ML20059M392
| ML20059M392 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 08/17/1993 |
| From: | AFFILIATION NOT ASSIGNED |
| To: | |
| References | |
| OLA-2-A-005, OLA-2-A-5, NUDOCS 9311190064 | |
| Download: ML20059M392 (49) | |
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'l b353-00 2- 'j EXHIBIT 5 ;;Lr.U U:
U5NFC PG&E'S LICENSE AMENDMENT REQUEST ,9 - OcF28; P5;40
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July 9,1992 PG&E Letter No. DCL-92-154 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk j -R
. Washington, D.C. 20555 Re: Docket No. 50 275, OL DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 License Amendment Request 92-04
) 40-Year Operating License Application Gentlemen:
, Pacific Gas and Electric Company applies for an amendment to Facility Operating J License Nos. DPR 80 and DPR-82 to change the expiration dates of the full power licenses for Diablo Canyon Units 1 and 2. PG&E's enclosed request would change the expiration date for the Unit 1 Operating License from April 23,2008, to
. September 22,2021, and the expiration date for the Umt 2 Operating License from
, December 9,2010, to April 26,2025. These proposed expiration dates would allow 1
) for 40 years of operation as permitted by 10 CFR 50.51. l The present operating license terms for Diablo Canyon are based on the NRC policy in effect prior to the 1982 determination by the Commission that the 40-year term of operation may begin upon issuance of the first operating license, rather than upon j issuance of the construction permit. Therefore, the present operating license terms for.Diablo Canyon commence with the dates of issuance of the construction permits for Units 1 and 2, April 23,1968, and December 9,1970, respectively.
Accordingly, the expiration date for the Unit 1 Operating License is April 23,2008, and the expiration date for the Unit 2 Operating License is December 9,2010.
) Since 1982, the Commission has accepted and approved requests to amend existing operating licenses to change the expiration dates and recover the time between the effective dates of the construction permit and the first operating license. More than 50 such license amendments have been granted by the Commission. Based on the enclosed request, the proposed 40-year term start dates for Diablo Canyon are J September 22,1981, for Unit I and April 26,1985, for Unit 2, which correspond to the effective dates of the fuel-load / low-power operating licenses for each unit.
The proposed license term changes do not affect the design, operation, or Technical Specifications of the plant. Based on a review of the Diablo Canyon Final Safety 3 Analysis Repon Update and the associated NRC Safety Evaluation Report and Supplements, PG&E concludes that the proposed changes do not involve significant hazard considerations.
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Document Control Desk July 9,1992 j i- PG&E lett:r No. DCL-92-154 i l 1 !
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j PG&E also has determined that the environment will not be adversely affected by the
- proposed license 'erm changes based on a review of the Diablo Canyon Environmental t Report and Suppsments and the NRC Final Environmental Statement and Addendum.
Therefore, pursuant to 10 CFR 51.30 through 51.35, PG&E believes the proposed changes ,
require preparation only of an environmental assessment and finding of no significant impact, i
, and that preparation of an environmental impact statement is not required. Dese actions are j consistent with the Commission's practice for similar amendment requests.
l Based on these safety and environmental reviews, PG&E requests that the Diablo Canyon
, operating license expiration dates be changed from April 23,2008, to September 22,2021,
! for Unit I and from December 9,2010, to April 26,2025, for Unit 2.
t Sincerely, i
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j Gregcry hi. Rueger j cc: Edgar Bailey, DHS i Ann P. Hodgdon l 1
John B. hfartin j Philip J. hforrill
- Harry Rocd
. CPUC
{ Diablo D;stribution Enclosure i
5408S/S5K/EhtG/2057
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l ENCLOSURE UNITED STATES OF AMERICA l NUCLEAR REGULATORY COMMISSION
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) Docket No. 50-275 i In the Matter of ) Facility Operating License !
) No. DPR-80 ~ j PACIFIC GAS AND ELECTRIC COMPANY ) i
) Docket No. 50 323 )
Diablo Canyon Power Plant ) Facility Operating License l Units I and 2 ) No. DPR-82 l
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License Amendment Request !
No. 92-04 !
Pursuant to 10 CFR 50.90, Pacific Gas and Electric Company hereby applies to amend its 1 Diablo Canyon Power Plant Facility Operating License Nos. DPR-80 and DPR-82. The l purpose of these license amendments is to change the term of the licenses to permit 40-year )
operation of each unit from the date ofissuance of the operating licenses, i
Information regarding the proposed amendments is provided in Attachments A and B.
l These changes have been reviewed and are considered not to involve a significant hazards consideration as defined in 10 CFR 50.92 and to have no significant environmentalimpacts.
j Further, there is reasonable assuran:e that the health and safety of the public will not be adversely affected by the proposed changes, i Sincerely,
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.% mr i Gregory M. Rueger L
Subscribed and sworn to before me Howard V. Golub f this 7th day of July 1992. Christopher J. Warner Richard F. I.ocke l Attorneys for Pacific Gas and Electric Company
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Mildred J. Williams, Notary Public Christopher J. Warser
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b ATTACHMENT A DIABLO CANYON POWER PLANT UNITS 1 AND 2 40-YEAR OPERATING LICENSE APPLICATION
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CONTENTS 1
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1.0 DESCRIPTION
OF AMENDMENT REQUEST . . ...................1-l l
2.0 BACKGROUND
.......................................... 1 3.0 JUSTI FI C ATI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 h
3.1 Baseload G eneration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3.2 El ectric Rat es ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3.3 A ir Emissions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3.4 State and Local Economy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
) 4.0 S AFETY EVA LUATI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 l
4.1 I nt rod u ct io n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 4.2 Licensing Basis Documents / Programs ........................4 l
4.2.1 FSAR and Technical Specifications . . . . . . . . . . . . . . . . . . . 4 4.2.2 Probabilistic Risk Assessment . . . . . . . . . . . . . . . . . . . . . . . 4 4.2.3 Surveillance and Maintenance Programs . . . . . . . . . . . . . . . . 5 l 4.2.3.1 ISI and IST Programs ...........................5 L 4.2.3.2 EQ Program ..,.............................. 6 j 4.2.3.3 Maintenance Program ........................... 7
! 4.3 Plant Operating History . . . . . . . . . . . . . . . . . . ........... 10 4.3.1 Operating Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . _ 10 f
4.3.2 Reliability / Safety.Related Plant Modifications . . . . . . . . . . . . 1I
) 4.3.3 Regulatory Performance . . . . . . . . . . . . . . . . . . . . . . . . . 13 4.4 Assurance of Continued Functional Capability of Safety.Related Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 4.4.1 Reactor Coolant System Pressure Boundary . . . . . . . . . . . . . . 13 1
4.4.1.1 General ................................... 13 4.4.1.2 R e a ct or Ves s el . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 4.4.1.3 Reactor Vessel Internals ! . . . . . . . . . . . . . . . . . . . . . . . . 15
) 4.4.2 Other Machanical Components . . . . . . . . . . . . . . . . . . . . . 15 l 4.4.3 Electrical Components . . . . . . . . . . . . . . . . . . . . . . . . . . 16 4.4.4 Structural Components . . . . . . . . . . . . . . . . . . . . . . . . . . 16 4.4.4.1 Primary Containment . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 y 4.4.4.2 Other Stru ctures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 l
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5.0 ENVIRONMEhTAL EVALUATION . . . . . . . . . . . . '. . . . . . . . . . . . . . . . 18 5.1 Introd u ction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 i 5.2 Systems and Programs for Environmental Control and Monitoring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 I
5.2.1 . Waste Processing System ........................18' 5.2.2 ALARA Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 -
5.2.3 Process and Area Radiation Monitoring System . . . . . . . . . . . 20 5.2.4 Radiological Environmental Monitoring Program . . . . . . . . . . 20 j 5.2.5 Nonradiological Surveillance Program .................21 5.2.5.1 Environmental Protection Plan . . . . . . . . . . . . . . . . . . . . . 21' 5.2.5.2 NPDES Permit . . . . . . . . . . .' . . . . . . . . . . . . . . . . . . . . 22 5.3 Environmental Impact During Normal Operation . . . . . . . . . . . . . . . . . 22 5.3.1 . Occupational Radiation Exposure . . . . . . . . . . . . . . . . . . . . 22; i 5.3.2 Offsite Radiation Exposure . . . . . . . . . . . . . . . . . . . . . . . 22 . )
5.3.3 Uraniu m Fu el . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 4 2 5.3.4 Spent Fuel Storage ............ ...............25 ,
5.3.5 S oli d Waste . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 5.3.5.1 Low Level Rad:ccctive Waste . . . . . . . . . . . . . . . . . . . . . . 26 S p ent Fu el . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6 .
i 5.3.5.2 5.3.5.3 Waste Shipping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
. 27 5.3.5.4 Solid Waste Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . 27 ;
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5.3.6 ihermal and Ecological Effects of the Cooling Water System . . 27--
5.3.7 Protection of Historic Properties . . . . . . . . . . . . . . . . . . . . 28
. 5.4 Exposure from Releases During Postulated Accidents . . . . . . . . . . . . . . . 29 l
) 5.5 Environmental Related Plant Modifications . . . . . . . . . . . . . . . . . . . . . 33 5.6 Decommissioning . . ............................. 34 6.0 NO SIGNIFICANT HAZARDS EVALUATION . . . . . . . . . . . . . . . . . . . . . 35
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TABLES ,
I 51 Diablo Canyon vs. INPO Industry Goal Average Annual Occupational Exposure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 -
i 5-2 Comparison of Offsite Appendix ! Radiation Exposure Limits -
and A ctu al D ata . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 j 5-3 Annual Volume of Low Level Solid Radioactive Waste Generated l at Diablo Canyon Compared with Median PWR . . . . . . . . . . . . . . . . . . . . 26 5-4 Summary of Population Projections for the Diablo Canyon Vicinity . . . . . . . . . . 30 FIGURES ,
i 5-1 Diablo Canyon Exclusion Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 l
52 Low Population Zone . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 -
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ATTACHMENT A DIABLO CANYON POWER PLANT UNITS 1 AND 2 40 YEAR OPERATING LICENSE APPLICATION 3
1.0 DESCRIPTION
OF AMENDMENT REQUEST This license amendment request (LAR) proposes to revise License Condition 2.1, ' Term of License," in the Diablo Canyon Power Plant (DCPP) Units I and 2 Operating Licenses.
The proposed amendment would change the expiration date for the Unit 1 Operating License DPR 80 from April 23, 2008, to September 22, 2021, and change the expiration date for the Unit 2 Operating License DPR 82 from December 9,2010, to April 26,2025.
Marked-up pages from the licenses are included in Attachment B.
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2.0 BACKGROLWD Both the Atomic Energy Act of 1954 and NRC regulations authorize issuance of facility operating licenses D for a period of up to 40 years. Commencement of the 40 year period from the date ofissuance of the operating license is allowed by 10 CFR 50.51, which states in part: *Where the operation of a facility is involved the Commission will issue the license for the term requested by the applicint or for the estimated useful life of the facility if the Commission determines that the estimated usefullife is less than the term requested." The NRC has established that requests for operation for a 40-year term from.the
} date of issuance of the first operating license may be granted, provided the utility demonstrates reasonable assurance of safety during 40 year facility operation and that the environmental ef% cts of 40 year operation are evaluated in the Environmental Report.
Operating licenses were issued for a term of 40 years starting from the effective date of the construction permit for plants licensed prior to 1982. Beginning in 1982, however, the NRC's policy has been to D issue operating licenses for a 40-year term commencing with the effective date of the first operating license. This revised policy includes not only newly licensed plants, but also earlier plants that may apply for amendments to their operating licenses to change the license expiration date to 40 years from the effective date of the operating license.
s DCPP Units 1 and 2 are currently licensed for 40 years commencing with the effective dates of the J
construction permits on April 23,1968, and December 9,1970, respectively. 'Ihus, the Unit I license expires at midnight on April 23,2008, and the Unit 2 license expires at midnight on December 9,2010.
This LAR proposes that the 40-year license terms, as permitted by 10 CFR 50.51, begin from the effective dates of the Unit I and Unit 2 first operating licenses. The initial fuel load / low-power operating D license for Unit I was effective on September 22,1981, and the initial fuel load / low power operating license for Unit 2 was effective on April 26,1985. Thus, the proposed license term expiration dates would be September 22,2021, for Unit I and April 26,2025, for Unit 2.
