ML20199M000

Forwards Response to Violations Noted in Insp Rept 50-454/97-13.Corrective Actions:Quantitatve Analysis Was Performed to Document Conclusion of No Impact

ML20199M000
Person / Time
Site:	Byron
Issue date:	11/24/1997
From:	Graesser K COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
50-454-97-13, BYRON-97-0279, BYRON-97-279, NUDOCS 9712020171
Download: ML20199M000 (10)

Text

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nyron. suoloimv 6 Tel MI'L2.4 4 5 6 tl November 24,1997

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Byron 97-0279 FILE:

.1.10 0101 U. S. Nuclear Regulatory Commission

. Washington, DC 20555 Attention:

Document Control Desk

Subject:

Byron Nuclear Power Station Umt 1 Response to Notice of Violation inspection Report No. 50-454/97013 NRC Docket Number 50-454

Reference:

John A. Grobe letter to Mr. Graesser dated October 30,1997, transmitting NRC inspection Report 50-454/97013 Enclosed is Commomvealth Edison Company's response to the Notice of Violation (NOV) which was transmitted with the referenced letter and Inspection Report. The NOV cited two (2)

Severity Level IV violations requiring a written response. CemEd's response is provided in the attachments.

This letter contains the following commitments:

1) As a result of the identification of two items with inadequate safety evaluations, additietal reviews have been initiated to focus on possible unidentified impacts of RSG changes on

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Safety Systems. %csc additional reviews will ensure that, as a minimum, the bases for

/2 conclusions of no impact are adequately documented.

2) Revise UFSAR Section 5.4.7, " Residual liert Removal System," to discuss the quantitative evaluation on RilR performed as part of the SGR project UFSAR update.

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3). Revise UFS AR Section 6.1.3.2 to reflect the sump pH response with the RSGs as part of the SGR project UFSAR update.

4) - Revise UFSAR Figure 6.1-1 to retket the containment sump water volumes with the RSGs as part of the SGR project UFSAR update.

5) RcGew all ulculations that utilized RCS volume as a design input and revise calculations, as necessary, to ensure acceptable results due to the RCS volume increase.

6) Include as part of the UFSAR update, a detailed table of RCS total and component volumes, hot and cold, for both smits.

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Byron Ltr. 97-0279 November 24,1997 Page 2 If your staff has any questions or comments concerning this letter, please refer them to Don Brindle, Regulatory Assurance Supervisor, at (815)234-544I ext. 2280.

Respec' fully, V

,[q K, L Grac s U Site Vice President Byron Nuclear Power Station KLG/DB/rp Attachment (s) cc:

A. B. Beach, NRC Regional Administrator - Rlli G. F. Dick Jr:, Byron Project Manager - NRR Senior Resident inspector, Byron M. J. Jordan, Reactor Projects Chief-Rlli F. Niziolek, Division of Engineering - IDNS (p 197Mhrs'9702792)

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D. L Farrar, Nuclear Regulatory Services Manager, Downers grove d

' Safety RcWew Dept, c/o Document Control Desk,3 Floor, Downers Grove DCD-Licensing, Suite 400, Downers Grove.

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(p/97byhrs 9702793)

ATTACilh1ENT I XIOl,ATION (50-454/97013 01a b) 10 CFR 50.59(a)(1) states, in part, that a licensee may make changes to the facility as described in the safety analysis report without prior Commission approval unless the proposed change involves an unreviewed safety question.

10 CFR 50.59(b)(1) states, in part, that the licensee shall maintain records of changes in the h

facility as described in the safety analysis report and that records must includc a written safety evaluation which provides the basis for the determination that the change does not involve an unreviewed safety question The Byron Updated Final Safety Analysis Report (UFS AR) Section 5.4.7.1 " Design Basis" stated that " _, the RilRS [ Residual lleat Removal System] is designed to reduce the temperature of the reactor coolant from 350 F to 140*F within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />."

The Byron UFSAR Section 6.1.3 " Post Accident Chemistry," Section 6.1.3.1 "Steamline Break c

L61e Containment"and Section 6.1.3.2 "hiain Feedwater Line Break inside Containment" wBed the efTect of a main steamline break (htSLB) and main feedline break (hiFLB) on mam. ment sump level and pit.

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Contrary to the above, as of September 9,1997, the licensee had not performed an adequate safety evaluation to determine whether the impact on the design basis of the J

RilR system for the replacement steam generator (RSG) modification, constituted an unreviewed safety question. Specifically, the es aluation was deficient because it failed to consider the effect of the increased heat load (associated with the increased reactor volume for the RSG modification) on the RilR system performance.

(50-454/97013-Ola(DRS))

b.

