IR 05000454/1997013

Insp Rept 50-454/97-13 on 970811-0918.Violations Noted.Major Areas Inspected:Engineering & Plant Efforts for Upcoming Unit 1 SG Replacement Project

ML20212F542
Person / Time
Site:	Byron
Issue date:	10/30/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
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ML20212F518	List: IR 05000454/1997013 ML20212F514 ML20212F525
References
50-454-97-13, NUDOCS 9711040324
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Text

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U.S. NUCLEAR REGULATORY COMMISSION REGION lil Docket No: 50-454 License No: NPF-37 Report No: 50-454/97013(DRS)

Licensee: Commonwealth Edison Company Facility: Byron Nuclear Power Station Unit 1 Location: 4448 N. German Church Re.ad [

Byron, IL 61010 Dates: August 11- September 18,1997 Inspectors: D. S. Butler, Reactor Engineer M. S. Holmberg, Reactor Engineer Approved by: J. A. Gavula, Chief Engineering Specialists Branch 1 Division of Reactor Safety-

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9711040324 971030 PDR ADOCK 05000454 0 PDR

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EXECUTIVE SUMMARY Byron Nuclear Power Plant, Unit 1 NRC Inspection Report 50-454/97013 This inspection included a review of the engineering and planning efforts completed for the steam generator replacement modification scheduled for the upcoming Unit 1 refueling outag Engineering

• The overall steam generator replacement project demonstrated good engineering efforts focused on safety. The design enhancements and material upgrades were signtilcant improvements over the originat steam generators (Section E2.1).

• The safety evaluation provided detailed bases regarding the impact of the replacement steam generators on design basis accidents (Section E2.1).

• Inspectors identified that the safety evaluation did not provide documentation regarding the impact of the replacement steam generators on the residual heat removal system's cooling performance and on the containment sump and pH levels (Section E2.1).

• The licensee replacement program met ASME Section XI requirements. The certified design specification was comprehensive and technically rigorous with respect to design, fabrication and materials of construction. However, several minor weaknesses were identified in the Code reconciliation effort (Section E2.2).

• The engineering change notices for the modification were technically accurate and with one minor exception in accordance with licensee administrative requirenients (Section E3.1).

e Inspectors identified errors in the calculations for the change in the reactor coolant system volume and for the condensate storage tank minimum water volume (Section E3.2).

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Report Details 111. Engineering E2; Engineering Support of Facilities and Equipment -

E2.1 L Reolacement Simam Generator (RSG) Safetv Evahiation - Insoection Scone (50001. 377%)-

Inspectois reviewed port'.ons of the steam generator replacement safety evaluation, associated documents, t nd supporting calculations to verify technical adequacy and accuracy, ; Observations and Findinoa Safety evaluation,6H-97-0047, consisted of an evaluation originated by Framatome

-_ Technologies Intemational (FTI) with additional sections originated by the licensee. The-description of the RSG (model Delta 47 fabricated by Babcock and Wilcox Intemational

--- (DWI)) in the safety evaluation was thorough and provided an in-depth, comprehensive -

comparison of the design and performance changes between the original steam generator (OSG) and the RSG Thirteen design improvements had been incorporated

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into the RSG design to minimize or preclude degradation of the "more important problem areas" which had been observed in the industry for older models. Materin improvements such as Incond i lloy 690 tubes were used in the RSG to increase reliability and reduce. maintenance cost b.1 - Safety evhluation 6H-97-0047 fully addressed the updated final safety analysis (UFSAR)_

Chapter 15 accidents impacted by the RSGs. The objective of the' evaluation was to demonstrate that the accident analysis presented in the UFSAR bounded operatioit of the Byron Unit 1 and Braidwood Unit 1 with the RSGs.1 For those cases which could not be demonstrated to meet the acceptance criteria using a simple evaluation, confirmatory computer analyses were performed, inspectors noted that for most accident scenarios, the predicted plant transient was demonstrated to be less severe with the RSGs than 1 with the OSGs (e.g. Increased margins to design temperature and pressure limits).' The analytical approach utilized in the main steamline break (MSLB) and large break loss-of-