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l 3.0 JUSTIFICATION I
PG&E has assessed the safety aspects of plant design and operation of each DCPP unit for the proposed 40 year operation. Based on this assessment, as provided in Section 4.0 of this attachment, PG&E concludes that plant safety will be maintained during the 40-year operation. This conclusion is supported by the following factors: (1) the DCPP facility has been designed and analyzed for at least 40 years of operation; (2) the equipment, structures, and materials were purchased or constructed based on operation of at least 40 years; and (3) inspection and maintenance programs were developed to be applicable for at least 40 years of plant operation.
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PG&E has reviewed the Diablo Canyon Final Environmental Statement and Addendum and the Environmental Report and its Supplements to ensure that 40 year operation of each unit, commenring with the effective date of the operating licenses, is consistent with the previously evaluated environmental effects. He results indicate that no additional significant environmental impacts beyond those originally l
addressed are involved with 40 year operation. A summary of the environmental evaluation is provided I in Section 5.0 of this attachment.
PG&E has also performed an assessment of the potential impact on historic properties in accordance with Section 106 of the Historical Preservation Act of 1966, as amended, and in accordance with the l provisions of 36 CFR 800, Protection of Historic and Cultural Properties. No significant impact on historic properties was identified that could be associated with the proposed license term changes.
Benents expected from the additional period of operation include: (1) continued availability of reliable baseload generation, (2) avoided increase in electric rates to consumers, (3) avoided air emissions, and (4) continued benefits to the state and local area economy. Each of these benefits is briefly summarized .
below.
3.1 BASELOAD GENERATION Diablo Canyon provides baseload generation to consumers throughout Northern and Central California, he plant has produced approximately 100 billion kilowatt hours (kWh) of electricity since beginning
) commercial operation in 1985 and is one of the most reliable nuclear power facilities in the country, with a lifetire capacity factor of approximately 77 percent. Accordingly, it is prudent and beneficial to keep this eliable source of power in operation, particularly in light of the projected growth of California's ele;tricity demand.
In 1990, PG&E's electricity demand was more than 85 billion kWh. It is expected to grow through
) 2025. De California Energy Commission in 1992 adopted a new forecast for the PG&E service area under which electricity demand is projected to increase at the rate of about 1.5 percent compounded ;
annually through 2011. Assuming the demand for electricity continues to increase at a 1.5 percent per year rate, the overall increase in demand (1990 to 2025) would be 57 billion kWh or 65 percent. For comparison, in the period 19801990, PG&E's actual electricity demand grew at an annual rate of about 2.6 percent.
i Given the projected electricity demand and associated increased capacity requirements through 2025, the construction of some new centralized power plants will probably be required, even assuming the term of -
Diablo Canyon's operating licenses are changed as requested in this LAR. Early retirement of Diablo Canyon would significantly exacerbate such requirements. Therefore, continued availability of Diablo -
) Canyon will be of substantial benefit to Northern and Central California.
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3.2 ELECTRIC RATES In addition to providing reliable baseload generation, Diablo Canyon power costs to the customer are i
anticipated to be competitive with PG&E's overall production costs by 2008. After 2008,Diablo Canyon costs are currently estimated to increase at less than the rate ofinflation or otherwise be competitive with new power plant costs. If the plant is not operated beyond 2008, it is likely that it will be necessary to construct new baseload capacity. Accordingly, by current estimates, continued operation of Diablo Canyon through the proposed license terms would reduce future electric rate increases.
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33 AIR ES11SSIONS Substantial reduction in air emissions would result from the proposed amendment because the air emissions from natural gas or other fossil fired replacement power source to replace a prematurely closed
) Diablo Canyon would be greater than from Diablo Canyon. If PG&E were to operate a gas fired generator instead of Diablo Canyon, an additional annual emission of approximately 14 billion pounds of carbon dioxide,2.5 million pounds of nitrogen oxides, and 70,000 pounds of sulfur-dioxide would occur. Continued operation of Diablo Canyon will avoid these emissions.
) 3.4 STATE AND LOCAL ECONOS1Y The Diablo Canyon plant is the largest private employer and tu source in San Luis Obispo County, California. This prominence is expected to continue through 2025. PG&E fully recognizes its role as a major economic base for the community. Currently, annual property and related taxes on Diablo
) Canyon tatal about $60 million. The majority of these tax revenues goes to San Luis Obispo County and represen s about 25 percent of the total county property tax collections. The remainder is distributed throughou: the rest of the state of California, where it represents about 0.2 percent of the total state property tax collections. PG&E currently estimates that Diablo Canyon property taxes will remain substantial throughout the operational lifetime of the plant.
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Equally important, Diablo Canyon expends approximately $100 million annually for vendor goods and services. The plant has over 1,500 full time employees as well as 500 contract employees on a year .d basis with an annual payroll of approximately S190 million. During refueling ouwges, which occur approximately every 18 months for each unit, nearly 1,000 additional contract personnel are employed.
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These benefits are anticipated to continue through the prop'osed license terms.
4.0 SAFETY EVALUATION D
4.1 INTRODUCTION
The purpose of this safety evaluation is to determine whether the proposed 40 year operating license terms would adversely affect the health and safety of the public. Most of the information that follows summarizes existing programs and activities that have been previously approved or provided to the NRC.
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1 Section 4.2 reviews documents and programs that assure continued safety through the operating lifetime of the plant. Section 4.3 summarizes the plant performance and safety record during the initial 7 years of operation. Finally, Section 4.4 reviews the assurances for continued functional capability of safety-related components, systems, and structures through at least 40 years of plant operation. l These activities and programs provide assurance that continued operation for a 40-year term will be consistent with protection of the health and safety of the public. ;
i 4.2 LICENSING BASIS DOCUMENTS / PROGRAMS 4.2.1 FSAR and Technical Speci0 cations The Diablo Canyon Final Safety Analysis Report (FSAR) was originally submitted to the NRC (then AEC) in September 1973 to support the operating license applications for Units 1 and 2. ~ The FSAR included facility and system design descriptions, site characteristics, analyses of design basis accidents, and descriptions of plant operations. The NRC review of the FSAR was documented in a Safety Evaluation Report (SER) issued in October 1974 and SER Supplements 1 through 34 issued periodically through June 1991. In 1984, a major update of the FSAR (termed the FSAR Update) was submitted to the NRC in accordance with 10 CFR 50.7)(e). The FS AR Update is revised annually to reDect changes in plant design, evaluations, and analyses. The latest revision (Revision 7) was submitted to the NRC in September 1991.
Separate Technical Specifications were issued for Unit 1'in 1981 and 1984 and for Unit 2 in 1985.
Subsequently, combined Technical Speci0 cations applicable to Units 1 and 2 were issued in August 1985 as Appendix A to the operating licenses. Since that time, PG&E has implemented a comprehensive Technical Speci0 cation improvement Program for Diablo Canyon designed to reduce the likelihood of reactor trips and forced outages, while maintaining a his level of operational safety. The program involves evaluation of plant systems and operations to identify desired improveinents to individual Technical Speci0 cations. The improvements have included eliminating unnecessary on line testing, increasing surveillance test intervals for enuipment, revising tolerances and serpoints, eliminating or relocating unnecessary speci0 cations, supporting plant betterment design changes, and allowing use of improved fuel designs. To date, over 60 amendments to the Technical Speci0 cations have been requested and issued.
4.2.2 Probabilistic Risk Assessment The Diablo Canyon probabilistic risk assessment (DCPRA) is a full scope, Level I risk assessment. It has recently been updated and extended to a Level 2 assessment addressing Individual Plant Examination (IPE) requirements, as specified in Generic Letter 88 20 and NUREG 1335.
In addition to using the DCPRA to meet IPE requirements, the DCPRA is used, as appropriate, for other goals and objectives, such as to: .
- Support Technical Specification changes, especially changes to allowed outage times and surveillance test intervals Evaluate continued plant operation under nonconformini conditions by quantifying effects on plant risk I
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- Evaluate certain proposed design changes that are intended to improve plant safety or reliability
- Provide insights to the Training Department for use in developing operator training plans For these applications, it is necessary to maintain an up to date PRA. Rus, a process has been developed and implemented to update the DCPRA on a periodic basis (every 18 months) to ensure it represents the current plant configuration. De update process will include revisions to:
- DCPP specific initiating event data
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- DCPRA models and databases based on a review of DCPP design changes and plant procedure I
changes In summary, the DCPRA will be used to continually enhance and improve operation throughout the proposed 40-year operating license terms.
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4.2.3 Surveillance and Maintenance Programs In accordance with the Technical SpecificatioJ and the requirements of Title 10 of the Code of Federal Regulations, Diablo Canyon has established programs for maintenance and surveillance of safety-related equipment. These programs include the Inservice Inspection (ISI) Program, inservice Testing (IST)
) Progran, Environmental Qualification (EQ) Program, and Maintenance Program. These programs assure that any significant degradation of plant equipment will be promptly identified and corrected throughout the praposed 40-year operating license terms.
l y 4.2.3.1 ISI and IST Programs The Diablo Canyon 151 and IST Programs were initiated in 1985 for Unit I and 1986 for Unit 2, corresponding to the start of commercial operation of each unit. The ISI and IST Programs comply with the requirements of 10 CFR 50.55a(b)(2) and 50.55a(g). The 151 and IST Programs also comply with the requirements in the Diablo Canyon Technical Specifications. The ISI and IST Programs include ;
) inspection, testing, and maintenance of pressure-retaining components (including their support structures) l as required by the American Society of Mechanical Engineers Boller and Pressure Vessel (ASME B&PV)
Code for Class 1,2, and 3 systems. Components that are within the scope of the IST Program are Class 1,2, and 3 pumps and valves that are required to perform a specific function in shutting down the reactor or mitigating the consequences of an accident.
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As part of the inservice inspection effort, a Preservice Inspection (PSI) Program for Class 1,2, and 3 systems was conducted on Unit I which, to the extent practicable, complied with the requirements of- 4 ASME B&PV Code,Section XI,1974 Edition including the Summer 1975 Addenda. For the PSI piping examinations, the examination technique of Appendix 111 and the acceptance criteria ofIWB 3514, both from the Winter 1975 Addenda of the ASME B&PV Code,Section XI, were used. For Unit 2, a PSI
) Program for Class 1,2, and 3 systems was conducted which, to the extent practicable, complied with the requirements of ASME B&PV Code,Section XI,1977 Edition including the Summer 1978 Addenda.
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The first 10-year inspection interval, which began in 1985 for Unit I and 1986 for Unit 2, meets the
When examination techniques differ due to code changes between the PSI and ISI examinations, the 151 '
exar-ination data are used as the new baseline.
The ISI Program includes visual, surface, and volumetric examinations. The surface examinations are l done with the liquid penetrant or magnetic particle methods. The volumetric examinations are done using !
the ultrasonic or radiographic examination methods. The objectives of these examinations are to:
i e Identify unexpected service-induced component degradation, which would be evidenced by surface cracks, wear, Corrosion, or erosion e Locate any evidence of componer t leakage during system pressure or functional tests
- Verify operability of components and integrity of component supports Records of inspections completed under the 15! Program are kept in accordance with the requirements of ANSI Standard N45.2.9 and ASME Section XI and are transmitted to the NRC following each inspection.
10 CFR 50.55a(g) requires revL.on of the ISI and IST Programs as necessary to comply (to the extent practical within the limitations of design, geometry, and materials for construction of components) with the edition of the ASME B&PV Code and Addenda in effect and adopted by the NRC 12 months prior to the start of each 10-year inspection interval. These programs ensure that pressure-retaining components will be adequately inspe:ted, tested, and maintained throughout the proposed 40-year operating license terms.
4.2.3.2 EG Progmm Environmental qualification (EQ) is a rigorous program of testing, analysis, and maintenan:e to confirm that electric equipment relied on in the event of an accident will be capable of performing its design safety function, despite exposure to the harsh environment resulting from the accident.
The EQ Program for DCPP complies with the requirements of 10 CFR 50.49, " Environmental Qualification of Ele:tri: Equipment important to Safety for Nuclear Power Plants." As applied to DCPP, 1 10 CFR 50.49 requires that electric equipment important to safety and located in a harsh environment be environmentally qualified, at a minimum, in accordance with IEEE Trial Use Standard 323-1971 and the Category 11 positions in NUREG-0588 ("For Comment" version, dated December 197.9).
