ML20212F525

Notice of Violation from Insp on 970918.Violation Noted: Licensee Had Not Performed Adequate SE to Determine Whether Impact on Design Basis of RHR Sys for Replacement SG Mod, Constituted Unreviewed Safety Question

ML20212F525
Person / Time
Site:	Byron
Issue date:	10/30/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20212F518	List: IR 05000454/1997013 ML20212F514 ML20212F525
References
50-454-97-13, NUDOCS 9711040321
Download: ML20212F525 (2)

Text

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z NOTICE OF VIOt.ATION Commonwealth Edison Company Docket No. 50-454 Byron, Unit 1 License No. NPF-37 During an NRC inspection completed on September 18,1997, violations of NRC requirements were identified in accordance with the " General Statement of Policy and Procedure for NRC -

Enforcement Actions," NUREG 10, the violations are listed below:

1.

10 CFR 50.59A)(1) states, in part, that a licensee may make changes to the facility as described in the safety analysis report without prior Commission approval unicss 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 facility as described in the safety analysis report and that records must include 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 (UFSAR) Section 5.4.7.1 " Design Basis" stated that "..., the RHRS [ Residual Heat 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 "Postaccident Chemistry," Section 6.1.3.1 "Steamline

' Break Inside Containment" and Section 6.1.3.2 " Main Feedwater Line Break Ins'de Containment" described the effect of a main steamline break (MSLB) and main feedline break (MFLB) on containment sump level and pH.

Contrary to the above:

a) As of September 9,1997, the licensoe had not performed an adequate safety evaluation to determine whether the impact on the design basis of the RHR system for the replacement steam generator (RSG) modification, con =tituted an unreviewed safety question. Specifically, the evaluation was deficie.,t because it failed to consider the effect of the increased heat load (associated with the increased reactor coolant volume for the RSG modification) on the RHR system performance.

' b)' . 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 pH for the RSG modification, constituted an unreviewed safety question.

- Specifically, the evaluation was deficient because it failed to consider the RSG increased secondary mass inventory and larger feedwater break area on the '

containment sump and pH level under a MSLB or MFLB.

This is a Severity Level IV Violation (Supplement 1).

9711040321 PDR 971030 0 ADOCK 05000454 PDR

Notice of Violation 2 2.

10 CFR Part 50, Appendix B, Criterion ill, " Design Control," requires in part, that design control measures shall provide for verifying or checking the adequacy of design.

- Contrary to the above; a) As of September 3,1997, licensee design control measures for verifying the adequacy of the replacement steam generator modification had been inadequate for BW1 Calculation 222-7720-A13 " Engineering Calculations - Byron /Braidwood RSG , Primary Fluid Volumes vs. Height," Revision 0, issued April 5,1995 in which the new reacto" coolant system volume had been inc b) As September 18,1997, licensee design control measures for verifying the adequacy of the replacement steam ge 1erator modification had been inadequate for FTl calcuiation 51- 1266158-01, "RSG AFW [ Auxiliary Feedwater) Cooldown Requirements" Revision 1, issued June 6,1997, in which the licensee had failed to consider the specific heat capacity of the replacement steam generators and the heat load of the main feedwater system.

This is a Severity Level IV Violation (Supplement 1).

Pursuant to the provisions of 10 CFR 2.201, Commonwealth Edison is hereby required to submit a written statement of explanation to the U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington D.C. 20555 with a copy to the Regional Administrator, Regica til, and a copy to the NRC Resident inspector at the facility that is the subject of this Notice of Violation (Notice), within 30 days of the date of the letter transmitting this Notice. This reply should be clearly marked as a " Reply to a Notice of Violation" and should include for each violation:

(1) the reason for the violation or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violatior', and (4) the date when full compliance will be achieved.

Your response may reference or include previous docketed correspondence,if the correspondence adequately addresses the required response. If an adequate reply is not '

received within the time specified in this Notice, an order or a Demand for Information may be issued as to why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.

Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR without redaction. However, if you find it necessary to include

!' such information, you should clearly indicate the specific information that you desire not to be placed in the PDR, and provide the legal basis to support your request for withholding the information from the public.

Dated at Lisle, Illinois, thigo day of ocfc4JL997 -