ML20203D013

Supplements TS Amend Re Primary Containment & RCS Volume,Due to Error Discovered in Current TS W/Regards to Total RCS Volume & Correction to Increase in RCS Volume Associated W/Unit 1 SGs Accounting for Hot Conditions

ML20203D013
Person / Time
Site:	Byron, Braidwood
Issue date:	12/09/1997
From:	Hosmer J COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20203D017	List: ML20203D013 ML20203D036
References
NUDOCS 9712160081
Download: ML20203D013 (8)

Text

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Commonwealth Edrum Company j

• p llraidwmed Generating Station

• %, , . Route et. Iku H 6 tiraceville. IL 60607@619 Tel H15-45&2He1 December 9,'1997 United States Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Document Control Desk

Subject:

Byron Nuclear Power Station, Units 1 & 2 Facility Operating Licenses NPF-37 & NPF-66 NRC Docket No. 50-454 and 50-455 Braidwood Nuclear Power Station, Units 1 & 2 Facility Operating Licenses NPF-72 & NPF-77 NBC Docket No. 50-456 and 50-451 Supplement to Technical Speification Amendment Pertaining to Primary Containment and Reactor Coolant System Volume -

References:

1. J. Hosmer (Comed) Letter to USNRC, Primary Containment and Reactor Coolant System Amendment, dated January 30,1997

2. USNRC Request for Additional Information legarding Primary Containment and Reactor Coolant System, dated March 20,1997

3. J. Hosmer (Comed) Letter to USNRC, Primary Containment and Reactor Coolant System RAI Response, dated May 23,1997

4. J. Hosmer (Comed) Letter to USNRC, Primary Containment and Reactor Coolant System RAI Response to Question M, dated August 8,1997

5. J. Hosmer (Comed) Letter to USNRC, Primary Containment and Reactor Coolant System Update to RAI Response, dated November 11,1997 In Reference 1, Comed submitted a request for a License Amendment in accordance with 10CFR50.90 regarding the revised calculated peak containment pressure, P., and the increased RCS volume. These changes are associated with ti.e replacement steam generators to be installed on Byror and Braidwood Units 1. Subsequent to that submittal, he NRC Staffissued a Request for Additional Information regarding the proposed change t Jeference 2). Comed responded to that request in References 3,4 and 5.

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p U.S. Nuclear Regulatory Commission December 9,1997 This supplement is needed due to an error diu: overed in the current technical ,

specifications with regards to total RCS volume and a correction to the increase in RCS volume associated with the Unit 1 Replacement Steam Generators (RSGs) accounting for hot conditions. These changes affect Unit I and Unit 2 at both Byron and Braidwood.

During the process of preparing this technical specification revision, Comed evaluated the validation of the accident analysis (Spurious Safety Injection) related to the reactor coolant volume change. A potential conflict between the UFSAR assumptions and the Emergency Operating Procedures (EOPs) was discovered. Specifically, the EOPs do not provide explicit direction for the manual action of the Power Operated Relief Valve (PORV) as stated in the FSAR. Comed will determine whether to revise the UFSAR to

, credit automatic actuation of the PORVs, or whether to revise EOPs to support manual action, and will inform the NRC cf the results of this determination by 12/19/97.

This determination will not change the content of the amendment request or the results of this supporting analysis.

Enclosed is:

Attachment A: Detailed Description of the Proposed Changes Attachmer.t B-1 A: Byron Marked-Up pages Attachment B-2A: Braidwood Marked-up pages

> Attachment C: Evaluation of Significant Hazards Consideration Attachment D: Environmental Assessment Please address any comments or questions regarding this information to this atlice.

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Sincerely, l

+ S hw 8 John B. Hosmer Vice Preside <t cc: Regional Administrator-RIII Byron /Braidwood Project Manager - NRR Senior Resident Inspector- Byron Senior Resident Inspector - Braidwood -

OfIice of N.sclear Safety -IDNS Kadshttmdsyp/pwnwdoc2

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I ATTACHMENT A 4 DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES FPF-37, NPF-66, NPF-72, AND NPF-77 A. DESCRIPTION OF THE PROPOSED CHANGE (SUPPLEMENTAL INFORMATION By letter dated January 30,1997, Commonwealth Edison (Comed) proposed to revise Technical Specifications (TS) 1.0, " Definitions," 3/4.6.1, " Primary Containment" and associated Bases, and 5.4.2," Reactor Coolant System Volume," for Byron Nuclear Power Station (Byron) and Braidwood Nuclear Power Station (Braidwood) to support steam generator replacement. Additionally, several editorial changes were also proposed to improve clarity and consistency of the TS. Comed will be replacing the original Westinghouse D4 steam generators (OSGs) at Byron and Braidwood with Babcock and Wilcox Intemational (BWI) steam generators. The replacement steam generators (RSGs) increase the Reactor Coolant System (RCS) volume which results in a higher calculated peak containment pressure (Pa ) value. Subsequent to the original submittal, issues penaining to the RCS volume have been raised which require that the original submittal be supplemented. The supplemental change affects only the RCS volume reported in TS Section 5.4.2.