In accordance with 10 CFR 50.49 (1), replacement equipment (for equipment that is required to be environmentally qualified) is required to be qualified in accordance with IEEE Standard 323-1974 and I the Category I positions in NUREG-0588 ("For Comment" version, dated December 1979), unless there l are sound reasons to the contrary. j The DCPP EQ Program is a continuing program. The master list of equipment required to be qualified is maintained as a controlled engineering drawing and is revised as plant design changes are implemented.
Detailed EQ files do:ument the results of the testing and analyses that substantiate that the equipment will perform as required in a:cident environments. Surveillance activities are performed to detect adverse trends in aging or performance. Maintenance procedures assure that the qualified configuration of 540SS 85K 6
l E equipment is restored after maintenance. Equipment that is not designed for the entire 40-year operating license terms is refurbished or replaced prior to exceeding its qualified life.
Supplements 15 (dated September 1981) and 31 (dated April 1985) to the Diablo Canyon Safety Evaluation Repor6 (SER) provide the NRC Staff's evaluation of the Diablo Canyon EQ Program. In Supplement 31 to the SER, the Staff concluded that the Diablo Canyon EQ Program is acceptable and that compliance with 10 CFR 50.49 has been demonstrated. Supplement 31 also noted that the EQ ;
program had been expanded to include Regulatory Guide 1.97 Category 1 and 2 instrumentation. De NRC Staff's findings were premised on the continuing nature of the Diablo Canyon EQ Program (e.g.,
replacement of equipment prior to expiration ofits qualified life), without regard to the remaining license period. Rus, no Staff firidings regarding the adequacy of the EQ Program are altered by this amendment request.
In summary, the EQ Program ensures that electric equipment important to safety within the scope of 10 CFR 50.49 will be adequately qualified and maintained, and thus capable of performing required safety functions throughout the proposed 40-year operating license terms.
4.2.3.3 Maintenance Program
! hiaintenance is the integrated means of maintaining the plant material condition throughout its 40-year i operating life. It consists of those maintenance, surveillance, engineering, and operations activities I necessary to control normal degradation that occurs with time and use.
PG&E has developed and implemented comprehensive programs to manage the effect of aging and service wear on Diablo Canyon's systems, structures, and components (SSCs) throughout the 40-year operating l life of the plant. The programs provide the methods for inspection, surveillance, and monitoring of the l plant to detect aging effects beftre the loss of system function and to effect maintenance and component replacement practices for mitigating the effects of degradation.
l It is recognized that SSCs can deteriorate as 'he service life of the plant increases. This recognition forms l the basis for regulatory requirements, industry codes and standards, and the development of PG&E's ,
l hiaintenance Program and its component categories. Design features have been incorporated into Diablo l l Canyon that provide the ability to test, inspect, and perform preventive and corrective maintenance on l these SSCs. As a rewit, hiaintenance Program practices provide assurance that any unexpected degradation in plant equipment will be identified and corrected.
1 The hiaintenance Program was developed with the underlying philosophy that it is necessary to have the requisite administrative and technical controls to ensure that maintenance is performed in a timely and
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safe manner consistent with applicable license and quality control criteria. This philosophy is important as the purpose of the hiaintenance Program is to maintain the continued functionality and safe operational performance of all plant equipment throughout the 40-year operating life.
The hiaintenance Program is implemented through procedures developed in conjunction with the Diablo Canyon Quality Assurance Program and PG&E's operational philosophy. These procedures incorporate relevant information from the Technical Specifications, design basis criteria, and NRC Safety Evaluation Reports. ney provide means to monitor, inspect, maintain, and test plant SSCs in a programmatic manner.
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hiaintenanse tasks are categorized as preven;ive maintenance or corrective maintenance. Preventive and corrective maintenance tasks were identified and initiated during the construction phase by PG&E General ?
Construction personnel. Maintenance continuity has been maintained throughout the construction phase, system turnover to plant staff. *he period from operational testing through the start of commercial operation, and during commerciau operation.
Maintenance tasks for equipment (and systems) are largely based on the applicable manufacturer maintenance recommendations, applicable industry standards and experience, plant maintenance history, and PG&E experience. Additionally, Westinghouse (the manufacturer of most major plant components) was present onsite during the construction phase and has provided preventive maintenance information and guidelines for the Nuclear Steam Supply System equipment it supplied. The Westinghouse guidelines include recommended chemistry controls and system layup conditions.
In addition to procedural guidance on speci0c maintenance activities, maintenance procedure' s also provide '
for scheduling, implementing, and documenting activities covered by the Maintenance Program. PG&E installed a computer-based Plant Information Management System (PIMS) prior to the start of commercial operations to assist plant maintenance and engineering personnel in these tetivities. PG&E also committed additional resources to develop state-of-the-art machine shops, maintenance training facilities (which include extensive laboratories for training technicians), spare parts inventories, and management systems. Staff resources and personnel training are provided to fully implement and use these support resources.
De 151, IST, and EQ Programs, discussed in Sections 4.2.3.1 and 4.2.3.2, are integral parts of the Mairmance Program. Interdisciplinary activities in the areas of surveillance, engineering, and
, operations, along with a description of program enhancements, are presented below.
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! Surveillance and Predictive Maintenance l Surveillance test praedure', (STPs) are implemented by the DCPP Maintenance Groups, Plant Engineering, Operations, Chemistry, Radiation Protection, and Plant Support Services groups. The STPs include those tests that are required by Technical Specifications, licenses, and other documents regulating the operation and maintenance of the plant, in addition, predictive maintenance programs include vibranon analysis, acoustic analysis, oil analysis, ferrography, thermography, and component monitoring (using both on-line and off-line performance monitoring systems).
Encineerine Engineering is integrally involved in all aspects of the Maintenance Program. Engineering support is integral to the maintenance process in assessing the effectiveness of maintenance, in-depth failure analysis, configuration control, and equipment qualification. Engineering provides the design and construction support for plant improvements and the retirement of equipment that is degraded or approaching obsolescence.
Engineering also provides evaluation and analysis support for SSCs with specific aging problems that are identified at the plant or through industry experience or research. Engineering has generated the programmatic solutions to mitigate these problems and incorporate solutions into maintenance and monitoring practices to permit management and control of the aging process to within specifications.
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Onerations Plant operations is also involved in the maintenance of plant equipment. Diablo Canyon procedures and practices emphasize the need to control the manner in which equipment is operated as well as its environment. Equipment life is prolonged thrcugh proper control of equipment environmental factors, i
rigid chemistry controls, and proper equipment layup when not in service. Operational procedures have j been developed to minimize the stress placed on equipment through transients and to prevent error-Induced stress initiators. Other operational practices contribute to prolonging the service life of equipment such as balancing the run time on redundant equipment, minimizing unnecessary operation, and monitoring and maintaining external environmental factors including temperature, humidity, moisture, ,
and foreign contaminants. I Enhancements to the Maintenance Procram The Maintenance Program requires continual adjustment as plant specific performance is evaluated, industry experience is gained, and new methods and technologies develop. The Maintenance Program at DCPP is a living program that has evolved continuously through the construction and operation phases, Maintenance Program task improvements have been developed as a result of the equipment failure analysis process (root cause analysis and corrective action program), component failure reports, evaluation of industry operating experience, and through better understanding of aging and failure mechanisms i
applicable to Diablo Canyon equipment. These program improvements are proceduralized and receive significant engineering support and evaluation.
Examples of program enhancements include:
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Predictive maintenance programs have been expanded and are managed by a dedicated staff. These programs allow greater reliance on condition monitoring to enablish preventive maintenance activities.
A Reliability Centered Maintenance Pr) ject has been implemented to optimize the maintenance !
program and improve plant / system reliability and availability. This program uses risk based methods
) for identifying the necessity and importance of maintenance on components, allows optimization of maintenance resources, and emphasizes the use of predictive maintenance methods.
Conduct of an additional review of tne latest revisions of component vendor manual maintenance recommendations to update maintenance tasks and validate the component programs.
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- Establishment of a computerized component database on PIMS that provides the basis for the performance of preventive maintenance tasks. This effort was an enhancement to the PIMS system, which already contained a component database and history section. l Implementation of a long-term master schedule to plan major component overhaul and replacement
) strategies.
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- Implementation of major preservation program activities, such as those for concrete structures.
Exa'mples of engineered solutions to improve equipment performance through improved maintenance and surveillance include:
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e Implementation of inspection and monitoring for piping deterioration in.the component cooling.
water / auxiliary saltwater (CCW/ASW) system.
e Implementation of a program to replace salt water piping with lined steel pipe in the intake area.
e Upgrading of bolting, e Performance of piping inspections (e.g., ASW system) with robot cameras.
I o Development and implementation of component diagnostic or inspection programs; for example: ;
- Motor-operated valve testing and diagnostics.
- Emergency diesel generator engine analyzer diagnostic testing.
- Check valve inspections.
l - Valve internal leakage evaluations using acoustic monitoring. i i - Reactor coolant pump seal inspections. -l e Improvement of steam generator maintenance and inspection practices and implementation of design modifications to inspect for and mitigate the effects of various aging mechanisms or mechanical problems related to generator design and construction.
1 e Implementation of a piping erosion / corrosion monitoring program in the condensate, feedwater, and '
extraction and heater / moisture separator reheater drain systems using EPRI-developed techniques.
l In summary, the comprehensive Maintenance Program at DCPP ensures that both normal and unexpected -
degradation of plant equipment will be promptly identined and corrected.
4.3 PLANT OPERATING HISTORY Diablo Canyon has been in commercial vperation for approximately 7 years. During this time, a significant amount of data and experience has been accumulated that demonstrates the safety and reliability of the plant. Numerous plant modifications have also been made to improve reliability and upgrade safety related equipment. Since beginning operation, PG&E has invested about 100 million dollars per year on an aggressive capital equipment upgrade and replacement program. Operating performance, plant modi 0 cations, and regulatory performance are reviewed in the following' sections.
4.3.1 Operating Performance
'Ihe lifetime capacity factor for Diablo Canyon through its first 7 years of commercial operation is approximately 77 percent. Achievement of reliable operating performance, in the form of a high capacity -
factor, has always been a PG&E goal for Diablo Canyon because PG&E is convinced that high capacity factors are the result of a well managed, safely operated nuclear plant. Keeping refueling outages as short as possible through sound outage planning, while safely and effectively performing all necessary work, has also contributed to high capacity factors for Diablo Canyon. Overall industry experience supports our conclusion that units with consistently high capacity factors, relatively short refueling outages, and low forced outage rates are well maintained, follow good operating practices, and thus can be expected to operate at high levels of safety.
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In addition to capacity factors, there are other performance indicators commonly used to monitor performance: unplanned automatic reactor trips, unplanned automatic safety system actuations, collective radiation exposure, significant events, safety system failures, forced outage rate, and equipment forced i
outages /1000 critical hours. As indicated in the NRC's quarterly publication of industry performance indicators, Diablo Canyon currently compares favorably with the industry 2verages for these indicators.
4.3.2 Reliability / Safety Related Plant Modifications
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A number of major plant modifications designed to improve reliability or upgrade safety related equipment have been made during the 7 years since Diablo Canyon began operation. Several others are planned in the near future. Some of the more significant mulifications include:
) moisture separator reheaters. These changes resulted in the removal of essentially all copper from the secondary side of the plant to increase the life of the steam generators.
- Condensate Polisher Addition - To further increase the life of the steam generators, a Condensate Polisher System was added to process secondary water by ion-exchange. Regeneration of the ion-exchange resin in the condensate polisher system results in additional solid waste and liquid effluents.
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- Ammonium Hydroxide Storage - To regenerate condensate polisher resin, a 6,000 gallon bulk storage tank for ammonium hydroxide was added.
- SG Blowdown Rate Increase - To improve secondary water chemistry and thus increase the expected
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> life of the steam generators, the blowdown rate has been increased.
- Control Room Upgrade - A detailed control room design review (DCRDR) was performed in accordance with the requirements specified in Supplement 1 to NUREG-0737. Weaknesses in the man machine interface between control room operators and equipment were identified in the DCRDR.
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Following review and approval by the NRC, control room equipment upgrades have been implemented.
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- High Density Spent fuel Pool Racks The original fuel racks in each unit's spent fuel pool were replaced with high density racks, increasing the capacity in each spent fuel pool from 270 to 1324 fuel assemblies.
- Improved Fuel Design - The reactor fuel used in each unit is being replaced with an improved VANTAGE 5 Westinghouse design. Most of the fuel in each reactor will be of the VANTAGE 5 design by early 1993.
