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Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-454/97-13
	ML20203G164
Person / Time
Site:	Byron
Issue date:	12/12/1997
From:	Grobe J; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	Graesser K; COMMONWEALTH EDISON CO.
References
	50-454-97-13, NUDOCS 9712180066
	Download: ML20203G164 (2)
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Deccrrber 12, 1997

Mr. K. Graesser

Site Vice President

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Byron Nuclear Power Station

Commonwealth Edison Company L

4450 North German Church Road

Byron,IL <1010 ;

SUBJECT: NOTICE OF VIOLATION (NRC INSPECTION REPORT NO. 50-454 97013(DRS))

Dear Mr Graesser:

This will acknowledge receipt of your letter dated November 24,1997, in response to

our letter dated October 30,1997, transmitting a Notice of Violation associated with the above

mentioned inspection report at the Byron Nuclear Power Station. We hcve reviewed your

corrective actions and have no further questions at this time, These corrective actions will be

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examined during future inspections.

Sincerely,

Original Signed by J. A. Grobo

John A. Grobe, Director

Division of Reactor Safety

Docket No. 50-454

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Enclosure: Ltr 11/24/97, K L. Graesser,

Comed, to USNRC w/ attachment k

See Attached Distribution lIlk !!,. ,

DOCUMENT NAME: G:DRS\BYR12097,DRS

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OFFICE Rlll/DRS lG Rlll/DR$ , . l A/ Rlll/DRP,- /q l/ Rill /DRS_ . , l

NAME Holmberg:s("*' Gavuta hW Jordan W/)/ GrobeQT)

DATE 12/6*/97 12/f/971 12/9/97- 7 12/1491

OFFICIAL RECORD COPY

971,2190066 971212

PDR- ADOCK 05000454

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K. L Graesser 2

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cc w/ encl: O. Kingsley, Nv itar Generation Group

President & Cniof Nuclear Officer

M. Wallace, Senior Vice President,

Corporate Services

H. G. StLnley, Vice President,

PWR Operations

Liaison Officer, NOC BOD

D. A. Sager, Vice President,

Generation Support

D. Farrar, Nuclear Regulatory

Services Manager

1. Johnson, Licensing Operations Manager

Document Control Desk - Licensing

K. Kofron, Station Manager

D. Brindle, Regulatory Assun! x

Supervisor

Richard Hubbard

Nathan Schloss, Economist,

Office of the Attorney General

State Llalson Officer

State Liaison Officer, Wisconsin

Chairman, Illinois Commerce Commission

QlstributioD

Docket File w/enci Rlli PRR w/enct Rlli Enf. Coordinator w/enct

PUBLIC IE-01 w/enci SRI, Byron w/ encl TSS w/ encl

LPM, NRR w/ encl J. L. Caldwell, Rlll w/ encl R. A. Capra. NRR w/ encl

DRP w/onci A. B. Beach, Rlll w/9nct DOCDESK w/enci

DRS w/onci CAA1 w/enci

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November 24,1997

LTR: Byron 974279

FILE: 1.10 0101

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U. S. Nuclear Regulatory Commission

Washington, DC 20$$$

Attention: Document Control Desk

Subject: Dyron Nuclear Power Station Umt i

Response to Notice of Violation

Inspection Report No. 50-454/97013

NRCJMLet Number 50-454

Reference; John A, Grobe letter to Mr. Graesser dated October 30,1997,

transmitting NRC Inspection Report 50-454/97013

Enclosed is Commonwealth 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. Comed's response is provided in the

attachments.

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This letter contains the following commitments:

1) As a result of the identification of two items with inadequate safety evaluations, additional

reviews has e been initiated to focus on possible unidentified impacts of RSG changes on

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

conclusions of no impact are adequately documented.

2) Revise UFS AR Section 5.4.7," Residual lleat Removal System," to discuss the quantitative

evaluation on RilR performed as part of the SGR project UFSAR update.

3) Revise UFSAR Section 6.1.3 2 to rc0cet the sump pH response with the RSGs as part of the

SGR project UFSAR update.