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coolant accidents (LOCA) for evaluating the plant response with the RSGs complied

with industry _and NRC accepted practice The Impact of the RSG modification on the balance of plant systems was evaluated in FTl document 51-1239285-03. However, the inspectors identified two examples where the RSG's affect was'not evaluated for certain plant system Byron UFSAR Section 5.4.7.1 " Design Basis" stated thet "..., the RHRS (Residual Heat Removal System] It designed to reduce the temperature of _the reactor coolant from

'350 F to .140*F with,,136 hours0.00157 days <br />0.0378 hours <br />2.248677e-4 weeks <br />5.1748e-5 months <br />." ' Additionally, UFSAR Figure 5.4-6 indicated that the RHR system could accomplish plant cool down in 34.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The reactor coolant

system (RCS) volume was increased by approximately 10 percent with the RSGs;

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however, the effect on the RHR system cooldown performance was not evaluated. The failure to provide a basis for the determination that the RSG modification did not involve an unreviewed safety question is an example of a violation of 10 CFR 50.59 (VIO 50-454/97013-01(a)(DRS)).

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Byron UFSAR Sections 6.1.3 "Postaccident Chemistry," 6.1.3.1 "Steamline Break Inside Containment" and 6.1.3.2 " Main Feedwater Line Break inside Containment" described the r,ffect of a main steamline break (MSLB) and main feedline break (MFLB)

on containme it sump level and pH. Since the RSGs operate with a higher secondary mass inven'.ory and also have a larger potential feedwater break area (0.86 square feet (ft') vice U.223 ft2), the postaccident chemistry would be potentially impacted by the RSGs in two ways. First, the larger secondary inventory and break area would result in increased sump fill rates and higher containment sump levels. Second, the increase in peak containment pressure may no longer temain t Aw the containment spray actuation setpoint, causing sump pH and fluid levels to dange. The potential effects or)

containment sump and pH levels were not evaluated by tne licensee. The failure to provide a basis for the determination that the RSG modification did not involve an unreviewed safety question is an example of a violation of 10 CFR 50.59 (VIO 50-454/97013-01(b)(DRS)). in Section 4.6.19 of the safety evaluation, the licensee evaluated the condensate storage tank (CST) water supply volume required to cool down the plant under complete loss of electric power. The licensee concluded that the minimum CST inventory required to cooldown the plant with auxiliary feedwater (AFW) system operation would be within the minimum value of 200,000 gallons as stated in UFSAR Section 10.4.9. However, the inspectors identified sources of error (see section E.3) in the supporting calculation (FTl calculation 51- 1266158-01) that appeared to invalidate this conclusio Additionally, a licensee evaluation had not been documented for the adequacy of the technical specification (TS) 3.7.1.3 minimum CST inventory (usable voiume) with respect to the minimum UFSAR design basis inventory of 200,000 gallons. TS 3.7. allowed a minimum CST volume of 40 percent, which equated to 200,000 gallon Calc'.;!ation BYR 97-273 " Condensate Storage Tank Level Error Analysis" Revision 5, concluded that the instrument accuracy for the condensate storage tank was approximately four percent (+/- 20,000 gallons). Thus, at the TS 3.7.1.3 minimum allowed tank capacity, (without positive level instrument setpoint bias adjustments) a usable tank volume of 180,000 gallons could exist, which would be an insufficient inventoiy to support plant cooldown. The inspectors considered this issue to be an unresolved item (URI) pending review of a licensee analysis that demonstrated that Section 4.6.19 of the safety evaluation and TS 3.7.1.3 as documented on September 18,1997, were adequate to meet 10 CFR T 59 requirements (URI 50-454/97013-02(DRS)). The RSG feedwater nozzle location is approximately 33 feet higher than the OSG and the upper lateral restraint (ULR) creates an interference for complete inservice inspection of the feedwater nozzle-to-thell weld. Section 2.2.11 of the safety evaluat;on stated that due to the existing design of the ULR, the ULR belly band will require removal in order to provide access to the main feedwater nozzle-to-shell weld for inservice inspection. The inspectors noted that this configuration could potentially be characterized as a hardship for the licensee's ISI program. However, because the current modification produced this situation, a relief request in the future would not be

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appropriate. Pending completion of an inservice examination of the RSG feedwater nozzle-to-shell weld, this is considercd an Inspection Followup Item (IFI)