- Boron Injection Tank (BIT) Removal-in response to industry experience and NRC recommendations,
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the BITS in both Diablo Canyon units were removed from service to reduce the potential for boric acid crystallization in ECCS piping and valves which could potentially have degraded safety-related equipment operability.
- Reduced Boric Acid Concentration - The boron concentration in the Boric Acid System has been
) reduced from 12 to 4 weight percent to reduce the potential for boric acid crystallization in safety-related components.
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e Digital Feedwater Control System - A digital feedwater control system was installed to improve'-
feedwater control performance and reliability. The enhanced feedwater control features provided by .
this system reduces the likelihood of steam generator level related reactor trips; Implementation of the' steam generator level median signal select feature prevents interaction between *he feedwater
) control and reactor protection systems, which allowed deletion of the low feedwater flow reactor trip ~ !
function. .
e Computer Replacement The original plant process computer was replaced with one having impro man-machine interface, greater capacity, faster response time, improved print and report capability, '
) improved retrieval of historical data, and complete redundancy to prevent loss ofinformation due single failure. ,
e A7WS Mitigation System Attuation Circuitry (AMSAC) - ne AMSAC System was installed in both Diablo Canyon units to ensure reactor protection during an anticipated transient without scram (ATWS) event that results in the loss of the secondary side heat sink. 'AMSAC is designed to. trip[
) the main turbine, initiate auxiliary feedwater flow, and isolate steam generator blowdown and sample .
lines during an ATWS with a low steam generator level condition.
e Chlorination System Modifications - Modifications to the Chlorine System include (1) the use ofliquid -
hypochlorite to control microbiofouling instead of gaseous chlorine, (2) implementat!on of continuou chlorination of the auxiliary saltwater system to control macrobiofouling (invertebrate marine life),
) and (3) possible use of intermittent injection of a chlorins/ bromine mixture to prevent macrofouling ,
in the Circulating Water System. .
e Fatigue Monitoring - PG&E is installing an on line fatigue monitoring system at D_lablo Canyon that j will continuously analyze plant operational data to track fatigue usage of critical reactor, coolant -
system components.
e Additional Diese. Generator Addition of a sixth diesel generator will provide each unit witn three ,
dedicated diesel generators. This will enhance reliability of the onsite power distribution system by .
eliminating dependence on a swing diesel generator and the associated procedural. complexities.
), Installation, testing, and tie in of the sixth diesel is scheduled for completion in the spring of 1993. ,
e Plant Process Protection System Upgrade - This project will upgrade the Process Protection System '
by replacing the existing HAGEN 7100 equipment with a Westinghouse Eagle 21 system. New steamline break logic and steam generator low low level trip time delay options will be included in the upgrade. These changes will improve the reliability and availability of the Process Protection .
j System. The digital microprocessor-based system with computer-enhanced testing will also m the likelihood of personnel error during surveillance testing. System installation is sch'eduled for the spring of 1994 for Unit 1 and the fall of 1994 for Unit 2.
l e R7D Bypass Elimination - This project will replace the resistance temperature detector (RTD) bypass loop piping with fast response RTD. t installed in the hot and cold legs of the Reactor. Coolant Syste
) Plant downtime and radiation exposures will be reduced and numerous snubbers can be eliminated.
l Installation is scheduled for the spring of 1994 for Unit 1 and the fall of 1994 for' Unit 2.s J
e Radiation Monitoring System (RMS) Upgrade - His project will upgrade the present RMS to improve performance, reliability, and capability. Most of the work is scheduled to be completed by 1l j
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-- . .-. - .. -_ -- ..- - .- - - . ~ -,
i l 4.3.3 Regulatory Performance The latest Systematic Assessment of Licensee Performance (S ALP) report for Diablo Canyon covered the l
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period of January 1990 through June 1991. The S ALP Board found the performance oflicensed activitie" at Diablo Canyon to be very good, in some cases to be superior, and clearly directed toward safe facility operation.
) 4.4 ASSURANCE OF CONTINUED FUNCTIONAL CAPABILITY OF SAFETY RELATED f COMPONENTS ]
Assurance of an acceptable level of safety throughout the proposed 40-year operating license terms is provided by the continued functional capability of safety related components. These are components ,
associated with systems designed to prevent or mitigate events that could cause a release of radioactivity j
) to the environment. The following discussion reviews such components at Diablo Canyon.
4.4.1 Reactor Coolant System Pressure Boundary l l
4.4.1.1 General The mechanical components associated with the Reactor Coolant System (RCS) pressure boundary include the rea: tor vessel, pressurizer, portions of the steam generators, piping, valve bodies, and pump casings.
The design of these components incluued consideration of potential effects of age-related phenomena such j as corrosion, thermal and vibration fatigue, and radiation indxed embrittlement. Consideration of these effects was also taken into account when the operating limits and surveillance requirements were established in the Te:hnical Specifications. In accordance with the latter requirements, the RCS is included in the Inservice Inspection Program (see Section 4.2.3.1). Components are located such that critical areas are sufficiently a:cessible for the required inspections and/or tests.
b The potential for corrosion was accounted for by using corrosion resistant materials in the design of the plant. All mechani:a! components that are in contact with reactor coolant, except the fuel, are either made of or clad with austenitic stainless steel or ni:kel-based corrosion resistant alloys. The fuel is clad with Zircaloy. The RCS water chemistry is managed to minimize corrosion. Analysis of the coolant is routinely performed to verify that it meets the specifications for chemistry, radioactivity level, and boron concentration.
Components of the RCS pressure boundary are designed to withstand the fatigue effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. During startup and shutdown, the heatup and cooldown rates are limited to less than 100'F/ hour, consistent with system design specifications and
) the Technical Specifications. Deviations from these conditions are evaluated to determine complian:e with code requirements.
Plant procedures provide for the ongoing collection of data relative to thermal transients. nese data are compared with original design requirements to assure the adequacy of the remaining fatigue life of RCS components. PG&E is in the process of installing an online fatigue monitoring system at DCPP that will
) continuously analyze plant operation data to tra:k fatigue ussge of the critical RCS components.
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l 4.4.1.2 Reactor Vessel 1 l
The Diablo Canyon reactor vessels were designed and fabricated in accordance with ASME Boiler and Pressure Vessel Code, Section 111, ' Nuclear Power Plant Components." hey were designed for transients considered to envelope design conditions over a 50 year operating period. To ensure the continued integrity of the vessels during operation, an ISI Program (see Section 4.2.3.1) has been in place since plant startup, ne 151 Program requires volumetric examination of each accessible pressure boundary weld at least once in each 10 year interval.
) De effects of neutron radiation embrittlement of the vessel beltline region are considered in the design and operation of the units. Compliance with all NRC regu:ations governing vessel integrity has been documented recently in PG&E's response to Generic Letter 92-01 (PG&E Letter No. DCL-92-150, dated ;
June 30,1992). In addition, PG&E has instituted an Embrittlement Management Plan to manage reactor !
vessel embrittlement throughout the entire operating life of Diablo Canyon. l
) Pressurized Thermal Shock Following Cycle 1 for each unit, the neutron fluence at the reactor vessel inner wall was reduced by installing increasingly lower neutron leakage cores, thus decreasing the reactor vessel rate of embrittlement and prolonging vessel life. The Diablo Canyon reactor pn.ssure vessel beltline materials have been evaluated according to the NRC's Pressurized Thermal Shock (PTS) screening criterion defined in 10 CFR 50.61. The Reference Temperature for Pressurized Thermal Shock, RTm, has been calculated for each weld metal and base metal in the DCPP beltline regions for neutron fluences corresponding to 40 operating years. The RTm for ali materials will not exceed the screening limit of 270*F for base metal and longitudinal welds and 300'F for circumferential welds. Since all materials a
meet the screening criterion, neither additional Oux reduction nor plant specine PTS analyses are required ;
) to comply with the PTS rule. Details of the PTS evaluation have been submitted to the NRC (PG&E L tier No. DCL-92-056, dated March 6,1992).
Based on a conservative Cuence projection for 40 operating years, DCPP will also meet the requirements of 10 CFR 50, Appendix G. Charpy Upper Shelf Energies were determined (FSAR Update Tables 5.2- I
) 19A,5.2-19B,5.2 21 A, and 5.2 21B)in accordance with Regulatory Guide 1.99, Revision 2. All Diablo Canyon beltline materials will remain above the 50 ft-lb Charpy Upper Shelf Energy fracture toughness requiremm for more than 40 operating years. In addhion, reactor vessel pressure temperature limits will meet 10 CfR 50, Appendix G requirements for 40-year opezation without requiring plant modi 0 cation or imposing operational restrictions.
i I
) Material Surveillance Procram l De toughness properties of the reactor vessel beltline material will be monitored throughout the proposed 40-year operating license terms with a material surveillance program that meets the requirements of 10 CFR 50, Appendix H.
The surveillance test program for DCPP Unit I complies with ASTM E 185-70, the standard in effect when the vessel was manufactured. Although the Unit I surveillance program was designed prior to the existence of 10 CFR 50, Appendix H, it does contain the signincant features required for later surveillance programs and will effectively monitor vessel embrittlement throughout the requested license period. The program includes a total of eight surveillance capsules. Three of the eight capsules contain
) the limiting weld metal and base metal, correlation monitor material, dosimeters, and thermal monitors.
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The remaining five capsules contain the limitmg base metal, but no wel charpy specimens in the capsules are longitudinally oriented.
The Unit 2 surveillance program includes six capsules and conforms . All capsules to ASTM
) contain the limiting weld metal, which is the limiting beltline material. The b capsules are not from the limiting plate, but were machined from an adj the same heat treatment, and similarlevel of embrittlementn (adjusted re at plate end.oflife as the limiting plate.
To date, one surveillance capsule from Unit I and two capsules from Un
) Analysis of these capsules confirms that the measured shifts,in nilm,ductility ref ,
predicted by Regulatory Guide 1.99, Rev. 2.for the limiting s materials In addition to those required surveillance programs, a supplemental surveillanc implemented for Unit 1 beginning with Cycle 6 in 1992. The supplemental prog
) the plant operating life. surveillance capsules that will provide additional da Neutron Dosimetry Program. Additional measures to monitor DCPP Unit consists ofirradiating and evaluating reactor cavity dosimetry, Results obtained are used to confirm and complement su .
The overall program to monitor reactor vessel beltline materials is thorough an degree of embrittlement of beltline materials over the p 1
) .
4.4.1.3 Reactor VesselInternals 4
h The design of the reactor vessel internals meets the intent of Section 111 of the Vessel Code. The material used for fabrication of most of the rea treated, unsensitized Type 304 austenitic stainless steel conforming toWe;J ASTM specific fabrication was done using procedures and personnel qualified in accordance w ASME Boiler and Pressure Code.
Evaluations performed prior to initial plant startup document the during normal operation, abnormal operational transients, an Periodic inspections performed under the 151 Program ensure that any signific vessel internals over the proposed 40-year operating license terms will be dete timely manner.
) 4.4.2 Other h!echanical Components i The passive mechanical components (tanks, pump casings, and valve bodies)
( systems are designed to meet the intent of Regulatory Guides 1.26 and 1.29. Consid d to possible aging effects including corrosion. erosion, and thermal cycling fatigue The 540SS/85K 15
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life of these passive components is greater than 40 years, as is the service life for the RCS pressure boundary. Nevertheless, such components are included in the plant ISI and Maintenance Programs, so that any unexpected degradation will be identified and corrected should it occur. Many of the active (moving or rotating) mechanical components, such as pumps and valves, are expected to wear and are h therefore periodically tested and maintained under the IST and Maintenance Programs. Degradation of these components will therefore be identified and corrected, and component functional capability will be maintained.
In summary, passive mechanical components are designed such that they are not expected to be replaced over the 40-year operating license terms, while the functional capability of active components will be
) maintained through inservice testing, maintenance, and/or periodic replacement.
l 4.4.3 Electrical Components '
) Electrical components that are required to function in a harsh environment (significantly more severe than the environment that would occur during normal plant operation, including anticipated operational occurrences) during a design basis event are covered by the EQ Program. The EQ Program complies with the requirements of 10 CFR 50.49, as discussed in Section 4.2.3.2.
