ML20213E005
| ML20213E005 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 02/10/1982 |
| From: | Houston R Office of Nuclear Reactor Regulation |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-0457, CON-WNP-457 NUDOCS 8203030207 | |
| Download: ML20213E005 (11) | |
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50-347 RWHouston FEB 101992 lEt'ORANnVM FOR: Robert L. Tedesco, Assistant Director i
for Licensing Division of Licensing FROM:
R. Wayne douston, Assistant Director for Radiation Protection f
Division of Systems Integration SUBJ EC T :
WASillhGT0!4 fiUCLEAR PLANT Uh!T 2 SER INPU A
ACCIDErlT EVALUATION BRAliCH g M
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Wf:P-2 N'
Licensing Stage: OL
&l Project Manager:
.1. Auluck, Licensing Branch i.'2 9
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Requested Cor'pletion Date:
February 12, 1982 5
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Hoview Status: AEB Review Complete h
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Enclosed are the revised pages to the draft SER input trans iltted to you in a ne',o dated October.21,1981. The changes in these pages are noted by a line in the right hand colunn.
The unchanced pages of the draft SER input shculd be considered as final.
The three potential open itens identified in the October 21, 1981 transmital have been resolved by the other dranches anr1 appropriate changes have been nade to our SER input.
The potential open issues have been resolved in the following nanner:
1.
The Siting Analysis Branch has indicated that its concern witn tne OnE railroad in the site vicinity has been resolved and AEB does not need to consider any potential adverse impacts on the control roon uperators f ron possible shipments of toxic materials, d.
The Containnent Systems Branch (CSB) has accepted the applicant's esti.nate of 3 scfn of Dypass leakage tnrough the feedwater lines. This increases the total containment bypass leakage f ro"i the 'ariginal U.74 scth to 3.74 scth and the revised doses in the LUCA evaluation of Chapter 15 reflect tnis cnange.
Ide results of our calculations indicate that altnough the doses are considerably greater tnan those in tne draf t SER tne radiological consequences 'are still less than the 10 CFR Part 100 guideline vetlues and the plant design remains accept.tDie. djpass leakage linits for the feedwater lines and other bypass paths will be incorporated into the plant Technical Specifications by the CSB.
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The Auxiliary Systems Branch (ASB) has indicated that the applicant will incorporate the 11.5 scfh per MSiv leak rate into the plant Technical Specifications. The original AEB analysis assumed tnis leak rate through the MSIV's and, therefore, no change was dictated by tne i
resolution of this issue by ASB.
Other pages have been changed to insert referecnes to the Standard Review Plan.
f This review was coordinated by Angela Chu and any questions regarding it inay be directed to her on x 23351.
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R. Wayne Housten, R. Wayne Houston, Assist. int Director for Radiation Protection Division of Systems Integration cc w/ encl:
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indi'.i Lal sections are in accordance with these appropriate NJREG-CE00 Sections.
2.3.1 Regional Climatology j
The Columbia Basin of western Washington has a mild, dry climate as a result of frequent incursions of maritime polar air from the Pacific Ocean into the continental steppe climatic area. Although maritime polar air masses are predominant over the basin, the air usually loses most of its moisture in as-cending the western slopes of the Cascades and is warmed in descending the eastern slopes.
Occasionally, continental air penetrates into the basin from the interior of the continent.
These continental air masses are responsible for the large annual range of temperatures in the region and, also, can cause large diurnal temperature ranges. Data collected at the Hanford Meteorological station show that annually the temperature may range between -33 C (-27 F) to 46 C (115 f).
Since the basin is in the "rainshadow" of the Cascades, pre-cipitation is sparse, averaging only 150 to 200 mm (6 to 8 inches) annually including wintertime snow.
l The regional meteorological conditions used as design and operating bases in-clude heavy rain, snow, and ice, thunderstorms and hail, tornadoes, strong winds, high air pollution potential, and dust storms.
The design snow load used for all WNP-2 structures is 98 kg/m (20 lbs/ft ).
Thunderstorms may occur at the site during any month, however, the thunderstorm season is from April to i
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trunderstorm is 0.55 inch in 20 ninutes. (Hail-stones of 1/2-inch diameter also fell during this storm.)
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The site is in the Class 111 tornado region identified in Regulatory Guide 1.76.
The plant has been designed for a maximum wind speed of 360 mph, a rotational speed of 300 mph, a maximum translationat speed of 60 mph a pressure drop of 3 psi and a rate of pressure drop of 1 psi /sec.
These values are closer to the more stringent Class I area design basis tornado charactaristics in Regulatory Guide 1.76, and are acceptable.
There is a high frequency of low-level inversions in winter over the site area; l
the occurrence of moderately and very staDie conditions between the surface and 60 meters in the winter is 66.57..
Two out of three winter seasons can be expected I
to produce a stagnation period of at least eight da,<s.
These stagnation episodes result in inhibited dispersion of atmospheric effluents, faust storms can occur at the site area during windy periods.
The information available from measurements made at the Hanford Reservation identifies a " worst case" dust storm with time integrated dust loading of 160 ug-hr/m.
The staff concludes that, pursuant to the requirements of 10 CFR Part 100.10, adequate consideration has been given to the regional climatology. The require-ments in 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 2, to corsider natural phenomena, and GDC 4, to determine the design basis tornado for the generation of missiles have been met for meteorological parameters.
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Therefore, the control room design design satisfies NUREG-0800 and meets GDC 4, " Environmental and Missile Design Bases" and GDC-19,
" Control Room" and is acceptable.
