ML20239A092
| ML20239A092 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 09/30/1987 |
| From: | YANKEE ATOMIC ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20238F813 | List: |
| References | |
| NUDOCS 8709170073 | |
| Download: ML20239A092 (76) | |
Text
- _ _ _ _ _ _ _ _ _ _ _ _ , _ . _ _ _ _ _ - . _ _ _
ATTACHM,II 2 ASSESSMENT OF BENEFITS AND POTENTIAL IMPACTS OF PROPOSED CHANGE TO EXPIRATION DATE l
)
0F YANKEE NUCLEAR POWER STATION OPERATING LICENSE i
I September 1987 i
I i
i i
k i
I 1
i Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701 5784R/4.70 8709170073 g70915 PDR ADOCK 05000029 P PDR L_______-----
i l
f .-
TABLE OF CONTENTS
(- Page l ' LIST OF TABLES................................................... v i
LIST OF FIGURES.................................................. vi. j 1.0
SUMMARY
.......................................................... 1 1.1 Introduction............................................... 1 1.2 Basis for Proposed Change.................................. 1~
2.0 BENEFITS ASSESSMENT.............................................. 7 2.1 Introduction............................................... 7 l 2.2 Continued Availability of Reliable Baseload Generation..... 7 )
2.3 Avoided Increase in Electric Rates to Consumers............. 8 2.4 Avoided Environmental and Health Effects................... 9 2.5 Continued Benefits to the Local Area Economy............... 10 2.6 Conclusion................................................. 11 ,
1 1 :
3.0 SAFETY ASSESSMENT................................................. 13 I l
3 .1. Introduction............................................... 13 .j 3.2 Licensing Basis Documents.................................. 13 j i
l 3.2.1 Final Safety Analysis Report....................... 13 l 3.2.2 NRC Systematic Evaluation Program.................. 14 L 3.2.3 Probabilistic Safety Study......................... 15 3.2.4 Technical Specifications........................... 16 3.2.5 Surveillance and Maintenance Program............... 17 3.2.5.1 In-Service Inspection Program............ 17 3.2.5.2 Environmental Qualification Program...... 19 3.2.5.3 Preventive Maintenance Program........... 20 3.3 Plant Operating History.................................... 21 1
3.3.1 Operating Performance.............................. 21 3.3.2 Component Integrity................................ 22 3.3.3 Plant Modifications................................ 23 3.3.4 Regulatory Performance............................. 25 3.4 Assurances for Continued Functional Capability of l Safety-Related Components.................................. 26
! 3.4.1 Mechanical Components.............................. 26 1
3.4.1.1 Main Coolant Pressure Boundary........... 26 3.4.1.2 Other Mechanical Components.............. 28 5784R/4.70
TABLE OF CONTENTS (Continued)
Page 3.4.2 Electrical Components.............................. 29 3.4.3 Structural Components.............................. 30 3.4.3.1 Vapor Container.......................... 30 3.4.3.2 Other Structures......................... 31 3.5 Conclusion................................................. 31 4.0 ENVIRONMENTAL ASSESSMENT......................................... 35 4.1 Introduction............................................... 35 4.2 Systems and Programs for Environmental Control and Monitoring................................................. 36 4.2.1 Environmental Control Systems and Programs......... 36 4.2.1.1 Waste Disposal System.................... 36 4.2.1.2 As Low As Reasonably Achieveable Program. 37 4.2.2 Environmental Monitoring Systems and Programs...... 38 4.2.2.1 Radiation Monitoring System.............. 38 4.2.2.2 Environmental Radiological Surveillance Program.................................. 39 4.2.2.3 Nonta41ological Surveillance Program..... 40 4.3 Assessment of Environmental Impact During Normal Operations 40 4.3.1 Radiological....................................... 40 4.3.1.1 Occupational Radiation Exposure.......... 40 4.3.1.2 Off-Site Radiation Exposure.............. 41 4.3.1.3 Solid Waste Generation................... 42 4.3.1.4 Uranium Fuel Cycle....................... 43 4.3.1.5 Spent Fue1............................... 44 4.3.2 Nonradiologica1.................................... 44 4.3.2.1 Thermal and Ecological Effects of the Circulating Water System................. 44
-111-5784R/4.70
TABLE OF CONTENTS (Continued)
Page ,
4.4 Exposure From Releases During Posulated Accidents.......... 47 4.5 Summary and Conclusions........................ ........... 49 REFERENCES....................................................... 52 TABLES........................................................... 55 i FIGURES........................................................... 70 i
i I
I i
1 1
l 1
l I~
l I
-iv-5784R/4.70
1.0
SUMMARY
I 1.1 Introduction Section 103.c of the Atomic Energy Act of 1954 authorizes the issuance.
of facility operating licenses for a period of time up to 40 years. The current license term for the Yankee Nuclear Power Station began with the date of issuance of the construction permit, November 4, 1957, and ends on November 4, 1997. Accounting for the two years and eight months required for plant construction, this represents an effective operating license term of only 37 years and 4 months.