.g Safety-related electrical equipment at Diablo Canyon that is not covered by the EQ Program is covered D by the plant Maintenance Program (Se: tion 4.2.3.3). As required by the Technical Specifications, equipment is tested periodically in a:cordance with surveillance test procedures. Equipment is replaced as required, j 4.4.4 Structural Components 4.4.4.1 Primary Containment The structural integrity of Diablo Canyon ** containment steel liner, the concrete shell, and internal concrete and steel structures is assured for a 40-year operating period for the following reasons:
)
- The containment liner, the con: rete shell, and internal concrete and steel structures were designed and constru:ted in a:cordance with Amerien Concrete Institute (ACI), American Institute of Steel Construction (AISC), and ASME Section Vill Codes to withstand a:cident load combinations, including postulated seismic and loss-of-coolant accident events. These load combinations have
) resulted in mu:h stronger primary containment structuren than required to support normal operating loads.
e Construction was carried out in accordance with strict construction and quality control procedures.
Acrual material testing results indicate a high level of quality in construction techniques and materials used to construct the containment structure. For example, the concrete strength tests for the
) containment structure indicate that the in situ concrete strength is approximately 25 percent higher than the nominal design values used in the design calculations.
l e Through the first 7 years of operation at Diablo Canyon, minor maintenance to concrete surfaces and protective coatings has been performed to ensure the continued structural integrity of the primary contamment. The need for this work has been detected through the following actions:
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Inspections and testing required by plant procedures. _
Periodic inspections of protective coatings for concrete and steel surfaces required by plant procedures.
Periodic performance oflocal leak rate and integrated leak rate tests required by the Technical
? Specifications.
The inspection and documentation process is designed to detect unacceptable concrete cracking, concrete spalling, or deterioration of protective coatings. Repairs are made as needed. Root cause evaluations are performed as required to ensure other potential problem areas are addressed.
I e A review of the performance of other principal concrete structures in the U.S. nuclear industry has.
not yielded experience data that would indicate an adverse impact on the structural integrity of the primary containment for the proposed 40-year operating license terms.
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p American Concrete Institute, January 1989) for nuclear power plant concrete structures, service lives j of 60 or more years are possible, provided that appropriate preservation activities are performed.
l 4.4.4.2 Other Structures The structural integrity of Diablo Canyon's other critical plant structures, such as the auxiliary and turbine buildings and the intake structure, is assured for the proposed 40 year operating license terms for the following reasons:
( e These critical plant structures consist of reinforced concrete and/or structural steel. "Ihese structures were designed and constructed in accordance with ACI and AISC codes to withstand abnormal design.
)
l load combinations. These load combinations have esulted in much stronger structures than required -
to support normal operating loads.
e Construction was carried out in accordance with strict construction and quality control procedures.
Actual material testing results indicate a high level of quality in construction techniques and materials
)
used to constmet these structures.
e These plant structures and associated protective coatings are periodically inspected and maintained -
- to ensure continued structural integrity.- Through the first 7 years of operation, the intake structure -
! has been the only concrete structure to show some degradition. An evaluation by PG&E showed that - i j minor refurbishment would restore the intake to its original condition. The intake structure is currently being refurbished to ensure its continued structural integrity, in additien, PG&E is ,
enhancing marine inspection and repair procedures to better preserve the integrity and serviceability l of the structure.
, e A review of the performance of other principal concrete structures in the U.S. nuclear industry has ;
) not yielded experience data that would indicate an adverse impact on the structural integrity of critical plant structures for the proposed 40 year operating license terms.
- Based on research presented in ACI-SP ll7 for nuclear power plant concrete structures, service lives i
of 60 or more years are possible, provided that appropriate preservative activities are performed for j these structures.
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5.0 ENVIRONMENTAL EVALUATION
5.1 INTRODUCTION
PC 2 has reviewed and assessed information contained in the " Final Environmental Statement related to the Nuclear Generating Station, Diablo Canyon -Units I and 2" (FES) and Addendum, the Environmental Report and the eight Supplemental Environmental Repons, the Final Safety Analysis Report Update, and other studies and data accumulated over the past years of operation to ensure thorough and complete evaluation of potential environmental issues related to the proposed 40-year operating license terms.
He environmental evaluation is divided into six additional sections. In Section 5.2, systems and .
programs for environmental control and monitoring are evaluated to ensure they meet applicable regulatory criteria and show evidence of continued enhancement and effectiveness. De systems and programs evaluated include the Waste Processing System, ALARA Program, Process and Area Radiation Monitoring System, Radiological Environmental Monitoring Program, and Nonradiological Surveillance Program.
Section 5.3 presents an assessment of the environmental effects of plant operation during the proposed 40-year license terms to ensure they remain within the bounds of the FES and applicable regulatory criteria and permits, or where appropriate, upper ilmits estab'ished from plant operation to date. Topics addressed include occupational and offsite radiation exposures, new fuel requirements and spent _ fuel storage, generation and transportation of spent fuel and other wastes, thermal and ecological effects of ,
the cooling water system, and the poter'ti9 impact on historic properties. I Section 5.4 considers the offsite exposures from releases during postulated accidents. These exposures i were previously evaluated in the original FSAR (presently the FSAR Update) where the results were l found to be within the guideline values in 10 CFR 100. Using projected population size tnd distribution in the vicinity of the plant for the pposed 40-year operating license terms, it is shom that 10 CFR 100 -
criteria will continue to be met.
The potential effects of population increases on emergency planning are evaluated in Section 5.5, and Section 5.6 summarizes the plant modifications that are environmentally related. Also, the effects of the proposed 40-year operating license terms on decommissioning are evaluated in Section 5.7.
5.2 SYSTEMS AND PROGRAMS FOR ENVIRONMENTAL CONTROL AND MONITORING 5.2.1 Waste Processing System De Waste Processing System is described in Chapter 11 of the FSAR Update. His system, which has been significantly upgraded since the original design, receives, processes, and safely discharges or packages radioactive waste for further offsite processing or disposal. De basic processes used are decay.
of radioactive isotopes, particulate filtration, selective ion exchange, dewatering or solidation of wet waste, filtration of gases by HEPA filters, release of low activity wastes, and compaction of dry active waste. As described in the FSAR Update, the system consists of components such as liquid and gas storage tanks, pumps, compressors, filters, ion exchange vessels, instruments, piping, and valves.
540SS/85K Ig
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i Effluent releases from the Waste Processing System have always remained well below the limits contained 1 in the Technical Specifications, and within the range projected in the Final Environmental Statement.
System performance is monitored, and design and operational enhancements have been made, where feasible, to minine effluent releases. Some examples of enhancements include optimizing the ion exchange resins ano filter pore sizes, modifying the spent resin storage tanks, and processing waste more frequently and to lower activity levels. A summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant is reported to the NRC in the Semiannual Radioactive Effluent Release Report.
This system will continue to be used throughout the proposed 40-year operating license terms to ensure wastes are safely and effectively processed.
5.2.2 ALARA Program PG&E's management is strongly committed to maintaining personnel exposures to ionizing radiation at ,
l DCPP as low as is reasonably achievable (ALARA). As a result of this commitment,'a formal l organization has been established to review, evaluate, and recommend ALARA actions for all radiological work at DCPP. PG&E has established a goal to maintain the performance indicator of plant collective
- radiation exposure, in person rem per year, to be within the Institute of Nuclear Power Operations (INPO) guidelmes.
Occur lonal exposures resulting from the proposed 40 year operating license terms will remain within the limits of 10 CFR 20. The DCPP ALARA Prcgram established in response to the requirements of 10 CFR 20.l(c) will contribute, in large part, to minimizing the already low levels of occupational j exposure at the plant.
Diablo Canyon's ALARA Program requires that an estimate of the total dose be provided for all Radiation Work Permits (RWPs). If the estimate is less than 1 person-rem, the ALARA controls are established by the person initiating the RWP. If the estimated total dose is at least 1 person-rem but less l than 10 person rem, the job receives an ALARA review by the Radiation Protection Work Planning
) Group. All jobs estimated at 10 person rem or greater require an ALARA review by the Radiation Protection Engmeering Group prior to thejob being initiated. Jobs estimated at 25 person rem or greater, or an indwidual dose of I rem or greater, require additional review and approval by the plant ALARA l Committee. '
) *ne Plant Staff Review Committee (PSRC) serves as the plant ALARA Committee, and is composed of ;
l personnel from the various Diablo Canyon departments. A separate group, the Joint ALARA Review l
l Committee (JARC), comprised of Diablo Canyon and corporate management, meets periodically to assess l the effectiveness of exposure control methods in keeping personnel exposures ALARA. Funhermore, l the JARC assists in developing ALARA policy and procedures, and monitors the implementation of ALARA measures.
b The ALARA Program is assessed after each outage in a report that lists exposures incurred on majorjobs and summarizes the ALARA lessons learned for future reference and application. The report allows the JARC to identify ALARA related inadequacies in designs or procedures used for equipment installation, operation, surveillance, and maintenance. The results of these post-job critiques provide knowledge that
) can be used to improve designs and reduce exposures on the same or similar jobs in the future.
Similarly, Diablo Canyon also sets an annual plant exposure goal based on the outage report and input 5408S/85K 19
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e
l from each plant department. Current exposures are periodically reviewed by management to identify adverse trends, thus assuring timely corrective action, when necessary. l l
Continued compliance with the 8' ARA Program at DCPP ensures that personnel exposures to ionizing radiation will be minimized over tne proposed 40-year operating license terms.
5.2.3 Process and Area Radiation Monitoring System i The Process and Area Radiation Monitoring System is described in Chapter 11 of the FSAR Update.
This system monitors radiation levels associated with process systems and areas at various locations in the containment and in the auxiliary and turbine buildings, it is designed for use during norma! operation or postulated accident situations and includes equipmeni for detecting, computing, indicating, and )
alarming. Periodic testing and inspection of the system assure its functional readiness.
l A number of radiation monitors and monitoring systems are also provided on process liquid and gas lines l
! that may serve as discharge routes for radioactive materials. The monitors include the following: )
e Main steam line radiation monitors ;
l
- Steam generator blowdown radiation monitor l
- Plant vent radiation monitors l
- Liquid radwaste radiation monitors l
- Fuel handling building radiation monitors j e Condenser off gas monitors PG&E is implementing a program to upgrade the Radiation Monitoring System to improve performance, reliability, and capability. These upgrades combined with surveillance test procedure requirements ensure l the system will provide effective radiation monitoring over the proposed 40 year operating license terms.
! 5.2.4 Radiological Environmental Munitoring Program i The Radiological Environmental Monitoring Program is described in Chapter 11 of the FSAR Update.
The program was established prior to the start of plant operation to determine preoperational background levels. The Radiological Environmental Monitoring Program is designed to validate the adequacy of safeguards inherent in plant design and the effectiveness of dose calculations, based on plant emission data and appropriate meteorological and aquatic dispersion models. Emphasis is placed on control at the source, with follow up and confirmation by environmental surveillance. *!his is accomplished by continuously measuring radiation levels and airborne radioactivity levels and periodically measuring-amounts of radioactivity in samples at various locations surrounding the plant. Results from the Radiological Environmental Monitoring Program are reported to the NRC in the Annual Radiological Environmental Operating Report.
The several types of sample media used correspond to the possible exposure pathways. These pathways are direct radiation, inhalation, and waterborne or airborne ingestion. Direct radiation is measured' 540SS/S5K 20
continuously by thermoluminescent dosimeters (TLDs). Airl.orne radioactivity is collected continuously by passing air through a fiber filter in series with a charcoal absorption media, ne filter collects particulate radioactivity, and the charcoal collects radiciodine. Waterborne radioactivity levels are 2
monitored by taking drinking water, surfac vater, and nearine samples. Ingested radioactivity is collected by obtaining samples of vegetation, milk, and fish.
In examining the distribution of radionuclides in the environment and lower trophic levels, comparisons are made with the preoperational data to determine if there are any biological or physical compartments
, in nature that are accumulating radioactivity. Similarly, external radioactivity measurements taken after plant operation are compared with the average and range of data obtained in the preoperational program.
A second comparison of external radiation measurements is made by selecting TLD stations considered 1 to be out of the radiological influence of the plant and using them as a measure of natural background exposure. Dese reference stations are compared with the remaining stations to make estimates of any increases in external exposure levels that can be attributed to operation of DCPP.
)
To ensure that the program continues to include environmental sample locations most likely to detect plant related radioactivity, a land use census is conducted annually. Changes in milk sampling locations may be required following the census based on relative potential doses or dose commitments and the availability of samples.
