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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20217G9961999-10-14014 October 1999 SER Accepting First 10-year Interval Inservice Insp Requests for Relief for Plant,Units 1 & ML20212A7441999-09-10010 September 1999 Safety Evaluation Concluding That Alternatives Contained in Relief Request 12R-07 Provide Acceptable Level of Quality & Safety ML20210U8111999-08-0404 August 1999 SER Granting Licensee Relief Requests VR-1,VR-3 & Portion of VR-2 Pursuant to 10CFR50.55a(a)(3)(ii).Relief Request VR-4 Does Not Require Explicit NRC Approval Because Relief Applies to Valves Not Required by 10CFR50.55a ML20210K9861999-07-30030 July 1999 Safety Evaluation Accepting Licensee 60-day Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs ML20207B6481999-05-25025 May 1999 SER Accepting Revised SGTR Analysis for Byron & Braidwood Stations.Revised Analysis Was Submitted to Support SG Replacement at Unit 1 of Each Station ML20205B5091999-03-26026 March 1999 SER Accepting Relief Requests 12R-24,Rev 0 & 12R-34,Rev 0, Related to Second 10-year Interval Inservice Insp for Byron Station,Units 1 & 2 ML20205C5101999-03-21021 March 1999 Revised Safety Evaluation Supporting Improved TS Amends Issued by NRC on 981222 to FOLs NPF-37,NPF-66,NPF-72 & NPF-77.Revised Pages Include Editorial Corrections ML20204G3831999-03-19019 March 1999 Safety Evaluation Accepting Second 10-yr Interval ISI Request for Relief 12R-11 ML20155B6711998-10-26026 October 1998 Safety Evaluation Accepting Requests for Relief Associated with Second 10-yr Interval ISI Program Plan ML20154D4401998-10-0202 October 1998 Safety Evaluation Authorizing Second 10-yr Interval ISI Program Request for Relief 12R-30 for Plant,Units 1 & 2 ML20238F3281998-08-31031 August 1998 SER Approving Second 10-year Interval Inservice Insp Program Request for Relief 12R-14 for Braidwood Station,Units 1 & 2 ML20238F6551998-08-28028 August 1998 SE Authorizing Licensee Request for Relief NR-20,Rev 1 & NR-25,Rev 0 Re Relief from Examination Requirement of Applicable ASME BPV Code,Section XI for First ISI Interval Exams ML20217K7171998-04-20020 April 1998 Safety Evaluation Accepting Requests for Relief NR-22,NR-23 & NR-24 for First 10-yr Insp Interval ML20217K6331998-04-20020 April 1998 Safety Evaluation Accepting Methodology & Criteria Used in Generating Flaw Evaluation Charts for RPV of Braidwood IAW Section XI of ASME Code ML20216F4921998-03-11011 March 1998 Correction to Safety Evaluation Re Revised SG Tube Rapture Analysis ML20212H1851998-03-0606 March 1998 SE Approving Temporary Use of Current Procedure for Containment Repair & Replacement Activities Instead of Requirements in Amended 10CFR50.55a Rule ML20197B7531998-03-0404 March 1998 SER Accepting License Request for Relief from Immediate Implementation of Amended Requirements of 10CFR50.55a for Plant,Units 1 & 2 ML20199G2591998-01-28028 January 1998 Safety Evaluation Rept Accepting Revised SG Tube Rupture Analysis ML20199H0031998-01-21021 January 1998 SER Accepting Pressure Temp Limits Rept & Methodology for Relocation of Reactor Coolant Sys pressure-temp Limit Curves & Low Temp Overpressure Protection Sys Limits for Byron Station,Units 1 & 2 & Braidwood Station,Units 1 & 2 ML20199C1401998-01-16016 January 1998 SER Accepting Request to Integrate Reactor Vessel Weld Metal Surveillance Program for Byron,Units 1 & 2 & Braidwood,Units 1 & 2 Per 10CFR50 ML20199C1231998-01-13013 January 1998 Safety Evaluation Granting Second 10-yr Inservice Insp Program Plan Relief Request ML20197G0041997-12-11011 December 1997 Safety Evaluation Accepting First 10-yr Interval Insp Program Plan,Rev 4 & Associated Requests for Relief for Plant ML20198H3211997-12-0303 December 1997 Safety Evaluation Re Licensee Submittal of IPE for Plant, Units 1 & 2,in Response to GL 88-02, IPE for Severe Accident Vulnerabilities ML20198R3061997-10-27027 October 1997 Staff Evaluation Rept Concluding That Licensee IPE Complete Wrt Info Requested by GL 88-20 & Process