone 75 rem 4.7 rem Total Accident Doses at. Exclusion Radius Thyroid 150 rem 17.4 rem Bone 75 rem 4.8 rem Whole Body
t a
s.
~l.
TABLE 4.3
~
p7' ACCIDENT ASSUMPTIONS e
PERMANENT LOSS-OF-FORCED-CIRCULATION CX)0 LING - DBA #1 ITEM STAFF APPLICANT Power Level 879 MWT 879 MWT Assumed duration of accident emissions used in dose calculations 180 days 180 days Halogens Released to PCRV 50% of Core
~
inventory) 5.5%
(
2 Halogen Plateout Factor Halogens Available for Release from PCRV 25% (of Core inventory) 4.6%
Strontium Available for Release from PCRV 1% (of Core inventory) 0.026%
Noble Gases Available for Release from 100% (of Core PCRV inventory) 99%
Reactor Building Filter Efficiencies 90%
90%
Halogens Particulates 95%
95%
PCRV Leak Rate 0.2%/ day 0.2%/ day inhalation Rates
~4 3 Most unfavorable 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period 3.47 x 10 m /sec 0-12 hrs 2.9 x 10~ m /sec
-4 Next most unfavorable 16' hours 1.75 x 10 m / cec
~4 Remainder of 180 day period 2.32 x 10 m /see 2.32 x 10 m /see W
Meteorological Dilution Factors
-6 Most unfavorable 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period 7.8 x 10 sec/m
-6 Next most unf avorable 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> 3.9 x 10 sec/m 1-12 hrs 4,7 x 10~ sec/m
-6 For the 1-4 day period 2.4 x 10 sec/m
~9 l
For the 4-30 day period 7.8 x 10~ sec/m All others 9.5 x 10 sec/m For the 30-180 day period 4.6x 10~ sec/m Strontium Dose Conversion Factor 3.6 x 10~ rem /Ci 1.07 x 10 rem /Ci Course of Accident dose at Low Population Zone
< 8 rem 66 mrem Thyroid Bone
< 5 rem 1.9 mrem Whole Body
< 330 mrem 0.56 mrem l
,