ML20010A130
| ML20010A130 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 07/24/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20010A126 | List: |
| References | |
| NUDOCS 8108110097 | |
| Download: ML20010A130 (10) | |
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JUL 2 41981 SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 2 TO CONSTRUCTION PERMITS CPPR-108 AND CPPR-109 Introduction On December 19, 1980, Georgia Power Company, acting on its own behalf and agent for Oglethorpe Electric Membership Corporation, Municipal Electric Authority of Georgia, and the City of Dalton, requested an amendment to the Construction Permits CPPR-108 and CPPR-109, for the Alvin W. Vogtle Nuclear Plant, Units 1 and 2, to reflect a modification in plant design. The modification would remove the enclosure building in the application for Vogtle licenses and add as a substitute a steel-framed, metal-siding equipment building from grade to the 270-foot level, and add a more restrictive primary con-tainment leakage rate of 0.2 weight percent (%) per day. These substitutions, together with onsite meteorological data, will ensure that offsite post-accident doses are less than doses judged to be acceptable for t'.esign and are lower than those calculated at the time construction was initially authorized.
la support of this application for ame.1dment, Ge 71a Power Company had previously submitted Supplement No. 6, dated August 21, 1979, to its application fc Construction Permit and Operating License, which included a description of the associated physical modifications. These modifications would entail removal of the enclosure building and its related equipment as described in Sections 1.2.6, 3.8.4.1.1, and 6.6 of the Vogtle PSAR, while retaining from grad 6 to the 270-foot level, an equipment building described in Sections 1.2.6 and 3.8.4.1.1 of Supplement No. 6.
To further support this request for amendment, Georgia Power Company has submitted Supplement No. 8 to the application, dated December 30, 1980, which contains additional information on the more restrictive primary containment leak rate and the resultant projected offsite doses for a design basis accident if the enclosure building and its related equipment were removed.
Georgia Power Company has committed to a containment leak rate of 0.2%
per day with the proposed modified enclosure building design.
Employing this leakage rate, dispersion factors using NRC methodology and the latest onsite meterological data, Georgia Power Company has submitted revised design basis accident offsite doses for the Vogtle Units.
Based on the information submitted in the Preliminary Safety Analysis Report (PSAR) and Supplement Nos. 6 and 8 to the PSAR, the NRC staff has completed its j
review of all safety-significant matters related to the issuance of the construction permit amendment as requested in the December 19, 1980 application.
This Safety Evaluation is issued in support of Amencment No. 2 to Construction 1
Permits CPPR-108 and CPPR-109, allowing the removal of the enclosure building and related equipment.
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0108110097 810724 PDR ADOCK 03000424 A
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. The purpose of this Safety Evaluation is to examine the impact of the l
proposed modifications to the enclosure building for Vogtle. Units 1 and 2.
Specifically, the Safety Evaluation addresses' the following safety significant items:
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53.0. Design criteria for structures, components, equipment and i
- systems, 2.
56.0. Engineered safety features (containment systems), and, 3.
515.0. Accident analyses.
l Evaluation i
We have reviewed this application for amendment submitted on December 19, 1980, and supplemented by letters of August 21, 1979, and December 30, 1980.
I Our review of safety-related matters and our conclusions concerning each item are described in the following subsections of this evaluation report.
Date: Jul 2 41981 l
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JUL 2 41981
$3.0 DESIGN CRITEF.IA FOR STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS We have reviewed the structural aspects pertaining to modifications of the enclosure building in relation to the primary containment integrity.
We
- ind that our prior reviews in Sections 3.3, 3.5, 3.7 and 3.8 of the Safety.
Evaluation Report are not affected by the removal of the enclosure bullding and its replaceraent with an equipment building from grade to the 270-foot level based on the following:
1.
there is no change in the design methods and design criteria of the containment or other structures due to the enclosure building modi fications.
No structural credit has been taken for the l
enclosure building in either the original or modified design case-L 2.
the enclosure building modifications do not affect the plant's i
susceptibility to tornado missiles since it is assumed that the metal siding of the enclosure building afforded no protection t
from missiles; l
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the dynamic response characteristics of the containment building l
under seismic loads would essentially be unaltered because the onclosure building modifications result in only 21, reduction in weight and a very small change in stiffness of the overall contain-I ment building.
Based on the foregoing, we conclude that the safety margin of the original l
structural design or containment integrity is not significantly affected; l
therefore, the proposed design modification is acceptable.
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l JUL'2 41981 56.0 EhGINEERED SAFETY FEATURES 6.2 CONTAIBMENT SYSTEMS 6.2.1 Containment Functional Design In Supple aent No. 6 to the Vogtle 1 & 2 PSAR, dated August 21, 1979, the Georgia Power Company revised its analysis of the radiological con-sequences of a postulated loss-of-coolant accident (LOCA) using assumptions that did not take credit for an enclosure building. This revised analysis is associated with the deletion of the enclosure building and its replacement with an equipment building from grade to the 270-foot level and the more restrictive containment leak rate l
of 0.2% per day.
I Because the enclosure building was designed only to treat leakages l
from the containment building, the functional performance of the con-l tainment building in the event of a LOCA is not affected by the presence i
or absence of the enclosure building. We conclude that since the removal of the enclosure building will have no effect on our prior review of the containment system, the removal of the enclosure building is acceptable with respect to containment functional design.