Meets Intent of Subj GL ML20211L2151997-10-0303 October 1997 Safety Evaluation Supporting Licensee Relief Request,Per 10CFR50.55a(a)(3)(i) ML20217C1681997-09-22022 September 1997 Safety Evaluation Accepting Request for Relief from ASME Code,Section Iii,Div 2 for Repair of Damaged Concrete Reinforcement Steel NUREG-1335, Staff Evaluation Rept Concluding That Licensee IPE Complete Wrt Info Requested by GL 88-20 & Associated Guidance, NUREG-13351997-08-28028 August 1997 Staff Evaluation Rept Concluding That Licensee IPE Complete Wrt Info Requested by GL 88-20 & Associated Guidance, NUREG-1335 ML20141L9321997-05-29029 May 1997 Safety Evaluation Accepting Use of ASME Boiler & Pressure Vessel Code,Section Ix,Code Cases 2142-1 & 2143-1 for Reactor Coolant Sys for Plants ML20141B5551997-05-13013 May 1997 SE Accepting First 10-yr Interval Inservice Insp Program Plan Request for Relief NR-29 for Braidwood Station,Units 1 & 2 ML20140H8871997-05-0808 May 1997 Safety Evaluation Supporting Request for Relief from ASME Code Repair Requirements for ASME Code Class 3 Piping Ceco ML20134L7811996-11-18018 November 1996 Safety Evaluation Granting Listed Relief Requests,Per 10CFR50.55a(f)(6)(i) Based on Impracticalities in Design of Valves That Limit IST in Traditional Manner Using Position Indicating Lights ML20129F9101996-10-25025 October 1996 Safety Evaluation Accepting Request to Apply LBB Analyses to Eliminate Large Primary Loop Pipe Rupture from Structural Design Basis for Plant,Units 1 & 2 ML20059E2871993-12-30030 December 1993 Safety Evaluation Supporting Amends 57,57,45,45,93,77,152 & 140 to Licenses NPF-37,NPF-66,NPF-72,NPF-77,NPF-11,NPF-18, DPR-39 & DPR-48 Respectively ML20056D4921993-07-27027 July 1993 Safety Evaluation Re Fuel Reconstitution ML20127N1851993-01-25025 January 1993 Safety Evaluation Accepting Inservice Testing Program for Valves,Relief Request VR-4 ML20059L3371990-09-14014 September 1990 SER Granting Interim Relief for 1 Yr or Until Next Refueling Outage to Continue Current Testing Methods While Licensee Investigates Feasibility of Acceptable Alternatives ML20059L4581990-09-14014 September 1990 Sser Supporting Util Changes to Inservice Testing Program ML20058M0001990-08-0606 August 1990 Safety Evaluation Denying Licensee Response to Station Blackout Rule.Staff Recommends That Licensee Reevaluate Areas of Concern Identified in SER ML20058L9961990-08-0606 August 1990 Safety Evaluation Denying Licensee Response to Station Blackout Rule.Staff Recommends That Licensee Reevaluate Areas of Concern Identified in SER ML20248D5911989-08-0707 August 1989 SER Accepting Util 881130,890411,27 & 0523 Submittals Re Seismic Qualification of Byron Deep Wells ML20247D1471989-07-18018 July 1989 SER Supporting Util Proposed Implementation of ATWS Design, Per 10CFR50.62 Requirements ML20244D8191989-06-13013 June 1989 SER Supporting Util ATWS Mitigating Sys Actuation Circuitry Designs ML20247B3281989-04-24024 April 1989 Safety Evaluation Re Mechanical Draft Cooling Tower Tests ML20244A7221989-04-11011 April 1989 Safety Evaluation Concluding That Rev 1 to First 10-yr Interval Inservice Insp Program Plan Constitutes Basis for Compliance w/10CFR50.55a & Tech Spec 4.0.5.Response to Items 2.2.2 & 2.2.3 of Inel Technical Evaluation Rept Requested ML20155F1591988-10-0606 October 1988 Safety Evaluation Re Mixed Greases W/Greater than 5% Unqualified Contaminant in Limitorque Valve Operators. Insufficient Info Presented to Draw Conclusions ML20236L2001987-10-30030 October 1987 Safety Evaluation Supporting Amends 11 to Licenses NPF-37 & NPF-66,respectively & Amend 1 to License NPF-72 ML20237G8561987-08-10010 August 1987 SER on Util 870303 & 0522 Ltr Re Optpipe Computer Code Used in Snubber Reduction Program.Code Acceptable for Piping Dynamic Analysis Using Both Uniform & Independent Support Motion Response Spectrum ML20210T2571987-02-0606 February 1987 SER Re Util 850802 Submittal Describing Design Details of Steam Generator Blowdown & Auxiliary Steam Sys to Detect & Isolate High Energy Line Breaks.Sys Design Acceptable, However,Two Deviations from IEEE-STD-297 Criteria Apparent ML20210R2061987-02-0606 February 1987 Safety Evaluation Supporting Util 850517,0802,0823,1211 & 860429 Submittals Re Environ Effects of High Energy Line Breaks in Auxiliary Steam or Steam Generator Blowdown Sys. Design of Blowdown Sys Acceptable ML20209C3571987-01-23023 January 1987 SER Supporting Facility Design,Per Generic Ltr 83-28,Item 4.5.2, Reactor Trip Sys Reliability,On-Line Testing 1999-09-10