Current Nuclear Regulatory Commission (NRC) policy is to issue operating licenses for a 40-year period, commencing with the date of issuance of the operating license (not the construction permit). For Yankee, this date was July 9, 1960 (Reference 1). Accordingly, it is proposed that the. Yankee operating license te amended to change'the expiration date to July 9, 2000, censistent with current NRC policy. This would permit an additional two years and eights months of plant operation.
1 1
l l 1.2 Basis for Proposed Change l
Sections 2.0, 3.0, and 4.0 of this document describe assessments that have been made to determine the benefits and potential impacts of an additional 2 years and 8 months of plant operation. The results of these assessments demonstrate that this additional plant operation: (1) would provide substantial benefits to New England and the local area and (2) would not have an adverse effect on public health and safety or the environment. l The principal reasons for these conclusions are the following:
o Reliability I
The Yankee plant is one of the most reliable nuclear plants in the country. The lifetime average capacity factor over 27 years of operation is 74 percent. This is substantially higher than the b average of all United States nuclear plants, which is 60 percent.
li 5784R/4.70 l
I'1 .
m, l Accordingly, it would be beneficial to keep this reliable power source in operation.
o Economic Benefits-
- - d The Yankee plantIs also one of the most economical sources of power in' the Miew England region. The capital cost of replacement '
with new base load, coal-fired capacity (which will'he necessary if Yankee is shut'down) would be approximately $1.830/kW or 320 million in 1987 dollars.
In addition,. Yankee's fuv1 costs would only be about 1.0d/kWh.
This is approximately 0.5d/kWh below the expected fuel cost for a new base load coal plant in New England, and even further below the fuel cost expected for any oil or gas-fired alternative.- o
-Estimated power costs in 1987 dollars are:
Coat (d/kWh)
Yankee Coal l
'l 0&M 3.1 ,1,0 Fuel 1.0 1.5 Capital u 0.8 5.5 j Total 4.9 8.0 l
Based on'an average annual generation of 1.13 billion kWh. this 1 q
means that a shutdown at Yankee would cost New England consumers l over $35 million per year in higher electric bills. Over the 2.7 ll year period covered by the proposed amendment, the total cost 2 increase would be nearly $95 million.
Such expenditures would not be justified as long as. Yankee continues to oprerate safely and reliably. (
, i 5784R/4.70
7,- --
y.
4& .'),', i
,s
' f ,'
' s >
- o Safety g i \.
- i ;
The Yankee plant has one of the best safety records in the nation.
This is demonstratedjey recent results of the NRC's Systematic Assessment of' Licensee Performance which represent perhaps the best
' measure of safe operations.at a nuclear plant:
Yankee SALP Ratings 2/85 to 9/83 to 5/82 to J 10/86 2/85 9/83 Plant Operations 1 1 1 I
Radiologic 41 Controls 1 2 2
' Maintenance and )
Modifichtien 1 1 1 1
Surveillance 1 1 1
.;' u Fire Protection and f Housekeeping ' 1 1 1 Emergency Preparedness 2 1 1 )
Security and Safeguards 2 2 2 l
Refueling and Outage Management 1 1 1 i 4 Assurance of quality 1 2 2 1 i
Training and Qualification Effectiveness 2 - -
Licensing Activities 1 1 1 There t.re a m2mber of inherent. safety f actors at Yankee, as well.
Thg plant operates at a relatively low power level, which means
>~ that there are less fission products available for release in any j accident. This means that the potential radiation from such an ,
! l event would be reduced by approximately a factor of five, compared i to the current generation of licensed power reactors. !
5784R/4.70 ;
I I
, s i E ,
The extremely low' population density around the plant is also a major, inherent safety factor. Only 60. people live within'a mile of the plant, and only 260 live within the entire Low Population Zone. This fact would make evacuation around Yankee substantially less difficult than at most licensed plants, in the unlikely event that evacuation ever became necessary.
Occupational radiation exposures are veey low at Yankee, well below NRC limits, and, indeed, well below most other nuclear plants.
During the most recent five-year period reported by the NRC in NUREG-0713, " Occupational Radiation Exposure of Ccmmercial Nuclear Power Reactors and Other Facilities," the average annual personnel
. exposure at Yankee was 281 person-rem. This is approximately one-half of the corresponding exposure at the average U.S. PWR plant.over this same period.
