ML20041C176

Determination of Acceptable Bypass Leakage of Secondary Containment Used in Design Basis Analysis of Radiological Consequences of Loca.

ML20041C176
Person / Time
Site:	Clinton
Issue date:	02/28/1982
From:	Green G ILLINOIS POWER CO.
To:
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ML20041C175	List: ML20041C174 ML20041C176
References
NUDOCS 8202260303
Download: ML20041C176 (13)

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Clinton Power Station Unit 1 Determination of an Acceptable Bypass Leakage of Secondary Containment used in the Design Basis Analysis of the Radiological Consequences of a Loss-of-Coolant Accident i

G. S. Green Illinois Power Company February, 1982 7 i

8202260303 820223 PDR ADOCK 05000461 E PDR

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PURPOSE i~

The-purpose of'this' report'is.co1 demonstrate

. the acceptability:of.a Technical 1 Specification. limit of.11 percen't unfiltered bypass .leakagelfrom the Clinton Unit 1 containment.' This acceptability is based on meeting 1the regulatory' requirements for the. ,

i analysis of the radiological consequencesJof a Loss-

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of-coolant accident, and is demonstrated'using'the con-

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servative dose conversion factors asLused-by the NRC.

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SUMMARY

OF DISCUSSION TOPICS The following topics will be presented in the DISCUSSION section of this report:

-1. Radiation safety benefits of.ll percent (vice 4%) bypass leakage limit.

2. 'Conservatisms in'the bypass leakage calcu-lations.

3. Conservatisms .:ba the MSIV' leakage calcula -

tions.

4. Conservatisms in the-dose conversion factor calculations.

5. Appropriate limit 'for offsite dose calculations.

6. Appropriate' operating power level' assumption.

7. Re-estimate of-0-2 hour thyroid dose. based ,

on 11% bypass leakage.

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BACKGROUND

.The Clinton FSAR'has_ proposed a maximum allowable containment leakage rate of 0.65%'per-24_ hours and a 12 percent unfiltered' bypass. The FSAR'also' presents a computed time of 194 seconds between isolation of the normal _ secondary containment ventilation system and the establishment of 1/4 inch WG subatmospheric pressure-within the secondary containment by the'SGTS.

In the Clinton SER, the NRC staff ~ states'that under provisions of the SRP, all containment leakage during the 0-194 see time period is considered as by'-

passing the SGTS, and because ofithis, "... the doses; computed for the applicant's proposed Technical-Speci-fications exceed 'the staff guidelines." The NRC then stated.

that a bypass leakage limit of 4 percent would be required ~

for the CPS Technical Specifications.

Illinois Power believes that the NRC position is more restrictive than NRC's own regulations and in fact detracts from optimum plant nuclear safety. Specifically a 4 percent bypass leakage limit will. contribute to a.

real' increase in plant personnel exposure, whereas it is not required for meeting the off site dose-limits resulting from a low probability lobs of coolant accident.

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DISCUSSION

'l. Radiation ~ safety benefits of 11' percent (vice 4%)-

bypass leakage limit The valves and penetrations in the designated bypass paths have a design leakage of about 1/3 of the NRC proposed 4% bypass leakage limit.

However, after several years of plant operation, it is likely that the leakage rate in these valves will approach the 4% limit. Therefore a'very rigorous surveillance and maintenance program will be required which would_ tend to increase-the radiation exposure of plant personnel.

The Clinton plant has committed (a) to comply with nuclear regulations that plant.personne1' doses be kept as low as reasonably achievable (ALARA) as well as (b) to meet regulations concerning calculated offsite' doses.

An 11 percent bypacc'lcakage limit would permit a more reasonable level of valve leakage surveillance and maintenance; this would therefore minimize un-necessary radiation exposure of plant personnel.

2. Conservatisms in the bypass leakage assumptions Bypass leakage can hypothetically occur via process. lines or penetrations which are routed be-tween the primary containment atmosphere and the atmosphere outside secondary containment. In order-for the bypass leakage to occur, piping or penetra-tion failure must occur outside of secondary con-

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-DISCUSSION (Cont'd.)

2. -(Cont'd.)

tainment. Often the failure must also occur inside the primary containment. The calculations-do not take credit.for the low probability of coincident failure of all bypass leakage paths.

