ML20199F363

Amends 97 & 97 to Licenses NPF-37 & NPF-66,respectively. Amends Revise TS & Associated Bases Re Primary Containment Pressure & Reactor Coolant Sys Volume

ML20199F363
Person / Time
Site:	Byron
Issue date:	01/22/1998
From:	Dick G Office of Nuclear Reactor Regulation
To:
Shared Package
ML20199F368	List: ML20199F363 ML20199F375 ML20199H786
References
NPF-37-A-097, NPF-66-A-097 NUDOCS 9802030151
Download: ML20199F363 (11)

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+,.....,o COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-454 BYRON STATION. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 97 License No NPF-37 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated January 30,1997, as supplemented on December 9,1997, complies with the standards and requirements of the Atomic Energy Act of 1954, er amendea (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activitias will be conducted in compliance wiih the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF.37 is hereby amended to read as follows:

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Technical So6cifications The Technical SpecM.ations contained in Appendix A as revised through Amendment No. 97 and the Environmental Protectbn Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accoroance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION lty L

George Dick, Senio Project Manager Project Directorate 1112 Division of Reactor Projects -lil/lV Office of Nuclear Reactor Regulation Attachmer.t:

Changes to the Technical Specifications Date of Issuance:

January 22, 1998 e

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\\.....l2 WASHINGTON, D.C. 20tio6-0001 COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-455 SXHON STATION. UNIT NO. 2 AMENDMENT TO FACIL!TY OPERATING LICENSE Amendment No.97 License No. NPF-66 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendmerit by Commonwealth Edison Company (the licensee) dated January 30,1997, as supplemented on December 9,1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assuran e (i) that the activitics authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requiremerits have been satisfied.

2.

Accordingly, the license is inmended by changes to the Technical Specifications as indicated in the attechment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-66 is hereby amended to read as follows:

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Technical Soecificatigns The Technical Specifications contained in Appendix A (NUREG 1113), as revised through Amendment No. 97 and revised by Attachment 2 to NPF 66, and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF 37, dated February 14,1985, are hereby incorporated into this license. Attachment 2 contains a revision to Appendix A which is hereby incorporated into thir license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment la effective as of the date of its issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION lM L

Geor

. Dick, Seni Project Manager i

Project Dires torate ill 2 Division of Reactor Projects. Ill/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: January 22, 1998 y

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ATTACHMENT TO LICENSE AMENDMENT NOS. 97 AND 97

' FACILITY OPERATING LICENSE NOS. NPF 37 /.ND NPF-66 DOCKET NOS. STN 50-454 AND STN 50-455 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

Remove Panes Insert Paaes 14 14 3/4 6 5 3/4 6-5 3/4612 3/4 6-12 B 3/4 6-1 B 3/4 6-1 B 3/4 6-2 B 3/4 6-2 5-4 5-4 4

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DEFINITIONS OFFSITE DOSE CALCULATION MANUAL 1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program.

The ODCM shall also contain (1) the Radioactive Effluent Controle and Radiological Environmental Monitoring i

Programs required by Sections 6.8.4e and f, and (2) descriptio1s of the information__that should be included in the-Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.7.

1 OPERABLE - OPERABILITY l

1.19 A system, subsystem, train, component or devico shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s),

i and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their re ated support function (s).

OPERATING LIMITS REPOkT 1.19.a The OPERATING LIMITS REPORT (OLR) is the unit-specific document that provides operating limits for the current operating reload cycle.

These cycle-specific operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9.

Plant operation within these operating limits is addressed in individual specifications.

OPERATIONAL MODE - MODI 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant. temperature specified in Table 1.2.

Ec 1.20.a P, shall be the maximum calculated primary containinent pressure (44.4 psig*, 47.8 poig**) for the design basis loss of coolant accident.

PHYSICS TESTS 1.21 PHYSICS TFSTS shall be those tests performed to measure the fundamental nuclear characteristics of the core and related instrumentation: (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE _ BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

'Amlicable to Unit 1 through cycle 8 and to Unit 2.

1

• • Wot aplicable to Unit 2.

Aplicable to Unit i af ter cycle 8.

BYRON - UNITS 1 & 2 1-4 AMENDMENT NO. 97

CONTAINMENT SYSTKlig ggtvEIt fuct RE00IREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a.

By conducting airlock seal leakage tests in accordance with Regulatory Guide 1.163, September 1995, and 10 CFR 50, Appendix J, Option B, by:

(1) Verifying that the door seal leakage is less than 0.0024La when the volume between the door seals is pressurized to l

greater than or equal to 3 psig by means of a permanently installed continuous pressurization and leakage monitoring j

system, or (2)

Verifying that the door seal leakage is less than 0.01La as determined by precision flow measurements when measured for at least 30 seconds with the volume between the seals at a constant pressure of greater than or equal to 10 psig; b.

By conducting overall air lock leakage tests in accordance with Regulatory Guide 1.163, September 1995, and 10 CFR 50, Appendix J, Option B.

c.

At least once per 6 months by verifying that only one door in each air lock can be opened at a time.

d.

By verifying that the airlock saal leakage tests are less than I

0.01 La as determined by precision flow maarurements when measured for at least 30 seconds with the volume between the seals at a constant pressure of greater than or equal to 10 peig in accortlance with Regulatory Guide 1.163, September 1995, and 10 CFR 50, Appendix J, option B.

BYRON - UNITS 1 & 2 3/4 6-5 AMENDMENT NO. 97 A

l EQMIAINMENT SYSTEMS SURVEILLANCE fTOUIREMENTS 4.6.1.7.1 Each 48-inch containment purge supply and exhaust isolation valve (s) shall be verified closed and power removed at least once per 31 days.