(50-454/97013-03 (DRS)). The mechanical design flow is the conservatively high flow used in the mechanical design of the reactor vesselinternals and fuel assemblies. The RSG had been designed for less recirculation system loop flow resistance than the OSG design, which resulted in an increased the best estimate loop flowrate. UFSAR Section 5.1 stated that the mechan: cal design flow is approximataly 3.7 percent greater than the best estimate flow. Safety evaluation Section 4.6.1 concluded that the best estimated flow, including a 2 percent margin for instrument error, would be just within the mechanical design flow limit. Based on the small margin available to the mechanical design flow limit, the inspectors will review the actual loop flowrate measurements during post modification testing (IFl 50 454/97013-04(DRS)). Conclusiom The overall steam generator replacement initiative demonstrated a good engineering effort focused on safety The design changes and material upgrades in the RSGs represented significant inarovements over the OSG The licensee's safety evalu ation was thorough with respect to the UFSAR design basis accidents impacted by the RSGs, Additionally, inspectors noted that for most accident scenarios, the predicted plant transient was demonstrated to be less severe with tho RSGs than with the OSG Inspectors identified two examples where the licensee had not performed a written evaluation of the impact of the RSGs on systems affected by this modification. The licensee had not considered the fullimpact of the RSG modification on the design basis plant cooling performance of the RHR system. Additionally, the effects of a main steamline break and main feedline break for the RSG modification on the containment sump and pH levels had not been fully evaluated. Inspectors considered that these errors indicated a less than rigorous engineering staff effort for evaluation of the scope of effected systems impacted by the RSG modification.

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E2.2 ASME Section XI Reolacement Prooram imolementation insoection Scooe(50001. 73753. 37700)

Inspectors reviewed procedures, specifications, and other documents that implemented the steam generator replacement program, to evaluate compliance with ASME Code, Section XI requirements: Observations and Findinos The RSG certified design specification, FTl 18-1229648, was comprehensive ano technically rigorous in addressing the design, fabrication and material requirement Additionaily, the RSG design, fabrication and construct!on materials as described in this document met or exceeded, industry codes, standards and NRC requirement __-____

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The Code reconciliation document, SL-4744, for the RSG identified Byron and

. Braidwood sites; however, the applicable station Unit and the original steam generator

(SG) construction Code had not been documented. The licensee issued Nuclear Design ilnformation Transmittal BB-EXT-1298 to correct this documentation error. Appendix A -

- of SL-4744, reconciled the specific technical changes to Code requirements which had '

occurred from the original SG construction code up through the 1986 Edition of Section -

lll of the ASME Code. Item'15 of Appendix A described a et,ange to the ASME Code, Section lil, Paragraph NB-4622.4, relating to attemate requirements for post-weld heat treating temperature and times. The technical reconciliation basis for this item stated that "this is not significant for steam generators since the vessels are normally P-3 materials." This justification did not bound the Byron RSGs, which contained P-1 as well

, _ as P-3 materials in the construction of the RSGs. Thus, the attemative requirements of Paragraph NB-4622.4 could have been invoked in construction of the RSGs. However, inspectors confirmed that the option to use attemate post weld heat treatment

. temperatures and times had been prohibited in the RSG procurement specification -

S-2002. Inspectors considered the documentation error and the weak technical basis

for Appendix A, item 15 of SL-4744, to indicate a lack of rigor in design code reconciliation of the RSG. -

ihe inspector verified that Code cases used in design and construction of the RSG had been accepted for use by the NRC. The ASME Code Sectioa XI, paragraph

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IWA-7210(c)(1), stated "The requirements affecting design, fabrication, and' examination of the item to be used for replacement are reconciled with the Owner's through the

' Stress Analysis Report, Design Report, or other suitable method that demonstrates the item is satisfactory for the specified design and operating conditions."_ The engineering

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staff reported that this requirement had been met by listing the effective Code cases on the Form N. " Certificate Holders' Data Report for Nuclear Vessels." Inspectors considered that the licensee's list and description of Code Cases used in RSG design or

_ coastruction, without comparison to the original construction Code requirements,

[ indicated a less than rigorous application of IWA-7210 requirements, c. Conclusions Overali, the licensee replacement program met ASME Section XI requirements. - The inspectors considered the RSG certified design specification to be comprehensive and technically rigorous with_ respect to design, fabrication and materials of constructio Additionally, the RSG design, fabrication and construction materials met or exceeded, industry Codes, Standards and NRC requirements. However, several weaknesses'wsre identified in the RSG Code reconciliation effor E3 Engineering Procedures and Documentation E Steam Generator Desion Chance Packan_e insoection Scone (50001. 37700)

Inspectors reviewed six of the twenty-three engineering change notices (ECNs)

associated with' design change package (DCP) 9500394 " Steam Generator Replacement Modifications."