)
Continued environmental monitoring and surveillance under 'his program ensures early detection of any increase in exposures over the proposed 40-year operating license terms.
5.2.5 Nonradiological Surveillance Program
)
5.2.5.1 Environmental Protection Plan The Environmental Protection Plan (EPP) was developed to provide for the protection of the environment during construction and operation of DCPP. The EPP is Appendix B to the Facility Operating License and has three principal objectives: (1) to verify that DCPP is operated in an environmentally acceptable
)
manner, as established in the FES and other assessments, (2) to coordinate NRC requirements and maintain consistency with other federal, state and local requirements for environmental protection, and (3) to keep the NRC informed of the environmental effects of facility construction and operation and of ,
actions taken to control those effects. Environmental protection activities required by the EPP are !
reported to the NRC in the Annual Environmental Operating Report. Any environmental concerns
) identified in the FES concerning water quality issues are deferred to the Regional Water Quality Control Board (RWQCB), as regulated under the National Pollutant Discharge Elimination System (NPDES) permit.
The environmental issues in the EPP are categorized as aquatic issues and terrestrial issues. The aquatic issues have been deferred to the RWQCB for resolution. Presently, the RWQCB protects the water
)
quality in the vicinity of DCPP through the provisions of the NPDES permit. Terrestrial issues include erosion control in the vicinity of the transmission line rights-of way, the controlled use of herbicides on ;
transmission rights-of way, and the preservation of areas !n the vicinity of DCPP of archeological significance and the provision of access to these areas by the Chumash Indian Tribe. The EPP contains specific requirements with regard to these issues.
)
540SS/85K 21
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5.2.5.2 NPDES Permit An NPDES permit, No. CA0003751, to discharge Diablo Canyon effluents into the Pacific Ocean has been issued through the RWQCB. This permit describes all linuid discharges from the plant to the Pacific-Ocean and Diablo Creek to include cooling water, steam ge.arator blowdown, makeup water systems, seawater reverse osmosis system blowdown, radioactive waste treatment system,~ intake screen wash, biology laboratory discharge, and storm water runoff. Specific effluent limitations and monitoring requirements are established for various parameters depending on the probable constituents. Dese limitations and requirements are based on Titles 111 and IV of the Federal Clean Water Act in accordance with 40 CFR 122125 and 423, the California Water Code, and California Water Quality Control Plans.
Biological investigations, which began in 1966 in the vicinity of Diablo _ Canyon, are used to support compliance with NEPA and the Clean Water Act. These studies are also used to ensure protection and propagation of the balanced, indigenous communities of shellfish, fish, and wildlife in the Pacific Ocean in and round the Diablo Canyon vicinity. NPDES permits are issued to.Diablo Canyon every 5 years.
These permits contain requirements, as appropriate, for contiaued monitoring to ensure protection of the environment. The extensive studies to date have not identified any additional concerns with relation to the cooling water discharge to the Pacith Ocean. Required monitoring studies have been reduced in scope as accumulated information supports the absence of any significant effects.
! 53 ENVIRON 51 ENTAL 15fPACT DURING NOR51AL OPERATION 5.3.1 Occupational Radiation Esposure Diablo Canyon's occupational exposure trend and comparison with the INPO 3-year industry average goal -
) for PWRs are shown in Table 51. Diablo Canyon's average _~ annual occupational exposures ^have consistently been below the INPO goal of 288 person-rem. , As shown, aggressive implementation of the Al .mA Program has resulted in a distinct reduction in exposures for the most recent 3-year. peri 3d compared with the previous two averaging periods.
Table 51 also shows the projected occupational exposure averages per unit through the year 2000. The I
projection is based on resistance temperature detector (RTD) bypass elimination in 1994, and continued operation with an 18 month fuel cycle. Assuming 3.5 person. rem per month for non-outage periods, and savings of 50 person-rem per outage following RTD bypass elimination, the levelized 3 year average dose -
is projected'to decrease to 150 person rem per year per unit by 1997. Subsequent . annual exposure averages can reasonably be expected to remain approximately at this level, and within the INPO 1995
) 3 year industry average goal. These projected exposures are significantly less than the 450 man rem per year per unit values estimated by the Staffin the FES Addendum for Diablo Canyon.
Given Diablo Canyon's current trend of decreasing refueling outage exposures and continued emphasis on effective refueling and outage management, it is expected that the base occupational exposures-established above will serve as a realistic estimate through the proposed 40 year period of operation.
5.3.2 Offsite Radiation Exposure The proposed 40-year operating license terms will have no significant impact on the capability of the plant -
) to maintain routine releases of radioactive materials (liquid.and gases) to the environment in compliance -
with 10 CFR 20.l(c) and 10 CFR 50, Appendix 1. This conclusion is based on the calculation of annual' 540SS/85K 22 4
)
2
doses to individuals and population groups over the past 5 years of plant operation as repone '
NRC in Semiannual Effluent Release Reports. Table 5-2 provides a comparison of Appendix I rad '
exposure limits and actual maximum exposures based on plant operating data.
.l TABLE 5 ;
)
Diablo Canyon'vs.1NPO Industry' Goal Average Annual Occupational Exposure .:
i l
a Total Dose -
Refueling (nerson-rem ner reactor unit)
Year Outages
.i DCPP 3 Yr Average :INPO 3-Yr Average Goal. '
o 1986 1 151 288 ,
1987 1 168' 288 1988 ;
2 :253 288 -i 1989 1 275 288 '
1990 1 269
.288.- =5 1991 2 214- 288 ,
1992* 1 199- 288 1993* 1 195 '
288 1994* 2 218- '
288 1995" l ,202- 185' L!
1996*
'1 188 185 i 1997" .2 150 1998" 185 '!
1999*
1 1
~- 150 -
150.
'I85 I85 l
2000* 2 150 185.
- Projected, based on:
e 18 month fuel cycle operation.
- 3.5 person rem per non-outage month e 1993 based on 80% of 1992 due to dose rate differences between units -
50 person rem savings per outage due to RTD bypass elimination in.1994 i
540SS/85K 23 .
4 w w &
TABLE 5 2 4
Comparison of Offsite Appendix I Radiation g Exposure Limits and Actuul Data DCPP 5-Year Percent of Parameter Appendix I Maximum Appendix I Dose Limits Individual Dose Dose Limit (mrem) (mrem)
L. quids s3 0.031 1.0-}
Gases s 10 0.212 2.16 lodines and s 15 0.027 0.18 Particulates With regard to the plant's liquid eftluent, a review of the actual liquid ef0uent since 1986 indicates that the maximum individual dose averages no more than 0.013 mrem per year. This represents only 0.43 percent of the Appendix I objective doses. From this operating experience, the plant has clearly demonstrated its capability to process waste to a degree that ensures continued compliance with 10 CFR 50, Appendix 1, design objectives.
Gaseous eftlaents from Diablo Canyon have also resulted in doses far below Appendix I objectives.
Based on recorded noble gas efCuents over the past 5 years, the calculated maximum whole body dose i was, on average, only 0.07 mrem per year, or approximately 0.7 percent of the Apper. dix 1 dose objective of 10 mrc.r. per year. Since this average maximum whole body dose occurred at the site br.mdary, the plar.t's contribution to the total 50 mile population dose is insigni0 cant in comparison to ther from background radiation.
For iodines and particulates, dose calculations based on the actual releases over the past 5 years indicate that the average maximum annual organ dose is only 0.016 mrem per year. This represents just 0.11 percent of the Appendix ! dose criterion of 15 mrem per year to any organ. As with the noble gases, the doses from iodine and particulates also represent an insignincant fraction of the total person-rem exposure the same 50 mile population receives each year from background radiation.
Based on the design and performance history of the Waste Processing System, it is expected that offsite radiation exposures will remain within the plant's ALARA criteria through the proposed 40-year operating license terms. The projected exposures are also well within the offsite exposures estimated by the Staff in the Diablo Canyon FES. It is important to note that the ALARA criteria are formally incorporated into the plant's operating Technical Speci0 cations. Furthermore, the plant's contribution to the local population dose within a 50-mile radius is expected to remain insignificant in comparison to that from background radiation.
5.3.3 Uranlum Fuel j The DCPP reactor fuel is in the form of sintered UO, pellets containing an initial enrichment of U-235 less than 4.5 percent by weight. The fuel pellets are contained in Zircaloy rods. The fuel parameters l 540SS/S5K 24 l
l
l
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1
- meet the screening criteria of 10 CFR 51.52(a)(2), except for fuel enrichment, which may be as awch as 0.5 weight percent higher in the DCPP fuel rods.
3 The environmental effects of extended fuel burnup and higher initial enrichment are addressed by the i NRC in a Notice of Environmental Assessment and Finding of No Significant Impact published in 11.
Federal Register on February 29,1988 (53 FR 6040). His notice stated that the NRC's environn' ental assessment of extended fuel burnup and higher enrichment fuel is' complete, and that the environmental l
impacts summarized in Tables S 3 of 10 CFR 51.51 and S-4 of 10 CFR 51.52 bound the corresponding j impacts for burnup levels up to 60 gigawatt days / metric ton uranium (GWD/MTU) and enrichments up _
to 5 weight percent U 235.
De environmental impacts of transportation resulting from the use of extended fuel burnup and higher enrichment fuel were funher addressed in the Federal Register on August 11,1988 (53 FR 30355 - NRC -
Assessment of Environmental Effects of Transportation Resulting From Extended Fuel Enrichment and
, Irradiation). This notice reiterated the conclusion stated in 5"i FR 6040 and further concluded that there i are no significant adverse radiological or nonradiological impacts associated with the use of extended burnup and/or increased enrichment, and that burnup levels to 60 GWD/MTU and enrichments to 5 ,
weight percent U-235 will not significantly affect the quality of the human environment. Moreover, pursuant to 10 CFR 51.31,
- Determinations Based on Environmental Assessment," the Commission i determined that an environmental impact statement need not be prepared for this action. ;
) i Each Diablo Canyon reactor contains 193 fuel assemblies. The assemblies consist of fuel rods in a 17x17 )
array. About 39 to 46 percent of the fuel assemblies are replaced every refueling. Since issuance of the j operating licenses, PG&E has adopted several fuel design changes and improved fuel management j schemes. These changes have significantly improved uranium utilization. ,
l
)
To date, both Diablo Canyon units have been operating with nominal 18 month refueling cycles. It is anticipated that the units will continue to operate on the 18 month refueling frequency through the end of the proposed 40-year operating license terms. ;
in the Diablo Canyon FES, it was assumed for purposes of estimating the amount of uranium required
) that the plant would operate for 40 years with an 80 percent capacity factor, it was further assumed that ,
the units would be refueled on approximately an annual basis. Since the Diablo Canyon units are refueled approximately every 18 months and improvements in uranium utilization have been made, the total amount of uranium required for the proposed 40 year operating license terms is expected to be less than the amount projected in the FES.
)
5.3.4 Spent Fuel Storage The Diablo Canyon spent fuel pools currently have onsite storage capacity for plant operation through about 2007 while maintaining the capability for a full-core off-load. After 2007, storage space would no 1 longer be available for a full core off load. The existing spent fuel storage racks will be filled by 2010.
PG&E has a contract with the U.S. Department of Energy for the removal and disposal of Diablo Canyon spent fuel, ne U. S. Department of Energy currently plans to begin accepting spent fuel for permanent disposal no later than 2010, and is considering providing interim storage in a monitored retrievable storage facility earlier than 2010. .
j 540BS/S5K 25 .
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5.3.5 Solid Waste ,
i 5.3.5.1 Low Level Radioactive Waste "Ihe volume of solid low level radioactive waste generated at DCPP has historically been among the lowest in the nuclear power industry. Table 5 3 compares the annual volume of solid waste generated at Diablo Canyon with the PWR industry median. These values show that Diablo Canyon has generated significantly less solid waste compared with most PWRs and has followed the industry trend of reduced volume in recent years. During future years of plant operation, reducing solid waste generation will
)
continue to be emphasized. The maximum solid waste volume for the proposed license terms should be about 75 cubic meters per year average for each unit, which is consistent with the recent rate of solid ,
waste generation.
