ML20198P786
| ML20198P786 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 01/09/1998 |
| From: | Dick G NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20198P791 | List: |
| References | |
| NUDOCS 9801220249 | |
| Download: ML20198P786 (7) | |
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7590-01.P MNITED STATES NUCLEAR REGULATORY COMMISSIO!!
COMMONWEALTH EDISON COMPANY t
DOCKET NOS. STN 50 454. STN 50-455. STN 50-456 AND STN 50-457 BYRON STATION. UNITS 1 AND 2. AND BRAIDWOOD STATION. UNITS 1 AND 2 ENVIRONMENTAL ASSESSMENT AND FINDING OF NO SIGNIFICANT IMPACT The U.S. Nuclear Regulatoty Commission (the Commission) is considering lasuance of an exemption from certain requirements of its regulations to Facility Operating Lloense Nos.
NPF.37, NPF.66. NPF.72 and NPF.77, issued to Commonwealth Edison Company (the licensee), for operation of Byron Station, Units 1 and 2, and Braidwood Station, Units 1 at/.d 2, located in Ogle County and Will County, Illinois, respectively.
fsNVIRONMENTAL ASSESSMENI identification of the Proposedj,qligg The proposed action would permit the licensee to use the 1996 Addenda to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, Appendix G, to determine the reactor vessel pressure temperature (P.T) limits and the low-temperature overpressu o pmtection (LTOP) system setpoints. By application dated April 3, 1997, as supplemented by letter dated June 19,1997, the licensee requested an exemption from certain requirements of 10 CFR Part 50.60, ' Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation." The exemption would allow application of an attomate methodology to determine the P T limits and LTOP system setpoints for Byron, Un:ts 1 and 2, and Braldwood, Units 1 and 2. The proposed attomato 9901220249 990100 PDR ADOCK 05000454 P
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methodology is consistent w6th guidelines developed by the ASME Working Group on Operating Plant CrMoria lo define pressure limits during LTOP events that avoid consin unnecessary oporptional testrictions, provide adequate margins against failure of the reactor pressure vessel, and reduce the potential for unnecessary activation of pressure relieving devloss used for LTOP.
These guidelines have been incorporated into the 1996 Addenda to the ASME Code, Sootion XI, Appendix G. However,10 CFR 50.65a, todos and Standards,' has not been updated to reflect the acceptability of the 1996 Addenda to the ASME Code.
The Need for the Pronosed Action:
Pursuant to 10 CFR 50.60, all lightwater nuclear power reactors must meet the fracture toughness requirements for the reactor coolant pressure boundary as set forth in 10 CFR i
Part 50, Appendix 0,10 CFR Part 50, Appendix G, defines P.T limits during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests to which the pressure boundary may be subjected over its service lifetime, and sseoirees that these P T limits r.ust be at least as conservative as the limits obtained by following the methods of analysis Lnd the margins of safety of the ASME Code,Section XI, Appendix G.10 CFR 50.55a requires that any reference to ASME Code,Section XI, in 10 CFR Part 50 refers to addenda j
through the 1988 Addenda and editions through the 1989 Edition of the Code, unless otherwise noted.10 CFR 50.60(b) specifies that shomatives to the described requirements in 10 CFR i
Part 50, Appendix 0, may be used when an exemption is granted by the Commission under 10 CFR 50.12.
To prevent transients that would produce excursions exceeding the P T limits while the
, reactor is operating at low temperatures, the licensee installed the LTOP system. The LTOP system includes pressure relieving devices called power-operated relief valves (PORVs) that are set to open at reduced pressure when reactor pressure and temperature are reduced. The
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PORVs prevent the pressure in the reactor vessel from exceeding the P T limits. However, to prevent the PORVs from lifting as a result of normal operating pressure surges, some margin is needed between the normal operating pressure and the PORV setpoint, in addition, when instrument uncertainty is considered, the operating window between the PORV setpoint and the min 6 mum pressure r6 quired for reactor coolant pump seals is small and presents difficulties for plant operation.
To prevent pressure from exceeding the P T limits, the PORVs would be set to open at a pressure very close to the normal pressure inside the reactor. With the PORV setpoint close to the normal operating pressure, minor pressure perturbations that typically occur in the reactor could cause the PORVs to open. This is undesirable from the safety perspective because after every PORV opening there is some concem that the PORV may not reclose. A stuck open PORV would continue to discharge prir%y coolant and reduce reactor pressurs until the discharge pathway was closed by operator action.
The licensee requested use of the 1996 Addenda to the ASME Code,Section XI, Appendix G. These addenda to the Code would perm!t a slightly higher pressure inside the reactor and a slightly higher PORV setpoint during low temperature, shutdown conditions. This would reduce the likelihood for inadvertent opening of the PORVs.
Appendix G of the ASME Code requires that the P T limits be calculated: (a) using a safety factor of twt M the principal membrane (pressure) stresses, (b) assuming a flow at the surface with a depth of one quarter (1/4) of the vessel wall thickness and a length of six (6) times Ms depth, and (c) using a conservative fracture toughness curve that is based on the lower bound of static, dynamic, and crack arrest fracture toughness tests on material similar to the Byron /Braidwood reactor vessel material.
