ML20207B648

SER Accepting Revised SGTR Analysis for Byron & Braidwood Stations.Revised Analysis Was Submitted to Support SG Replacement at Unit 1 of Each Station

ML20207B648
Person / Time
Site:	Byron, Braidwood
Issue date:	05/25/1999
From:	NRC (Affiliation Not Assigned)
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ML20207B640	List: ML20207B636 ML20207B648
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NUDOCS 9906020052
Download: ML20207B648 (8)

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066-0001 4

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REVISED STEAM GENERATOR TUBE RUPTURE ANALYSIS COMMONWEALTH EDISON COMPANY BYRON STATION. UNIT 2. AND BRAIDWOOD STATION. UNIT 2 DOCKET NOS. STN 50-455 AND STN 50-457

1.0 INTRODUCTION

By letter dated November 13,1996, Commonwealth Edison Company (Comed, or the licensee) submitted a revised analysis for Steam Generator Tube Rupture (SGTR) in support of steam generator (SG) replacement at Byron Station, Unit 1, and Braidwood Station, Unit 1. Physical differences between the original steam generators (OSGs), which are Westinghouse D-4 models, and the replacement steam generators (RSGs), which are made by Babcock & Wilcox International, affect the plant response to a SGTR and, in particular, reduce the margin to overfill. As a result of the reduced margin to overfill, Comed revised certain operator actions to allow the operators to isolate the ruptured SG earlier if a tube rupture is suspected. Comed '

revised the SGTR analysis, using the revised operator actions, to verify that a margin to overfill exists, and the offsite dose does not exceed a small fraction of 10 CFR Part 100 limits or the acceptance criteria of Standard Review Plan (SRP) 15.6.3. Because the revised operator actions are the same regardless of the type of SG, Comed re-analyzed the SGTR for all four units. Comed provided additional!aformation by letters dated March 20, June 24, August 19, I November 3, November 26, and December 19,1997.

The staff documented its review of Comed's SGh an:'yses by letter dated January 28,1998, as supplemented by letter dated March 11,1998. In those letters, the staff concluded that 1) the operator action times during a SGTR are satisfactory,2) the analysis performed in calculating SG margin to overfill used an approved methodology with appropriately conservative j inputs and assumptions,3) there is an acceptable margin to SG overfill, and 4) the use of the above methodology for input into the dose consequence analysis is acceptable. Those conclusions are applicable to Units 1 and 2 of both stations. Also documented in those letters is the staff's confirmatory dose analysis for Unit 1, which determined that 1) the doses would not exceed the dose guidelines presently contained in the SRP for the SGTR accident,2; in no case would the offsite doses exceed the specific fraction of 10 CFR Part 100 ilmits, and 3) the doses to operators in the control room would not exceed the limits in General Design Criterion (GDC) 19 of 10 CFR Part 50, Appendix A.

However, the licensee's November 3,1997, submittal only provided information related to the dose analyses for Unit 1, and the staff's letter dated January 28,1998, as supplemented by letter dated March 11,1998, only performed a confirmatory dose analysis for Unit 1 at each station. To support completion of the staff's review of the Unit 2 dose analysis, the licensee ,

provided additional information by letters dated April 13 and June 10,1998, and April 20,1999. 1 Doh $8g "

ENCLOSURE.

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2.0 EVALUATION 2.1 Licensee Assessment j

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The licensee evaluated the consequences of a postulated SGTR accident. Two cases were l analyzed. The first case assumed an iodine spike occurred prior to the SGTR. This is referred l to as the pre-existing spike case. During the SGTR, primary to secondary leakage was assumed to be occurring at the technical opecification rate of 150 gallons per day (gpd) for each SG. In addition, primary to secondary leakage was occurring through the ruptured tube.

For the pre-existing spike case, the reactor coolant system (RCS) iodine specific activity was i assumed to be at the Technical Specifications Figure 3.4.16-148-hour, full power limit of 60 l microcurie / gram (pCl/gm) of dose equivalent iodine-131 ('8'l). The secondary coolant iodine i specific activity was assumed to be at the Technical Specification normal operation secondary coolant specific activity limit of 0.1 pCi/gm.

l The second caso, referred to as the accident-initiated spike case, assumed the SGTR event itself initiated an iodine spike concurrent with the accident. Immediately prior to the accident, the RCS activity level was assumed to be at the Technical Specifications long-term RCS activity limit of 1 pCi/gm of dose equivalent '8'I. The secondary system activity was assumed to be at the Technical Specifications normal operation limit of 0.1 pCl/gm dose equivalent '8'l. The SGTR was assumed to initiate an iodine spike which results in a release of iodine from the fuel gap to the reactor coolant at a rate which is 500 times the normal iodine release rate necessary to maintain the reactor coolant activity level at 1 pCligm of dose equivalent '8'l.

