ML20217D942

Forwards Info Which Completes SGTR Dose Evaluations for Byron & Braidwood Stations,In Response to NRC Rai.Dose Evaluations for Main Steamline Break,Rod Ejection & Locked Rotor Event Do Not Credit Operator Response

ML20217D942
Person / Time
Site:	Byron, Braidwood
Issue date:	04/13/1998
From:	Helwig D COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9804270144
Download: ML20217D942 (8)

Text

l Commonw ealth Ediv>n Company

• l 100 0 pus Ptate Downers Gnn c, II. 60515-5701 N

l April 13,1998 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555

Subject:

Revised Steam Generator Tube Rupture Analysis Dose Evaluations for Byron and Braidwood Stations Unit 2 Byron Nuclear Power Station Facility O )erating License NPF-37 and NPF-66 NRC Doc cet Numbers: 50-454 and 50-455 Braidwood Nuclear Power Station Facility Operating License NPF-72 and NPF-77 NRC Docket Numbers: 50-456 and 50-457

References:

1.

USNRC Letter to Comed, dated October 3,1997, " Request for Additional Information Regarding the Revised Steam Generator Tube Rupture Analysis; Removal of Steam Generator Repair Methodologies; and Restoration of Previous Dose Equivalent Iodine Limits - Byron and Braidwood Stations 2.

J. Hosmer Letter to USNRC, " Response to Request for Additional Information Regarding the Revised Steam Generator Tube Rupture Analysis", dated November 3,1997 3.

USNRC Letter to Comed, " Revised Steam Generator Tube Rupture Analysis - ByTon Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, dated January 28,1998 4.

USNRC Letter to Comed, " Revised Steam Generator Tube Rupture Analysis - ByTon Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, dated March 11,1998 I

Reference 1 requested information regarding dose evaluations for the replacement steam generator project. Reference 2 contained the response to the request for additional l

mformation (RAI). References 3 and 4 approved the revised steam generator tube l

(SGTR) analysis for Byron and Braidwood Stations, Unit 1. This letter provides

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U.S. Nuclear Regulatory Commission April 13,1998 information in order to complete the SGTR dose evaluations for Byron and Braidwood Stations Unit 2.

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l Please direct any questions to this omce.

Sincerely, 1

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l David R. Helwig Senior Vice Presiden Nuclear Generation Support Attachment k

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Regional Administrator, Rill Byron Project Manager, NRR Braidwood Project Manager, NRR Senior Resident Inspector, Byron Station l

Senior Resident inspector, Braidwood Station l

Omce of Nuclear Safety, IDNS I

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Response to RAI Regarding SGTR Dose Evaluations for Byron /Braidwood Unit 2 The dose evaluations for the main steamline break, rod ejection and locked rotor events do not credit operator response. These evaluations remam unchanged for Unit 2 with the Westinghouse original steam generators. Since operator response times are applicable for the Steam Generator Tube Rupture (SGTR) event only, only the sections related to this event are included as part of the response.

1.

For the SGTR accident, provide the following information:

a. Mass ofliquid and steam released from the faulted SG as a function of time.

l As a minimum, releases should be designated as those within two hours and

{

those after two hours.

j There is no liquid release from the ruptured SG.

Steam Release Faulted SG (Ibs) 0-2 hours 9.19E4

> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

• 0.0

• The B/B SGTR methodology approved by the NRC (Ref.1) does not calculate release during plant cooldown.

b. Mass of steam released from the intact SG as a function of time. As a l

minimum, releases should be designated as those within two hours and those l

after two hours.

Total releases from three intact SGs (Ibs) 0-2 hours 1.51E5

> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

• 0.0

• The B/B SGTR methodology approved by the NRC (Ref.1) does not calculate release during plant cooldown.

c.

Flashing fraction in the intact and faulted SG.

See Table 1 for the ruptured SG. The primary coolant that flows through the i

ruptured tube flashes to steam and is conservatively assumed to carry with it, the respective fraction of break flow nuclide activity concentration into the SG steam space.

For the intact SG, it is assumed that the leaked primary coolant is completely mixed in the secondary liquid with no flashing occurring.

d. Scrubbing fraction in the intact and faulted SG.

0.0 t

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e.

Primary bypass fraction for the intact and faulted SG.

0.0 f.

Time to isolate faulted SG.

1715 seconds.

The B/B SGTR methodology approved by the NRC (Ref.1) does not calculate release during plant cooldown, including plant cooldown that occurs while the i

i ruptured S/G flow is being isolated.

g. Duration of plant cooldown by the secondary side.

The B/B SGTR methodology approved by the NRC (Ref.1) does not calculate release during plant cooldown to RHR cut in. Therefore, this time is not provided.

h. Primary to secondary release rate from the ruptured tubc as a function of time.

See Table 2.

i. Indicate if overfill conditions do or do not exist. If they do exist, appropriate mass release data should be provided as a function of time for the faulted SG.