Contrary to the above, as of September 9,1997, the licensee had not performed an adequate safety evaluation to determine whether the impact on the containment sump level and pil for the RSG modification, constituted an unreviewed safety question.

Specifically, the evaluation was deficient because it failed to consider the RSG increased g

secondary side mass inventory and larger feedwater break area on containment sump and pil level under a htSLB or h1FLB. (50-454/97013-Olb(DRS))

This is a Severity Level IV Violation (Supplement 1)

REASON FOR Tile VIOL,ATION a.

Inadequate Safety Evaluation - RIIR Performarace The RHR sy stem b capable of reducing the tert perature of the reactor coolant per UFSAR Figures 5.4-6 and 5.4-7 for dual and single train operation, respectively (UFS AR Section 5.4.7). The initial review of system impacts did not identify impact on RilR performance as requiring quantitative analysis because: 1) the integated decay heat is much larger than the added heat due to the increased RSG volume and metal mass, and 2) qualitatively the RSG impacts are small compared to the existing margin between (p W7httrv970279 4)

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UFSAR Section 5.4.7.1 " Design Basis" and the calculated system performance. There -

vas a failure to document engineering judgeriera, and therefore, a documented qtatitative analysis was not performed to support the conclusion of no impact.

b, Inadequare safety Evaluation -.,ntainment Sump Level & Sump pil Calculations were perfortred to assess the impact of the RSG design. an UFSAR Section 6.1.3," Post-Accident Chemistry". Both the minimum and maximum pil calculations were performed for the appropriate limiting accident conditions. The results were verified to be within the acceptable pli band as specified in the plant Technical Specifications. Ilowever, UFSAR Sections 6.1.3.1 and 6.1.3.2 discuss the safety system response to the MSLB and MFLB, respectively. Changes in the MFLB transient response (Containment Spray (CS) actuation) were not specifically identified and documented in the safety evaluation regarding containment sump pli values.

Calculations were also performed to determine the maximum volume of water in the containment following an accident. Ilowever, only the limiting case for containment maximum flood level following a Large Break Loss of Coolant Accident (LB LOCA) was determined. UFSAR Sections 6.1.3.1 and 6.1.3.2 discuss the safety system response to the MSLB and MFLB, respectively. The specific containment water volumes for these -

transients were not determined. Cl anges in the MFLB transient response (CS actuation) l were not specifically identified and documented in the safety evaluation regarding containment sump water volumcw

%c initial review of these transients did not identify the potential impact as requiring quantitative analysis for containment sump level and pli because qualitatively the RSG impacts are bounded by the analysis performed for the LB LOCA transient. Therefore, due to a lack of attention to detail, a documented quantitatise analysis was not performed to support the conclusion of no impact.

CORRECTIVE STEPS TAKEN AND RESULTS ACillEVED a

Inadequate Safety Evaluation - RIIR Performance A quantitative analysis was performed to document the conclusion of no impact. The RSGs contain more primary liquid mass and more metal mass than the original Stean Generators (OSGs). There are also differences in the relative amounts of water and steam on the secondary side. A calculation that accounts for these differences was -

performed based on the assumptions and methodology used to generate the UFS AR curves. Single train and dual train RiiR cooldown were analyzed to evaluate the effects on RilR system performance. For RilR operations with only a single operable RiiR train, the measure of RilR performance is based on the duration required to cool the RCS from 350 F to 200'F (UFSAR Figure 5.4 7). In this case the RSGs cause an increase of approximately 0.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to the ex sting 39 hour4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> single train R11R cooldown i

time.

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For RilR operations with two operable RHR trains, the measure of RilR performance is based on the duration required to cool the RCS from 350 F to 140 F (UFSAR Figure 5.4-6). In this case the RSGs cause an increase of approximately 0.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to the existing 30.3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> dual train RilR cooldown time. This is within the design basis ofless than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

b.

Inadequate Safety Evaluation - Containment Sump Level & Sump pH A quantitative evaluation of the A1FLB transient was performed to determine the impact of the revised hiain Feedwater/ Auxiliary Feedwater configuration on the containment sump pH. The current UFSAR evaluation (Section 6.1.3.2,"htain Feedwater Line Break") states that the h1FLB transient does not initiate CS since containment pressure remains lower than the CS actuation setpoint. The sump pH, therefore, is considered to be the pil of the condensate fluid assumed in the UFSAR. The evaluation performed for the RSGs demonstrates that a h1FLB could elevate containment pressure sufficiently to actuate CS, thus causing the event to respond similar to the A1SLB transient. However, the mass and energy release for the htFLB with the RSGs is less than that of the htSLB transient, therefore, the htSLB remains the more limiting transient for the containment environmental conditions. The current UFS AR evaluation for h1SLB (Section 6.1.3.1) indicates a constant pH for the CS fluid that is higher tnan in condensate fluid (Section 6.1.3.2). Therefare, the impact of the RSGs on the h1FLB is that the containment sump pH could increase to a value consistent with t'

'1SLB transient due to the actuation of CS. The amount of additional RSG secondr je mass is not sufficient to appreciably reduce the pH of the sump fluid. This char -

4crefore, has no safety significance since the sump pH remains within the limits acceptable for post-accident conditions specified in the Technical Specifications.

