ML20094C628
| ML20094C628 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-15.6.4, NUREG-75-87, NUREG-75-87-15.6.4, SRP-15.06.04, NUDOCS 9511020289 | |
| Download: ML20094C628 (4) | |
Text
E NUREG 75/087
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f U.S. NUCLEAR REGULATORY COMMISSION 1
STANDARD REVIEW PLAN gr O
OFFICE OF NUCLEAR REACTOR REGULATION SECTION 15.6.4 RADIOLOGICAL CONSEQUENCES OF MAIN STEAM LINE FAILURE OUTSIDE CONTAINMENT (BWR)
REVIEW RESPONSIBILITIES Primary - Accident Analysis Branch (AAB)
Secondary - Site Analysis Branch (SAB)
ReactorSystemsBranch(RSB)
I.
AREAS OF REVIEW The AAB review covers the following areas:
1.
The release of primary coolant due to a main steam line failure outside of containment in a boiling water reactor (BWR) plant.
2.
The calculation of whcle body and thyroid doses at the site boundary due to the releases resulting from these accidents.
The purposes of the review are to calculate the whole body and thyroid doses resulting from a postulated failure of a main steam line outside containment, and to assure that releases due to the failure are adequately limited by the technical specifications on primary coolant activity. The RSB will notify the AAB if this accident is predicted to cause fuel failures. The RSB may also be requested to confirm the assumption of instant-aneous reactor coolant release. The SAB provides the reviewer with the accident condition wind speed at the nearest exclusion area boundary.
II.
ACCEPTANCE CRITERIA The plant is considered adequately designed against the radiological consequences of a main steam line failure outside containmente and the primary coolant activity appropriately limited by the technical specifications if calculations show that the resulting doses at the nearest exclusion area boundary are small fractions of the 10 CFR Part 100 exposure guidelines.
III.
REVIEW PROCEDURES The reviewer selects and emphasizes aspects of the areas covered by this review plan as may be apnropriate for a particular case. The judgment on areas to be given attention and emphasis in the review is based on an inspection of the material presented to see whether it is similar to that recently reviewed on other plants and whether items of special safety cignificance are involved.
USNRC STAND ARD REVIEW PLAN
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4 The detailed review of the radiological consequences of a main steam line failure outside containment is done at the.op2 rating license (OL) stage when system parameters and accident analyses are fully developed. At the construction permit (CP) stage. the review is limited to a brief survey of the pertinent portions of the plant design and the applicant's discussion of the accident to detemine that there are no uhusual features that would prevent limitation of doses to acceptable levels by appropriate limits on coolant activity concentrations.
- The AAB review of main ' steam line failure accidents at' the OL stage consists of the following steps:
1.
Review of the applicant's descriptions of the steam line failure accidents.
2.
Perfonnance of an independent analysis of the radiological consequences of the failure of the main steam line, using the assumptions of Regulatory Guide 1.5. (Ref. 2).
At present, the following conservative assumptions are used to simplify the analysis:
a.
The mass of reactor coolant released to the environment is 140,000 lbs. The release is assumed to occur instantaneously. (These assumptions are made unless notified otherwise by RSB.)
b.
The. initial fission product concentrations in the primary coolant are those given as the maximum values in the technical specifications. The RSB reviews, on a generic basis, the effect on core thermal margins of a steam line failure in a BWR. If this event is predicted to cause fuel failures. RS8 notifies AAB so that the predicted magnitude and extent of fuel failures can be properly considered in the evaluation of the radiological consequences. No decontamination 4
factors, or other reductions in the fission product concentrations are assumed.
3.
Determination of the values of meteorological parameters for the dose calculations.
The SAB provides the reviewer with the accident condition (5 percentile) wind speed at the nearest exclusion area. boundary. The X/Q value for the calculation of the two-hour doses is obtained from Regulatory Guide 1.5 (Ref. 2) and corrected for wind speeds differing from 1 m/sec (inverse ratio).
4.
Review of the results of the dose calculations. The resulting doses from these accidents should be small fractions of the 10 CFR Part 100 limits. If this-is not the case, the primary coolant concentration limits in the technical specifications should be reduced.
IV.
EVALUATION FINDINGS The reviewer verifies that sufficient information has been provided and that the review and calculations support conclusions of the following type, to be included in the staff's safety evaluation report at the operating license stage:
"The applicant's analysis of a failure of a main steam line has been reviewed, and an independent analysis of this accident has been performed by the staff. The results of 15.6.4-2 11/24/75 p i page
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E the staff's calculations are presented in Table The technical specifications for the primary coolant activity concentrations limit the consequences of is accident to a small fraction of 10 CFR Part 100 exposure guidelines."
At the construction permit stage, the following paragraph is included in the sta s safety evaluation report:
"On the basis of our experience with the evaluation of steam line failure accidents for boiling water plants of similar design, we have concluded that the consequences of these accidents can be controlled by limiting the permissible primary coolant radio-activity concentrations so that potential offsite doses are small. We will include appropriate limits on the primary coolant activity concentrations in the technical specifications."
V.
REFERENCES 1.
10 CFR Part 100, " Reactor Site Criteria."
2.
Regulatory Guide 1.5, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors."
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