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Cl:arup Systems" and GDC 43, " Testing of Contair. c-nt Atraspnere Cleanup Systems." Therefore, v.e conclude that the system design r.eets the acceptance criteria of SRP Section 6.5.3 and is acceptable.
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the radiological consequences of such an accident are within the guicelines set forth in 10 CFR Part 100.11 (a)(1) and (2).
The analysis has included the following sources and radioactivity transport paths to the atmosphere:
4 (1) contribution from containment. leakage:
(2) contribution from pcst-LOCA leakage from ESF systems outside i
containment; and (3) contribution from Main Steam Isolation Valve leakage.
The staff's review confirms the applicant's finding based upon the following:
(1) The applicant's provisions for and design of the containment system-and the Standby Gas Treatment System are acceptable as identified in Chapter 6 of this report; and l
(2) The staff's independent analysis of the radiological consequences i
of a hypothetical design basis LOCA as described below.
A. Staff Evaluation
- 1. Containment Leakage Contribution The staff's calculation of the consequences of the hypothetical LOCA used the conservative assumptions of SRP Section 15.6.5, Appendi x A (NUREG-0800), and positions C.1.a through C.1.e of Regulatory Guida 1.3 (Revision 2), " Assumptions Used for Evaluating i
the Potential Radiological Consequences of a Loss-of-Coolant Accident for Boiling Water Reactors." The primary containment was assumed to leak to the secondary containment at a constant rate of 0.5 percent of
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This leakage rate will be incorporated in the Technical Specifications and be used in the leak rate tests for 10 CFR Part 50 Appendix J.
The pressure within the reactor building is maintained pt a negative pressure i
of 0.25 inches water gauge during normal operation by exhausting the reactor building air through the normal ventilation system.
Upon receipt of a Safe'ty Features Actuation Signal, the normal ventilation system is automatically switched off and the Standby Gas Treatment ventilation system is actuated.
The applicant's analysis indicates that during this changeover a pressure transient occurs within the reactor building, such that the pressure increases to a maximum of atmospheric pressure and then returns to a value of -0.25 inches i
water gauge.
The total time for the system to return the secondary containment to a pressure of -0.25 inches water gauge was estimated by the applicant to be i
as long as two minutes.
Thus, during this transient, the pressure within the reactor building is not expected to become positive. Maintenance of a negative pressure is required as a criterion for precluding direct outleakage.
A value of -0.25 inches of water gauge is typically used to assure that no local positive pressure exists in portions of the secondary containment building volume due to various atmospheric effects such as winds and temperature differences.
In the staff's analysis no mixing of the primary containment leakage was assumed in the reactor building air during this period.
The assumptions used in calculating the Design Basis LOCA doses are summarized in Table 15.6.
The calculated doses resulting from the LOCA are summarized in Table 15.1.
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LeMage f rca Dgi esced F3 '-:y F:-en res (ESF) components outside the primary contair. ment releases iodines to the secondary containment, then the iodines are mixed within the secondary containment with activity from the primary containment leakage, t
Releases to the envircncent are filtered by the Standby Gas Treatment System.
The staff has assumed a value of 1 gpm for calculation of the ESF component leakage contribution to the LOCA dose.
The results of the staff calculations are summarized in table 15.1.
Because the applicant has provided a ESF grade filtration system which will filter the reactor building exhausts containing airborne iodines released from ESF components passive failure leakage, the staff has not calculated the contribution to the LOCA dose resulting from ESF component leakage due to passive failures.
4 B. Staff Findings The staff concludes that the distances to the exclusion area boundary and to the low population zone boundary of the WPPSS-2 site, in conjunction with the engineered safety features of the WPPSS-2 plant, are sufficient to provide reasonable assurance that the total radio-logical consequences of such an accident will be within the exposure guidelines set forth at 10 CFR Part 100, paragraph 11. This conclusion is based on the staff review of the applicant's analysis and on the independent analysis by the staff which confirms that the calculated total doses are within these guidelines, l
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- -ect i Ln 15.5.2 ar.d a re l i s ted i n T 5bl e 15.4 a.aif:u; doses are shcan in Table 15.1.
These doses are well within the 10 CFR Part 100 guideline values, as specified both in Regulatory Guide 1.11 and the acceptance criteria of SRP 15.6.2.j Therefore, the staff concludes that the design of the WNP-2 plant will be effective in controlling the release of potential radioactivity to the environment folicwing a postulated instrument line break.
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.3 i e feeling crane ir.to the reactor c; i n.tcr vessel 1
l bead is off.
This results in the release of gasecus fission products to the refueling water and eventually to the reactor building at-mosphere.
The applicant states that no airborne activity is expected 4
to be released through the reactor building ventilation system, because the system design is such that the isolation valve will be fully shut prior to the airborne activity reaching the vent.
The staff has, however, evaluated the fuel handling accident according to SRP Section 15.7.4 and assumed activity was released from the plant.
The staff used the applicant's number of damaged fuel rods (124 fuel rods i
total; 105 rods failing on the first impact, 19 rods failing on the secondary impact).
The rods were assumed to be damaged during fuel j
movement 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter reactor shutdown.
The staff's evaluation of the l
radiological consequences of this accident used these assumptions as well i
as the assumptions given in Regulatory Guide 1.25 (Positions C.1'.a through C.1.h), " Assumptions used for Evaluating the potential Radiological consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactor."
All of the radioactivity is assumed to be released af ter processing by the Standby Gas Treatment System filters.
Our assumptions are I
listed in Table 15.3.
The resultant salculated doses are shown in l
Table 15.1 and are 0.5 rem to the thyroid and 0.2 rem to the whole body at the exclusion boundary and at low population zone are less than 0.1 rem to the thyroid and to the whole body.
These doses are l
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