Comed's submittal indicated that a SGTR accident did not result in any melted fuel nor any additional releat,e of fuel gap inventory to the reactor coolant. For both cases, it was assumed that a primary to secondary leak occurred hi the intact SGs at a rate of 150 gpd per SG for the duration of the accident. In both cases, it was assumed that offsite power was lost and the main condenser was unavailable for the r%am dump. The licensee's analysis did not calculate the consequences to the control room Oprator because the licensee indicated that the results of the loss-of-coolant accident (LOCA) were bounding.

2.2 Staff Assessment The staff performed an assessment of the licensee's analyses. From Figure 15 of NFSR-0114, Revision 0, of the licensee's SGTR analysis, the staff assumed that break flow continued until approximately 3763 seconds after the tube ruptures with some momentary perturbations out to 4200 seconds. These perturbations arise from RCS pressure fluctuations. However, no steam release would occLr from the faulted SG after 1715 seconds follo,ving the SGTR. This was noted in Figure 16 of this report. Figure 16 shows that the release of steam occurred from the faulted SG from approximately 500 seconds following the tube rupture until 1700+ seconds following the rupture. As noted above, both the Comed report and the April 20,1999, letter, indicated that there would be no further release of steam from the faulted SG during the remainder of the accident even though break flow continues past 1715 seconds.

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~ The licensee's original analysis for the SGTR provided neither the cooldown late of the core I

using the intact SGs nor the duration of the cooldown nor the radiological dosa contribution from the steam releases associated with the cooldown. Previous licensee analyses had not included the dose contribution from the cooldown releases. The licensee indicated that such contribution had been ignored because the releases associated with the cooling using the intact SGs provided a small contribution relative to the total dose. However, this calcu!ation methodology was inconsistent with the Westinghouse methodology presented in WCAP 10698-P-A, Supplement 1. In the April 20,1999, letter the licensee provided cooldown rates, cooldown duration and a commitment to incorporate the radiological dose contribution of the cooldown releases associated with the intact SGs the next time Comed performs the SGTR analysis. The staff's assessment of the consequences of a SGTR incorporated the cooldown rates and duration information from the licensee's April 20,1999, letter.

For the purpose of this astisssment, the staff did not calculate the'whole body dose associated with ths release of noble gasen, because the thyroid dose is limiting with respect to compliance with GDC 19 and 10 CFR Part 100 limns. n vfdition, the whole body doses were expected to be less than the staff's previously calculated doses iuc a rod ejection accident which involves fuel failures.

Table 1 presents the assumptions utilized by the staff in their assessment of the Byron, Unit 2, and Braidwood, Unit 2, SGTR. The potential dose consequences of a SGTR accident at Byron and Braidwood are presented in Table 2. The staff's calculations confirmed the licensee's i conclusions that both the on-site doses (GDC 19) and the off-site doses (10 CFR Part 100) )

were found to be acceptable.

l 3.0 COMMITMENTS The licensee made two commitments related to the SGTR analysis. By letter dated June 10, 1998, the licensee committed to modify the remaining auxiliary feedwater (AFW) flow control c!ves to ensure AFW flow is limited to 464 gallons per minute (gpm) consistent with the revied SGTR analysis;. By letter dated April 20,1999, the licensee committed to include in the SGTR offsite dose calculation, the next time the analysis is performed, the dose contribution from steam r# eased from the intact SGs during the 2 to 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> RCS cooldown period. The NRC staff finds that reasonable controls for the implementation and for subsequent evaluation ci peporei changes pertaining to the above regulatory commitment (s) are best provided by the licensee's administrative processes, including its commitment management program. The abtve regulatory commitments do not warrant the creation of regulatory requirements (items requiring prior NRC approval of subsequent changes). The staff notes that pending industry and regulatory guidance pertaining to 10 CFR 50.71(e) may call for some information related to ,

the above commitments to be included in a future update of the facility's final safety analysis j report.

4.0 CONCLUSION

S The staff has confirmed that the licensees' calculations associated with a SGTR accident for Byron, Unit 2, and Braidwood, Unit 2, showed that the potential consequences would not result in doses which would exceed the dose guidelines presently contained in SRP Section 15.6.3 for

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the SGTR, and that in no case wou!d the offsite dose exceed the specific fraction of 10 CFR Part 100 limits, nor would the doses to the operators in the control room exceed the limits in j GDC 19 of Appendix A to 10 CFR Part 50. The staff concludes that the proposed changes in 1 operator actions to mitigate the consequences of a SGTR are acceptable.