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217G9961999-10-14014 October 1999 SER Accepting First 10-year Interval Inservice Insp Requests for Relief for Plant,Units 1 & BW990066, Monthly Operating Repts for Sept 1999 for Braidwood Station, Units 1 & 2.With1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Braidwood Station, Units 1 & 2.With ML20217H5221999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Byron Station, Units 1 & 2.With ML20217P6351999-09-29029 September 1999 Non-proprietary Rev 6 to HI-982083, Licensing Rept for Spent Fuel Rack Installation at Byron & Braidwood Nuclear Stations ML20217A1691999-09-22022 September 1999 Part 21 Rept Re Engine Sys,Inc Controllers,Manufactured Between Dec 1997 & May 1999,that May Have Questionable Soldering Workmanship.Caused by Inadequate Personnel Training.Sent Rept to All Nuclear Customers ML20212A7441999-09-10010 September 1999 Safety Evaluation Concluding That Alternatives Contained in Relief Request 12R-07 Provide Acceptable Level of Quality & Safety BW990056, Monthly Operating Repts for Aug 1999 for Braidwood Station, Units 1 & 2.With1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Braidwood Station, Units 1 & 2.With ML20212B9261999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Byron Station,Units 1 & 2.With ML20210R6421999-08-13013 August 1999 ISI Outage Rept for A2R07 ML20210U8111999-08-0404 August 1999 SER Granting Licensee Relief Requests VR-1,VR-3 & Portion of VR-2 Pursuant to 10CFR50.55a(a)(3)(ii).Relief Request VR-4 Does Not Require Explicit NRC Approval Because Relief Applies to Valves Not Required by 10CFR50.55a ML20210R3431999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Byron Station, Units 1 & 2.With BW990048, Monthly Operating Repts for Jul 1999 for Braidwood Station, Units 1 & 2.With1999-07-31031 July 1999 Monthly Operating Repts for Jul 1999 for Braidwood Station, Units 1 & 2.With ML20210K9861999-07-30030 July 1999 Safety Evaluation Accepting Licensee 60-day Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs ML20210E2251999-07-21021 July 1999 B1R09 ISI Summary Rept Spring 1999 Outage, 980309-990424 M990002, Revised NFM9900022, Braidwood Unit 2 Cycle 8 COLR in ITS Format & W(Z) Function1999-07-12012 July 1999 Revised NFM9900022, Braidwood Unit 2 Cycle 8 COLR in ITS Format & W(Z) Function ML20216D3841999-07-12012 July 1999 Revised NFM9900022, Braidwood Unit 2 Cycle 8 COLR in ITS Format & W(Z) Function ML20209G1751999-07-0808 July 1999 SG Eddy Current Insp Rept,Cycle 9 Refueling Outage (B1R09) ML20209H3711999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Byron Station, Units 1 & 2.With BW990038, Monthly Operating Repts for June 1999 for Braidwood Station, Units 1 & 2.With1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Braidwood Station, Units 1 & 2.With ML20207H8071999-06-30030 June 1999 Rev 0 to WCAP-15178, Byron Unit 2 Heatup & Cooldowm Limit Curves for Normal Operations ML20207H7771999-06-30030 June 1999 Rev 0 to WCAP-15177, Evaluation of Pressurized Thermal Shock for Byron,Unit 2 ML20207H7851999-06-30030 June 1999 Rev 0 to WCAP-15183, Commonwealth Edison Co Byron,Unit 1 Surveillance Program Credibility Evaluation ML20207H7941999-06-30030 June 1999 Rev 0 to WCAP-15180, Commonwealth Edison Co Byron,Unit 2 Surveillance Program Credibility Evaluation ML20207H7561999-06-28028 June 1999 Pressure Temp Limits Rept (Ptlr) ML20207H7621999-06-28028 June 1999 Pressure Temp Limits Rept (Ptlr) BW990029, Monthly Operating Repts for May 1999 for Braidwood