ML20210S897
| ML20210S897 | |
| Person / Time | |
|---|---|
| Site: | Satsop |
| Issue date: | 02/03/1976 |
| From: | Vollmer R Office of Nuclear Reactor Regulation |
| To: | Deyoung R Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-1653 NUDOCS 8605290498 | |
| Download: ML20210S897 (12) | |
Text
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DISTRIBUTION:
7Ib d Central Files NRR/ Reading File g
AAB/ Reading File Richard C. DeYoung, Assistant Director for Light Water Reactors, DFM WPFSS 3, 5 SER INPUT FBOM AAB PLANT MAME: WFPSS 3, 5 LICENEING STAGE: CP
'BOCIET WMBER: 58 g 5_08 _
RESPONSIBLE BRANCH: LNR #3; A. Bournia, LPH MILESTONE NUMBER: 24.31 REQUESTED COMPLETION DATE: January 5, 1976 REVIEW STATUS: AAB SER Input Complete Enclosed are Sections 6.2.3, Containment Air Purification and Cleanup Systens (Spray), and 15.0, Radiological Consequences of Accidents, for the WPPSS 3, 5 SER from the Accident Analysis Branch. The information contained in Amendment 28, dated January 20, 1976, was included in our review.
Richard H. Vollmer, Assistant Director for Site Analysis Division of Site Safety and Environmental Analysis Office of Nuclear Reactor Ear,ulation
Enclosure:
As stated cc: w/o enclosure R. Boyd R. Haiamman Y. Stallo l
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l 6.2.3 Containment Air Purification and Cleanup Systems The containment. spray system is used for fission product scrubbing following a postulated LOCA. Sodium hydroxide is added to the spray solution to enhance the system's effectiveness for elemental iodine removal.
The method of NaOH injection chosen by the applicant differs from conventional designs in that a pre-pressurized additive tank is used as the motive force for additive injection, as opposed to additive pumps or eductors. It is essential to demonstrate adequate performance of this design by a pre-operational test program. The pre-operational test-ing described in the PSAR is inadequate for this purpose, as it is designed to demonstrate the operability of individual components of the system, as opposed to an integrated test of the spray and spray additive systems, to verify that the system, as installed, is capable of producing the pH values assumed in the system evaluation. The applicant has made a verbal commitment to devise a system. test to verify the adequacy of the design. We find this acceptable, subject to documentation in a PSAR amendment.
Subject to verification of the system design by test, we find that the pH values quoted in the PSAR meet our requirements, i.e., a spray solution pH between 9.0 and 11.0 during the additive injection phase, and a long term containment sump solution pH above 6.5.
Spray solutions with these pH values have been shown ef fective for elemental iodine removal and recention. We calculate removal coefficients of 10.0 and
~1 0.7 hrs for the elemental and particulate forms of iodine, respectively, in an estimated sprayed region comprising 81" of the total free volume 1
of the containment.
15.0 Radiological Consequences of Accidents The postulated design basis accidents analyzed by the applicant to determine the offsite radiological consequences are the same as those analyzed for previously licensed PWR plants. These include a loss-cf-coolant accident (LOCA), a steam line break accident, a steam generator tube rupture, a fuel handling accident, a rupture of a radio-active gas storage tank, and a control rod ejection accident. We have reviewed these accidents and further evaluated the LOCA and the fuel handling accident. The offsite doses we calculated for these accidents mad the assumptions used in the analyses are given in Sections 15.1 and 15.2 of this report.
On the basis of our experience with the evaluation of the steam line break and the steam generator tube rupture accidents for PWR plants of similar design, we have concluded that the consequences of these accidents can be controlled by limiting the permissible reactor coolant system and secondary coolant system radioactivity concentrations.
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At the operating license stage of review, we will include limits in the technical specifications on the reactor coolant system and secondary coolant system activity concentrations such that the potential two-hour doses at the exclusion radius, as calculated by the staff for these accidents, will be small fractions of the guideline doses of 10 CFR Part 100. Similarly, we will include limits in the technical specifi-caticus on gas decay tank activity so that any single failure such as L.
g.,
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, 15.0 Radiological Consecuences of Accidents (Cont.)
lif ting or failure to close of a relief valve will not result in doses that are more than a small fraction of the 10 CFR Part 100 guideline values.