Also, the impact of the RSGs on the containment sump water volume was evaluated for both the h1SLB and the h1FLB. In the case of the h1SLB, the differe m for the RSG case is an increase in secondary side mass (approximately 22 000 lbs) Relatne to the total volume of fluid pumped from containment spray in 30 minutes, this increase is negligible and the sump water volume with the RSG remains consistent the quantities indicated in UFSAR Figure 6,1 1. UFSAR Figure 6.1-2 provides the containment sump water volume following a h1FLB accident. Since Unit I actuates CS with the RSGs, the containment water volume in Figure 6.1-2 no longer applies to Unit 1. The sump water volume shown in Figure 6.1-1 which account for CS actuation would apply for the Unit I h1FLB. Both the h1SLB and h1FLB transients have containment sump water volumes below the limiting case for containment flooding (LB LOCA) documented in UFSAR Attachment D3,6. This change, therefore, has no safety significance since the sump water volumes remain within the maximum volume acceptable for post-LOCA conditions.

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l-CORRECTIVE STEPS TilAT Wil.1, HE TAKEN TO AVOID 11J.RTilER VIOI.ATION a, b.

Inadequate Safety Evaluation - RilR Performance and Containment Sump Level

& Sump pil As a result of the identincatk. vt two items with inadequate safety evaluations, additional reviews hav. ocen initiated. These re-reviews focus on possible unidentined impacts of RSG changes on Safety Systems. They include: 1) A ic-review of select i

UFS AR and Technical SpeciGcation sections, and 2) A re-review of FTI's "NSS and HOP Systems Review"(document 51 1239285-03). These additional resiews will ensure that, e a minimum, the bases for conclusions of no impact are adequately documented.

The SGR safety evaluation will be augmented, if required, to reflect the review results so that the bases of conclusions are readily accessible to the reviewer. This action will be tracked by NTS item # 454100 97 01301-01.

UFSAR Section 5.4.7," Residual lleat Removal System," will be revised to discuss the quantitative evaluation performed as part of the SGR project UFSAR update.

Additionally, the SGR safety evaluation will be revised to reflect the results of this analysis. This action will be tracked by NTS item # 454 100-97-01301-02.

UFSAR Section 6.1.3.2 will be revised to reflect the sump pil response with the RSGs as part of the SGR project UFSAR update. The SGR safety evaluation will be revised to renect the results of this evaluation. This action will be tracked by NTS item # 454 100-97-01301-03, UFSAR Figure 6.1 1 will be revised to reacct the containment sump water volumes with the RSGs as part of the SGR project UFSAR update. 'the SGR safety evaluation will be revised to re0cct the results of this evaluation. This action will be tracked by NTS item # 454100 97-01301-04.

A thorough understanding of and strict adherence to the requirements of the 10CFR50.59 process is necessary to ensure an adequate safety evaluation. Initiatives underway by Comed include advanced training to provide a depth of understanding for those performing and reviewing safety evaluations. This training focuses on the need to identify potential impacts associated with changes, This training stresses the requirement for adequate research and documentation to provide the bases of safety evaluation conclusions. Appropriate individuals from the SGR project engineering organization have successfully completed this training.

DATE WilEN FUI 1, COMPI,I ANCE WILI, BE ACillEVED Full compliance will be achieved on December 31,1997 with the 6nal issuance of the RSG safety evaluation.

(p 97byhrs 9702717)

ATTACHMENT 11 VIOL,ATION (50-454/97013-06a.b) 10 CFR Part 50, Appendix B, Criterion 111, " Design Control," requires in part, that design control measures shall provide for verifying or checking the adequacy of design.

a.

Contrary to the above, as of September 3,1997, licensec design change control measures for verifyin the adequacy of the replacement steam generator modification had been inadequate tor BWI Calculation 222 7720-A13 " Engineering Calculations -

Byron /Braidwood RSG - Primary Fluid Volumes vs. Height," Revision 0, issued April 5, 1995 in which the new reactor coolant system volume had been incorrectly determined.

(50-454/97013-06a(DRS))

b.

