ML20203D036

Proposed Tech Specs Pages Correcting Errors Discovered in Current TS W/Regards to Total RCS Volume & Correction to Increase in RCS Volume Associated W/Unit 1 Replacement SGs Accounting for Hot Conditions

ML20203D036
Person / Time
Site:	Byron, Braidwood
Issue date:	12/09/1997
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20203D017	List: ML20203D013 ML20203D036
References
NUDOCS 9712160089
Download: ML20203D036 (9)

Text

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DESIGN FEATURES

. 's, 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain Ig3 fuel assemblies with each fuel assembly l containing 264 fuel rods clad with Zircaloy-4 or ZIRLO, except that limited substitution of fuel rods by filler rods consisting of Zircaloy-4, ZIRLO, or stainless steel or by vacancies may be made if justified by a cycle specific reload analysis. Each fuel rod shall have a nominal active fuel length of 144 inches. The initial core loading shall have a maximum enrichment of less than 3.20 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading or previous cycle loading. The enrichment l of any reload fuel design shall be determinii to be acceptable for storage in either the spent fuel posi or the new fuel vault. Such acceptance criteria

' shall be based on the results.of the CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEi STORAGE RACKS.

1 CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. All control rods shall be hafnium, silver-indium-cadmium, or a mixture of uoth types. All control rods shall be clad with stainless steel tubing.

5.4 REACTORC00LkNTSYSTEM DESIGN PRESSURE AND TEMPERATURE 4

5.4.1 The Reactor Coolant System is designed and shall be maintained:

a. In accordance with the Code requirements specified in Section 5.2 of the UFSAR, with allowance for normal degradation pursuant to the l applicable Surveillance Requirements,

b. For a pressure of 2485 psig, and

c. For a temperature of 650*F, except for the pressurizer which is 680'F.

YOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is '

M,257 tubic feet at a nominal T ft 3@ of 588.4*F. (

J.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

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p PDR e BYRON - UNITS 1 & 2 5-4 Amendment No.78

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f-ATTACHMENT B-2 MARKED UP PAGES FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-72, AND NPF-77

' BRAIDWOOD STATION UNITS 1 & 2 REVISED SUPPLEMENT PAGE:

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DESIGN FEATURES e

5.3 REACTOR CORE L -

FUEL ASSEMBLIES 5.3.1

The core shall contain 1g3 fuel assemblies with each fuel assembly i containing 264 fuel rods clad with Zircaloy-4 or ZIRLO, except that limited l substitution of fuel rods by filler rods consisting of Zircaloy-4, ZIRLO, or stainless reload analysis. steel or by vacancies may be made if justified by a cycle specific i

144 inches. The initial Each fuel rod shall have a nominal active fuel length of

! core loading shall have a maximum enrichment of less than 3.20 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading or previous cycle loading. The enrichment of any reload fuel design shall be determined to be acceptable for storage in l either the spent fuel pool or the new fuel vault. Such acceptance criteria l shall be based on the results of the CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS.

C0KTROL R00 ASSEMB(lLi

-5.3.2 The core shall contain 53 full-length and no part-length control rod

! assemblies. The full-length control rod assemblies shall contain a nominal t

142 inches of absorber material.

indium-cadmium, or a mixture of both types.All control All control rods rods shall shall bebe hafnium, silver-clad i

with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM

DESIGN PRESSURE AND TEMPERATURE

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4 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a.

( In accordance with the Code requirements specified in Section 5.2 of the UFSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, l i b. For a pressure of 2485 psig, and c.

For a temperature of 650*F, except for the pressurizer which is 680 F.

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1 YOLUME 5.4.

( N I Q S cubic feet at a nominil TThe total water and steam volume of the Reuctor C

, of 588.4*F. p~

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

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BRAIDWOOD - UNITS 1 L 2 5-4 Amendment No. [

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9 ATTACHMENT C 4

EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF-FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72, AND NPF-77

Comed has evaluated the originally proposed amendment dated January 30,1997 and this supplement and determined that it involves no signincant hazards considerations.

According to Title 10 Code of Federal Regulations Section 50 Subsection 92 Paragraph c i (10 CFR 50.92 (c)), a proposed amendment to an operating license involves no signincant hazards considerations if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any

[ accident previously evaluated; or i 3. Involve a significant reduction in a margin of safety.

A. INTRODUCTION Commonwealth Edison (Comed) proposes 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. Comed will be replacing the original Westinghouse D4 steam ge nerators 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. The RCS volume for both Unit I and Unit 2 is being changed to correct a calculational error, however, the Pa value for Unit 2 remains unchanged. Additionally, several editorial changes are being made to improve clarity and consistency of the TS.

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H. NO SIGNIFICANT HAZARDS ANALYSIS

1. The proposed change does not involve a signiGcant increase in the probability or consequences of an accident previously evaluated.