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I The control rod ejectien accident will also be evaluated at the operating l
license stage of our review. A technical specification will limit the allowable operational leakage of reactor coolant into the steam generator secondary side to assure that the radiological consequences of this accident will be well within the dose guidelines of 10 CFR Part 100.
15.1 Loss-of-Coolant Accident Dose Analysis The WPPSS 3, 5 pressurized water reactors are each surrounded by a double -
containment structure consisting of a low leakage steel containment vessel and an outer reinforced concrete shield building to mietmize the offsite radiological consequences of the design basis loss-of-coolant accident
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(LOCA). The applicant has specified a design leak rate for the pri=ary containment of 0.2% per day for the first 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> following the LOCA and 0.1% per day for the duration of the accident. For dose evaluation I
purposes, radioactive materials that leak from the primary containment following a postulated LOCA can take any of the following pathways to the environment:
1.
Leakage to the annulus between the primary and secondary containment structures (the shield building annulus) which l
will be treated by the shield building ventilation system.
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15.1 Loss-of-Coolant Accident Dose Analysis (Cont.)
2.
Leakage to the ECCS area of the reactor auxiliary building which will be treated by the ECCS area exhaust system.
3.
Direct bypass leakage which will not be treated.
Both the shield building ventilation system (SBVS) and the ECCS area exhaust system are engineered safety features.
The applicant has determined the leakage pathway percentages to be 40%
of the primary containment leakage to the shield building annulus, 50*
co the ECCS area of the reactor auxiliary building, and 10* direct by-pass leakage.
'a'e have used these leakage pathway percentages in our calculation of the design basis LOCA doses. The results of our calcu-lations are shown in Table 15.1 and the assumptions used in the analysis are listed in Table 15.2.
The doses we calculate for the LCCA are well w'ichin the guideline dose values of Regulatory Guide 1.4 for a plant at the construction permit review stage (150 rem thyroid and 20 rem wnole body).
In modeling the releases through the shield building annulus pathway, we conservatively assumed that the $3VS operates at full exhaust throu ghout the course of the accident following the initial pressure transient in the annulus. In actuality, the S3VS will switch from full exhaust to full recirculation in numerous short ti=e steps in order to control the pressure buildup in the shield building annulus following the accident. This = ode of operation will provide additional holdup time for the fraction of the primary containment leakage which enters the shield building annulus.
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. As part of the loss-of-coolant accident (LOCA), we have considered the consequences of leakage of containment sump water which is circu-laced by the ECCS outside the containment after a postulated LOCA.
We have assumed the sump water to contain a mixture of iodine fission products in agreement with Regulatory Guide 1.7.
At the time of the recirculation mode of operation, about 2200 seconds after the accident, the sump water is circulated outside of the containment to the Reactor Auxiliary Building to be cooled. If a source of leakage should develop, such as from a pump seal, a portion of the iodine would becooe gaseous and would exit to the atmosphere. As the ECCS area is served by an engineered safety feature air filtration system, we find that the offsite doses from possible equipment leakage would be within the guidelines of 10 CFR Part 100, even for substantial amounts of leakage.
The applicant will provide redundant hydrogen recombiners for the purpose of controlling any accumulation of hydrogen within the primary containment after a design basis LOCA. In the event both recombiners fail, the applicant has provided a backup purge system which discharges to the shield butiding annulus and subsequently to the atmosphere through the SBVS filters. Assuming operation of the SBVS at full exhaust and with no credit for mixing or holdup La the annulus, we have computed the additional dose an individual mignt receive due to purging the contain-ment after the accident. The calculated doses are shown in Table 15-1
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. 15.1 Loss-of-Coolant Accident Dose Analysis (Cont.)
and the assumptions used in the analysis are listed in Table 15.3.
We find that the calculated doses at the low population zone distance from purging, when added to the LOCA doses, are well within the guide-lines of 10 CFR Part 100.
15.2 Fuel Handling Accidsnt For the analysis of the fuel handling accident, we have assumed that a fuel assembly was dropped in the fuel pool during refueling operations and that all of the fuel rods in the assembly were damaged thereby re-leasing the volatile fission gases from the fuel rod gaps into the pool.
