ML19308A207
| ML19308A207 | |
| Person / Time | |
|---|---|
| Issue date: | 12/29/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-15.1.5, NUREG-75-87, NUREG-75-87-15.1.5, SRP-15.01.05, NUDOCS 7901020069 | |
| Download: ML19308A207 (6) | |
Text
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TRANSMITTAL SHEET
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REVISION TO
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NUREG-75/ 087
" Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (LWR Edition)
Section No.
15.1.5 Revision No.
1 Filing Instructions Pages to be removed New pages to be inserted Page Number Date Page Number Date 1S.1.5-8 11/24/75 15.1.5-8 Rev. I 15.1.5-9 11/24/75 15.1.5-9 Rev. 1 J
15.1.5-10 11/24/75 15.1.5-10 Rev. 1 15.1.5-11 11/24/75 15.1.5-11 Rev. I 15.1.5-12 11/24/75 15.1.5-12 Rev. 1
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APPENDIX STANDARD REVIEW PLAN SECTION 15.1.5 RADIOLOGICAL CONSEQUENCES OF MAIN STEAM LINE FAILURES OUTSIDE CONTAINMENT (PWR)
REVIEW RESPONSIBILITIES Primary - Accident Analysis Branch (AAB)
Secondary - Hydrology and Meterology Branch (HMB)
Reactor Systems Branch (RSB)
Ef fluent Treatment Systems Branch (ETSB)
I.
AREAS OF REVIEW The AAB reviews (a) the sequence of events in the applicant's description of the steam line failure accident outside containment, with and without offs'ite power, to assure that this sequence of events represents the most severe case from the standpoint of release of radioactive materials, and (b) the coolant activity concentration technical specification limits to assure that potential doses resulting from this accident are adequately limited.
The HMB provides acceptable atmospheric dispersion X/Q values for this accident. The ETSB provides acceptable models and assumptions for iodine spiking and their effects upon coolant activity. The RSB determines the acceptability of the applicant's description of events, including operator actions, for this accident.
II. ACCEPTANCE CRITERIA Standard Technical Specification (STS) limits on PWR primary and secondary coolent activity concentrations and primary-to-secondary leak rate have been issued (References 4, 5 and 6). The plant is considered adequately designed against the consequences of a main steam line failure outside containment if calculations show that the resulting doses at the exclusion area and low population zone boundaries, based on STS limits, are: (a) small fractions (less than 10%) of the 10 CFR Part 100 exposure guidelines, and (b) within 10 CFR Part 100 guidelines for the cases of a preaccident iodine spike or one rod held out of the core.
If the doses are not within these guidelines using the S.TS limits, the technical speci'ica-tion limits on coolant concentrations and/or primary-to-secondary leak rate are reduced until the calculated doses are within these guidelines.
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l III. REVIEW PROCEDURES The reviewer selects and emphasizes aspects of the areas covered by this appendix as may be appropriate 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
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safety significance are involved, i
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At the construction permit stage, there is generally insufficient infomation available to j
make meaningful radiological consequence calculations for these accidents. At this stage, the review is limited to a brief review of the applicant's discussion of the main steam line failure accidents to detennine that there are no unusual design features that would 4
preclude the limitation of radiological consequences by appropriate limits on coolant con-centrations and primary-to-secondary leakage. The detailed review of radiological con-sequences of the main steam line failure accidents is done at 7.he operating license stage when system parameters are fully developed.
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The Standard Technical Specifications for each of the three PkR vendors' NSSS include limits on the primary and secondary coolant activities and primary-to-secondary leak
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rate. If the applicant proposes to use these standard limits and his plant is one of the standard NSSS/ BOP plants for which the steam line failure accident has been evaluated generically with the standard coolant activity and leakage limits, the reviewer need not reevaluate the offsite doses from this accident if the X/Q's for the site under review are lower than the limiting X/Q used in the generic review of the standard plant steam line failure.
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l The AAB review of main steam line failure accidents at the operating license stage consists
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of the following:
1 1.