Each of the RSGs has a larger RCS primary side volume than the original steam generators (OSGs). As a result of the RCS volume increase, the mass and energy release during the blowdown phase of the large break loss of coolant accident (LBLOCA) is increased. Additionally, the heat transfer rate of the RSGs is greater than the OSGs, and the RSGs will operate at a slight; higher pressure than that for the OSGs Consequently, the steam enthalpy exiting the break during the reflood period, for the R5Gs, will be greater than for the OSGs. This results in an increase in the containment building peak pressure, Pa-The proposed revisions to the Technical Specifications involve the corrected value of the current Unit I and Unit 2 RCS volume and the incremental change in RCS volume for the RSGs. The proposed revisions also involve the defined value of Unit 1 Pa following installation of the RSGs. Several editorial changes are also being made to improve clarity and consistency of the TS.

RCS volume is not an initiator for any event and an increase in volume does not affect any operating margin or requirements. Therefore, increasing the primary volume does not increase the probability of any event previously analyzed.

The current value of Pafor Unit 2 is unchanged due to conservatism in the original analysis. The uvised value of Pa for Unit I continues to be less than the design basis piessure for the containment structure. The change represents only a revision to the containment test pressure for containment leakage testing. Such testing is only performed with the affected unit in the shutdown condition. Therefore, the proposed change in Pa for Unit I does not involve a signincant increase in the probability of an accident previously evaluated.

All accidents in the Updateo Final Safety Analysis Report (UFSAR) were evaluated to determine the effect of an increase in primary volume on accident conseque..ces. The er:nts identined that may be impacted by an increase in primary volume are the Waste Gas System Leak or Failure and LBLOCA. For the Waste Gas System Leak or Failure, the activity of the decay tank is controlled to Technical Specification limits which are unaffected by RCS volume. Therefore, an increase in RCS volume would not increase the offsite dose.

The offsite dose calculation for the LDLOCA is unaffected by the propowd change. The license basis offsite dose calculation is in accordance with NRC Reg Guide 1.4

" Assumptions Used for Evaluating The Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors." This Regulatory Guide states,in part, K.n!Abyrbwd/Sgrp/p.vicup doc-1 l

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"...a number of appropriately conservative assumptions, based on engineering judgment and on applicable experimental results from safety research programs conducted by the AEC." These conservatisms include (but are not limited to) the following assumptions:

Twenty five percent of the equilibtium full power radioactive iodine inventory is immediately available for leakage from the primary containment.

100 (7c of the equilibrium full power radioactive noble gas inventory is immediately <

available for leakage from the primary containment.

The primary containment should be assumed to leak at the (maximum) leak rate specified -

in the technical specifications for the Erst 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at 50% of this value for the remaining 29 days of the accident duration.

The design basis leakage corresponding to a peak contair. ment pressure of 50 psig utilized in '.he design basis accident analysis is 0.10% per day of the containment free air mass.

Therefoie, the offsite dc,3e calculation was performed with a leakage of.1 % per day for day one and .05 % per day for days 2 through 30. Isotopic inventories are unaffected by the increase in reactor coolant volume. Tht.3, the offsite dose is unaffected by the increase in the peak containment pressure. Therefore, this proposed change to Pa does not involve a significant increase in the consequences of an accident previously evaluated.

The editorial changes proposed are for clarity and consistency within the Technical Specifications and do not affect either the probability or consequences of an accident previousi) evaluated.

2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change in RCS volume is a change in a plant parameter within the " Design Fe .tures" section of the Technical Specifications, increasing the RCS volume does not create any new or different failure modes. The existing RCS design requirements cominue to be met.

The revised value of Pa for Uni l following replacement of steam generators coatinues to be less than the design basis pressure for the containment building structure. The change represents only a revision to the test pressure for containment leakage testing.

Such testing is only performed with the affected unit in the shutdown condition.

Therefore, no new or different failure modes are being introduced by modification of the testing parameters.

The editorial changes proposed are for clarity and consistency within the Technical Specifications and do not result in any physical changes to the facility or how it is operated No new or different failure modes are being introduced by these changes.

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Therefore, the proposed change does nct create the possibility of a new or different kind of accident from any accident previously evaluated.

3.' The proposed change does not involve a significant reduction in a margin of safety.

Changing the RCS volume in the Technical Specifications does not reduce the margin of safety. RCS volume is e design feature. An evaluation of all UFSAR accidents was performed to determine the effect of an inercase in RCR volume. This evaluation is summarized as follows:

l An evaluation of the Chemical and Volume Control System Malfunction was performed to determine the effect of the increased RCS volume. The larger RCS volume reduces the reactivity insertion for a given dilution flow rate. Therefore, the UFSAR analyses remain bounding for Byron and Braidwood and there is no '

reduction in the margin of safety.