The low copper / nickel content in the reactor vessel material is another inherent safety feature at Yankee. This makes the reactor vessel less susceptible to radiation embrittlement than many reactors. Recent analysis have indicated that the Yankee vessel will not reach the NRC screeting criterion of 270 F for RT PTS until March 2020, well beyond the July 9, 2000 date proposed herein.
I Finally, probabilistic analyses indicate that the overall risk of '
radiation exposure to the general pub.lic from Yankee is well within l
NRC safety goals. Indeed, the risk levels et Yankee are substantially lower than at most nuclear plants operating toda). I o Environmental Benefits Substantial environmental benefits would accrue from the proposed 4 i
amendment. The burden on the environment f*om a fossil-fueled
]
replacement power source would be much greater than from Yankee. I Sulphur and particulate emissions from such plants are severe problems in the northeast (see discussion in Section 2.4). Yankee i
does not contribute to these problems, whereas eny replacement !
power source would exacerbate them.
I l
5784R/4.70 !
l
y , .g- - .,
l.
Q4 ,
j LiquM cnd gaeous radioactivity reieasch from normal plan operation up are , Jarefully controlled ind imo,aitored contim;.ously. Actual; ,
J 'r s :t .. , , r.
measurements indicate that off-sit'e doso due to discharges f rom b <
Yankee'are only 'a small fraction (le;s than %) of the ALARA design y ', ) dose objectives established in 10CFR50, Appendix I. '
c' r .,. l l
1
- Any nuclear. plant also prodisces come low-level, solid' radioactive l 't
) waste. Considerable effont is expended af. Yanbe to limit the i
- I amount of Uarit'e produced. Over. the past sixJyears, lde arrount of. 5 W1 '
-,t
, s i clow-level Oste produced at Yankee was less thau' oneLhalf oC the amount produced by the average U.S. PWR plant. Also, be:atse / '
l', Yanpee gerates at a relatIvely low power level, the plant only ,
- produces 'about one-fif th ed the high 1cvel waste produced by one;of the large power re.actorr.. currently being licenned. J 4
b' ,' Therefore, Yankee does not represent an undue burden,on the environment; indeed the plant' is much less of a burho . than any j y baseload source of power that wot1d be built. to replace it.
i o Plant Improvements '
The Yankee plant has been upgraded continuously since it went ort line in 1960. A few examples are the addition of:
o An improved Emergency Core Cooling System o Three emergency diesel generators .
o Two new emerg;ency feedwater pumps
,, o A Safety Parameter Di.4 play System (SPDS)
<l o A new independent Safe Shutdown System o A new solid-state Reactor Protection System and Feedwater t 1 Control System I
- l The status oi-the lankee pfant has recently been reviewed in detail f by the NRC under the Systematic Evaluation Program (SEP). The results of the program show the the plant design either raeets e
current safety stande ds or 115 provide an equivalent levet of 57841i/4.70 $
f
J safety.once certain defined modifications are completed. Many of
(. .these modifications have already been. completed.
l These upgrades have kept the Yankee plant equivalent, from a safety and reliability standpoint, to the newest plants in the country.
Indecd, in many; instances, the smaller, simpler design makes 'iankee even more reliable than current plants.
1 i
1 l
i l
t
(- 5784R/4.70 .!
1 I
l 1
2.0 BENEFITS ASSESSMENT 2.1 Introduction The purpose of the following discussion is to provide an assessment of the benefits expected from the additional two years and eight months of operation that would be permitted by the proposed amendment. The benefits considered include: (1) continued availability of reliable baseload generation (2) avoided increase in electric rates to consumers, (3) avoided environmental and health effects of a fossil-fueled replacement power source, and (4) continued benefits to the local area economy.
2.2 Continued Availability of Reliable Baseload Generation The Yankee plant provides baseload generation to the New England region. This plant is one of the most reliable nuclear plants in the country. The lifetime average capacit, factor over nearly 27 years of operation is 74 percent. This is substantially higher than the average for all U.S. plants, which is 60 percent. Accordingly, it would be beneficial to keep this reliable source of power in operation, particularly in a period of severe capacity shortages such as exist in the New England region today and are likely to continue throughout the 1990's.
The present New England load is about 100 billion kWh annually. This load is expected to grow substantially by the late 1990's. Although the amount of growth is uncertain, reasonable estimates can be made by looking at the recent record. Over the past 14 years (which includes both the 1974 and 1979 oil crisis years), the load growth has averaged 2.4 percent, compounded annually. Over the past 4 years, it has averaged 4.3 percent (Reference 2).
