ML20216F169

Rev 3 to CR Thyroid Dose from MSLB Event

ML20216F169
Person / Time
Site:	Braidwood
Issue date:	08/29/1997
From:	Bergner J COMMONWEALTH EDISON CO.
To:
Shared Package
ML20216F120	List: ML20216F114 ML20216F135 ML20216F162 ML20216F169
References
95-011, 95-011-R03, 95-11, 95-11-R3, NUDOCS 9709110176
Download: ML20216F169 (9)

Text

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1 Braidwood Calculation No. 95411 Control Room Thyroid Dose from a Main Steam Line Break Revision 3 i August 29,1997

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Exhibit C NEP.12 02 Revisi;tg CALCULATION TITLE PAGE Page No.1 Calculation No: 95-011 Comed O'% . A^M Discipline Code: ._

Draidwood STATION UNIT (S)1&2 System Code: VC.MS TITLE: __ContteLRoom TiwroidEms_&eni a Main Stemn_Linc31rakfyrnt X Safety Related - - - _ Augmented Quality Non Safety Related REFERENCE NUMilERS Type Number Type Number COMPONENT EPN: DOCUMENT NUMilERS:

EPN Compt Type DELpe/Sub Tmc D2cmnent Number

1. Calc / Eng DRW 97-078 M Rev/0

2. Nureg-0800

3. UFSAR Tables 15.6 9,6.41 and figurc 7.2-1 (Sht. 8)

4. Tf D USAEC TID 14844 5 Calc / Eng DR VC-02, Rev/2 6 Correspondent Westinghouse letter CAE 97171 REMARKS:

REVISING APPROVED ORGANIZATION REY. NO. PRINT / SIGN DATE O Braidwood SEC Druce Acas 3/01/95 1 Braidwocxi SEC Bruce Acas 3/17/95 2 Braidwocxl SEC Bruce Acas 4/13/97 3 Braidmxxl SEC 7 j gygyg h fg '

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F8008 00Qt9 HF5 F A tt H 0 . 8 680 663 9110 09*03 9P 908Wid Q E hibit E NEP 12-02 Revision 5 COMMONWEALTH EDISON COMPAST CALCULATION REVISION PAGE __

CALCULATION NO. 95.oi 1 PROJECT NO. N/A PAGE NO.: 2 REVISION SUMMARIES REY: 3 REVISION

SUMMARY

This calenlation is being rtytsed to mclude the followint :

1)Updme the Calculation Title

2) Recalculate the conuol reorn thyroid dose using the actnity relcaic data from calculadon BRW 97 0798 M. Rev. 0

)) Include a calculation to determine the Unit I control dose based on the Unit 1 Cycle 7 projected end of cycle primary to secondary leakage and oore activity This revision supersedes all previous revisions of this calculabon Electamic Calculadon Data Files: None (Proywn Nwe, Vmion, Fde name eWeddaem/. nun)

Prepared by. Jan1;. Im  % $ '

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Print / Sign Reviewed by: GG4L* 8 l. A H T

%* J.i S fo,j f /f f y Print / Sign Date Type of Review fM Detailed ( l Alternate i 1 Test DO AN ASSUMPT10NS IN DES CALCtR.ATION REQUIRE LATER VERIFICATION [ ] YES ( X 1 NO Truked br _

REV:

REVISION

SUMMARY

Elecunnic Cakulation Data Files:

(Pngram Name, Vw% rile same entwomew min)

Prepared by: _

Print / Sign Date Rmewed by:

Print / Sign Date Type of Review I 1 Detmled I 1 Ahernate [ ] Test DO ANY ASSUMmONS IN THIS CALCULATION REQUIRE LATER VERITICATION [ ] YES [ } NO Tracked by:

SEP 03 '97 08:59 630 663 7118 PAGE 02

Eshibit C NEP.12-02 Revision 5 COMMONWEALTil EDISON COMPANY CALCULATION TABLE OF CONTENTS PROJECT No. N/A CALCULATION NO. 95411 REV.NO.3 PAGE NO. 3 SECTION PAGE NO, SUB PAGE NO.

