ML20199H786
| ML20199H786 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 01/22/1998 |
| From: | Dick G NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20199F368 | List: |
| References | |
| NUDOCS 9802050121 | |
| Download: ML20199H786 (11) | |
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UNITED STATES g
j NUCLEAR REGULATORY COMMISSION
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COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-456 BRAIDWOOD STATION. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 88 License No. NPF 72 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, compues with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rutes and regulations set forth in 1
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 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 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 Ucense No NPF-72 is hereby amended to read as follows:
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Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 88 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with thr: Technical Specifications and the Environmental Proteci.lo, Plan.
3.
This license cmendment is effective as of the date of its issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION s/e
'l Georg(p r u
F. Dick, eni Project Manager Project Directorate lll 2 Division of Reactor Projects Ill/lV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:.lanuary 22,19%
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UNITED STATES 4j
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NUCLEAR REGULATORY COMMISSION
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COMMONWEALTH EDISON COMPANV, DOCKET NO. STN 50-457 BRAIDWOOD STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.88 L; cense No. NPF 77 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendmert 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 assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of th3 public, and (ii) thSt 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 s,afety of the public; and E.
The lesuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable reqairements 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 77 is hereby amended to read a' follows:
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2-(2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 88 and the Environmental Protection Plan contained in Appendix B, both of which were attached to Ucense No. NPF 72, dated July 2, 1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date if itt issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/ /1 Geor
. Dick, Seni Project Mansger Project Directorate lil 2 Division of Reactor Projects Ill/IV l-Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
January 22, 1998
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ATTACHMENT TO LICENSE AMENDMENT NOS 88 AND 88 FACillTY OPERATING LICENSE NOS. NPF 72 AND NPF 77 DOCKET NOS. STN 50-456 AND STN 50-457 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain verticallines indicating the area of change,
-Remove Pace _1 Insert Paoes 1-4 1-4 3/4 6-5 3/4 6 5 3/4 6-12 3/4612 B 3/4 6 1 B 3/4 6 1 B 3/4 6-2 B 3/4 6-2 5-4 5-4 1
g, IONS OFFSITE DOSE CALCULATION _M&liMAL 1.18 The OFFSITE DOSE CALCULATION KANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-l active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the 2nviron-mental Radiological Monitorlag Program.
The ODCM shall also contain (1) the Radioactive Effluent Controle and Radiological Environmental Monitoring Programs required by Sections 6.8.4.e and f, and (2) descriptions of the information that should be included in the Ar.nual Radiological Environmental Operating and Radioactive Ef fluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.7.
OPERABLE - OPERABILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s),
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 related support function (s).
OPERATING LIMITS REPORT 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 - MODE 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.
Ea 1.20.a P, shall be the maximum calculated primary containment pressure (44.4 peig*, 47.8 peig**) for the design basis loss of coolant accident.
PHYSICS TESTS 1.21 PHYSICS TESTS 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) otherwiss approved by the Commission.
PRESSURE BOUNDARf LEAKACE 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.
- Applicable to Unit 1 through cycle 7 and to Unit 2.
- Not applicable to Unit 2.
Applicable to Unit 1 after cycle 7 BRAIDWOOD - UNITS 1 & 2 1-4 AMENDMENT NO. 88
4 CONTAIMMENT SYSTEMS
- SURVEIt.t AMCE REQUIREMENTS l
4.6.1.3 Rach contair. ment 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 poig by means of a permanently.
installed continuous pressurisation and leakage monitoring system, or (2) Verifying that the door seal leakage is less than 0.01La as l
determined by precision flow measurements when measured for i
at lonec 30 seconds with the volume between the seals at a constant pressute of greater than or equal ~to 10 poig;
-b.
By conducting overall air lock letkage tests in accordance with Regulatory Guide 1.163, September 1995, and 10 CFM 50, Appendix J, j
option B.
c.
At least once per 6 months by verifyir.g that only one door in each ir lock can be opened at a time.-
d.
By verifying that the airlock seal leakage tests are less than 0.01 La as determined by precision flow measurements when measured l
for at least 30 seconds with the volume between the seals at a constant pressure of greater than or equal '.o_10 psig in accorde.nce-with Regulatory Guide 3.163,-September 1995, and 10 CFR 50,.
Appendix J, Option B.
BRAIDWOOD - UNITS 1 & 2 3/4 6-5 AMENDMENT NO. 88
CONTAINMENT SYSTEMS i
SURVEILLANCE REQUIREMENTS 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.
4.6.1.7.2 Each 8-inch containment 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.
4.6.1.7.3 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 contain.7,cnt purge supply and exhaust penetration shall be demonstrated OPERABLE by.erifying that the measured leakage rate is *,ess 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 OPEr.ABLE by verifying that the measured leakage rate is less than 0.01 L, when pressurized to at least P,.
d BRAIDWOOD - UNITS 1 & 2 3/4 6-12 AMENDMENT NO.88
.3/4.6 CONTAINMENT SYSTEMS mamma 3/4.6.1 PRIMARY CONTAINMENT J/4.6.1.1 CONTAINMENT INTEGRITY.