By the first 'ull year of the proposed license amendment period in 1998, even the most conservative 2.4 percent growth rate would require an j additional 33 billion kWh annually. If all of this additional generation were provided by baseload plants, with a capacity factor of 75 percent, an add!:fonal 5,000 megawatts of generation would be needed. More capacity would be required if peaking units with lower capacity factors were used to provide some of the generation, or if higher rates of load growth were experienced. A 5784R/4.70
_ _ _ _ _ _ _ - - . _ _ _ .--- J
3 i
4'.3 percent growth rate would require nearly 10,000 rather than 5,000 )
\
additional baseload megawatts. j Existing plans for additions to New England's generating capacity between now and 1998 total less than 5,000 megawatts and their successful completion is uncertain. Seabrook would provide 1,150 megawatts, but so far 1 has been unable to obtain a license. The Hydro-Quebec Phase II tie to Canada would provide the equivalent of 1500 megawatts of capacity, but its availability and price must be considered uncertain in view of the recent decision of the Canadian National Energy Board to deny Hydro-Quebec's s
application to export the power to the United States (Reference 3).. The 1 proposed Ocean State gas-fueled plant would provide 200 megawatts, but has-not yet received a license. Cogeneration will provide some additional capacity, but the amount is highly uncertain.
'These projections and uncertainties combine to indicate the possibility of a severe shortage of generating capacity in the late 1990's. Under these circumstances, continued availability of the Yankee plant, with its proven record of reliable operation, will be especially important. Therefore, assuring its availability by amending the license term to allow operation during the period November 1997 through July 2000 would be of substantial benefit to the New England region.
2.3 Avoided Increase in Electric Rates to Customers In addition to providing reliable baseload generation, Yankee is also one of the most economical plants in the New England region. If the plant is not operated beyond November 1997, it will be necessary to replace it with 1
I substantially more expensive baseload capacity. Accordingly, continued )
operation of the Yankee plant through the proposed amendment term would avoid !
increased electric rater ( to New England consumers.
1 i
A recent Burns and Roe study provides an analysis of alternative l 1
generation technologies for the New England region (Reference 4). Based on a !
review of this study, a 400 MWe atmospheric fluidized bed coal plant is considered to be the most likely alternative for replacement of Yankee's capacity. This is because this technology appears to be the best choice for a 1
l 5784R/4.70
small baseload, coal-fired plant having to meet the air quality standards
.likely to be in effect in New England in the 1990's.
Table 1 summarizes the characteristics of a new 400 MWe atmospheric fluidized bed coal plant located in New England. The table indicates that the capital cost of replacement'of Yankee's 175,000 KWe would be $1,830/KWe (including AFUDC) or $320 million in 1987 dollars. In addition, it indicatas that the coal fuel cost would be 1.5 cents /kWh (1.6 x 100 x 9,450/10 ), 6 which is about 50 percent higher than the nuclear fuel cost for Yankee.
Table 2 fr>vides a comparison and breakdown of estimated 1998-2000 power costs for Yankee and the assumed replacement fluidized bed coal plant.
The amounts given are in 1987 dollars and, thus, neglect inflation. For Yankee, the costs were obtained from a corporate model. For the fluidized bed coal plant, the costs were calculated from the unit costs summarized in Table 1. However, the latter costs were only multiplied by 175,000 KWe so that the amounts given in Table 2 refer to the same capacity, regardless of whether Yankee or the fluidized bed coal plant are being considered.
I The comparison in Table 2 shows that the estimated power cost is 3.1 cents /kWh lower for Yankee than for the fluidized bed coal plant.
Assuming an average capacity factor of 74 percent, the total annual generation of each plant would equal 1.13 billion kWh. This means that continued operation of the Yankee plant would save New England consumers over $35 million per year in lower electric bills. Over the additional two years and eight months that would result from the proposed amendment, the total. savings would be nearly $95 million.
2.4 Avoided Environmental and Health Effects Substantial environmental benefits would also result from the proposed amendment. This is because the burden on the environment from a fossil-fired replacement power source would be much greater than from Yankee. Sulphur and I particulate emissions from fossil-fired generation are particularly severe problems in the Northeast region because of acid rain and public health concerns. Yankee does not contribute to these problems, but a fossil-fueled replacement power source would exacerbate them.
_9_
5784R/4.70
__ , . m r
a.
LIfLthe; assumed' replacement fluidized bed coal plant is operated'instead-i F Lof Yankee eit:would resultiin annual emission of:
m 2.1'million pounds'of sulfur,.
T..
6.0 million pounds of NO ,X and.
o 0.3 million pounds 'f. particulate.