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! Observations and Findinas  !

Of the twenty-three ECNs issued for design control package 9500394, as of September 15,1997, eight were on hold. The ECNs on hold were typically awaiting final licensee approval and verification of vendor supporting calculat!ons or analysi Inspectors identified that the approved ECNs reviewed did not contain the UFSAR and/or technical specifications (TS) affected on the approved drawing list for the EC This was contrary to NEP 08-01 " Engineering Change Notices" Revision 1 requirements. Inspectors considered that this would not have resulted in failure to update the UFSAR and TS, which were initiated and controlled by administrative 5 procedures referenced in the modification approval letter. The licensee issued problem identification form (PlF) B1997-02818 and initiated field change request (FCR) 970064 to correct DCP 9500394. This issue was of minor significance and was considered a Non-Cited Violation (NCV), of 10 CFR Part 50, Appendix B, Criterion V, consistent with Section Vll.B.1 of the NRC Enforcement Policy (NCV 50 254/97013-05(DRS)).

Inspectors considered the overall administrative quality of the ECNs reviewed to be good, in light of the large amount of engineering work and coordination with interfacing organizations involved for a project of this magnitude, Conclusions The ECNs supporting design control package 9500394 for the RSGs were technically accurate and, with one minor exception, were in accordance with licensee administrative requirements, which inspectors considered to indicate a good on-site engineering effor E3.2 Steam Generator Reolacement Calculations Insoection Scoce(50001. 37700)

Inspectors reviewed selected calculations, analyses, and computer codes supporting the replacement steam generators to verify technical adequacy, accuracy, and compliance with NRC requirements and licensee commitment Observations and Findinas Overall, the calculations reviewed contained a clear description of their intended purpose and used a technically acceptable calculational approach or methodology, in particular, inspectors considered the additional input margins and thorough supporting documentation in Commonwealth Edison Calculation 3C8-1280-001 " Post LOCA Hydrogen Concentration," to demonstrate a conservative and rigorous approach to design basis calculation The licensee submitted a technical specification (TS) amendment for a revised P, containment pressure on January 30,1997. As part of this amendment, TS 5.4.2, associated with the reactor coolant system volume, was being changed to account for the increased coolant volume with the RSGs. Engineering staff reported that BWI calculation 222-7720-A13 had been used as a basis to determine the 1251 cu ft total increase In volume recorded in this amendment. On September 3,1997, inspectors identified that this calculation did not account for additional RCS volume from:

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increased tube diameter in the tubesheet region due to hydraulic expansion of the tubes (estimated to be 2.0 cubic feet (ft') per SG); thermal expansion of tubes at normal operating temperature (estimated to be 4.4 ft* per SG (this estimate did not include the volume for thermal expansion of the U bend region of the tut;a bundle)); and thermal expansion of the primary SG head (estimated to be 3.0 ft5 per SG). These errors totaled 37.6 ft' of RCS vo!9me (four RSGs) and potentially impacted any RSG analysis /

calculation which used BWI calculation 222 7720-A13 as an input for the RSG primary volume. Failure to ensure that design basis information had been correctly incorporated into this calculation was considered an example of a violation of 10 CFR Part 50 Appendix B, Criterion 111 (VIO 50-454/97013-06a(DRS)).