Review of the applicant's descriptions of the steam line failure accident (with and
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without offsite power). This includes a review of the time sequence of occurrence l
of events.
2.
The RSB should be contacted for a determination of the acceptability of the applicant's description of events including operator actions. The AAB reviews the sequence of events to assure that this sequence of events represents the most severe case from the l
standpoint of release of radioactive materials and calculated doses.
3.
Determination of coolant activity concentrations. The reviewer assumes the primary and secondary coolant activity concentrations allowed by the technical specifications (SAR Chapter 16 or the Standard Technical Specificauons given in References 4) 5 or6)asequilibriumconditionspriortotheaccident.
4.
Determination of_ the iodine spiking effects. Two cases are analyzed, one with an j
iodine, spike assumed to begin because of reactor trip or primary system depressuriza-I tion when the steam line break occurs. and one with an iodine spike assumed to have begun well before the accident due to a previous. reactor transient (Refs. 3 and 7).
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The ETSB provides the AAB with the iodine spiking model. If the applicant proposes to use a spiking model different than the model described below, the ETSB should be requested to review the model.
The current spiking model assumes that at the time the steam line break occurs, the iodine release rate from the fuel rods to the primary coolant (expressed in unit of curies per unit time) increases by a factor of 500 relative to the release rate cal-culated assuming that the primary coolant iodine concentration is at the equilibrium concentration technical specification limit.
For the case with an iodine spike which already exists (due to a previous transient),
the iodine concentration is aswmed to be that allowed by figure 3.4-1 of References 4, 5 or 6, or that which is proposed in Chapter 16 of the SAR.
5.
Review of the effects of possible fuel damage during the accident on exclusion area boundary and LPZ doses. Additional coolant activity may become available for release if fuel failures result fran the accident. The RSB reviews the effect of a steam line break (with the most reactive control rod stuck at its fully withdrawn position) on core thermal margins. If this event is pre;1cted to cause fJel failures, RSB notifies AAB. The reviewer assumes the applicant's calculations of fuel damage are correct unless informed otherwise by RSB. If fuel damage does occur, calculations should be perfomed in order to assure that 10 CFR Part 100 guidelines are not exceeded (with-out a preaccident iodine spike).
6.
Determination of the leakage into the steam generators. Normal operating primary-to-secondary leakage is assumed to exist in the steam generators. The leakage rate should be the maximum allowed by the technical specifications. This value is 1 gpm in the STS but may be lower if required because of the consequences of a rod ejection accident or an anticipated transient without scram (ATWS). The leakage should be apportioned between affected and unaffected steam generator (s) in such a manner that the calculated dose is maximized.
7.
Determination of iodine transport. During periods of steam generator dry-out, all iodine transported to the secondary side by primary coolant leakage is assumed to be released to the environment. During periods of total submergence of the tubes, the fraction of iodine lost is equal to the flash fraction of the primary coolant ledage; appropriate credit for scrubbing by the secondary coolant may also be claimed (Reference 2). Any todine transferred to the secondary side coolant will become air-borne at a rate which is a function of the steaming rate and iodine partition coefficient. An iodine partition coefficient of 100 between steam generator water and I
steam phases may be conservatively assumed, unless the applicant presents reasonable evidence that the use of some other value is justified.
8.
The AAB provides the HMB with the release location and release conditions. The HMS then provides the AAB with appropriate X/Q values for this accident.
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9.
Calculation of tha exclusion area boundary (E't) and low population zona (LPZ)
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boundary doses. The reviewer computes the doses for the steam line break accident.
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both with and without preaccident fodine spiking. A breathing rate of 3.47 x 10~4 3
m /sec is used in the calculation of thyroid doses for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the steam line break.