The ruptured SG does not overfill.

j. Additionalinformation which should be provided is contained in Attachment 2.

See responses to Attachments 2 and 3.

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Table 1 - Flashing Fraction for the Ruptured SG Time (sec)

Flashing Fraction (lb flashed /lb break flow)

O 0.00E+00 45 7.29E-02 105 7.23E-02 145 7.25E-02 185 7.29E-02 225 7.34E-02 265 7.30E-02 305 7.29E-02 l

345 7.36E-02 385 7.39E-02 425 7.40E-02 465 7.37E-02 500 8.93E-03 540 1.49E-02 580 2.28E-02 620 2.96E-02 660 3.26E-02 700 3.43 E-02 740 3.53E-02 800 3.62E-02 820 3.62E-02 860 3.64E-02 900 3.63 E-02 1

940 3.62E-02 980 3.61 E-02 1020 3.58E-02 1060 3.56E-02 1

1100 3.53 E-02 1140 3.50E-02 l

1180 3.47E-02 1220 3.45E-02 1260 3.41 E-02 j

1300 3.38E-02 1340 3.34E-02 1380 3.30E-02 1420 3.26E-02 1460 3.22E-02 1500

3. I 7E-02 1540

3. I 2E-02 1580 3.07E-02 1620 3.02E-02 1660 2.98E-02 1700 2.94E-02 1716 2.92E-02 i

i Table 2 - Ruptured Tube Flow Time (sec)

Break Flow (Ib/sec) 0 1.00E-06 45 6.17E+01 105 6.03 E+01 145 5.95E+01 185 5.87E+01 225 5.79E+01 i

265 5.70E+01 305 5.61E+01 345 5.54E+01 385 5.46E+01 l

425 5.38E+01 l

465 5.28E+01 l

500 4.29E+01 540 4.73 E+01 i

580 5.08E+01 620 5.29E+01 660 5.40E+01 700 5.48E+01 740 5.53E+01 l.

800 5.57E+01 820 5.57E+01 l

860 5.57E+01 l

900 5.57E+01 l

940 5.57E+01 980 5.57E+01 1020 5.57E+01 1060 5.57E+01 1100 5.58E+01 1140 5.58E+01 1180 5.58E+01 1220 5.59E+01 1260 5.60E+01 1300 5.61E+01 l

1340 5.62E+01 l

1380 5.63E+01 1420 5.64E+01 l

1460 5.66E+01 1500 5.67E+01 i

1540 5.68E+01 1580 5.70E+01 1620 5.71E+01 i

1660 5.72E+01 1700 5.74E+01

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1716 5.74 E+01

Re'erence:

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1. " Byron, Units I and 2, and Braidwood, Units 1 and 2 - Steam Generator Tube Rupture Analysis (Tac Nos. M57080, M63247, M64026 and M64053)," letter from R. M. Pulsifer to T.J. Kovach, April 23,1992.

ATTACHMENT 2 INPUT PARAMETERS FOR EVALUATION OF SGTR BYRON AND BRAIDWOOD STATIONS UNIT 2 I3I

1. Primary coolant concentration of TS value for dose equivalent I.

Pre-existing Spike Value (uCi/g) 131 1 38.7 132 = 43.4 1=

1 = 61.9 4

1 9.3 135 = 34.0 1=

2. Volume of primary coolant and secondary coolant.

3 Primary Coolant Volume (fl )

10086, excluding pressurizer Primary Coolant Temperature ( F) 567 3

Secondary Coolant Steam Volume Total (fl )

17227*

Secondary Coolant Mass Total (Ibs) 347220*

Primary Coolant Pressure (psia) 2293 Primary Coolant Massfibs) 477740 Pressurizer Volume (fl )

1150**

Pressurizer Temperature ( F) 657 Pressurizer Pressure (psia) 2293 Secondary Coolant Liquid Mass /SG (Ibs) 79836*

Secondary Coolant Steam Mass /SG (Ibs) 6969*

Secondary Steam Temperature ( F) 509 Secondary Liquid Temperature ( F) 509

• Based on nominal-5% narrow range level

• • Based on 65% level I3I

3. TS limits for DE l in the primary and secondary coolants:

Maximum insgntaneous in primary coolant (pCi/g) 60.0 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> DE I in primary coolant (pCi/g) 1.0 Secondary Coolant (pCilg) 0.1

4. TS value for the primary to secondary leak rate (include reference temperature and pressure).

Any SG (gpd) 150.0 Total all SG (gpd) 600.0 Reference temperature ( F) 70.0 Reference pressure (psia) 14.7

5. Primary coolant activity (Ci) due to a pre-existing spike:

131 1 = 8.4E3 1321=

133 = 9.4E3 1

1.3E4 1341=

135 = 2.0E3 1

7.4E3 I

An inconsistency in the Byron /Braidwood UFSAR regarding the activity source term for the primary and secondary coolant has recently been identified. The values presented here correspond to the correct concentrations (Attachment 2, item 1). The incorrect activities used in the dose calculation have been evaluated to show negligible impact on dose releases.