1These amounts were calculated assuming: -(1) use of' Appalachian bituminous coal with a nominal Btu content of 12,300. Btu /lb and a stifur content of
'2.4 percent, .(2) fuel usage corresponding to the power level, heat. rate, and
. capacity factor.given in Table 1, and~(3) emissions equivalent to the limits allowed by the' June 1979 Nas Source Performance Standards (Reference 5).
Estimates of the health effects of sulfur emissions from a coal-fired z power plant have been previously reported in the 1977 Ford-MITRE study-(Reference 6). These health effects include: chronic respiratory disease, aggravated heart-lung disease symptoms, asthma attacks, and children's respiratory' disease' Although~the health effects of particulate and other.
emissions such na NOX ,. carcinogenic hydrocarbons, and carbon monoxide appear to be much.less than for' sulfur-related emissions, they undoubtedly add to health' impairment and fatalities (Reference 6).
2.5 Continued Benefits to the Local Area Economy The Yankee plant has 170 full-time employees and an annual payroll of approximately'7 million dollars. Ninety-five percent of these employees live within twenty miles of the plant. Therefore, most of this money is spent or investedLin the local area.: The tax payments to the Town of Rowe and purchases of equipment,' materials, and services from 50 local suppliers add another 3 million dollars per year to the area economy. Therefore, the total benefit to thu local area economy approaches 10 million dollars per year.
Over the proposed additional-two years and eight months of plant operation,
.this benefit would' total nearly 27 million dollars. This amount is significant, since the population density of the local area is the lowest in
-Massachusetts.
5784R/4.70 cy . - .
2.6 Conclusions The preceding discussion shows that continued operation of the Yankee plant through the proposed amendment term would provide significant benefits to the New' England region and local area. These benefits include:
o Continued Availability of Reliable Baseload Generation The Yankee plant is one of the most reliable nuclear plants in the country. It has a cumulative capacity factor of 74 percent over nearly 27 years of operation. Projection of the need for capacity in New England indicates the potential of a severe shortage of generating capacity in the late 1990's. Shutdown of the Yankee plant in November 1997 would worsen this shortage.
o Avoided Increase in Electric Rates to Consumers The Yankee plant is also one of the most economical sources of power in the New England region. If the plant is not operated beyond November 1997, it will be necessary to replace it with substantially more expensive baseload, coal-fired generation.
Based on an average annual generation of 1.13 billion kWh/yr, the estimated cost in higher electric bills would be more than $35 million per year. Therefore, the additional two years and eight months of plant operation that would be allowed by the proposed amendment would save New England consumers nearly $95 million.
1 o Avoided Environmental and Health Effects The burden on the environment from a fossil-fueled replacement power source would be much greater than from Yankee. Sulfur and particulate emissions from fossil-fired generation are particularly severe problems in the northeast region (see Section 2.4). Yankee j does not contribute to these problems, but any replacement power 1
- j. source would exacerbate them.
5784R/4.70 E____________.-----. ___
o Continued Benefits to'the Local Area Economy i
l If operation of the Yankee plant is continued beyond November 1997, t approximately 10 million 1987 dollars per year would continue to be provided to the local area economy in the form of tax dollars, salaries to employees, and purchases from local suppliers. Over the proposed additional two years and eight months, this benefit would total nearly $27 million.
l i
i i
l l
J i
l 5784R/4.70 i
3.0 SAFETY ASSESSMENT-3.1 Introduction The purpose of this assessment is to demonstrate that the proposed license amendment to permit an additional 2 years and 8 months of plant operation would not adversely affect the public health and safety. Most of the information that follows summarizes material previously provided to the NRC. 1 1
-I Section 3.2 provides a review of reference documents which describe the j 1
basis for assuring continued plant safety through its licensed operating !
lifetime. Section 3.3 gives a summary of the plant performance and safety record over 27 years of operation. Section 3.4 provides a review of the assurances for continued functional capability of safety-related components through at least 40 years of plant operation. The assessment conclusion is provided in Section 3.5.
l 3.2 Licensing Basis Documents !
i 1
Ihe Yankee plant is designed, constructed, operated, and maintained to assure continued safety through its licensed operating lifetime. Documents which provide the basis for this assurance are listed in Table 3. They are l the primary references used for this assessment. The following discussion gives a brief review of the origin and content of each document.