The proposed changes associated with this supplement are discussed in detail in Section E of this attachment. Affected TS pages showing the proposed volume changes for this supplement are included in Attachments B-1 and B-2 for Byron and Braidwood, resnectively. Improved Technical Specifications (ITS) are unaffected by this supplement since the value for RCS volume is not retained in ITS.

B. DESCRIPTION OF THE CURRENT REQUIREMENT The description of the current TS requirements remains unchanged from the .~anuary 30, 1997 submittal. Only the section pertaining to this supplement is provided here for reference.

TS 5.4.2 indicates 12,257 cubic feet for the total water and steam volume of the Reactor Coolant System at a nominal Tavg of 588.4 0F for each unit. This information is provided as part of the " Design Features" section of the Byron and Braidwood Technical Specifications and does not represent a limiting condition for operation.

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N C.' BASES FOR THE CURRENT REQUIREMENT The bases for the current TS requirements remains unchanged from the January 30,1997 submittal. Only the section pertaining to this supplement is provided here for reference.

TS 5.4.2 is a statement of the volume of the reactor coolant system with the plant in its original configuretion, which includes the Westinghouse Model D4 or D5 steam generators.

D. NEED FOR REVISION OF THE REQUIREMENT (SUPPLEMENTAL INFORMATION)

Each of the RSGs has a larger primary side volume than the OSGs. TS 5.4.2 provides information on the total RCS volume and requires a revision to reflect the volume increase associated with the Steam Generator Replacement Project. The original submittal dated January 30,1997 addressed only the increase associated with the RSGs.

Two issues have been identified which make the original submittal of the proposed revision to TS Section 5.4.2 in need of revision.

First, the B&W calculation for the RSG volume was reviewed as part of an NRC Region III Inspection.' As a result, it has been determined that the change in volume previously proposed for the RSGs is based on a cold volume and did not properly account for

expansion factors at operating conditions (Tav3 at 588.4 F). As a result, the RSG 3

primary side volume reported in the original submittal (1251 ft ) must be increased by an additional 29 ft3, Second, Comed was notified by Westinghouse that the current TS value for Unit I and 2 total RCS volume (12,257 cubic feet) is not correct. The correction to the current TS value is an increase of 83 ft 3 to 12,340 ft 3. This increase applies to both units.

O As a result of the RCS volume increase with the RSGs, the mass and energy release during the blowdown phase of the large break Loss of Coolant Accident (LOCA)is increased for Unit 1. Additionally, the heat transfer rate of the RSGs is greater than the OSGs, and the RSGs will operate at a slightly higher secondary side pressure than that for the OSGs. Consequently, the steam enthalpy exiting the bre_al during the reflood period, for the RSG, will be greater than that for the OSG. This results in an increase in the containment peak pressure, Pa. The January 30,1997 submittal identified an increase in Pa from the current value of 44.4 psig to a value of 47.8 psig. This increase was calculated in the Containment LOCA Analysis performed by Framatome Technologies, Inc. (FTI) in support of the steam generator replacement project.

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S Margin is available in the Pa determination for the RSGs and the OSGs to offset the small volume increases addressed by this supplement. This margin will be explained in Section F of this attachment.

E. DESCRIPTION OF TIIE REVISED REQUIREMENT (SUPPLEMENTAL INFORMATION)

All descriptions of the revised requirements presented in the January 30,1997 submittal remain unchanged except for TS 5.4.2. The levisions to TS 5.4.2 are as follows:

Technical Specification Design Features Section 5.4.2 will be revised to incorporate the corrected total RCS volume and the corrected additional RCS volume associated with the RSGs.

For both Byron and Braidwood, the proposed changes to TS Section 5.4.2 are:

Revise the current RCS total volume from "12,257 cubic feet" to "12,340 cubic feet."

- Revise the additional RSG volume to "1280 cubic feet at a nominal Tavg of 588.4 F."

F. BASES FOR THE REVISED REQUIREMENT (SUPPLEMENTAL INFORMATION)

The January 30,1997 submittal provided the bases for the proposed change in Pa . The bases centered around the impact of the increased RCS volume associated with the RSGs.

This supplement further increases the RCS volume with the RSGs. Additionally, this supplement increases the total RCS volume for both Byron and Braidwood units to account for a calculational error in the determination of the TS value for the original RCS

. volume. The bases information provided in this section explains the proposed change in the RCS volumes and why the values of P a (current value and proposed Unit i value with RSGs) remains unchanged.