Primary CONTAINMENT INTEGRITY ensures that the release of radioective materials-from the containment atmosphere will be restricted to those leakage l
paths and associated leak rates assumed in the safety analyses.
This restriction, in conjunction with the leakage rate limitation, will limit the SITE DOUNDARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions.
3/4.6.1.2 CONTAIMMENT LEAKAGE I
The limitations on containment leakage ratos ensure that the total containment leakage volume will not exceed the i alue assumed in the accident analyses at the peak accident pressure, P,.
As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L, during performance of the periodic test to account for possible degradation of the containment leakage barrines betweer.. leakage tests.
The surv6111ance testing for measuring leakagt rates are coneiatent with the requirements of Appendix J of 10 CFR Part 50, Option B, Regulatory Guide 1.163, September 1995, Muclear Energy Instituto document NEI 94-01, and ANSI /ANS-56.8-1994.
3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations.on closure and leak rate for the containment air locks are required to meet' the restrictions lon 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 containment internal pressure ensure thatt (1) the containment structure is prevented from exceeding its design. negative pressure differential with respect to the outside atmosphere of 0.1 psia, and (2) the
' containment peak pressure does not exceed the design pressure of 50 peig 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 peig.for initial positive containment pressure will mit the total pressure to P,
which is higher =than the UFSAR Chapter 15 accident analysis calculated peak pressure assuming a limit of 0.3 psig for initial positive containment pressure, but is considerably less than the. design pressure of 50 peig.
BRAIDWOOD - UNITS 1 & 2 B 3/4 6-l' AMENDMENT NO. 88
CONTAINAENT SYSTEMS RAsts 3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial temperature 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 ensuras that the structural integrity of the containment will be maintained comparable to the original design ntandards for the life of the facility.
Structural integrity is requited to ensure that the containment will withstand the maxinam 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 visuR4 examination of tendons, anchorages and exposed interior and exterior surfaces of the containment, and the Type A leakage test are sufficient to demonstrate this capability.
The Surveillance Requirements for demonstrating the containment's etructural integrity are in compliance with the recommendations of proposed Rev. 3 to Regulatory Guide 1.?S, " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Contaj.nent Structures," April 1979 and proposed Regulatory Guide 1.35.1, " Determining Prestressing Forces for Inspection of Prestressed concrete Containments," April 1979.
The required Special Reports from any engineering evaluation of containment abnormalties shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedure, the tolerances on cracking, the results of the engineering evaluation and the corrective actions taken.
3/4.6.l.7 CONTAINMENT PURGE VENTILATION SYSIEM The 48-inch containment purge supply and exhaust isolation valves are required to be pealed closed (power removed) during plant operations since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident.
Maintaining these valves sealed closed during plant operation ensures that excessive quantities of radioactive material will not be released via the containment Purge System.
To provide assurance that the 48-inch contain-ment 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 is restricted to the 8-inch purge supply and exhaust isolation valves since, unlike the 48-inch valveu, the 8-inch valves are capable of closing during a LOCA or steam line break accident.
Therefore, the SITE BOUNDARY dose guideline values of 10 CTR Part 100 would not BRAIDWOOD - UNITS 1 & 2 B 3/4 6-2 AKENDMENT NO.88
DEBIGM FEAT 3mma i
5.3 REACTOR CORE FUEL AssthaLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rode clad with tircaloy-4 or IIRLo, except that limited substitution of fuel rode by filler rode consisting of tircaloy-4, EIRLO, or stainless steel or by vacancies may be made if justified by a cycle specific
-reload-antlysis.
Each fuel rod shall have a nominal active fuel length of 144 inches.
The initial core loading shell have a maximum enrichment of less
.than S 20 weight percent U-235.
Reload fuel shall be similar in physical i
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-either the spent fuel-pool or the new fuel vault.
Such acceptance criteria l
shall be based on the results of the " Byron and Braidwood spent Fuel Aack Criticality Analysis Using soluble Boron Credit," May 1997, CAC-97-162 and j
" Criticality Analysis of the' Byron /Braidwood Fresh Fuel Racks," June 1989.
CONTROL ROD AE8EMELIE8 l '
assemblies.
The full-longth control rod assemblies shall contain a nominal 5.3.2 The core shall centain 53 full-length and no part-length control rod
-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
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with stainless steel tubing.
I 5.4 REACTOR COOLANT Si' STEM 4
DESIGN. PRES 5URd AND TEMPERATURL 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, b.
For a pressure of 2485 peig, and c.
For a temperatura of 650'F, except for the pressurizer which is 680'F.
VOLUME 4
5.4.2 The total water and steam volume of the Reactor Coolant System is 12,340cubicfeetatanominalT,ledto.theUnit1totalReactorCoolant of 588.4'F.
An additional 1,280 cubic feet at a nominal T of 588.4'F is ad ag system. volume as a result of the four replacement steam generators installed after cycle 7.-
5.5 METEOROLOGICAL TOWER LOCATION
- 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
BRAIDWOOD - UNITsL1 E-2 5-4.
Amendment No. 88
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