An evaluation of the inadvertent Actuation of the Emergency Core Cooling System During Power Operation Event was performed to determine the effect of the increased RCS volume due to the RSGs. For this emnt, the injection of borated water causes a negative reactivity insertion, which increases DNBR. For a given Refueling Water Storage Tank (RWST) boron concentration, the larger RCS volume will cause a reduction in the negative reactivity insertion rate as compared to the current UFSAR analysis. Ilowever, negative reactivity would still be inserted and no fuel pins would experience DNB. Additionally, the increased RCS volume was evaluated to determine the effect on pressurizer level following the inadvertent actuation of ECCS and was found to be acceptable.

Therefore, there is no reduction in the margin of safety.

An evaluation of the Small Bieak LOCA was performed to determine the effect of

, increased RCS volume. The additional RCS volume will cause a delay in the loop seal clearing which in tum delays the core uncovery as compared with the UFSAR

, a alysis. A delay in core uncovery reduces the amount of core heatup which ,

results in a lower peak clad temperature (PCT) because the core decay heat would

be less than in the UFSAR analysis. The benefit is considered small, but there is

still a benefit. Therefore. the increased RCS volume does not result in a reduction in the margin of safety.

An evaluation of the Large Break LOCA was performed to determine the effect of increased RCS volume for the RSGs. For a LB LOCA, the increased RCS volume causes the blowdown phase of the event to be longer. Increased blowdown phase, alone, could potentially result in a higher PCT. Hswever, the RSGs also have less resistance to flow due to increased primary side steam generator flow area, which results in a higher blowaown flow compared to the OSGs. The increased blowdown flow will compensate for the longer blowdown phase associated with the increased RCS volurue. The net effect is that the K:nidbyrbw&sgtp'parrcup cke:13

. 's' blowdown time (end of bypass) for the RSG will be the same or decrease compared to the OSG. Reduced resistance to break flow for the RSG compared to the OSG will result in a lower PCT for the RSG compared to the OSG.

The increase in the current value of RCS volume in Unit 2 is significantly less than the increase associated with the repiccement of the steam generators in Unit

1. The small increase in the RCS volume willlikely result in a slight increase in the blowdown period. This slight increaic in the blowdown period will have no significant impact on the peak clad temperature (PCT) calculation for Unit 2. Any small changes in the PCT due to this small increase in the RCS volume can be easily accornmodated for Unit 2 because of the significant margin in the PCT (over 100 degree) available to the Appendix K 10CFR50.46 acceptance criteria of 2200 'F. Therefore, there is no reduction in the margin of safety.

An evaluation of the Gas Waste System Leak or Failure was performed to determine the effect of the increased RCS volume. Because the activity of the decay tank is controlled within Technical Specification limits, an increase in RCS volume would not change the results of the event. Therefore, there is no reduction in the margin of safety.

An evaluation was performed to determine the effect of the increased RCS volume (associated with the RSGs) on the peak containment pressure following a LBLOCA. The increased RCS volume caused the peak containment pressure to increase to 47.8 psig. This is still below the containment design pressure of 50.0 psig. Therefore, there is no reduction in the margin of safety. The increase in RCS volume for the existing units (without RSGs) remains within the conservative volume used in the calculation of the current peak containment pressure value of 44.4 psig. Therefore, there is no reduction in the margin of safety.

This proposed change involves testing requirements designed to demonstrate acceptable Irakage rates are maintained, if acceptable leakage rates are maintained as outlined in the Technical Specifications, there will be no reduction in the margin of safety. In the event of degradation of a containment seal that results in unacceptable leakage, plant shutdown will occur as required by Te:hnical Specifications and administrative requirements in accordance with approved p ant procedures. Therefore,:his proposed change does not involve a significant reduction in a margin of safety.

The editorial changes proposed are for clarity and consistency within the Technical Specifications and do not result in any physical changes to the facility or how it is operated. Therefore, the changes have no effect on the margin of safety.

Thus, this amendment request does not result in any decrease in a margin of safety.

Based on the above evaluation, Comed has concluded that these changes involve no significant hazards considerations.

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9 A'ITACHMENT D  !

ENVIRONMENTAL ASSESSMENT FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-37, NPF 66, NPF-72, AND NPF-77 Cornmonwealth Edison Company (Comed) has evaluated the originally proposed License Amendment Request dated January 30,1997 and this supplement against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with Title 10, Code of Federal kegulations, Part 51, Section 21 (10 CFR l

51.21). Comed has determined that this proposed Licen:e Amendment Request meets '

the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9). This l determination is based upon the following: )

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1. The proposed licen: ting action involves the issuance of an amendment to a j license for a reactor pursuant to 10 CFR 50 which changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or which changes an inspection or a surveillance requirement. This proposed License Amendment Request will allow Comed to revise the RCS volume, to revise the Unit 1 Pa value, and to make editorial changes; i

2. this proposed License Amendment Request involves no significant hazards considerations;

! 3. there is no significant change in the types or significant increase in the amounts of any effh at that may be released offsite; and

4. there is no significant inercase in individual or cumulative occupational j radiation exposure.

Therefore, pursuant to 10 CFR 51.22(b). neither an environmental impact statement nor  ;

an environmental assessment is necessary for this p uposed License Amendment Request.

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