ML20211H321

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Rev 4 to CR Thyroid Dose from MSLB Event
ML20211H321
Person / Time
Site: Byron, Braidwood  Constellation icon.png
Issue date: 09/30/1997
From: Christiana D
COMMONWEALTH EDISON CO.
To:
Shared Package
ML20211H271 List:
References
95-011, 95-011-R04, 95-11, 95-11-R4, NUDOCS 9710060323
Download: ML20211H321 (8)
v • d • e


Text

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u Attachment D -

Braidwood Control Room Calculati >n l:

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. Exhibit C NEP 12-02 Revision 5 CALCULATION TITLE PAGE Page No.1 Calculation No: 95-011

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Comed =L Discipline Code:

A2 M

Braidwood STATION UNIT (S) 1&2 System Code: VC.MS TITLE: Control Room Thyroid Dose from a Main Steam Line Break Event l

X Safety Related __

Augmented Quality Non Safety Related

REFERENCE NUMBERS ,

Type Number Type Number i

COMPONENT EPN: DOCUMENT NUMBERS:

EPN Compt Type Doc Tspe/Sub Tme Document Number

1. Calc / Eng BRW-97-0798-M Rev/2
2. Nureg-0800

! 3. UFSAR Tables 15.6-9,6.4-1 and figure 7.2-1 (Sht. 8)

4. TID USAEC TID-14844

, 5 Calc / Eng BR-VC-02, Rev/2 6 Correspondent Westinghouse letter CAE 97-171 REMARKS:

REVISING APPROVED REY. NO.

ORGANMION PRINT / SIGN DATE

$ 0 Brai&. cod SEC Bruce Acas 3/01/95 1 Braidwood SEC Bruce Acas 3/17/95

2 Braidwood SEC Bruce Acas 4/13/97 3 Braidwood SEC Jeff Bergner jn , fj ,_ 8/29/97 4 Braidwood SEC b , [ C HAis W e [ [ [ 7[Jofil

Exhibit E NEP-12-02 Revision 5 COMMONWEALTil EDISON COMPANY CALCULATION REVISION PAGE CALCULATION NO.95-01I PROJECT NO. N/A PAGE NO.: 2 REVISION SUMMARIES REV: 3 REVISION

SUMMARY

This calculation is being revised to include the following :
1) Update the Calculation Title
2) Recalculate the control room thyroid dose using the activity release data from calculation BRW-97 0798 M, Rev. 0
3) 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 core activity This rnision supersedes all previous revisions of this calculation Electronic Calculation Data Files: None (Program Name, Version File name ext'siwdatshourh mm)

Prepared by: Dan K. Lee 8/29/97 Print / Sign Date l Reviewed by: Gerald P. Lahti __ __8/29/97 Print / Sign Date Type of Resiew I x l Detailed i 1 Alternate i 1 Test DO ANY ASSUMPTIONS IN THIS CALCULATION REQUIRE LATER VERIFICATION [ ] YES [ X l NO Tracked by:

I REV: 4 REVISION

SUMMARY

This calculation is being revised to reflect changes in calculation BRW-97-0798-M, Rev. 3 and to make the control room dose comparison based on eight hour releases Electronic Calculation Data Files: None (Program Name, Version. File name ext size /datehourh mm)

Prepared by: _ Dan K. Lee /fw # a 9/30/97 Print / Sign Date Reviewed by: -Gerald P. Lahti

, , /4]M, , , g - - - 9/30/97 Print / Sign Date Type of Resiew I x 1 Detailed i 1 Alternate i 1 Test DO ANY ASSUMPTIONS IN THIS CALCULATION REQUIRE LATER VERIFICATION { l YES [ x] NO Tracked by:

i Exhibit C NEP 12-02

' Revision 5 l COMMONWEALTH EDISON COMPANY CALCUL,ATION TABLE OF CONTENTS PROJECT NO. N/A CALCULATION NO.95-011 REV. NO. 4 PAGE NO. 3 SECTION PAGE NO. SUB-PAGE NO.

TITLE PAGE I 2

REVISION

SUMMARY

TABLE OF CONTENTS 3

PURPOSE / OBJECTIVE 4 4

METHODOLOGY AND ACCEPTANCE CRITERIA ASSUMPTIONS 4 DE-*"N INPUT '4 4

REFERENCES CALCULATIONS 5-8

SUMMARY

AND CONCLUSIONS ATTACHMENTS N/A

Exhibit E NEP-12-02

.. Revision P COMMONWEALTH EDISON COMPANY

_ l CALCULATION NO. : 95-011 PROJECT NO. N/A PAGE NO. 4 ll PURPOSE / OBJECTIVE:

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

METHODOLOGY AND ACCEPTANCE CRITERIA:

The approach is to demonstrate that the consequences of the MSLB 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 MSLB. The basis for this assumption is discussed in CALCULATIONS section.

l DESIGN INPUT:

The iodine releases and the maximum acceptable steam generator leak rates are from calculation BRW-97-0798-M, Rev 3 (Reference 1). The releases from the containment and the resulting control l room dose are from the UFSAR (References 3 and 7).

REFERENCES:

1. Comed Calculation BRW-97-0798-M, Rev. 3, " Allowable Leakrate Calculation for Steam Generator Interim Plugging Criteria," September 30,1997
2. USNRC, Standard Review Plan, NUREG-0800, Section 15.6.3, " Radiological Consequences of Steam Generator Tube Failure (PWR),"Section II, Acceptance Criteria, Rev. 2, July 1981.
3. B/B UFSAR, Table 15.6-9
4. USAEC Technical Information Document TID-14844, Calculation ofDistance Factors for Power and Test Reactor Sites, March 1962.