- 10. Review of the results of the dose calculations. The calculated doses are considered acceptably low if: (1) the doses calculated without assum..ig the existence of a pre-accident iodine spike, but assuming that iodine spiking occurs as a result of the accident, are less than a small fraction of 10 CFR Part 100 guidelines; and (2) the doses calculated assuming preaccident iodine spiking or additional fuel failures occur as a result of the accident (assuming the most reactive control rod remains stuck in its fully withdrawn position) are less than the dose guidelines of 10 CFR Part 100. If the doses are not within these guidelines, the technical specification limits on equilibrium and/or spiked primary coolant activity, or the technical specification limit on primary-to-secondary leak rate, should be reduced until the calculated doses are within these guidelines.
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IV.
EVALUATION FINDINGS L
The reviewer verifles that sufficient information has been provided and the review and calculations support conclusions of the following type, to be included with the RSB find-ings in the staff's evaluation report at the operating license stage:
"The radioactivity release has been evaluated according to a conservative description of the plant respcase to the accident. Our calculated doses are presented in j
Table and our assumptions are listed in Table
" Technical specification limits on primary and secondary cool nt activities will limit potential doses to small fractions of the 10 CFR Part 100 exposure guidelines. The potential doses are within the 10 CFR Part 100 exposure guidelines even if the acci-dent should occur with a preaccident iodine spike or assuming additional fuel failures occur during the accident as a result of the most reactive control rod remaining fully withdrawn."
The following paragraph is added if fuel damage is found to be a possible consequence of the accident:
"The evaluation of the main steam line failure outside containment has been evaluated with
% fuel damage in the core (as a result of the most reactive control rod remaining fully withdrawn). The resulting doses are within the guidelines of 10 CFR Part 100 provided the normal operating primary-to-secondary leakage is limited to gpm."
At the construction permit stage, the following paragraph is included with the RSB findings in the staff's safety evaluation report:
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"On thm basis of our experi:nce with the tvaluation of steam line and steam ginirator tube failure accidents for PWR plants of similar design, we have concluded that the consequences of these accidents can be controlled by limiting the permissible primary and secondary coolant system radioactivity concentrations and/or primary to secondary leak rates so that potential offsite doses are small. At the operating license stage, we will include appropriate limits on these parameters to be included in the plant technical specifications."
If the pla-t is a standard plant which has been reviewed before, the following paragraph may be uss "The radiological consequences of a steam line break in the (vendor's name) standard NSSS/(A/E's name) BOP were analyzed generically by the staff and the results reported in the staff SER dated (NSSS/B0P SER issue date). The offsite doses were found to be 3
acceptable for sites with X/Q's equal to or less than sec/m, if the coolant activities and steam generator primary-to-secondary leak rate are limited to the values in the (vendor's name) Standard Technical Specifications (NUREG-(issue date)). The technical specifications the applicant will use include these coolant activity and steam generator leak limits. The staff has estimated the 0-2 hour X/Q at the exclusion area boundary of the site is sec/m, as discussed in Section (2.3.4) in this SER. This X/Q is lower than the limiting X/Q for this standard NSSS/ BOP plant. Therefore, we conclude that the offsite doses from a steam line break accident in the plant would be less than the current dose guidelines and are acceptable, although a specific dose calculation for this accident has not been performed."
V.
REFERENCES 1.
10 CFR Part 100, " Reactor Site Criteria."
2.
A. K. Postma and P. S. Tam, " Iodine Behavior in a PWR Cooling System Following a Postulated Steam Generatur Tube Rupture Accident," NUREG-0409. USNRC, 1978.
3.
R. R. Bellamy, "A Regulatory Viewpoint of Iodine Spiking During Reactor Transients."
Trans. Am. Nucl. Soc., 28,(1978).
4.
Standard Technical Specifications for Combustion Engineering PWRs, NUREG-0212.
5.
Standard Technical Specifications for Westinghouse PWRs, NUREG-0452.
6.
Standard Technical Specifications for Babcock and Wilcox PWRs, NUREG-0103.
7.
W. F. Pasedag, " Iodine Spiking in BWR and PWR Coolant Systems," CONF-770708, 3-217 (1977).
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