TABLE 5-3
)
Annual Volume of Low Level Solid Radioactive Waste Generated nt Diablo Canyon Compared with hiedian PWR Diablo Canyon ' Median PWR 1 Year (cubic meters / unit) (cubic meters / unit) l i 1986 47 198
) 1987 83 156 1988 105" 128 1989 94 147 1990 42 95
{ 1991 95 " 85 Source: INPO, March 1992 Two refueling outages occurred in these years i
)
5.3.5.2 Spent Fuel }
The reactor core thermal power for DCPP is 3338 megawatts for Unit I and 3411 megawatts for Unit 2,
) satisfying the screening criterion in 10 CFR 51.52(a)(1), which requires core thermal power to be less than 3800 megawatts.
The average level of irradiation of the DCPP discharged fuel assemblies is less than 45,000 megawatt-days per metric ton uranium (MWD /MTU). The decay period after fuel is discharged prior to transportation will be greater than 270 days. The screening criteria of 10 CFR S t.52(a)(3) are an average
)
level of irradiation not to exceed 33,000 MWD /MTU and no shipments until at least 90 days after discharge from the reactor.
5408S/S5K 26
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3
)
Although the average burnup level of the DCPP fuel assemblies exceeds the screening criterion in 10 CFR 51.52 (a)(2), the environmental effects for fuel burnups to 60,000 MWD /MTU have been -
evaluated and shown to be acceptable by the NRC. As discussed in Section 5.3.3., the NRC has
) concluded that there are no significant radiological or nonradiological impacts associated with th'e use of extended burnup fuel and that fuel burnups to 60,000 MWD /MTU will not significantly affect the quality -
of the human environment. Furthermore, improved fuel cycle designs should result in less total spent fuel over the proposed 40-year operating license terms than was projected in the FES.
)
5.3.5.3 Waste Shipping <
Radioactive waste shipped from the DCPP site is packaged in solid form in accordance with the requirements of 10 CFR 61 and the screening criterion of.10 CFR 51.52(a)(4). Unirradiated and '
irradiated fuel from the DCPP site will be shipped in accordance with the requirements of 10 CFR
) 51.52(a)(5). De screening criteria require truck shipment for unirradiated fuel and truck, rail, or barge shipment for irradiated fuel.
Depanment of Transponation and NRC regulations provide protection of the public and transpon workers from excessive radiation exposure. Protection is achieved by standards and requirements applicable to packaging, limitations on the contents and radiation levels of packages, and procedures to limit the
) exposure of persons under normal and postulated accident conditions.
Primary reliance for safety in transport of radioactive material is placed on the packaging. The packagi ,
must meet regulatory standards (10 CFR 71 and 49 CFR 173) established according to the type and form of material for containment, shielding, nuclear criticality safety, and heat dissipation. The standards-
) require that the packaging prevent the loss or dispersal of the radioactive contents, retain shielding ef6ciency, assure nuclear criticality safety, and provide adequate heat dissipatica under normal conditions .
of transpon and 'mder specified accident damage test conditions < The contents of packages not designed to withstand accidents are limited, thereby limiting the risk from releases that could occur in an accident, and to ensure standards for radiation levels, .emperature, pressure, and containment are. met.
)
5.3.5.4 Solid Waste Conclusions The amount of nuclear fuel and volume of solid waste resulting from the proposed 40 year operating license terms will continue to be within the limits assumed for the original licensing basis. Because of
) improved fuel cycle designs, the total amount of spent fuel produced over the proposed 40-year operating license terms is expected to be less than that originally projected in the FES for DCPP.
Based on the above, PG&E concludes that the radiological and environmental impacts from the storage ,
and transponation ofirradiated fuel and solid radioactive waste are consistent with the impacts set fonh ,
in Table S-4 of 10 CFR 51.52, and the environmental costs will not be significantly affected during the
) proposed 40-year operating license terms.
5.3.6 Thermal and Ecological Effects of the Cooling Water System
) The DCPP cooling water system is a once-through system discharging directly into Diablo Cove of the Pacine Ocean. The potential ecological effects of the cooling water system are: (1) those resulting from elevated water temperatures in ponions of Diablo Cove, (2) entrainment of organisms in the cooling water 540BS/85K 27 *
)
l
\
system, (3) impingement of organisms on the intake traveling screens, and (4) scouring effects of the discharge in the intenidal zone at the point of discharge.
Dese effects have been extensively studied and the study results were considered in issuance of the NPDES Permit and renewals (see Section 5.2.5). The NPDES Permit is conditional upon the discharge complying with provisions of Division 7 of the California Water Code and of the Clean Water Act (as amended or as supplemented by implementing guidelines and regulations) and with any more stringent effluent limitations necessary to implement water quality control plans, to protect beneficial uses, and I to prevent nuisance. Two of the findings of the California Regional Water Quality Control Board, j Central Coast Region (Board) as noted in the current NPDES Permit (Board Waste Discharge l 1
Requirements Order No. 90-09) are: !
- Thermal effects on the receiving water, actual temperature i screases, and actual temperatures of the discharge are monitored and results correlated and evaluated hs part of ongoing studies. These studies are being performed as required in Monitoring and Reporting Program No. 90-09. This monitoring
, is developed and reviewed jointly by Board and Department of Fish and Game staff. The results of the studies are used to evaluate thermal limits.
- Section 316(b) of the Clean Water Act requires that the location, design, construction and capacity of cooling water intake structures reflect the Best Technology Available (BTA) for minimizing adverse environmental impact. -
An April 28, 1988, study of the cooling water intake structure was submitted to the Board which concluded the facilities at Diablo Canyon Power Plant reflect BTA. Funher, the Monitoring and Repong Program requires PG&E to continue ecological studies as approved by the Executive Officer in order to evaluate changes in distribution and abundance of marine plants and animals within the
{ vicinity of the discharge. These operational studies have indicated that the effects of the discharge are consistent with the preoperational studies and modelling predictions; i.e., that the discharge would not signmcantly affect the marine ecology in the vicinity of DCPP. The Board and Depanment of Fish and Game have found the observed chr.nges (mainly in relative abundance of species) to be acceptable.
Additional discharge and thermal effects are not anticipated based on operational data collected since 1984. Accordingly, the basis for the Board's order is expected to remain valid when the NPDES Permit is renewed in 1995 and thereafter.
5.3.7 Protection of Historic Properties In 1966, Francis Riddell coriducted a survey for PG&E of approximately 250 acres to be used as the site I for DCPP. In 1968, Greenwood and Associates undenook subsurface investigations for PG&E at six sites within the construction areas for the DCPP facilities and a proposed access road from the plant site to Avila Beach. Based on these excavations, Greenwood in'1972 suggested that an area of the DCPP site, SLO-2, was a major village which figured prominently in the social, economic, and political life of the
> indigenous occupants of the area. The 1968 excavations at the site resulted in the identification of one of the oldest culturally stratified sites identified to date in San Luis Obispo County, nese 1968 studies by Greenwood are discussed in the Diablo Canyon FES where it mentions three archaeological sites in the plant area were investigated by archaeologists.
In 1978, Greenwood and Associates reponed on a survey of 90 acres ofland thought to be the areal extent of site SLO-2. This report contained information to be used for the nomination of SLO-2 to the j 4
540SS/85K 28
i I
)
l National Register of Historic Places. The SLO-2 site, which was determined to be eligible for the j National Register of Historic Places in December of 1982 by application from the Nuclear Regulatory Commission, includes the locations of recorded sites SLO-2, SLO-3, and SLO-8.
)
In accordance with 36 CFR 800, a determination of no-effect was made by PG&E Jn May 1984 in conjunction with the California Historic Preservation Officer. The NRC agreed with the no-effect determination in a transmittal to PG&E on June 25,1984, stating that operation and maintenaace of the i
plant will not affect cultural resources (SLO-2) listed in or eligible for inclusion in the National Register !
of Historic Places.
)
l l
Since November 1983, photographs have been taken at regular intervals from 23 stations within the site l
to monitor any physical changes to the site caused by natural or other processes. Funher, a Cultural l Resources Management Procedure Plan has been drawn up which provides specific guidelines regarding i future management of SLO-2 by PG&E. The Archaeological Resources Management Plan (ARMP) and i these procedures have been approved by the State Historic Preservation Officer as pan of PG&E's
)
compliance with the National Historic Preservation Act (36 CFR 800). l In the spring of 1986, unexamined ponions (an area of approximately 495 acres) of the DCPP property were investigated for cultural resources. A total of six prehistoric sites were recorded and/or relocated during this study. This included three previously unrecorded sites, SLO Il61, SLO-ll62, and SLO ll63
) and one previously recorded site, SLO-61. In addition, site forms were completed for Riddell sites 3 (SLO-Il59) and 5 (SLO-ll60) which had not been recorded with the California Archaeological Survey, PG&E watinues to manNe and protect the histc ric propenies at DCPP in consultation with the California State Historic Preservation Office and the local Native American communities. As a result of this aggressive management, PG&E concludes, as did the NRC in 1984, that operation of DCPP throughout
)
the 40-year operatir.g license terms will not adversely affect any known historic sites.
5.4 EXPOSL'RE FROM RELEASES DURING POSTULATED ACCIDENTS
) The offsite exposure from releases during postulated accidents has been previously evaluated in the DCPP FSAR Update. The results are acceptable when compared with the criteria defined in 10 CFR 100. This type of evaluation is a function of four parameters: (1) the types of accidents postulated, (2) the radioactivity release calculated for each accident, (3) the assumed meteorological conditions, ana (4) population distribution versus distance from the plant. On the basis of the safety assessment in Section 4.0, it can be concluded that neither the types of accidents nor the calculated radioactivity releases
)
will change through the proposed 40 year operating license terms. Funhermore, the site meteorology, as defined in the FSAR Update, is essentially constant. Thus, population is the only time-dependent parameter. It is important to note that there is no expected change in land usage during the license terms that would affect offsite dose calculations.
) The population size and distribution in the vicinity of the plant have been reviewed several times since the construction permit was issued: in the FES in 1973 (through the year 2000), in the original FSAR in 1974, in the 1985 FSAR Update using 1980 census data, again in late 1991, and in a review of the California Department of Finance projections through 2025 performed for the purpose of the proposed license terms. The Depanment of Finance projections indicated that a compound average growth rate of 2.15 percent is expected for the 50-mile radius area around Diablo Canyon through the year 2025.
) Table 5-4 summarizes the projected population size and distributions.
54085/S5K 29
)
TABLE 5 4
)
Summary of Population Projections for the Diablo Canyon Vicinity -
) ,
Area Original FSAR Revised FSAR _
(miles) (1974) (1985) Current Current 2010 2010 2010- 2025:
) -
O - 6* 29 26 100' 100' ,
6 - 10 18,992 36,126 36,403 46,480' 0 - 10 19,021 36,152 36,503 46,580' 10 50 508,130 438,035 -555,108 730,566.
)
0-50 527,151 474,187 591,611 777,146
- Reflects Low Population Zone The plant exclusion area depleted in Figure 51 will remain uninhabited through the proposed license terms. De activities within the exclusion area only pertain to normal plant operation. No part of the -
eulusion vea will be sold, and no structure will be located within it except those o'.vned by PG&E or a related company and used in conjunction with normal utility functions. No residences will be permitted
, on the site.
The Low Population Zone (LPZ) for DCPP is the area included within a 6-mile radius of the site, as; !
show n in Figure 5 2. This land area was selected because it meets the requirements of 10 CFR 100 with -
respect to proximity to the nearest population center.: ,
j The major population centers (with populations of 25,000 or more) currently within 50 miles of Diablo ' ;
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Canyon are Lompoc (1990 population 37,649), about 45 miles to the south-south-east; Santa Maria (1990 population 61,284),29 miles to the south east; and San Luis Obispo (1990 population 41,958),10 miles : y to the east north east. Accordingly,10 miles is currently the Population Center Distance (PCD).
However, population projections indicate that the community of Baywood Los Osos (8 miles to the north) h will reach a population of 25,000 in approximately the year 2020. De 2020 projected population for Baywood-Los Osos is 26,844. Accordingly, the PCD would become 8 miles in approx'imately 2020.-
Federal Regulation 10 CFR 100.ll(a)(3) provides that the PCD be at least 1-1/3 times the distance from the reactor to the outer boundary of the LPZ. De community of Baywood Los Osos, at 8 miles north, will satisfy this rule as the projected PCD in the year 2020, thus requiring no change in the definition.-
of the current LPZ now or during the proposed 40-year operating license terms.