In addition to the January 30,1997, TS amendment for P., the inspector identified the

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following analyses which used an input of RCS volume change based on, or derived from, calculation 222-7720-A13: FTl document 51-1239285-03 "NSS [ Nuclear Steam Supply] and DOP [ Balance of Plant) Systems Review" Revision 3, calculation 3C8-1280-001 " Post LOCA Hydrogen Concentration" Revision 21, FTl analysis 51-1244429-01

"OSG-RSG Comparison" Revision The licensee's engineering staff stated that, for the volume error estimates discussed above, the final corrected volume added would be lower due to allowances made for thermal expansion in the OSG calculation. Also, the volume errors would not impact the P, pressure calculated in the technical specification amendment, since the REl AP5 computer code factored in the increase in coolant volume from thermal expansion of primary components. At the conclusion of the inspection the licensee staff had initiated PIF C1997-00392, placed the modification on hold, and were investigating the scope of affected analyses supporting the RSG modification, In FTl calculation 51-1266158-01 "RSG AFW Cooldown Requirements" the licensee calculated that 199,851 gallons of cooling water stored in the CST was the minimum

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inventory required to support cooldown of the plant to hot standby under station blackout conditions with the RSGs. Thic calcuiation used an erroneous RCS volume increase (1250.72 ft') input for the RSGs derived from the calculation discussed above, inspectors estimated that correcting the calculation for volume errors discussed above would increased the cooling water - aired by less than an additional 100 gallons, in addition, inspectors identified that the, ilcensee had assumed the same specific heat capacity for the RSGs as had been used for the OSGs, which had not been verified or demonstrated to be conservative. Due to differences in the types and amounts of materials used in the RSG, vice those used in the OSGs, the specific heat capacity was not likely to be the same. A change in the specific heat capacity for the metal mass of the RSG impacts the amount of c cling water required to cooldown the plant for this calculatio To support the RSG modification, the licensee had planned to reconfigure the AFW system to connect into the 16 inch main feedwater header outside containment. With this configuration, inspectors estimated that several hundred feet of 16-inch piping, with associated metal mass must be reduced in temperature from 440 *F to 350 *F under a station blackout using the auxiliary feedwater system. In calculation 51-1266158-01 the licensee had not accounted for this design basis heat load input. Inspectors calculated that this additional heat load combined with errors discussed above would increase the required minimum CST inventory above the 200,00'., gallons currently described in the 8 )

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UFSAR. The licensee issued PIF C1997-00394, placed a hold on the modification package and initiated a revised calculation. The failure to verify the adequacy of calculation 51 1266158-01 was considered an example of a violation of 10 CFR Part 50 Appendix B, Criterion lli (VIO 50-454/97013-06b(DRS)).

) Safety evaluation 6H 97-0047, Section 4.2.3.15, stated, " Critical parameters that effect system and core responses to the steamline break are heat transfer surface area and break size." Break size was limited by the secondary flow restrictor (SFR) at the outlet of the RSG which had been designed to limit flow in the event of a MSLB such that an excessive plant cooldown transient would not occur. Inspectors calculated a cioss sectional area for the SFR of 1.138 sqaure feet (ft') based on dimensions and tolerances in BWI drawing 7720G017, Revision 2. The 1.138 ft' number included design allowances for corrosion and allowance for thermal expansion to normal operating temperature. Inspectors identified that FTl 1239270-02 " Main Steamline Break for Core Response AIS", Revision 2, used a 1.1 ft' cross section as an input to the MSLB analysis and therefore was nonconservative by 3.4 percent. Additionally, inspectors calculated that due to thermal expansion of the RSG tubes to normal operating temperature an additional 311 ft' of secondary side heat transfer surface area would exist. Engineering staff stated that this additional surface area had not been incorporated into the secondary side heat transfer surface area of 79,800 ft' recorded in FTl document 51-1244429-01 and used in the MSLB analysis. Additionally, 51 1244429-01 incorrectly identified the 79,800 ft' secondary side surface area as a hot plant value. These issues were conside ed an unresolved item pending licensee demonstration that SFR cross sectional area and secondary side heat transfer surface area inputs utilized for the MSLB analysis were adequate and bound by other quantifiable calculational conservatism (URI 50-454/97013-07(DRS)). However, the licensee staff reportedly intended to demonstrate that cold plant values for the SFR area and RSG tube heat transfer surface area had been accepted and approved by the NRC for inputs to the MSLB acciden Safety and performance analysis computer codes (e.g. RELAPS/ MOD 2-B&W, CRAFT 2, REFLOD3B, BEACH, and CONTEMPT) used in the RSG design and analysis were documented in Framatome laternational document No. 77-123706-02 as meeting the Framatome International %U+u ^ ruce Progrem These computer codes had also been docw _ a cumwu R pw as NRC accepted Babcock & Wilcox topical