A11/itional Unit I and 2 RCS Volyme The TS states that the total volume of steam and water in the RCS is 12,257 ft3 at nominal operating conditions. Westingho"se notified Comed that there is an error in this number and that the correct value i,hould be 12,340 3ft . This applies to all four units at Byron and Braidwood. Westinghouse reviewed the original calculation of Pa for impacts of the higher volume. As a result of conservatisms, the original containment analysis calculation used an RCS volume of 12.619 ft3. The original containment analysis bounds the increase in RCS volume identified by Westinghouse and the current value of 44.4 psig for Pais unaffected.

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l Additional RSG Volume Ti e January 30,1997 submittal reported a chinge in RCS volume v.1251 ft 3 to account for the larger primary side volume of the RSGs. inis change in volume along with thermal hydraulic differences was used to calculate a change in Pa due to the RSGs.

However, the 1251 ft3 value was calculated as a cold volume, not at operating conditions.

This cold volume was used to calculate the changes in mass and energy release due to the RSGs and the result was added to the previous Westinghouse mass and energy release for the total RCS volume. A detailed discussion of this approach is presented in the May 23, 1997 Comed response to an NRC Request for Additional Information concerning the January 30,1997 submittal. The RCS volume basis in the earlier submittal for the determination of Pa was 13,870 ft3 (12,619 ft3 + 1251 ft3),

The additional 29 ft 3 increase in RCS volume associated with the RSGs is due to expansion factors. These expansion factors include: 1) thermal growth of the Inconel 690 tube material,2) pressure boundary dilation due to the primarf to secondary differential pressure, and 3) increase in ID of the steam generator tubes in the tubesheer stea due to the hydraulic expansion during the manufacturing process. The correcsd value for the increase in RCS volume due to the RSGs at operating conditions is 1280 ft3.

As previous!isuted, the corrected value for the existing RCS volume (at normal operating conditions)is 12 340 ft 3. The total RCS volume (at normal operating conditions) for the RSGs, then,is 13,620 ft3. The approv:d Westinghouse methodology adds an uncertainty of 1.4% for conservatism, resulting in a corrected value of 13,811 ft3 for calculatic tal purposes. Since this value is less than F.e 13,870 ft3 used to determine Pa in the January 30,1997 submittal, the 47.8 psig value is bounding and does not need to be changed as a result of the corrected RSG volume.

The bases discussions provided in the original January 30,1997 submittal remain unchanged except for the RCS volume section, TS 5.4.2. The revised basis is provided below.

The re.ised TS Design Features, TS 5.4.2 accounts for 1) the Unit 1 RSGs addition of a total of 1,280 cubic feet of vater volume to the RCS, and 2) a correction to the current RCS volume value to 12,340 cubic feet at nominal operating conditions. The effect of the increase in total RCS volume due to the RSGs was evaluated for all UFSAR, Chapter 15 accidents. The only impact on the Technical Specif, cations resulted from the increased mass and energy release following a LBLOCA event. The increased RCS volume was a cor.tributoi to the increase in the maximum calculated primary containment pressure, Pa , value. Therefore, the basis for acceptabiiity of the revised RCS volume is addressed by the basis for the Pa increase.

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G. IMPACT OF THE PROPOSED CHANGE (SUPPLEMENTAL INFORMATION) 1 This supplement to the January 30,1997 submittal addresses the changes necessary to l account for increased RCS volumes in both Units 1 and 2 at Byron and Braidwood. As l 2xplained in Section F of this attachment, the increases in RCS volume do not alter either 4

the proposed change in Pa for ute RSGs nor the current OSG value of Pa. Therefore, the impact of the proposed change as discussed in the original submittal remains unchanged.

Any impacts on Unit 2 from an increase in RCS volume, other than Pa, are bounded by the evaluations presented for Unit I with the RSGs since the total increase in volume with the RSGs is significantly greater than the proposed change in the Unit 2 RCS volume.

The potential impact of increase RCS volumes on post-LOCA conditions has also been evaluated for both units The impact on maximum 'looding levels, post-accident sump pH, and post-accident hydrogen generation has be erformed. These parameters remain within allowable values.

H. SCHEDULE REQUIREMENTS The Byron Unit 1 Steam Generator Replacement Outage (SGRO)is scheduled during the eighth refuel outage (BiR08). The Braidwood Unit 1 SGRO is scheduled during the

, seventh refuel outage (A1R07). Approval of this change (as supplemented by this attachment) is requested as soon as possible to support the current outage schedule for the l lead steam generator replacement station which is Byron Unit 1. I i

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l ATTACHMENT B-1 l MARKED UP PAGES COR .

?- PROPOSED CHANGES TO AP?ENDIX A TECHNICAL SPECIFICATIOdS OF FACILITY OPERATING LICEhJES l- NPF-37, NPF-66 BYRON STATION UNITS 1 & 2 REVISED SUPPLEMENT PAGE:

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