3.2.1 Final Safety Analysis Report 1 1
The plant design and safety analyses were initially documented in the Final Hazards Summary Report (FHSR). This report was submitted to the (thon)
Atomic Energy Commission (AEC) in September 1959 in support of initial plant construction and operation. The report included facility and system design descriptions, site characteristics, analyses of accidents and hazards, and plant operating procedures.
l When the initial version of the plant Technical Specifications were issued in July 1960, major sections of the FHSR were incorporated by i l 5784R/4.70 t
reference. This situation existed until the Technical Specifications were changed to their present standardized format in January 1977 (see Section 3.2.4). Yankee, in response to the AEC/NRC, or on its own initiative, requested modification to the plant Technical Specifications approximately 100 times during this approximate 16-year period. This required the AEC/NRC to re-review and reapprove major portions of the FHSR. As a result, a large part of this report was kept up-to-date until January 1977.
In December 1977, tne NRC began its Systematic Evaluation Program (SEP). Yankee was exempt from the periodic updating of an FSAR (FHSR) as required by 10CFR50.71(c) because of its ongoing SEP effort. In July 1983, after completing the major portion of SEP (see Section 3.2.2), Yankee was asked by the '&C to: update its FHSR, resubmit it as a Final Safety Analysis Report (FSAR), and then maintain it as a controlled document in accordance with 10CFR50. 71(c). This was done, and the initial FSAR was submitted in July 1985 (Reference 7). It has been updated annually since that time and presents an accurate and up-to-date description of the plant.
3.2.2 NRC Systematic Evaluation Program The Systematic Evaluation Program was initiated to review the designs of older operating nuclear power plants to reconfirm and document their safety. The review included: (1) an assessment of how the designs of these plants compare with current safety requirements, (2) a basis for deciding how differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety.
The review compared Yankee's as-built design with current design criteria in 137 topical areas. During the review, 48 of these topics were deleted from consideration under the SEP because: (1) they were generic issues which the NRC planned to evaluate separately from the SEP, or (2) they were not applicable to the Yankee plant. Of the 89 topics that renmined, 51 were resolved by determination that the plant met current criteria or was acceptable on another defined basis.
The review of the remaining 38 topics found that certain aspects of the plant design differed from current criterie. These topics were considered in ,
1 1
5784R/4.70
- - _ - _ _ - - _ _ _ _ i
the integrated assessment of the plant, which consisted of evaluating the safety. significance and other factors associated with the identified differences to decide whether backfitting was necessary. To arrive at these decisions, a series of rigorous engineering analyses and assessments were used as well as the results of a probabilistic safety study (see Section 3.2.3).
The results of the program were documented by the NRC in their final Integrated Plant Safety Assessment, NUREG-0825, (Reference 8) and subsequent Safety Evaluation Reports (References 9-16) for each SEP topic.
3.2.3 Probabilistic Safety Study A Probabilistic Safety Study (Reference 17) was completed in December 1982 and submitted to the NRC to support certain decisions being made under SEP and other regulatory areas. The study methodology was consistent with the NRC's PRA Procedures Guide (NUREG/CR-2300). A spectrum of internal events, ranging from a turbine trip to a large break loss-of-coolant accident, were examined. Event and fault trees were developed using plant-specific information and data. Operator intervention was considered at both the top event and fault tree levels. The study did not address external events.
Subsequent studies have been since completed, however, for high wind and tornado events.
The results presented in Reference 17 demonstrate that plant operation poses a smaller risk to public health and safety than larger plants, such as were considered in WASH-1400. The reasons for this are the following:
o The likelihood of an event sequence resulting in core melt is much lower. This is because: (a) a p1&nt operating history that shows that the frequency of off-normal events is lower, (b) the plant generally has diverse systems available to maintain or restore critical safety functions, and (c) these systems are simpler in design with larger margins of safety.
1 I
i 5784R/4.70 ]
l
n y_
. o. "The; reactor' cores i sLapproximately a factor lof five smaller than; newer,. larger plants. Thus, the' amount.of. radioactivity that could-
~
! be. released,'if'a: core' melt were to occur, is correspondingly les's.:
- o. The' Vapor Containertis passively-cooled and, thus, does not depend 3 on active systems-to. provide this function following a postulated closs-of-coolant accident. =Therefore,-the likelihood of.its failure
~
p due to overpressurization following'such an event is lower.
.o The population ~ density in the vicinity of the plant is'very low.
,Approximately 60 people live within a mile of.the plant boundary.
and only-260 live within the Low Population Zone (LPZ). .See discussion in Section 4.3.2.2.
3.2.4 Technical Specifications The plant Technical' Specifications were.first issued in July 1960_as
" Appendix A" to the; license for initial operation. This document included sections on: (1) site, (2) design specifications, (3) performance specifications, (4) initial startup, and (5) operating procedures.and restrictions. It was unique in that.it incorporated by reference complete
. sections of the (then) Final Hazards Summary Report (FHSR). Con'sequently, each time a proposed. change was requested, it was necessary.for the AEC/NRC to re-review.and reapprove major portions of the FHSR.