Stations, Units 1 & 2.With1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Braidwood Stations, Units 1 & 2.With ML20195J8001999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Byron Station,Units 1 & 2.With ML20209H7481999-05-31031 May 1999 Revised Monthly Operating Repts for May 1999 for Braidwood Station,Units 1 & 2 ML20207B6481999-05-25025 May 1999 SER Accepting Revised SGTR Analysis for Byron & Braidwood Stations.Revised Analysis Was Submitted to Support SG Replacement at Unit 1 of Each Station ML20195B2591999-05-19019 May 1999 Rev 66a to CE-1-A,consisting of Proposed Changes to QAP for Dnps,Qcs,Znps,Lcs,Byron & Braidwood Stations ML20206R6991999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Byron Station Units 1 & 2.With BW990021, Monthly Operating Repts for Apr 1999 for Braidwood Station, Units 1 & 2.With1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Braidwood Station, Units 1 & 2.With ML20195C7961999-04-28028 April 1999 Rev 2 to NFM9800233, Byron Station Unit 2 COLR for Cycle 8A (BY2C8A) M980023, Rev 2 to NFM9800233, Byron Station Unit 2 COLR for Cycle 8A (BY2C8A)1999-04-28028 April 1999 Rev 2 to NFM9800233, Byron Station Unit 2 COLR for Cycle 8A (BY2C8A) ML20205P7001999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Byron Station,Units 1 & 2.With BW990016, Monthly Operating Repts for Mar 1999 for Braidwood Generating Station,Units 1 & 2.With1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Braidwood Generating Station,Units 1 & 2.With ML20205B5091999-03-26026 March 1999 SER Accepting Relief Requests 12R-24,Rev 0 & 12R-34,Rev 0, Related to Second 10-year Interval Inservice Insp for Byron Station,Units 1 & 2 ML20205C5101999-03-21021 March 1999 Revised Safety Evaluation Supporting Improved TS Amends Issued by NRC on 981222 to FOLs NPF-37,NPF-66,NPF-72 & NPF-77.Revised Pages Include Editorial Corrections ML20204G3831999-03-19019 March 1999 Safety Evaluation Accepting Second 10-yr Interval ISI Request for Relief 12R-11 ML20206A8831999-03-17017 March 1999 Rev 0 to NFM9900043, Byron Unit 1,Cycle 10 COLR in ITS Format W(Z) Function M990004, Rev 0 to NFM9900043, Byron Unit 1,Cycle 10 COLR in ITS Format W(Z) Function1999-03-17017 March 1999 Rev 0 to NFM9900043, Byron Unit 1,Cycle 10 COLR in ITS Format W(Z) Function ML20196A0721999-03-16016 March 1999 Cycle 8 COLR in ITS Format & W(Z) Function ML20207M9231999-03-12012 March 1999 Amended Part 21 Rept Re Cooper-Bessemer Ksv EDG Power Piston Failure.Total of 198 or More Pistons Have Been Measured at Seven Different Sites.All Potentially Defective Pistons Have Been Removed from Svc Based on Encl Results ML20207J4371999-03-0808 March 1999 ISI Outage Rept for A1R07 ML20204H9941999-03-0303 March 1999 Non-proprietary Rev 4 to HI-982083, Licensing Rept for Spent Fuel Rack Installation at Byron & Braidwood Nuclear Stations BW990010, Monthly Operating Repts for Feb 1999 for Braidwood Generating Station,Units 1 & 2.With1999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Braidwood Generating Station,Units 1 & 2.With ML20204C7671999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Byron Station,Units 1 & 2.With ML20206U9011999-02-15015 February 1999 COLR for Braidwood Unit 2 Cycle 7. Page 1 0f 13 of Incoming Submittal Was Not Included BW990004, Monthly Operating Repts for Jan 1999 for Braidwood Generating Station,Units 1 & 2.With1999-01-31031 January 1999 Monthly Operating Repts for Jan 1999 for Braidwood Generating Station,Units 1 & 2.With ML20205M7061998-12-31031 December 1998 Unicom Corp 1998 Summary Annual Rept. with 1999-09-30
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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.
PDR
2-I l
2.0 EVALUATION 2.1 Licensee Assessment j
~
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.
1 I
~ 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
e 4
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
'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
"'I = 5,300 4
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
5