TITLE PAGE 1 l 2

REVISION

SUMMARY

TABLE OF CONTENTS 3 4

PURPOSE /0BJECTIVE 4

MFT110DOLOGY AND ACCEPTANCE CRITERIA i

. ASSUMPTIONS 4 4

DESIGN INPUT 4

REFERENCES 5-7 CALCUIATIONS

SUMMARY

AND CONCLUSIONS ATTACilMENTS N/A

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Exhibit E NEP 12 02 j Revision 5 I

COMMONWEALTil EDISON COMPANY j l cal,CUI.ATION NO. t 95-011 PROJECT NO. N/A PAGE NO. 4 l PURPOSE /OIUECTIVE: ,

The purpose of this calculation is to usess the post accident radiological dose to control room 3 occupants following a main steam line break (MSLB) accompanied by primary to secondary coolant i leakage.

METilODOLOGY AND ACCEPTANCE CRITERIA:

The approach is to demonstrate that the consequences of the MStil accident will be less than the design basis loss of coolant accident (LOCA) assessment and are, therefore, bounded by the assessment included in the Updated Final Safety Analysis Report (UFSAR).

ASSUMPTIONS:

The control room ventilation is assumed to be in operation following the MSLil. The basis for this

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assumption is discussed in CALCULATIONS section.

DESIGN INPUT:

j The iodine releases and the maximum acceptable steam generator leak rates are from calculation BRW 97 078 M, Rev. 0 (Reference 1). The releases from the containment and the resulting control room dose are from the UFSAR (References 3 and 7).

REFERENCES:

1. Comed Calculation IIRW 97 078 M, Rev. O," Allowable Leaktate Calculation for Steam Generator Interim Plugging Criteria," August 29,1997.

2. USNRC, Standard Review Plan, NUREG-0800, Section 15.6.3," Radiological Consequences of Steam Generator Tube Failure (PWR)," Section 11, Acceptance Criteria, Rev. 2, July 1981.

3. B/B UFSAR, Table 15.6 9

4. US AEC Technical Information Document TID-14844, Calculation of Distance Factors for Power and Test Reactor Sites, March 1962.

5. Sargent & Lundy Calculation BR VC-02, Rev. 2, " Radiation liabitability for the Control Room ,"

May 9 1988.

- 6. B/B UFSAR, Figure 7.21 (Sheet 8)

7. B/B UFSAR, Table 6.41

8. . Westinghouse Letter CAE 97 171, dated July 21,1997, pertaining to the RCS water density used in determining Byroa and Braidwood Alternate Tube Plugging Limit.

l REVISION NO.: 3 l

Exhibit C NEP 12 02 Revision 5 COh!A10NWEALTil EDISON COAIPANY

l cal,CULATION NO. 95-Oll PROJECT NO. N/A PAGE NO. 5 l 4

CALCULATIONS:

Comparison of Releases:

In order to determine control room doses for the MSLB case from doses calculated for the LOCA case, the radionuclide releases in each case must first be identified. Since the thyroid dose is the governing limit, releases of radioactive iodines will be compared.

• MSLB Release Reference 1 considers three release sources:

a. Primary coolant with a pre-accident iodine spike

b. Prirnary coolant with a post accident iodine spike

, c. Secondary coolant with iodine concentrations at a predetermined limit Reference 1, then, evaluates these to obtain a limiting primary to secondary lenkrate as follows. For a one gallon par minute primary to secondaiy leak, and a RCS I 131 activity of lpCi/gm we have:

Dose Equivalent Exclusion Area Source 1-131 Release Boundarv Dose a 15.9 Ci 4.60 rem b 10.6 Ci 3.05 rem c 5.31 Ci 1.53 rem NRC acceptance criteria considers the following combinations of sources and acceptance criteria:

Source Acceptable Dose Basis, Reference 2 at EAB a+c 300 rem 10CFR100 limit ,

b+c 30 rem 10% of10CFR100 limit l REVISION NO.: 3 l

Exhibit C NEP 12 02 Revision 5 COMMONWEALTil EDISON COMPANY I CALCULATION NO. : 95-011 PROJECT NO. NT PAGE NO. 6 l Using this criteria, Reference I calculates a limiting primary to secondary leak rate of 9.3 gallons per minute (Scenario b4 c). )

Using this leaktate, the iodine releases for the different sources are as follows :

Source Dose Equivalent I 131 Release a _

148 Ci b 99 Ci e 6 Ci Under these conditions, the worst case combination of(a+c) or (b+c)is (a+c), which results in a release of 154 curies of dose equivalent 1 131.