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i For determining the P-T limits, Comed proposed to use the safety margins based on the 1998 Addenda to the ASME Code in lieu of the 1989 Edition. When compared to the 1989 i
Edition of the ASME Code, the 1996 Addenda permits the use of a lower stress intensity factor
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for determining the apphed stress intenbity due to pressure and thermal stresses. This results in l
e slight reduction in the applied stress intensity and a corresponding shift in the allowable pressure at a 9 ven temperature in the non-conservative direction; however, this difference is 1
f mir.or when compared to the explicit conservatisrn incorporated into the Code, and the changes in the stress intensity factor are supported by the work performed by J. A. Keeney and T. L.
Dickson at Oak Ridge National Laboratory (ORNL) for the NRC, and others.
1996 Addenda to the ASME Code require that the system pressure is maintained below i
the P T limits during nonnel operation, but allows the pressure that may occur during LTOP 4
events to exceed the P.T limits, provided acceptable margins are maintained during thess events. This approach protects the pressure vesse! from LTOP events, and maintains the P.T limits applicable for normal heatup and cooldown in accordance with 10 CFR Part 50, e
i Appendix G, and Sections lli and XI of the ASME Code, in determining the PORV setpoint for LTOP events the licensee proposed to use the safety margins of the 1996 Addenda to the ASME Code,Section XI, Appendix G. This attomate methodology allows determination of the setpoint for LTOP events such that the maximum pressure in the vessel will not exceed 110 percent of the P-T limits that are developed using the l
1996 Addenda to the ASME Code,Section XI, Appendix G, methodologies described above.
l-This results in a safety factor of 1.8 on the principal membrane stresses. All other factors, includeg the assumed flaw size and fracture toughness, remain the same. Although this methodology would reduce the safety factor on the principal membratw stresses, use of the
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i,E;::d crHerta will pmvide adequate margi.3s of safety for the reactor vessel during LTOP events.
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Use of the 1996 Addenda to the ASME Code,Section XI, Appendix G, safety mergins will i
reduce operational challenges during low temperature, low pressure operations. In terms of 1
overall safety, the safety benefits derived from simplified operations and the reduced potential for i
undesirable opening of the PORVs will more than offset the reduction of the principal membrane 4
j stress safety factor that may occur during LTOP events. Reduced operational challenges will reduce the potential for undesirable impacts to the environment.
H should be r:Med that the provision to set the PORV setpoint such that r, protects 110 percent of the P-T limits is already part of the Byron and Braldwood licensing basis. This provision was approved in the exemption to 10 CFR 50.60 granted to Byron on November 29, 1996, and to Braidwood on July 13,1995, and December 12,1997, for Units 1 and 2, respectively, to allow the use of ASME Code Case N 514. Therefore, while H represents a change from the 1989 Edition of the ASME Code, it is not a change to the licensing basis for I
these faciilths.
Environmentalimonets of t% Proposed Action:
i The Commission has completed its review of the proposed action and concludes that the 1
proposed action involves features located entirely within the protected areas as defined in 10 CFR Part 20.
j The proposed action will not resuit in an increase in the probability or consequences of accidents or result in a change in occupational or offsite dose. Therefore, there are no raciological impects associated with the proposed action.
l The proposed action will not result in a change in nonradiological plant effluent and wl:1 l
have no other nonradiological environmental impact.
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i 6-0 Accordmgly, the Commission concludes that there are no environmentalimpacts l
associated with this action.
l Altematives to the Proposed Action:
Since the Commiselon has concluded there is no measurable environmentalimpact associated with the proposed action, any ahomatives with equal or greater environmental impact need hot be evaluated. A* mn ahomative it "~ roposed action, the staff considered denial of the proposed action, '
- a. of the application would resuh in no change in current environmental 4
impacts. The environmei 'limpsets of the proposed action and the shomative action are similar.
Ahemative Use of Resources:
s This action does not involve the use of any resources not previously considered in the l
Final Environmental Statement for the Byron Station or the Braidwood Station.
Aoencies and Persons Consulted:
i In accordance with its stated policy, on January 9,1998, the staff consulted with the Illinois State official, Frank Niziolek of the lilinois Department of Nuclear Safety, regarding the environmentalimpact of the proposed action. The State official had no comments.
FINDING OF NO SIGNIFICAN IMPACT i
Based upon the environmental assessment, the Commission concludes that the 4
proposed action will not have a significant effect on the quality of the human environment.
Accordingly, the Commission has determined not to prepare an environmentalimpact statement for the proposed action.
For further details with respect to the proposed action, see the licensee's letter dated April 3,1997, as supplemented by letter dated June 19,1997, which are av611able for public inspection at the Commission's Public Do wnt Room, The Gelman Building,2120 L Street, W
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7 NW., Washington, DC, and at the Local Public Document Room located: For B;m, the Byron Public Ubrary District,10g N. Franklin, P.O. Box 434, Byron, Illinois 61010; for Braidwood, the Wilmington Public Ubrary,201 S. Kankakse Street, Wilmington, Illinois 60481.
Deted at Rockville, Maryland, this 9th day of January 1998.
FOR THE llVCLEAR REGULATORY COMMISSION j
Geor F. Dick, Jr., Senior Project Manager i
Project Directorate ill 2 l
Division of Reactor Projects. lil/IV Office cf tN$::: P.sactor Regulation s
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