In the Clinton SER, the NRC' staff states that all containment leakage during the 0-194 second time period is considered as bypassing the SGTS. (The 194 seconds is the computed time for establishment of 1/4 inch WG-subatmospheric pressure within secondary containment). Such a requirement is ultra-conserva-

~ tive:

a. Little, if any, transfer of' radioactivity from the reactor to the primary containment, or from the primary to secondary containment-would occur during the first 194 seconds.

b. For the relatively small amount of. radioactivity.

transferred to the secondary containment during the first 194 seconds, very little would be released to the atmosphere during the first 194 seconds.

Illinois Power Co. believes that the acceptability of an 11 percent bypass leakage limit can.be demonstrated, even without taking' credit for these bypass leakage conservatisms.

3. Conservatisms in the MSIV leakage calculations The NRC staff has not allowed credit for the transport

DISCUSSION (Cont'd.)

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3. (Cont'd.)-

delay:related to the~MSIV Leakage Control System-(LCS) inucalculating the off-site doses. ' Illinois Power Co.

believes that'such credit.is appropriate and supported

-by RG.l.96, Revision 1 (Section C). In~ fact, RG 1.96 states that " Staff analyses of the contribution of main steam isolation valv'e. leakage.to total calculated offsite doses in postulated design--

~ basis accidents made with conservative allowances for transport delay effects show that the-2-hour site boundary dose is not affected by th'e 'subj ect- leakage."

Allowing such credit would reduce the calculated total off-site dose, and would permit a higher bypass leakage.

4. Conservatisms in the' dose conversion factor 1

calculations NRC staff has indicated:in1 discussions with.GE-and IP, that dose conversion factors (DCF)'(Rem /CI) from the Task Group' Lung Model'(TGLM) (Ref. NUREG-CR-0150)-

are preferred.but.that DCF's from TID-14844 are~also.

acceptable. GE's calculation uses DCF's from NUREG-0172

-(RG 1.109). The DCF's from TID-14844'are derived from ICRP'#2 whereas the.DCF's from NUREG-0172 are derived from ICRP #10. The factors in' ICRP # 2 and ICRP #10 are identical except that ICRP #10 takes into account the.different half lives of the various iodine isotopes.

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~NRC will accept TID-14844 (ICRP #2) because its.DCF's DISCUSSION (Cont'd.)

4. (Cont'd.)

are similar to the TGLM. The NRC prefers the TGLM.

because it conservatively assumes'all' iodine to be in the.

particulate form. In absence of data to the contrary, NRC staff does not allow credit'for precipitation of~-

some particulates prior to' reaching exclusion area bound-

'ary. Also in absence of data ~to the contrary, NRC re -

quires that the most restrictive particulate sizesito be assumed.

For comparison purposes, the DCF's from TID-14844 and Reg Guide 1.109 are summarized below:

DCF's (Rem / Curie) '

Isotope TID-14844 RG 1.109 I-131 1.5x10 6 1.5x10 6-I-132 5.4x10 4 1.4x10 4 I-133 4.0x10 5 -2.7x10 5 I-134 '2.5x100 3.7x10 3:

I-135 - 1.2x10 5 5.6x10 4

5. Appropriate limits on offsite dose calculatio'ns Regulatory Guide 1.3, Rev. 2, states that "It should be shown that the offsite dose consequences will be within the guidelines of 10CFR Part 100."

The guidelines of 10CFR Part 100 provide the following limits:

a. Exclusion Area boundary 0-2 hr dose

1. Whole body - 25 Rem

2. Thyroid from iodine inhalation - 300 Rem

DISCUSSION (Cont'd.)

5. (Cont'd.)

b. Low Population Zone boundary dose (during entire. period of radioactive cloud-passage)

1. Whole body - 25 Rem

2. Thyroid from iodine inhalation - 300 Rem

6. Appropriate operating power level assumption The Clinton SER (Table 15-2) assumes an operating power level of 3039 Mwt'(105% of Rated). Appare'ntly this is based on SRP 15.65 App A (Rev 1), Paragraph III.1.which requires the reviewer to assume that the core has operated at design power level for about -3 years. The'105% of rated is the instantaneous power level for' designing sys--

tems but should not be considered the design value for determining fission product inventory due to long term operation.