This situation existed until the early 1970's when Yankee began work'on revising the original Technical Specifications into a standardized version.
, , This revision was completed and approved by NRC as Amendment No. 27 to the
, operating license in July 1976 and became effective January 1977
~
(Reference 18)., The new Technical Specifications were written to conform.as Lelose as' practical to the format of the Westinghouse Standard Technical Specifications.. They include sections on: '(1) safety limits and limiting safety settings, (2). limiting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls. Of these sections, the'first three are most pertinent to this assessment and will be described briefly.
-5784R/4.70
Safety limits and' limiting safety settings, together, provide a " margin of safety" to protect the integrity of the reactor core and main coolant
= l pressure boundary during plant operation. The safety limits are chosen to maintain plant operating parameters to values that are.well below ]
~
conservatively-chosen design basis failure limits. The limiting. safety settings are chosen to assure that automatic protective action will prevent operating parameters from exceeding the safety limits during an abnormal !
situation.
Limiting Conditions for Operation (LCOs) establish minimum conditions )
necessary to assure the required functional capability of safety-related i
' components. Examples are: (1) limits on age-related material parameters such as the number of thermal fatigue cycles to which the Main Coolant System has been subjected over the plant lifetime, and (2) operability requirements for redundant safety system components.
Surveillance requirements are established to assure early detection of unexpected degradation or failure of safety-related components. These. include requirements for component monitoring, inspection, and/or functional testing.
l Monitoring. requirements focus on operating parameters which are indicators of component performance. Requirements for inspections focus on mechanical integrity of component materials. Testing requirements focus on assuring the operability of components associated with standby systems.
3.2.5 Surveillance and Maintenance Program In accordance with the Technical Specifications and requirements of the Code of Federal Regulations (10CFR), Yankee has established a Surveillance and Maintenance Program. This program includes: an In-Service Inspection (ISI)
Program and an Environmental Qualification (EQ) Program which are both complemented by a Preventive Maintenance Program.
3.2.5.1 In-Service Inspection Program The ISI Program was initiated in 1971. This program was developed and !
is being implemented in accordance with: (1) 10CFR50.5S(a), (2) Section XI of the ASME Boiler and Pressure Vessel Code, and (3) the plant Technical 5784R/4.70 L. __ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _
Specifications. The purpose of the program is'to assure continued maintenance of-the integrity and functional' capability of safety-related mechanical component's (including their' structural supports). Such components include:
pressure vesselsi tanks, piping, pumps,.and valves.
, The program has evolved as shown in Table 4. As' indicated in the
, table,' Yankee is presently in the process of' implementing a 10-year Program Plan which was issued'in July 1981 (Reference 19). The components.within the
- scope of this-plan include pressure retaining components (including.their
. support structures) classified as Safety Class 1, 2,.and 3 in accordance with ANSI Standard N18.2, " Nuclear. Safety Criteria for the Design of Stationary
', ' Pressurized Water Reactor. Power Plants", and included within the examination tables of'ASME Section.XI.
The inspections include visual, surface, and volumetric examinations.
.The' surface examinations are 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:
- 1) Identify unexpected service-induced component degradation, evidenced by surface cracks, wear, corrosion, or erosion;
- 2) Locate any evidence of component-leakage during system pressure or !
functional tests; and
- 3) Verify operability of components and integrity of their supports. j l
Records of-inspections completed under the ISI Program are kept in accordance with the requirements of ANSI N45.2.9 and ASME Section XI, and transmitted to the NRC.
The ISI Program Plan will be revised as necessary to comply with the
(. edition of the ASME Code and Addenda in effect 12 months prior to the start of each required 10-year interval, to the extent practical within the limitations of design, geometry, and materials for construction of components.
I 5784R/4.70 ;
?3.2.5.2 Environmental Qualification Program.
7 The Environmental Qualification Program was established in June'1984.
< This program was developed and. organized in'accordance with'the. requirements of.10CFR50.49. Its purpose is.to document that safety-related electrical-components.will' perform as required under all environmental' conditions anticipated or postulated to occur.during their specified service life. The
-program is described.in the Environmental Qualification Program Manual (Reference'20). It includes: an Environmental Qualification (EQ) Master . .
List,-Qualification Worksheets,' Qualification Documentation Reports (QDRs),
Environmental Qualification Assessments-(EQAs). and an EQ Maintenance and Surveillance Program.
The EQ Master. Lists are provided in Reference 20.. In developing these lists, equipment was included if it: (1) wa's relied upon to function during ;
and.following anticipated transients, postulated accidents or external events
- and (2) would be subjected to a " harsh" environment (significantly more severe j than during normal operation). Such equipment also includes all' electrical equipment whose failure'could prevent the required equipment from performing its intended function.