(.6.1.7.2 Each 8-inch containnent purge supply and exhaust isolation valve shall be verified to be positioned in accordance with Specification 3.6.1.7b at least once per 31 days.

4.0.1.7.3 At least once per 6 months on a STAGGERED TEST BASIS, the inboard and outboard valves with resilient material seals in each closed 48-inch containment purge supply and exhaust penetration shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than 0.05 L, when pressurized to at least P,.

4.6.1.7.4 At least once per 3 months, each 8-inch containment purge supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than 0.01 L, when pressurized to at least P,.

t BYRON - UNITS 1 & 2 3/4 6-12 AMENDMENT NO. 97 s

t 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY j

Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the. containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses.

This restriction, in conjunction with the leakage rata limitation, will limit the SITE B0UNDARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions.

3/4.6.1.2 CONTAI M,NT LEAKAGE The limitations on containment leakage rates ensure that tne total l

containment leakage volume will not exceed the value assuned in the accident measured overall integrated leakage rate l. As an added conservatism, the analyses at the peak accident pressure, P s further limited to less than or during performance of the periodic test to account for equal to 0.75 L, tion of the containment leakage barriers between leakage possible degrada tests.

The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50, Option B, Regulatory Guide 1.163, September 1995, Nuclear Energy Institute document NEI 94-01, and ANSI /ANS-56.8-1994, h 6.1.3 CONTAIMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks arc required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate.- Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

3/4.6.1.4 INTERNAL PRESSURE The limitations on e r tainment internal pressure ensure that: (1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 0.1 psig, and (2).the containment peak pressure does not exceed the design pressure of 50 psig during steam line break conditions.

The maximum increase in peak pressure expected to be obtained from a cold leg double-ended-break event is defined as P,

The limit of 1.0 psig for initial positive containment pressure will llmit the total pressure to P,

-which is higher than the UFSAR Chapter 15 accident analysis calculated p k ea pressure assuming a limit of 0.3 psig for initial positive containment pressure, but is considerably less than the design pressure of 50 psig.

BYRON - UNITS 1 & 2 B 3/4 6-1 AMENDMENT N0.97

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QQETAINMENT SYSTEMS BASES 3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature doos not exceed the initial temperatuse condition assumed in the accident analysis for a steam line break accident. Measurements shall be made at all of the listed running fan locations, whether by fixed or portable instruments, to determine the average air temperature.

3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original deoign standards tor the life of the facility.

Structural integrity is required to ensure that the containment will withstand the maximum pressure of P, in the event of a cold leg double-ended break accident.

The measurement of containment tendon lift-off force, the tensile tests of the tendon wires or strands, the visual examination of tendons, anchorages and exposed interior and exterior surfaces of the containment, and the Type A lockage test are sufficient to demonstrate this capability.

The 8urveillance Requirements for demonstrating the containment's structural f.ntegrity are in compliance with the recommendations of proposed Rev. 3 to Regulatory Guide 1.35, " Inservice Surveillance of Ungrouted Tendons in Prestressed concrete Containment Structures," April 1979 and proposed Regulatory Guide 1.35.1,

" Determining Prestressing Forces for Inspection of Premtressed Concrete Containments," April 1979.

The required Special Reports from any engineering evaluattu of containment abnormalties shall include a description of the tendon condition, the condition of the concrete (especielly at tendon anchorages), the inspection proceduro, tha tolerances on cracking, the results of the engineering evaluation and the corrective actions taken.

3/4.6.1.7 CONTAINMENT PURGE VENTILATION SYSTEM The 48-inch containment purge supply and exhaust isolation valves are required to be sealed closed (power removed) during plant operations since these valves have not tren demonstrated capable of closing during a LOCA or steam line break accident. Maintaining these valves sealed closed during plant operation ensuces that excessive quantities of radioactive material will not be released via the Containment Purge System.

To provide assurance that the 48-inch containment valves cannot be inadvertently opened, the valves are sealed closed in accordance with Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or prevents power from being supplied to the valve operator.

The use of the containment purge lines in restricted to the 8-inch purge supply and exhavet isolation valves since, unlike the 48-inch valves, the 8-inch valves are capable of closing during a LOCA or steam line break at:1 dent.

Therefore, the SITE BOUNDARI dose guideline values of 10 CFR Part 100 wculd not BYRON - UNITS 1 & 2 B 3/4 6-2 AMENDMENT NO. 97

DESIGN FLATURIS 5.3 REACTOR CORE EVEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4 or EIRLO, except that limited substitutie, of fue' rode by filler rode consisting of zircaloy-4, z!RLO, 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 fusi shall be similar in phycical design to the initial cora loading or previous cycle loading.

The enrichment of any reload fuel design shall be determined to be acceptable for storage in either the spent fuel pool or the new fuel vault.

Such acceptance criteria shall be based on the results of the " Byron and Braidwood Spent Fuel Rack Criticality Analysis Using Soluble Boron Credit," Hay 1997, CAC-97-162 and

" Criticality Analysis of the Byron /Braidwood Fresh Fuel Racks," June 1989.

CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 51 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 both types.

All control rods shall be clad with stainless steel tubing.

W REACTOR COOLANT SYSTEM DESIGN PRESSUAE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained a.

In acevrdance with the Code requirements specified in Section 5.2 of the UFSAR, with allowance for normal degrada* ion pursuant to the applicable Surveillarce Requirements, b.

For a pressure of 2485 peig, and c.

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

VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 12,340cubicfeetatanominalT,y$edtotheUnit1totalReactorCoolant of 588.4'F.

An additional 1,280 cubic feet at a nominal T,g af 588.4'r is ad System volume as a result of the four replacement steam generators installed after cycle 8.

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

BYRON - UNITS 1 & 2 5-4 Amendment No. 97

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