Table 1_- Assumptions for Byron /Braidwood SGTR Table 2 - Byron /Braidwood Thyroid Doses from SGTR Accident (Rem)

Principal Contributor: J. Hayes Date: May 25, 1999 l

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TABLE 1 ASSUMPTIONS FOR BYRON /BRAIDWOOD STEAM GENERATOR TUBE RUPTURE (SGTR) lodine Partition Factor 0.01 l l

Steam Release from Defective Steam i Generator (SG) 1 0-2 hours (Ibs) 9.19E4 i

> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (Ibs) 0 l Steam Release from Intact SGs (lbs) 0-1.05 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 1.51 E5 1.05-2 hours 3.64E5 2-8 hours 1.18E6 Estimated Break Flow to Faulted SG 9.56E4 (Ibs)

Primary to Secondary Leak Rate 150 (gpd/SG) {

Time to Isolate Faulted SG (sec) 1716 Flashing Fraction Variable with respect to time. Provided in Comed letter dated 4/13/98. I Scrubbing Fraction 0 Primary Bypass Fraction for intact 0 -

SGs Duration of Plant Cooldown (hrs) 8 Primary coolant concentration of 60 pCi/g of dose equivalent '8'l.

l Pre-existino Soike Value (uCi/o) l

'8'l = 46.2

'821 = 51.7

'881 = 73.9

• l = 11.1 ,

'851 = 40.6 l

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i' ' TABLE 1 ASSUMPTIONS FOR BYRON /BRAIDWOOD STEAM GENERATOR TUBE RUPTURE (SGTR) (Continued)

Volume and Mass of primary coolant and secondary coolant.

I Primary Coolant Volume (ft*) 10,086 @567 'F Primary Coolant Temperature ('F) 567 Mass of Primary Coolant (Ibs) 477,740

Primary Coolant Pressure (psla) 2,293 Pressurizer Temperature ('F) 657 Pressurizer Pressure (psia) 2,293 8

Pressurizer Volume (ft ) 1,150 Secondary Coolant Steam Mass /SG (Ibs) 6,969 Secondary Coolant Liquid Mass /SG(Ibs) 79,836 Secondary Coolant Steam Temperature (*F) 509 Secondary Coolant Feedwater Temperature ('F) 440 l

Technical Specification Limits for DE *l in the primary and secondary coolant.

Primary Coolant DE I concentration (pCi/g)

Maximum Instantaneous Value 60 48 Hour Value 1.0 Secondary Coolant DE "'I concentration (pCl/g) 0.1 Technical Specification Limits for the primary to secondary leak rate.

Primary to secondary leak rate, any SG (gpd) 150 Primary to secondary leak rate, total (gpd) 600 Maximum primary to secondary leak rate to the faulted and intact SGs.

Faulted SG (gpm) 150

' intact SGs (gpm/SG) 150 Letdown Flow Rate (gpm) 75 Equilibrium Release Rate from Fuel for a Spiking Factor of 500 times the Release Rate for 1 pCi/g of Dose Equivalent " l Ci/ day

• l = 2,040 n2 1= 5,300 "81 = 5,330

• l = 7,370

"'I = 5,300 4

• o TABLE 1 ASSUMPTIONS FOR BYRON /BRAIDWOOD STEAM GENERATOR TUBE RUPTURE (SGTR) (Continu9d)

Control Room Free Volume (ft') 4.05ES Filtered Recirculation Flow 4.45E4 I (cfm)

Recirculation Efficiency for 90 j all forms of lodine (%) l Makeup Filter Efficiency for 99 .

all forms of lodine (%)

Makeup Air Filtration Rate 5400 (cfm)

Unfiltered Air Infiltration Rate 89 (cfm)

Occupancy Factors 0-1 day 1.0 1-4 days 0.6 Atmospheric Dispersion Factors 3

(sec/m ) Byron Braidwood Control Room 0-8 hours 4.05E-3 6.2E-3 8-24 hours 1.9E-3 3.2E-3 1-4 day 5.7E-4 8.4E-4 4-30 days 3.8E-4 1.4E-4 l l

Atmospheric Dispersion Factors (sec/m8) Byron Braidwood EAB 6.8E-4 7.7E-4 LPZ l 0-8 hours 2.3E-5 7.9E-5 i 8-24 hours 1.5E-5 5.2E-5  :

1-4 days 6.4E-6 2.1 E-5 4-30 days 1.4E-6 5.6E-6 l

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TABLE 2 BYRON /BRAIDWOOD THYROID DOSES FROM SGTR ACCIDENT (REM)

Accident Byron Braidwood  !

I Coincident Spike I EAB 4.4 5.0 LPZ 0.16 0.55  ;

Control Room 0.102 0.078 {

Pre-existing Spike EAB 32 36 LPZ 1.1 3.8 l Control Room 0.70 0.53 l

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