Contrary to the above, as of September 18,1997, licensee design control measures for-verifying the adequacy of the replacement steam generator modification had been inadequate for FTl calculation 51-1266158-01,"RSG AITV (Auxiliary Feedwater)

Cooldown Requirements,: Revision 1, issued June 6,1997, in which the licensee had failed to consider the speci6c heat capacity of the replacement steam generators and the heat load of the main feedwater system. (50 454/97013-06b(DRS)).

This is a Severity Level IV Violation (Supplement I)

REASfN FOR TIIE VIOI,ATION a.

Inadequate Design Control - Reactor Coolant System (RCS) Volume Calculation BWI calculation 222-7720-A13, Revision 0, was non-conservative with regard to the calculation of the RSG primary volume. The calculation did not account for hydraulic expansion of the tubes into the tubesheet and also did not calculate or address the increase in volume due to thermal expansion. The reason for the violation can be attributed to lack of attention to detail, b.

Inadequate Design Control-AFW Cooldown Requirements Framatome Tecimologies, Incorporated (FTI) Calculation 51-1266158-01, Revision 1, calculates the additional AFW required to meet Technical Specifications and UFSAR requirements for cooldown. Additional water is required since the RSGs have increased stored energy in: 1) additional primary coolant reass, 2) additional steam generator metal mass, and 3) additional feedwater piping metal rnass. Also, the speci6c heat capacity value used in the FTl calculation was not adjusted for the materials and conditions in the RSG.

Because the calculation was performed using overall conservative assumptions, individual assumptions and non-conservatisms were not documented. The reason for the violation can be attributed to lack of attention to detail and failure to document with i

sufficient detail decisions / assumptions utilized in calculations. The new calculation demonstrates the overall conservative natum of the original calculation with the conclusion that the differential auxiliary feedwater equirement decreased.

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' CORRECTIVE STEPS TAKEN AND RESUI,TS ACillEVI;ll i

a.

Inadequate Design Control-RCS Volume Calculation BWI calculation 222-7720-A13, Revision 0 was revised to address the violation concerns. %c calculation was revised to account for three expansion factors not previously considered that are associated with the RSGs at operating conditions: 1) thermal expansion of the material (tubes and primary head),2) pressure boundary dilation due to the difTerential pressure, and 3) hydraulic expansion of the tubes in the tube sheet.

The calculation was acceptance reviewed by Comed. The revised calculation has been transmitted for use as design input where applicable for other SGR related calculations.

b.

Inadequate Design Control-AFW Cooldown Requirements A new calculation, 32 1266253, has been prepared that rigorously addresses the impact of the RSG on the design and license basis for AFW cooldown. The new calculation demonstrates that the original calculation result was conservative This calculation has been acceptance reviewed by Comed. This result confirms the earlier evaluation that the total AFW requirements for the RSG iemain below the design basis value of 200,000 gallons.

CORRECTIVE STEPS Til AT WII.l BE TAKEN TO AVOID FURTIIER VIOL.ATION a.

Inadequate Design Control - RCS Volume Calculation A reviuv of all project documents was performed to identify all calculations that utilized RCS volume as a design input This review covered RCS volume inputs from all sources notjust BW1 calculation 222-772B A13, Revision 0. These calculations will be revievred and revised, as necessary, to ensure acceptable results due to the RCS volume increase.

This action will be traclsed by NTS ite# 454-100-97-01306-01.

The revised calculation results also support a supplement to an existing Technical Specification amendment request. The amendment request primarily deals with the change in P., but also includes the change to the " Design Features" for the RCS, Technical Specification Section 5.4.2 which specifics the RCS volume. The supplement mformation corrects the value for the RSG RCS volume and documerts the analysis of impacts.

Comed also conducted an additional review of a sampling of B&W calculations to ensure technical accuracy. No deficiencies were identified that impact calculation conclusions.

As part of the UFSAR update, a detailed table of RCS total and component volumes, hot and cold. for both units will be included, This table will provide clear design basis parameters for utilization in future applications. This action will be tracked by NTS item # 454-100-97-01306-02.

(p '97by hrs 970279 9) l

1 b.

Inadequate Design Control AFW Cooldown Requirements An additional review of UFSAR non-chapter 15 calculations was performed by FTl ne additional review did not identify any deficiencies that impacted calculation conclusions.

DATE WilEN FIII,1, COMPI, LANCE WII,1, HE ACIIIEVED a.

Inadequate Design Control - RCS Volume Calculation Full compliance was achieved on 10/21/97 when the BWI calculation 222 7720-A13, Revision I was reviewed and accepted by Comed.

b inadequate Design Control-AFW Cooldown Requirements Full compliance was achieved on 11/21/97 when the affected calculation had been replaced with the new calculation, reviewed and accepted by Comed and the results demonstrated that the previous calculation was conservative.

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