A more appropriate criteria'for the. assumed power level is given by SRP 15.6.5-(Rev 2) Paragraph III.4.a which requires a power level of 102% for evaluating ECCS performance. It is logical to conclude that if 102% of rated power is appropriate for ECCS performance analysis, then 105%_of rated is excessive for fission product inventory calculations.

In summary,-some conservatism in the assumed power

-level is appropriate. However an assumed pouer level of

- 102% of rated should be adequate to account for

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. DISCUSSION - (Cont ' d. )--

6. '(Cont'd.)

uncertainties'in power levelLmeasuremer.ts, and is supported by SRPE15.6.5, Revision 2, Paragraph III.4.a.

'7. Re-estimate of 0-2 hour thyroid dose based on 11%. bypass:

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The Clinton SER tabulates the.following;NRC calcu-laced radiological doses as-a consequence lof a'de-sign basis. loss of coolant accident:

0-2 hour doses,, exclusion 0-30 day-doses,-low' area boundary, rems. population. zone, rems Lossiof . .

Coolant Accident thyroid whole-body- thyroid whole body Bypass 131' 1 71' O.2 SGTS 28 10 27 3.4 LCS 16 6 16 2.0-Total 176 17 114 5.6 The above values are based on an~ assumed bypass leakage

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of 112 scfh. For assumed increases.in bypass leakage, the 0.2 hr exclusion area boundary thyroid dose is clearly _

limiting.

Based on NRC's very conservative assumptions, the 0-2 hr thyroid dose is re-estimated for an 11 percent bypass leakage. First, estimate the portion of the 131 Rem

a. :due-to the 100% bypass during 0-194 sec and

b. due to the 4% bypass during 194 sec-2 hour

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. 0-DISCUSSION (Cont'd.)'

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(a). = 100% :x 194 sec 100% x 194-scc + 4% x 17200-194) sec x 131 Rem-

'(a) = 53.6 Rem.

(b) =.131 - 53.6 (b) = 77.4~ Rem l

Next,-estimate the dose for a bypass leakage of 11%:

Bypass (194.sec - 2'hr) = 11 x~77.4 = 212.9: Rem Leakage T Dose

) = 26 Rem SGTS=28Remxfl"0 y_ 0 J

4 Next, adjust the doses for an assumed operating power' level of 2952 Mwt-(102%.of rated) 100% Bypass Dose'(0-194 sec) =hhhh:x53.6=52.1 11% Bypass. Dose (0-2 hr) = 29 2 x 212.9 = 206.8 952-SGTS- =

0 x 2 6 = 25'. 3 MSIV =

.x-16 = 15.5 Next,- adjust the doses for no dose contribution from s MSIV LCS.

e' DISCUSSION (Cont'd.)

~7. (Cont'd.)

The re-estimates of the 2 hr thyroid dose are summarized below:

NRC Calculated Dose Re-estimates of NRC Dose 4% Bypass 11% Bypass 11% Bypass 11% Bypass 3039 Mwt 3039 Mwt 2952 Mwt 2952 Mwt LCS Dose LCS Dose LCS Dose No LCS Dos Bypass 131 265.6 258.9 260.3 0-194 see ( 53.6) ( 52.7) ( 52.1) 194sec-2hr ( 77.4) ( 212.9) ( 206.8)

SGTS 28 26 25.3 25.3 MSIV LCS _

16 16 15.5 0 Total 176 Rem 307.6 Rem 299.7 Rem 285.6Re In summary, a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exclusion area boundary thyroid dose of 285.6 Rem is estimated, using the following adjustments to NRC's very conservative assumptions:

(1) Increase bypass leakage from 4 percent to 11 percent (2) Reduce the assumed reactor operating' power level from '

105% to 102% of rated.

(3) Eliminate the dose contribution.from MSIV leakage by taking credit for fission product transport delay provided by the Leakage Control System.

The 285.6 Rem is less than the 300 Rem limit and therefore should be acceptable.

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-' CONCLUSION-Based on-the previous discussion, it-is concluded-that the Clinton Power Station Unit: 1 can _be operated safely.with Technical Specification Limits of

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a. 0.65%/24 _hr containment integra'ted leak rate-r.nd

b. 11% unfiltered bypass leakage

- We would hope that NRC will; find our position

- acceptable'and;that such a finding will be included'in a4 i

supplement to the Clinton SER.

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