)
ne changes projected for the population distribution through 2025 will not impact the boundaries used for existing accident analyses. De current exclusion area boundary, LPZ, and nearest population center -
distance will continue to meet the requirements of 10 CFR 100.11(a)(3) for the proposed 40 year license terms. Accordingly, the proposed 40 year operating license terms will not affect previous conclusions
) greached in the Staff's SER and FES on the potential environmental effects of offsite releases from ~
postulated accidents.
540SS/85K 30
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5.5 ENVIRONMENTAI RELATED PLANT MODIFICATIONS Several environmental related plant modifications have been made since issuance of the FES and Addendum. These changes further reduce the environmental impacts associated with DCPP operation and include the following:
1 e Wastewater Holding and Treatment (% HAT) System - ne WH AT System includes storage tanks and )
associated equipment for improved storage and treatment of wastewater. De tanks are placed in a concrete retention basin for secondary containment.
i e Wastewater Routing - Pumps and piping were added to route wastewater from the turbine building l
sump to the WH AT System, allowing for treatment prior to discharge. l e Hacardous Waste Storage - A building was provided for the storage of containerized hazardous waste prior to offsite shipment. Other areas in the plant are designated for hazardous waste storage. Dese storage areas are provided with fire protection equipment and secondary containment features to prevent the release of hazardous waste to the environment.
- OilSpill Prevention - A spill prevention pond was constructed at the 230 kV switch yard to prevent oil from transformer leaks from entering Diablo Creek.
e Seawater Reverse Osmosis System - A Seawater Reverse Osmosis System was provided as an
)
additional source of makeup water for the plant, effectively replacing the seawater evaporator described in the FES.
Discharge Structure Modifcations - To reduce the, peak temperature of cooling water being discharged from the plant, modifications were made to the discharge structure. These modifications
) allow for better mixing of water betsveen the two units and included cutting openings in the central wall of the discharge structure and modification to the weirs.
e intake / Discharge Temperature Monitoring - An improved temperature monitoting system was installed to measure the cooling water temperature at the intake and discharge structures. His system
) initiates an alarm in the plant control room when a set temperature difference is exceeded.
1 e Erpanded Sewage Treatment - To allow for increases in plant staffing, an expanded sewage treatment l system was constructed that provides for secondary treatment of effluent. The system discharges to !
the cooling water system andI is monitored as required by the NPDES permit. An expanded leach field was provided as a backup for the secondary treatment system. j e Chlorination System Modifcations - Modifications to the Chlorine System include (1) the use ofliquid hypochlorite to control microbiofouling instead of gaseous chlorine, (2) implementation of continuous chlorination of the auxiliary saltwater system to control macrobiofouling (invertebrate marine life),
and (3) possible use of intermittent injection of a chlorine / bromine mixture to prevent macrofouling j in the Circulating Water System.
- Dechlorination Process - Dechlorination, by injection of sodium bisulfite, is now used to reduce free chlorine levels in discharged cooling water. .
- Additional Steam Equipment - To provide for additional steam needs during plant startup testing, a second auxiliary boiler and two underground fuel tanks were added. While the combined air
! emissions from both boilers exceed the projections in the FES, both have been permitted by the local
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Air Pollution Control District and operate within the District's emission limits.
540ES/85K 33 .
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i e Ligvid Radwaste Modyications - The DCPP Liquid Rad' w aste System has been modified by providing additional processing capability via ion exchange and filtration. Based on operational problems j experienced at other plants, the radwaste concentrator was never operated and has been abandoned in-place. Piping modifications were made to allow for the use of mobile processing equipment if required.
e Solid Radwaste Processing - Regulatory changes to allcwable solid radioactive waste forms resulted in removal of the installed radwaste solidification equipment. Presently, contractor-provided mobile equipment is used to solidify or dewater radioactive waste.
i e Solid Radwaste Volume Reduerlon - To reduce the volume of solid radioactive waste being produced, two radioactive waste compactors were installed.
e Hazardous Materials Storage - A warehouse was provided for storage of hazardous materials needed in plant operation.
e Makeup Water Treatment - Contractor-supplied equipment is used to treat makeup water used at the l plant.
Most of these plant design modifications and changes have had a direct positive impact on the environment; for example, chemical discharges have decreased and spill prevention has improved.
Changes have been conducted in accordance with approved procedures, current license conditions, l Technical Specification requirement:,, and environmental permits. Additional plant modifications and changes may be %plemented during the proposed 40-year operating license terms. For example, the Radiation Monitoring System is presently being upgraded to improve performance, reliability, and capability. Based on past experience, future changes are r.ot expected to have a significant negative impact on the environment and should Provide improved raeans to monitor potential environmental effects.
5.6 DECOMMISSIONING In accordance with 10 CFR 50.33(k), PG&E submitted a decommissioning report to the NRC that provided reasonable assurance that sufficient funds will be available to decommission the plant.
Additionally, PG&E prepared a preliminary decommissionirq cost study for DCPP. This study was last updated in May 1991 to reflect recent changes in costs and schedules.
The decommissioning study provides cost, schedule, waste generation / disposition, and radiation exposure estimates associated with the decommissioning of the DCPP nuclear units following cessation of operations. The alternatives evaluated were DECON (prompt removal / dismantling) and SAFSTOR (mothball with delayed dismantling). The plant retirement dates were taken as 30 years following l commercial operation. This timeframe was used as input in scheduling analysis.
The decommissioning study did not determine the incremental impact on decommissioning due to the proposed 40-year license terms. However, PG&E anticipates that the proposed 40-year operating license terms will have a negligible incremental environmental and cost impact on decommissioning the DCPP units. A reactor's activation product inventory rapidly builds up and reaches an equilibrium level, so that after 10 years of operation the inventory is approximately 90 percent of the total expected after 40 years of operation. As a result, most of the factors affecting costs for decontamination and decommissioning, 540'S!85K 34
as well as occupational exposure, arise early in a reactor's operating ilfe. Thus, the proposed 40-year i I
operating license terms will not add signincantly to the cost and effort to decommission the plant.
Moreover, it was concluded in NUREG-0586, " Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," that decommissioning is not exp *ed to significantly impact the environment. The NUREG also categorically excludes power reactors from me mandatory Environmental Impact Statement requirement for decommissioning. The other key conclusions of this report were: I l
I e Decommissioning at the present time can be performed safely and at reasonable cost.
e Decommissioning of nuclear facilities is not an imminent health or safety problem.
e Decommissioning of a nuclear facility generally has a positive environmental impact. ;
-l n erefore, PG&E concludes that any changes in the environmental impact associated with I decommissioning after expiration of the proposed 40-year license terms will be negligible.
6.0 NO SIGNIFICANT HAZARDS EVALUATION j PG&E has evaluated the no significant hazard considerations involved with the proposed amendment, l focusing on the three standards set forth in 10 CFR 50.92(c) as quoted below:
The Commission may make a Onal determination, pursuant to the procedures in 150.91, that a proposed amendment to an operating license for a facility licensed under i 50.21(b) or 5 50.22 or for a testing facility involves no signincant hazards consideration, if operation of the facility in accordance with the proposed amendment would not:
- 1. Involve a signi0 cant increase in the probability or consequences of an accident .
previously evaluated; or l l
- 2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or
- 3. Involve a signincant reduction in a margin of safety.
The following evaluation is provided for the no signincant hazards consideration standards.
- 1. Does the change involve a signincant increase in the probability or consequences of an accident previously evaluated?
- The proposed 40-year operating license terms do not affect the probability or consequences of an
! accident previously evaluated since the requested extensions entail no physical change in the plant l equipment or operating procedures and the FSAR Update safety analyses are based on 40. year plant operation. Surveillance and maintenance practices, as well as other programs such as environmental quali0 cation of equipment, ensure timely identification and correction of any degradation of safety-related plant equipment. The long term integrity of the reactor vessels has been recently reevaluated using currently acceptable NRC calculational methods and best available DCPP specific data. The evaluation results demonstrate, as before, that both reactor vessels are safe for normal operations l
l 540BS/85K 35
in excess of 40 years. Also, the offsite radiation exposures resulting from postulated accidents have '
been reanalyzed using population projections for the propo!ed 40-year sperating license terms. The '
calculated exposures are not significantly different from those documented in the FSAR Update and are well within 10 CFR 100 guideline values.
1 Therefore, the proposed changes do not involve a significant . increase in the probability or consequences of an accident previously evaluated.
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- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
De possibility of a new or different kind of accident is not created by the proposed 40-year operating license terms since at least 40 years operation was assumed in the design and construction of DCPP Units I and 2. The plant Main'tenance Program is designed to both maintain and determine the need to replace safety related components. Rus, any degradation that might possibly create a new or different kind of postulated accident would be detected and corrected before the-occurrence of such an event.
Therefore, the proposed changes do not create the possibilityof a new or different kind of accident from any accident previously evaluated.
- 3. Does the change involve a Jignificant reduction in a margin of safety?
The proposed 40-year operating license terms do not involve a significant reduction in a margin of- I safety since degradation of safety related equipment will be identified and corrected by ongoing-surveillance and maintenance practices. Existing programs, routine maintenance, and compliance with Technical Specificaticns assure that an adequate margin of tafety is maintained. . These activities will remain in effect for the duration of the operating licesses. .
Therefore, the proposed changes do not involve a significant redu; tion in'a margin of safety. -
i In conclusion, based on the above safety and environmental evaluations, PG&E submits that the activities _
associated with this license amendment request satisfy the no significant hazards consideration standards J of 10 CFR 50.92(c), and, accordingly, a no significant hazards finding is justified. Such a finding is ;
consistent with NRC Staff approval of numerous applications of this type.- PG&E requests that, if !
necessary in accordance with the provisions of 10 CFR 50.91(a)(4), the NRC rnake a final determination .l that no significant hazards considerations are involved and issue a license amendraent in accordance with i this request.
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5408S/85K 36 l
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ATTACHMENT B DIABLO CANYON POWER PLANT UNITS 1 AND 2 l MARKED UP LICENSE PAGES~
Remove Pare Insert Pare :
I 9, Unit 1 License 9, Unit 1 License )
7, Unit 2 License .7, Unit License l
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540SS/85K
. . - , . , ._ v e- wry,f
Unit 1 License- !
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, Nuclear Plant Physical Security Plan," Revision 11 dated May 27, e 1982, as revised July 19, August 12, September 17, 1982; February 4, August 3,1983; January ll, February 6, March 19, April 19. August 29, 1984; "Diablo Canyon Nuclear Plant Guard Training and Qualtfication .
Plan," Revision 2 dated February 4, 1983 as revised August 29, 1984; l
- Diablo Canyon Nuclear Plant Safeguards Contin Revision 2 dtted February 4,1983 as revised August'983, 3 pency Plan August "
29, 1984.-
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F. Antitrust Pacific Gas and Electric Company shall: comply with the antitrust' cor.ditions in Appendix C to this license.
O G. Reporting PG&E shall report any violations of the requirements contained in e Sections 2.C(3) through 2.C(10), 2 E and 2.F. of this License within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Initial notification shall be made in accordance with the provisions of 10 CFR 50.72 with written follow-up in ..
d accordance with the procedures described in 10 CFR 50.73 (b), (c),
(d) and (e). .
H. Financial Protection PG&E shall have and maintain financial protection of such type and-h in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover pubFc liability claims.
- 1. Term cf License ,
This License is effective as of the date of issuance and shall expire at midnight on Ag u 20, 2000, fe.pfew J,er Tt.7 2.oz.l.
FOR THE NUCLEAR REGULATORY COMMISSION
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O R t =
aroldR.Denton, director ;
Office of Nuclear Reactor Regulation Attachments:
J 1. Appendix A - Technical Specifications ,
- 2. Appendix B - Environmental Protection Plan l
- 3. Appendix C - Antitrust Conditions )
Date of Issuance: November 2, 1984 O
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Unit 2 License
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i f H. Financial Protection PGLE sha'11 have and maintain financial protection of such type and !
in such amounts as the Comission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover !
) public liability claims. ;
- 1. Term of License This License is effective as of the date of issuance and shall expire at midnight on her;Mr 0, 0010. Ape:/ E4, 202f,
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FOR THE NUCLEAR REGULATORY COMMISSION
/ %.
) Harold R. Denton, Director !
Office of Nuclear Reactor Regulation Attachments: j
- 1. Appendix A - Technical Specifications (NUREG-1151) i
- 2. Appendix B - Environmental Protection Plan i
- 3. Appendix C - Antitrust Conditions Date of Issuance: August 26, 1985 -
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