6. Primary coolant activity levels (pCi/g) for accident initiated spike.

131 = 2.5 1

! = 2.8 I = 4.0 134 1

135 = 0.6 1 = 2.2

7. Primary coolant concentrgon at maximum instantaneous value of 60 pCi/g dose equivalent I.

131 1 = 38.7 132 1

133 = 43.4 1 = 61.9 1341= 9.3 135 = 34.0 1

8. Primary Coolant Activity (Ci) for Accident Initiated Spike.

131 = 5.42E2 1

132 133 = 6.07E2 1=

1 8.67E2 134I= 1.30E2 1351 = 4.77E2

9. Iodine Partition Factor Faulted SG 0.01 Intact SG 0.01 Condenser not credited

10. Steam Released to the environment as a function of time:

Faulted SG (Ibs) 0-2 hours 9.19E4

> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

• 0.0 Total releases from three intact SG (lbs) 0-2 hours 1.51E5

> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

• 0.0

• The B/B SGTR methodology approved by the NRC (Ref.1) does not calculate release during plant cooldown.

I1. L,etdown Flow Rate (gpm) 0.0.

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12. Atmospheric Dispersion Factors (sec/m ):

Byron Braidwood

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EAB (0-2 hours) 5.7E-4 7.7E-4 i

LPZ (0-8 hours) 1.7E-5 7.1E-5 Control Room (0-8 hours) 4.05E-3 6.24E-3 t

13. Control Room:

Byron Drpidwood r

Emergency Makeup Flow (cfm) 6.0E3 6.0E3 Makeup Filter efficiency (%)

99 99 l

Unfiltered Inleakage (cfm) 78.75 25.0 Recirculation Filter Flow Rate (cfm) 4.5E4 4.5E4 Recirculation Filter Efficiency (%)

90 90 Occupancy Factor (0-1 day) 1.0 1.0

14. For the Accident Initiated Spike Case gease Rate (Ci/hr)

LOOX Release Rate (Ci/hr)*

1 = 41. I8 2.06E4 1321 = 62.71 3.14E4 l

133 I = 92.16 4.61E4 134I=

135 = 108.0 5.40E4 I

83.52 4.18E4 l

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• The B/B SGTR methodology approved by the NRC (Ref.1) calculates the l

release rates based on escape coefTicients. The resultant release rates are different l

than from those used in the main steamline break dose calculations.

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15. Flashing Fraction, Primary Bypass and Scrubbing Fraction as a function of j

time.

l Flashing fraction for the ruptured SG is in Table 1. The primary coolant that flows through the ruptured tube flashes to steam and is conservatively assumed to carry with it, the respective fraction of break flow nuclide activity concentration into the SG steam space.

1 For the intact SG, it is assumed that the leaked primary coolant is completely mixed in the secondary liquid with no flashing occurring.

Primary bypass is 0.0 throughout the transient.

l Scrubbing fraction is 0.0 throughout the transient.

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16. Mass release rate through the ruptured tube as a function of time.

See Table 2.

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ATTACHMENT 3 STEAM GENERATOR TUBE RUPTURE THYROID DOSE ASSESSMENT (BYRON UNIT 2)

Case Involving Pre-existing Spike EAB LPZ Control Room Calculated doses (rem) 19.72**

0.59**

not analyzed

• Regulatory Guidelines (rem) 300 300 30 Case involving Accident Initiated Spike EAB LPZ Control Room Calculated doses (rem) 18.46**

0.55**

not analyzed

• Regulatory Guidelines (rem) 30 30 30

• The LOCA control room dose case bounds the steam generator tube rupture control room case.

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• • An inconsistency in the Byron /Braidwood UFSAR regarding the activity source term for the primary and secondary coolant has recently been identified. The incorrect activities used in the dose calculation have been evaluated to show negligible impact on dose releases.

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STEAM GENERATOR TUBE RUPTURE THYROID DOSE ASSESSMENT (BRAIDWOOD UNIT 2)

Case involving Pre-existing Spike E@

l PZ Control Room Calculated doses (rem) 26.64**

2.46*

• not analyzed

• Regulatory Guidelines (rem) 300 300 30 CazJnvolving Accident initiated Spike EAB LPZ Control Room Calculated doses (rem) 24.93**

2.30*

• not analyzed

• Regulatory Guidelines (rem) 30 30 30

• The LOCA control room dose case bounds the steam generator tube rupture control room case.

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" An inconsistency in the Byron /Braidwood UFSAR regarding the activity source term for the primary and secondary coolant has recently been identified. The incorrect activities used in the dose calculation have been evaluated to show negligible impact on dose releases.

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