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y;; CM40!64F n., f1W hW. NA, BAW-10171P-A, BAW-10166P-A, and MG95Ai !rupciots m o W the N c?. >afety evaluation accepting BAW 10164P

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' W Adva m d Cer W Pro / LT,: Vr .er Reactor LOCA and Non-LOCA u.~.~m mg." c.r s.~ 2 nrc y mio oncluded that the RELAP5 computer code

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sutm De RSC- % i eatimshN tner iccepted by the NRC. The structural Mm cv op m cWoc %g. PANSPANr,VSCAN, BWSPEC, RESPECT, COMPAR2)

W w tw avawcd by M NRC g. an NRC approved topical report had not been eview) hy M W;\ Ne ..ese codes were documented as bench-marked ogstr m cem! sch . . s ..uor industry accepted codes, c. !h&drions y '

Overall, the calculations reviewed contabed a clear description of their intended purpose and used a technically acceptable calcu'ational approach or methodolog However, inspectors identified errors for calculational inputs for reactor coolant volume

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changes associated with the RSG modification that impacted several other calculations and analysis.- Additionally, a design basis heat load input was missed in the calculation supporting the minimum CST water volume inventory, inspectors considered that the -

calculation input errors and nonconservatisms identified, indicated a lack of attention to detall for control of design inputs for calculations supporting the RSG modificatio V. Manaamment Meetings X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on September 17,1997. The licensee acknowledged the findings presented and did not identify any of the potential report input discussed as proprietar _ _ _ _ _ - _ _ _ _ _ _

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, 1 PARTIAL LIST OF PE'tSONS CONTACTED Commonwealth Edison - ,

Kl. Kofron Station Manager B. Moravec Steam Generator Replacement (SGR) Site Project Manager D. Shamblin - SGR Project Manage M. Leutloif . Byron SGR Project Engineer-D. Wozniak Byron Station Engineering Mar.ager D. Rogowski SGR Project Engineer _ .

T.- Schuster Quality Assurance Manager P. O'Neill SGR Quality Supervisor R. Goetzke SGR Mechanical Engineer R. Colglazer- NRC Coordinator S. Mullins Braidwood SGR Project Engineer M. Inserra SGR Engineer -

H. Kim Pressurized W -ter Reacter Analysis Supervisor D. Saccomondo Senior Pressurized Water Reactor Licensing Admhistrator

- M. Lesniak Nuclear Licensing S. Elch SGR !nstrument and Controls Engineer Bechtel A. Motyos Project Engineer C. Weaver Project Manager -

Saraent & Lundy S. Bertheau Project Manager NBC S. Burgess Senior Resident inspector N. Hilton Resident inspector INSPECTION PROCEDURES USED IP 50001 STEAM GENERATOR REPLACEMENT INSPECTION IP 73753 INSERVICE INSPECTION IP 37700 DESIGN CHANGES AND MODIFICATIONS

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ITEMS OPENED, CLOSED or DISCUSSED DDDD

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~ VIO 50-454/97013-01(a)(DRS) Failure to evaluate the impact of the RSG modification on

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the RHR system (Section E2.1).

VIO 50-454/97013-01(b)(DRS) Failure to evaluate the impact of the RSG modification on containment sump and pH levels in a MFLB and MSLB accident (Section E2.1),

i URI 50-454/97013-02(DRS) TS 3.7.1.3 and minimum CST inventory safety evaluation conclusions potentially inadequate for the RSG (Section E2.1).

IFl 50-454/97013-03(DRS) Accessibility of the RSG feedwater nozzle-to-shell weld for

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inservice inspection (Section E2.1).

IFl 50-454/97013-04(DRS) Review of actual loop flowrates to verify mechanical design flow margin (Section E2.1).

NCV 50-25 97013-05(DRS) RSG ECNs failed to follow NEP 08-01 requirements to list effected UFSAR and TS on the approved drawing list (Section E3.1).