4) Revise UFSAR Figure 6.1-1 to re0cet the centainment sump water volumes with the RSGs :,4

part of the SGR project UFSAR update.

5) Review all calculations 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 units.

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B) Ton Ltr. 974279

November 24,1997

Page 2

If your staff has any questions or comments concerning this letter, please refer them to Don

Drindle, Regulatory Assurance Supervisor, at (815)234 5441 cxt. 2280.

Respectfully,

. U

./n K. L Grac s

U Site Vice President

'lyron Nuclear Power Station

KLG/DD/rp

Attachment (s)

cc: A. D. Deach, NRC Regional Administrator - Rill

G. F. Dick Jr., Byron Project Manager - NRR

Senior Resident inspector, Dyron

M. J. Jordan, Reactor Projects Chief- Rlli

F. Niziolck, Division of Engineering - IDNS

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M: D. L Farrar, Nuclear Regulatory Services Manager, Downers grove

Safety Review Dept, c/o Document Control Desk,3'd Floor, Downers Grove

DCD-Licensing, Suite 400, Downers Gro.e.

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ATTACHMENT I

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VIOLATION (50-454/97013-01a.b)

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10 CFR 50.59(a)(1) states, in part, that a licensee may make changes to the facility as described in i

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the safety analysis repon without prior Commission approval unless the proposed change !'

involves an unreviewed safety quecion.

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

facility as desenbed 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. i

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uc Byron Updated Final Safety Analysis Report (UFSAR) Section 5.4.7.1 " Design Basis" stated

that " .., the RilRS l 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 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ."

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nc Byron UFSAR Section 6.l.3 " Post Accident Chemistry," Section 6.1.3.1 "Steamline Break ,

lnside Containment" and Section 6.1.3.2 " Main Feedwater Line Break Inside Containment"

desenbcd the effect of a main steamline break (MSLB) and main feedline break (MFLB) on

containment sump level and pH.

a. 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

RHR system for the replacement steam generator (RSG) modification, constituted an

unreviewed safety question. Specifically, the evaluation was deficient because it failed to

consider the cfTeet of the increased heat load (associated with the increased reactor

volume for the RSG modification) on the RHR 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 pH for the RSG modification, constituted an unreviewed safety question.

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

secondary side mass inventory and larger feedwater break arca on containment sump and

pit level under a MSLB or MFLB, (50454/97013 Olb(DRS))

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His is a Severity Level IV Violation (Supplement 1)

REASON FOR THE VIOLATION

a, Inadequate Safety Evaluation RilR Performance

%c RHR system is capable of reducing the temperature of the reactor coolant per

UFSAR Figures 5.4 6 and 5.4 7 for dual and single train operation, respectively (UFSAR

Section 5.4.7), ne initial review of system impacts did not identify impact on RHR

performance as requiring quantitative analysis becausc: 1) the integrated decay heat is

much larger than the added heat due to the increased RSG volume and metal mass, and 2)

quahtatively the RSG impacts are small compared to the existing margin between

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

was a failure to document engineering judgement, and therefore, a documented ,

quantitative analysis was not performed to support the conclusion of no impact.

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b. Inadequate Safety Evaluation Containment Sump Level & Sump pH

Calculations were performed to assess the impact of the RSG design on UFSAR Section

61.3," Post Accident Chemistry" Both the minimum and maximum pH calculations

were performed for the appropriate limiting accident conditions. He results were

veri 0cd to be within the acceptat>le pil band as specified in the plant Technical

Specifications, liowever, 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 specincally 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. However, only the limiting case for containment i

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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. Changes in the MFLB transient response (CS actuation)

were not specifically identined and documented in the safety evaluation regarding

containment sump water volumes.

The initial review of these transients did not identify the potential impact as requiring

quantitative analysis for containment sump level and pH 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 quantitative analysis was not performed

to support the conclusion of no impact.

CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED

a. Inadequate Safuy Evaluation RHR Performance

A quantitative analysis was performed to document the conclusion of no impact. He

RSGs contain more primary liquid mass and more metal mass than the original Steara

Generators (OSGs). %cre 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 UFSAR

curves. Single train and dual train RHR cooldown were analyzed to evaluate the effects

on RHR system performance. For RHR operations with only a single operable RHR

train, the measure of RHR performance is based on the duration required to cool the

RCS from 350'F to 200'F (UFS AR Figure 5.4-7). In this case the RSGs cause an

increase of approximately 0.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months to the existing 39 hour4.513889e-4 days 0.0108 hours 6.448413e-5 weeks 1.48395e-5 months single train RHR cooldown

time.

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For RilR operations with two operable RilR trains, the measure of R11R 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 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months to the existing

30.3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months dual train RiiR cooldown time. His is within the design basis ofless than 36

hours..

b. Inadequate Safet) Evaluation - Containment Sump Level & Sump pli

A quantitative evaluation of the MFLB transient was performed to determine the impact

of the revised Main Feedwater/ Auxiliary Feedwater configuration on the containment

sump plL ne current UFSAR evaluation (Section 6.1.3.2," Main Feedwater Line

Break") states that the MFLB transient does not initiate CS since containment pressure

remains lower than the CS actuation setpoint. The sump pli, therefore, is considered to

be the pil of the condensate Guid assumed in the UFSAR. He evaluation performed for

the RSGs demonstrates that a MFLB could elevate containment pressure sufficiently to

actuate CS, thus causing the event to respond similar to the MSLB transient. liowever,

the mass and energy release for the MFLB with the RSGs is less than that of the MSLB

transient, therefore, the MSLB remains the more limiting transient for the containment

environmental conditions. Ec current UFSAR evaluation for MSLB (Section 6.1.3.1)

indicates a constant pil for the CS fluid that is higher than in condensate fluid (Section

6.1.3.2). Herefore, the impact of th: RSGs on the MFLB is that the containment sump

pil could increase to a value consistent with the MSLB transient due to the actuation of

CS. De amount of additional RSG secondary side mass is not sufficient to appreciably

reduce the pil of the sump fluid. This change, therefore, has no safety significance sinca

the sump pil remains within the limits acceptable for post accident conditions spaified

in the Technical Specifications.

Also, the impact of the RSGs on the containment sump water volumes was evaluated for

both the MSLB and the MFLB. In the case of the MSLB, the difference for the RSG case

is an increase in secondary side mass (approximately 22,000 lbs). Relative to the total

volume of Guid 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 Figurc 6.1 1. UFSAR Figure 6.12 provides the containment sump water

volume following a MFLB accident. Since Unit I actuates CS with the RSGs, the

containment watet volume in Figure 6.12 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 1

MFLB. Both the MSLB and MFLB transients have containment sump water volumes

below the limiting case for containment flooding (LB LOCA) documented in UFSAR

Attachment D3.6 His change, therefore, has no safety significance since the sump water

volumes remam within the maximum volume acceptable for post-LOCA conditions.

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CORRECTIVE STEPS Til AT Wii.1, BE TAKEN TO AVOID

FURT11ER VIOI.ATION

a, b. Inadequate Safety Evaluation - RiiR Performance and Containment Sump Level

& Sump pil

As a result of the identification of two items with inadequate safety evaluations,

additional reviews have been initiated. Rese re-reviews focus on possible unidentified

impacts of RSG changes on Safety Systems. They include: 1) A re-review of select

UFSAR and Technical Specification sections, and 2) A re-review of FTI's "NSS and

BOP Systems Review"(document $1 1239285-03). These additional reviews will ensure

that, as a minimum, the bases for conclusions of no impact are adequately documented.

He SGR safety evaluation will be augmented, if required, to tellect the review results so

that the bases of conclusions are readily accessible to the reviewer. His action will be

tracked by NTS item # 454 100 97-01301-01.

UFSAR Section 5.4.7, " Residual licat 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. His action will be tracked by NTS item # 454-100-97-01301-02.