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JUL 2 41981 515.0 ACCIDENT ANALYSES 15.1 General, l
We have reviewed additional n:eteorological information submitted by the applicant on December 7,1979, to determine the potential offsite doses l
calculated for the postulated loss-of-coolant accident (LOCA) and the fuel handling accident. The calculated doses are presented in Table 15.1 and the assumptions used are discussed in 515.2. A quantitative compari-l son of LOCA doses _for the original and modified designs is presented in l
Table 15.2. All potential doses calculated by the applicant and by the staff for both the nriginal and modified design cases are within the 10 CrR Part 100 guideline values. We concluue that tne enclosure butiding modifications are acceptable.
l TABLE 15.1 CALCULATED DOSES DUE TO DESIGN BASIS ACCIDENTS l
l Exclusion Radius Low Population Zone (1098 meters)
(3220 meters) l 0-2 hours 0-30 days Thyroid Whole Body Thyroid Whole Body l
Loss-of-coolant (rem) 08 2.6 69 1.3 Fuel hcadling (rem) 8 0.7 1.3 0.2
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JUL 2 41901 TABLE 15.2 COMPARISON OF LOCA DOSES FOR THE V0GTLE NUCLEAR STATION LOCA Doses per SER LOCA Doses without (March 1974) with Enclosure Building Enclosure Building 0.2%/ day Leak Rate Primary Con. Leak
& X/Q Values as of Pate of 0.3%/ day 1/80 Exclusion Area Boundary Thyroid (rem) 122 98 Whole body (rem) 7 2.6 Low Population Zone Bo'undary Thyroid (rem) 70 69 Whole body 8
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15.2 DESIGN BASIS ACCIDENT ASSUMPTIONS 15.2.1 Loss of Coolant Accident We have reviewed the information supplied by the applicant in Supplement 8 to the PSAR. Based upon our review of this information, we have modelled the primary containment as two control volumes, consisting of the region covered by the containment spray and the remaining unsprayed containment volume. The spray region was assumed to be 78 percent of the total contalament free volume of 2.62 x 10-6 cubic feet.
The staff also assumed that a 30 second delay exists from the initiation of the accident to the time spray injection into the containment begins. Further, the sprays, enhanced by the l
addition of sodium hydroxide, were assumed to operate until a reduction factor (DF) of 100 was reached in the sprayed region; af ter which the sprays did not remove any additional radioiodine.
The single containment was assumed to leak at a design leak rate of 0.2*. per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at 0.17, per day thereaf ter.
The meteorological conditions and other important parameters used in our analysis of the consequences of a loss of coolant accident are tabulated in Table 15.3 and 15.4, and the calculated doses are given in Table 15.1.
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JUL 2 41981 TABLE 15.3 l
ASSUMPTIONS USED TO ESTIMATE RADIULOGICAL CONSEQUENCES DUE TO A PO5TULATED LOSS OF COOLANT ACCIDENT AT VUGTLE UNIT 5 1 & 2 Power level, megawatts thermal 3565 Operating time, years 3
Primary containment leak rate, 0.2 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> percent per day 0.1 greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> l
Fraction of Core Inventory Available for Leakage from Containment, percent:
Noble Gases 100 Iodine 25 Primary Containment Free Volume, cubic feet 2.75 x 106 Iodine form fractions, percent Elemental 91 Organic 4
Particulate 5
Spray removal rates, per hour Elemental 10 Particu: ate 0.45 Fraction of primary containment unsprayed, percent 22 Relative concentrations, second per cubic meter 0-2 hours at 1060 meters 1.8 x 10-4 0-8 hours at 3218 meters 3.3 x 10-5 I
8-24 hours at 3218 meters 2.2 x 10-s 24-96 hours at 3218 meters 9.2 x 10-6 96-720 hours at 3218 meters 2.7 x 10-6 i
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l TABLE 15.4' Refueling ~ Accident Calculation Input Parameters Shutdown Time 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> Total Number of Fuel Rods in the Core.
50,952 1
Number of Fuel Rods Involved in the Refueling Accident 264 Power Peaking Factor 1.65 l_
Iodine Fractions Released from Pool Elemental 75%
Organic 25%
Filter Efficiencies Elemental 90%
j Organic 70%
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X/0 values, sec/m 0-2 hours 0 1098 meters 1.8 x 10 0-2 hours 0 3220 meters 3.3 x 10 l
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jut. 2 41981 Conclusion _
We find the enclosure bJilding modifications acceptable for the following reasons:
l 1) the safety margin of the original structural design or containment integrity is not significantly affected; l
2) tne functional performance of the containment building in the event of a LOCA is not affected by the presence or absence of the enclosure building; and 3) the calculated dose consequences are less than the guideline values of 10 CFR Part 100 in both the original and modified design cases.
We have concluded, based on the considerations discussed above, that with l
respect to the facility design changes authorized by Amendment Nos. 2 to Construction Permit.s CPPR-108 and CPPR-109, (1) there is reasonable l
assurance, taking into consideration the criteria contained in 10 CFR Part 100, that the proposed facilities can be constructed and operated at the proposed location without undue risk to the health and safety of the public, and (2) the issuance of these Amendments will not be inimical to the common defense and security or to the he71th and safety of the public.
l Approval of this action doas not preclude future changes as deemed necessary by l
accident analyses on the mitigation of more severe accident sequences than I
postulated design basis accidents which will be performed during the operating license review for the entire Vogtle plant.
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