VIO 50-454/97013-06a(DRS) Design basis RCS volume miscalculated / misapplied effecting other RSG analysis (Section E3.2).

VIO 50-454 '97013-06b(DRS) Design basis inputs missed or nonconservative for CST minimum inventory calculation (Section E3.2).

URI 50-454/97013-07(DRS) MSLB analysis inputs appear nonconservative (Section E3.2),

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i LIST OF DOCUMENTS REVIEWED Safety evaluation GH-97-0047 for design control package 9500394 " Steam Generator .

Replacement Modifications."

FTl 51 1239285-03 "NSS and BOP Systems Review."

FTl 1239270-02 " Main Steamline Break for Core Response AIS," Revision FTl 51-1239256-01 "Analvtical input Specification for Large Break Loss-of-Coolant Accident,"

Revision BAP 1600-5, "ASME Section XI Repair / Replacement Requirements," Revision " Addendum to the Overpressure Protection Report for Byron /Braidwood Nuclear Power Plant Units 1," Revision SL-4744 " Steam Generator Replacement Code Reconciliation to ASME Boller & Pressure

. Vessel Code Section 111-1965 through 1986 Editions," Revision ASME Section 111 Data Report - Form N. " Certificate Holders' Data Report for Nuclear Vessels,"

(Draft).

FTl 18-1229648 " Certified Design Specification," Revision " Replacement Steam Generator Steam Generator Replacement Group Specification S-2002,"

Revision ECN BYR000763M SG Vessel Replacemen ECN BYR000758S Outside Containment lift syste ECN BYR000764M Primary Pipin ECN BYR000767M FW Piping (Inside Containment).

ECN BYR000769M FW Piping (Outside Containment).

ECN BYR000786S Upper Lateral and Lower Lateral Restraint Desig FTl document No. 51-1257351-01 "CDS /OSG /RSG Design Transient Comparison."

1 FTl calculation 51-123751-01 " Loading Specification for Steam Generator Replacement, Byron /Braidwood Unit 1," Revision BWI Calculation 222-7720-A13 " Engineering Calculations - Byron /Braidwood RSG - Primary

- Fluid Volumes vs. Height," Revision Comed Calculation 3C8-1280-001 " Post LOCA Hydrogen Concentra' ion," Revision 2 . l

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FTl document 51-1244429M1, ."OSG-RSG Comparison," Revision FTl calculation 51 1266158-01, "RSG AFW Cooidown Requirements," Revision 1.--

FTl calculation 32-1239262-01, "CorhEd LOFW," Revision NED I-EIC-0008, " Reactor Coolant Flow Channel Error Analysis," Revision 1

BYR-96-275, " Steam Generator Wide Range Level Indication Error Analysis," Revision 1~-

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.6IC-1 FW-001, " Steam Generator Narrow Range Level Transmitter Scaling," Revision /61C-1-FWOO8, Unit 1 Steam Generator Narrow Range Level Channel Error Analysis (Steam

' Generator Replacement Project)," Revision 1,

BYR-97-273 " Condensate Storage _ Tank Level Error Analysis," Revision LTemporary alteration TA 97-1-27 " Electrical Feed from C RCP," Revision Draft coordination curve for the Reactor Coolent Pump C electrical penetration.

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LIST OF ACRONYMS USED-AFW Auxiliary Feedwater -

ASME American Society of Mechanical Engineers BWI Babcock and Wilcox Intemational

'CDS Certified Design Specification CST Condensato Storage Tank DCP(s) Design Control Package (s)

ECN(s) _ Engineering Change Notice (s)

FTl ' Framatome Technologies Inteniational FW- Feedwater IFl_ inspection Followup Item ISI Inservice inspection

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LBLOCA Large Break Loss of Coolant Accident

_ LOFW Loss of Feedwater MFLB Main Feedline Break MSLB Main Steamline Break NSS Nuclear Steam Supplier OSG(s) Original Steam Generator (s)

PIF Problem Identification Form RCP- Reactor Coolant Pump RCS Reactor Coolant System RHR Residual Heat Removal RSG(s) Replacement Steam Generator (s)

SFR Secondary Flow Restrictor SG Steam Generator

.TS Technical Specification UFSAR Updated Final Safety Analysis Report URI Unresolved item VIO Violation

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