• LOCA Release The releases of radioactive iodines in the design LOCA case are given in Reference 3. These may be converted to dose equivalent I 131. In order to compare fairly with the control room dose assessment (because oflong term X/Q and other parameter variations), consider only the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of release (where all control room habitability parameters are constant).

Release, Curies Rem per Curie 0-2 hr 28 Total Dose Total X hr Conversion DCF Factor (Ref 4) 1 131 214 545 759 1.48E6 1.12 E9 l132 257 217 474 5 35E4 2.54E7 1133 469 ,1060 1529 4.00E5 6.12E8 1134 320 62 382 2 50E4 9.56E6 1135 401 684 1085 1.24E5 1.35E8 Sum 1.90E9 The total dose equivalent 1 131 release is then 1.90E9/1.48E6 or 1290 curies. Note that these are ICRP-2 dose conversion factors rather than the ICRP-30 dose conversion factors used for the MSLB dose calculation. This is done to be consistent with the LOCA control room dose, which also uses ICRP-2 dose conversion factors.

l REVISION NO.: 3 l

Eshibit C NEP.12 02 Revision 5 COMMONWEALTII EDISON COMPANY l CALCULATION NO. : 95-011 PROJECT NO. N/A PAGE NO. 7 l Evaluation of Control Roorn Ventilation System Configuration:  !

The proposed control room dose estimation is valid only if the control room cmergency filtration system (VC system) is in operation. 'Ihc VC system is assumed to be in operation for the LOCA assessment (Reference 5). A review of applicable functional diagrams (Reference 6) shows that, in the event of a low steam line pressure, the VC system is activated. This assures that the proposed dose estimation is valid.

Estimate of Control Room Dose:

Reference 7 gives the control room dose calculated for a LOCA. "Ihis calculation is documented in Reference 5.

From intermediate results included in Reference $ (page 16), one can obtain the cumulative dose for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in the LOCA case is 7.21 rem. This would be the resulting dose from the 1,290 curies of done equivalent I.

131 released in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as noted above. A normalized dose for this period can be calculated as:

7.21 rem /1290 Ci = 5.59E-3 rem /Ci 1 131 equivalent Using the above LOCA results, then, one can say that, all other things being equal, the control room dose for the MSLil is in proportional to the radioiodine release and would be:

154Ci x 5.59E 3 rem /Ci = 0.9 rem Since the ICRP.30 dose conversion factors are smaller than the ICRP.2 dose conversion factors, use of the

normalized dose based on ICRP 2 is conservative.

Estimate of Control Dose for Braidwood Unit 1 Cycle 7 :

Per Ref.1 (page 23), the Unit I Cycle 7 allowable leakage is 66.3 gpm (at room temperature conditions) based on the proposed reduced RCS Dosc Equivalent 1 131 limit of 0.1 pCilgm for Unit I Cycle 7. The 66.3 gpm is obtained by dividing the leakage rate calculated at RCS temperature and pressure conditions by 1.406 (Ref. 8) to account for RCS density difTerences.

Per Ref. I (page 23), the predicted Unit 1 Cycle 7 cnd of cyc!c leakage is 62.4 gpm (at room temperature conditions). The control room dose due to Unit I cycle 7 projected leak rate of 62.4 gpm can be determined based on the ratio of unit I cycle 7 leakage to the allowable leakage as follow:

= (62.4 gpm / 66.3 gpm) x 0.9 rem

= 0.85 rem l REVISION NO.: 3 l

4 Exhibit C NEP 12-02 "

Revision 5 COMMONWEALTH EDISON COMPANY l CALCUl.ATION NO. : 95-011 PROJECT NO. N/A PAGE NO. 8 l

SUMMARY

AND CONCLUSION:

Based on the Braidwood maximum primary to secondary leak rate of 9.3 gpm and the primary coolant activity of 1.0 pCi/gm, the 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> control room dose from a Main Steam Line Break (MSLB) event was determined to be 0.90 rem.

Based on the Braidwood Unit 1 Cycle 7 predicted end of cycle primary to secondary leak rate of 62.4 gpm and the prirnary coolant activity of 0.1 pCi/gm, the 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> control room dose from a MSLB cvent was determined to bc 0.85 rem.

Based on these results, it can be concluded that the post accident control room dose due to the MSLB accident is less than that due to the design basis LOCA already included in the UFSAR (Reference 7).

FINAL l REVISION NO.: 3 l I

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