The radioactive material that escaped from the fuel pool was assumed to be released to the environment over a two-hour time period with the iodine activity reduced by filtration through the fuel building exhaust system. The dose results are shown in Table 15.1 and the assumptions and parameters used in the analysis are shown in Table 15.4.
The dose model and dose conversion factors employed in the analysis are in agreement with those given in Regulatory Guide 1.25.
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TABLE 15.1 RADIOLOGICAL CONSEQUENCES OF DESIGN BASIS ACCIDENTS Exclusion Area 1/
Low Population Zone /
2 2-Hour Dose, Rem 30-Dav Dose, Rem Accident Thyroid Whole Body !
Thyroid Whole Bodv2!
2 Loss-of-Coolant 74 15 25 5
30 1
Hydrogen Purge Fuel Handling 3
3 1.
Exclusion area boundary distance = 1,310 meters 2.
Low population zone distance = 4,830 meters (3 miles) 3.
Dose from low penetrating beta radiation considered as a skin dose and not included in whole body dose.
TABLE 15.2 ASSUMPTIONS USED IN THE CALCULATION OF LOSS-OF-COLLA.Vr ACCIDENT DOSES Power Level
~4100 MWt Operating Time' 3 Years Fraction of Core Inventory Available for Leakage 25%
lodines Noble Gases 100%
Initial Iodine Composition in Containment Elemental 91%
Organic 4%
Particulate 5%
Shield Building Annulus Volume 3etween Upper and 4.3 x 10' ft3 Lower Elevation of SBVS Headers Mixing Fraction in Annulus 50%
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-Primary Containmen't Leak Rate 0-36 hours ~
0.2% per day rs
-> 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />'s 0.1% per day
' Primary Containment Leak Paths To Shield Building Filters 40%
To ECCS Area Exhaust Filters 50%
Direct to Atmosphere (Bypass) 10%
Shicld Building Ventilation S -tam Flow Distribution Tire Step
' ted t ulation Flow, cfm.
Exhaust Flow, cfm 0-30 sec 0
0 30-161 sec
.,000 0
161 see - 30 days 0
10,000 Shield Building and ECCS Area Iodine Filter Efficiencies
' Elemental Iodine 99%
Organic Iodine 99%
Particulate Iodine 99%
Primary Containment Volumes 6
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' Sprayed Volume 2.75 x 10 f t-5 3
Unsprayed Volume 6.5 x 10 g I
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-z-TABLE 15.2 (Cont'd.)
Containment Spray System Removal Coefficients Elemental Iodine 10 hr'1 Organic Iodine 0
Particulate Iodine
,1 0.7 hr Mixing Rate Between Sprayed and Unsprayed Volumes 21,500 cfm Elemental Iodine Spray Decontamination Factor 100 Minimum Exclusion Area Boundary Distance 1,310 m Low Population Zone Distance 4,830 m X/Q Values
-3 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> @ EAB 1.0 x 10 sec/m 0 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> @ LPZ 1.0 x 10 sec/m
-5 3
8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> @ LPZ 6.7 x 10 sec/m
-5 3
1 - 4 days @ LPZ 2.8 x 10 sec/m
-6 3
4 - 30 days @ LPZ 7.7 x 10 sec/m P
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' TABLE 15.3 HYDROGEN PURCE DOSE INPUT PARAMETER $
Power Level 4100 MWt Containment Volume,
3.4 x 10 ft Holdup Time in Containment Prior to Purge Initiation 16 Days
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4 Purge Duration 30 Days Purge Rate 50 sefs
-SBVS Filter Efficiency for Iodir.-
99%
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4-30 Day X/Q at 4,830 meters 7.7 x 10 sec/n I
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1 TABLE 13.4 ASSUMPTIONS USED IN THE FUEL HANDLING ACCIDENT ANALYSIS Power Level 4100 MWe Number of Fuel Rods Damaged 236 4
l Total Number of Fuel Rods in Core 56,876 i
Radial Peaking Factor of Drnaged ?.ods 1.65 Shutdown Time 72 Hours Inventory Released From Damaged Rods 10%
(Iodines and Noble Gases)
Pool Decontamination Factors Iodines' 100 Noble Gases 1
Iodine Fractions Released From Pool Elemental 75%
Organic 25%
Filter Efficiency for Iodine Removal 99%
0-2 Hour X/Q Value at 1,310m 1.0 x 10-3,,cf,3 es hni i i