UFSAP c-dion 6.1.3.2 will be revised to reflect the sump pli response with the RSGs as

part of the SGR project UFS AR update. The SGR safety evaluation will be revised to

reflect 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 reflect the containment sump water volumes with

the RSGs as part of the SGR project UFSAR update. De SGR safety evaluation will be

revised to ecflect the results of this evaluation. This action will be tracked by ;4TS .

item # 454-100-97-01301-04.

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 inch ie 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. Appropnate individuals from the SGR project engineering organization

have successfully completed this training.

DATE WilEN FUI.1,' MPl.l ANCE Wil.1, BE ACillEVED

Full compliance will be a aeved on December 31,1997 with the final issuance of the RSG safety

evaluation.

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ATTACllMENT II

VIOLATION (50-454/97013-06a,b)

10 CFR Part 50, Appendix B, Criterion 111, "Desiga 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, licensee design change control measures

for verifying the adequacy of the replacement steam generator modification had been

inadequate for BWI Calculation 222-7720-A 13 " 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 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. (50-454/97013-06b(DRS))

This is a Severity Level IV Violation (Supplement 1)

REASON FOR TIIE VIOLATION

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 f at 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

attnbuted to lack of attention to detail.

b. Inadequate Design Control- AFW Cooldown Requirements

Framatome Technologies, 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 mass, 2) additional steam gnerator metal

mass, and 3) additional feedwater piping metal mass. Also, the specific heat capacity

value used in the FTl calculation was not adjusted for the matenals and conditions in the

RSG.

Because the calculation was performed using overall conservative assumptious,

individual assumptions and non-conservatisms were not documented. The reason for the

violation can be attnbuted to lack of attention to detail and failure to document with

sufIicient detail decisions / assumptions utilized in calculations. The new calculation

demonstrates the overall conservative nature of th: original calculation with the

conclusion that the differential auxiliary feedwater requirement decreased.

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CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED

a. Inadequate Design Control - RCS Volume Calculation

HW1 calculation 222-7720-A13, Revision 0 was revised to address the violation

concerns. De calculation was revised to account for three expansion factors not l

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 differential pressure, and 3) hydraulic expansion of the tubes in the tube sheet.

Tric calculation was acceptance reviewed by Comed. He revised calculation has been

transmitted for use as design input where applicabic for other SGR related calculations. ,

b. Inadequate Design Control AFW Cooldown Requirements

A new calculation,32 12 6253, 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. his calculation has i

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been acceptance reviewed by Comed his result connrms the earlier evaluation that the

total AFW requirements for the RSG remain below the design basis value of 200,000

gallons.

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CORRECTIVE STEPS THAT WILL BE TAKEN TO AVOID FURTIIER VIOI,ATION

a. Inadequate Design Control - RCS Volume Calculation ,

A review of all project documents was performed to identify all calculations that utilized

RCS volame as a design input. This review covered RCS volume inputs from all sources

notjust BWI calculation 222 7720 A13, Revision 0. These calculations will be reviewed

and revised, as necessary, to ensure acceptable results due to the RCS volume increase.

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

De revised calculation results also support a supplement to an existing Technical

Specincation amendment request. De amendment request pnmarily deals with the

change in p., but also includes the change to the " Design Features" for the RCS,

Technical Soccification Section 5.4.2 which specines the RCS volume. The supplement

information corrects the value for the RSG RCS volume and documents 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 umts will be included. This table will provide clear design basis

paramete s for utilization in future applications. This action will be tracked by NTS

item # 454100 97-01306-02.

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b. Inadequate Design Control- AFW Cooldown Requirements

An additional review of UFS AR non-chapter 15 calculations was performed by Ffl He

additional review did not identify any deficiencies that impacted calculation conclusions.

DATE WilEN FUI,I, COMPI IANCE Wil,L llE ACIIIEVED

a. Inadequate Design Centrol RCS Volume Calculation

Full com,,liance 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 Al%' Cooldown Requirements

Full complians 'ws achieved on i1/21/97 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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ML20203G164

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