2 i

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

May 9,1988.

6. B/B UFSAR, Figure 7.2-1 (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 Byron and Braidwood Alternate Tube Plugging Limit.

l REVISION NO.: 4 l

Exhibit C NEP-12 02 Revision 5 COMMONWEALTH EDISON COMPANY l CALCULATION NO. : 95-011 PROJECT NO. N/A PAGE NO. 5 ll

9. Federal Guidance Report i1, " Limiting Values of Radionuclides Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988.

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 I considers three release sources:
a. Primary coolant with a pre-accident iodine spike
b. Primary coolant with a post-accident iodine spike
c. Secondary coolant with iodine concentrations at a predetermined linut Reference 1, then, evaluates these to obtain a limiting primary to secondary leakrate as follows. For a one gallon per minute primary to secondary leak, and a RCS I-131 activity of I Ci/gm we have:

Dose Equivalent Exclusion Area Source I-131 Release Boundary Dose a 15.9 Ci 4.60 rem b 10.6 Ci 3.05 rem c 5.20 Ci 1.50 rem l 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% of 10CFR100 limit l REVISION NO.: 4 l

- Exhibit C NEP 12-02 Revision 5 COMMONWEALTII EDISON COMPANY RALCULATION NO. : 95-011 PROJECT NO. N/A PAGE NO. 6 ll Using this criteria, Reference I calculates a limiting primary to secondary leak rate of 9.3 gallons per minute (Scenario b+c).

For the period 2-8 hours after the accident, the weighted activity released (IRi DCFi) due to a 9.3 gallons per minute primary to secondary leak rate can be calculated from the information from Tables 2.f.2 and 3.f.2 in Reference I as indicated below.

Source Weighted Activity Released I-131 Dose Conversion Dose Equivalent IR iDCFi (rem) Factor (rem /Ci) I 131 Released (Reference 1) (Reference 9) (Ci) a 7.17E+8 1.08E+6 664 b 8.52E+8 1.08E+6 789 l

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

Source Dose Equivalent I-131 Release a 9.3 x 15.9 Ci + 664 Ci = 812 Ci b 9.3 x 10.6 Ci + 789 Ci = 888 Ci c 5.2 Ci Under these conditions, the worst case combination of(a+c) or (b+c) is (a+c), which results in a release of 893 curies of dose equivalent I-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 ccntrol room habitability parameters are constuit).

4

- l REVISION NO.: 4 l 4

{'. -

Exhibit C L

. NEP-12 02 Revision 5 LCOMMONWEALTH EDISON COMPANY i

l CALCULATION NO. : 95-011 - PROJECT NO. N/A PAGE NO. 7 ll_ j

Release,' Curies - Rem per Curie 0-2 hr 2-8
Total - Dose Total X - i' hr Conversion DCF

' Factor (Ref 4)

I-131 214 545= 759 148E6 1.12 E9  :

I-132 257 217 474 5.35E4 2.54E7 1-133 469 1060 1529 4.00E5 6.12E8

I-134 320 62 382 2.50E4 9.56E6 I-135 -401- 684 1085 1.24E5 1.35E8 Sum 1.90E9- -

4 The total dose equivalent I-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 4

uses ICRP-2 dose conversion factors.

Evaluation of Control Room Ventilation System Configuration:

He proposed control room dose estimation is valid only if the control room'c:ncrgency filtration system (VC .

i system) is in operationJ The VC system is assumed to be in operation for the LOCA assessment (Reference 5). A l review of applicable fhac*ianal diagrams (Reference 6) shows that, in the event of a low steam line pressure, the

VC system is activated. His assures that the proposed dose estimation is valid.

Estimate of Control Room Dose:

Reference 7 gives the control room dose calculated for a LOCA. The total 30 day dose is 16.40 rem thyroid. This l calet ation is documented in Reference 5. From intermediate results included in Reference 5 (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 dose 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 -

- 1 period can be calculated ast 7.21 rem /1290 Ci = 5.59E-3 rem /Ci l Using the above LOCA results, then, one can say that, all other things being equal, the control room dose for the -

U MSLB is in proportional to the radioiodine release and would be:

893 Ci x 5.59E-3 rem /Ci = 4.99 rem l 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.

l REVISION NO.: 4 1 4

Exhibit C NEP-12-02

' Revision 5 COMMONWEALTII EDISON COMPANY l CALCULATION NO. : 95-011 PROJECT NO. N/A PAGE NO. 8 ll Estimate of Control Dose for Braidwood Unit I Cycle 7 :

Per Ref.1 (page 23), the Unit I Cycle 7 allowable leakage is 66.4 gpm (at room temperature conditions) based on the proposed reduced RCS Dose Equivalent 1 131 limit of 0.1 pCi/gm for Unit I Cycle 7. The 66.4 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 differences.

Per Ref. I (page 23), the predicted Unit I Cycle 7 end of cycle 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:

l

= (62.4 gpm / 66.4 gpm) x 4.99 rem l

= 4.69 rem l

SUhthtARY AND CONCLUSION:

Based on the Braidwood maximum primary to secondary leak rate of 9.34 gpm and the primary coolant actisity of 1.0 pCilgm, the 0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> control room dose from a Main Steam Line Break (htSLB) event was determined to be 4.99 rem.

Based on the Braidwood Unit 1 Cycle 7 predicted end of cycle primary to secondary leak rate of 62.4 gpm and the primary coolant activity of 0.1 pCi/gm, the 0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> control room dose from a htSLB cvent was determined to be 4.69 rem.

Based on these results, it can be concluded that the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> post-accident control room dose due to the MSLB accident is less than the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 7.21 rem due to the design basis LOCA.

FINAL l REVISION NO.: 4  !

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