The Qualification Worksheets specify for each component the most severe environmental conditions under which the component is expected to perform.
]
The environment conditions include consideration of: temperature,. pressure, ]
humidity, chemical effects, radiation, aging, and submergence. Consideration is'alsogiven to synergistic effects.
The QDRs and EQAs provide, for each piece of- equipment, evaluations of test data and/or analyses as necessary to demonstrate qualification for the environmental service conditions specified by the worksheets. The EQAs only address equipment subject to conditions outside the Vapor Container. Together these documents. provide the evidence that EQ has been established for each item on the EQ Master Lists.
The purpose of the EQ Maintenance and Surveillance Program is to maintain EQ components in a qualified status. The program consists of corrective maintenance, preventive maintenance, periodic testing, and 5784R/4.70 ;
L
l
- surveillance.. The QDR or EQA for a given component specifies-special
' installation, maintenance, and surveillance required in order to maintain its qualification. This.part of the EQ Program is implemented in accordance with 10CFR50, Appendix'B and is part of the overall plantiPreventive Maintenance Program (see Section 3.2.5.3).
.The EQ Program is a. continuing program. The Qualification Worksheets, QDRs, EQAs, and the EQ Surveillance and Maintenance Prodram,Lfor a given component, will all be updated as required throughout its service life.
In some cases, this aervice life may be specified such that it would be reached prior to the plant operating license termination date. Such components would either be requalified to a longer service life, replaced, or upgraded.
3.2.5.3 Preventive Maintenance Program The Preventive Maintenance Program was established at the.beginning of plant operation. -The purpose'of the program is to maintain the continued functional capability of all important plant components, including both safety-related and nonsafety-related. components. The. program complements the EQ and ISI Programs--in that it covers safety-related components not necessarily' included under those programs.
The program is implemented through procedures which have been developed in accordance with the Yankee Operational Quality Assurance Program, the Technical Specifications and Yankee operational philosophy. These procedures .I are contained in the Plant Procedures Manual (Reference 21). They specify requirements for scheduling, implementing, and documenting all activities _
within the program scope. These activities include: (1) component !
inspections and/or tests, (2) trending, (3) failure or root cause analysis, (4) preventive or correct maintenance, and (5) record keeping. 1 a
Surveillance and maintenance records are kept for each component covered by the program. These records include component operating and maintenance specifications, date of installation, subsequent maintenance or repair history, parameter trends obtained from past surveillance, and future surveillance or maintenance schedule.
5784R/4.70 m________--------------...-..----_-------- . - - - - - - - - - - - - - - --------------2 i
3 3 _ Plant Operating History
.The Yankee Nuclear Power Station has.been in operation for 27 years.
During that time, a' substantial amount of data and experience have been accumulated which demonstrate that plant aging has not adversely affected the safety and reliability of the plant. This data and experience is reviewed briefly-in the following discussion. This discussion considers: (1) operating performance,'(2) component integrity, (3) plant modifications, and (4) regulatory performance.
3.3.1. Operating Performance H Figure 1 shows the variation of the plant capacity factor with time j from the beginning of plant operation through 1986. This data demonstrates ~a f consistent performance. The lifetime average capacity. factor through December 1986 was 74.2 percent. Since 1982, the annual capacity factor has averaged 84.5 percent.- In 1985, during its 25th year of operation,--the plant operated for 336 consecutive days, the longest continuous run of its history. The t capacity factor.during 1985 was 80.8 percent. During 1986, it was 95.2 )
percent. This performance is among-the best in the industry.
t l Figure 2 shows the variation with time of the refueling outage length. l The refueling outage length for 17 outages has averaged approximately 10 weeks. There has been a 2-week increase in the average refueling length over the past six outages, as compared to the previous 10. This increase has been '1 primarily due to increased surveillance and modifications, not age-related problems.
Figure 3 shows the variation of plant trips with time. During the j ' plant's first 26 years of operation, the number of trips per year has averaged
'4.3. Over the past 12 years, the number of trips per ye'ar has averaged 3.8.
~
l I
Recent modifications to the plant tripping circuits made in 1983 has reduced
'the number of trips. During 1984-1986, the number of trips per year has averaged 3.0.
L 15784R/4.70
3.3.2 ? Component Integrity-As discussed'previously in Section 3'.2.5.1, an In-Service Inspection (ISI)' Program to assure continued component integrity has been active at
. Yankee since the first edition of Section XI of the ASME Code was published in 1970. To date, inspections have.been performed on'all Class'l and high-energy
-- Class. 2.. components. The following-is a brief description of some of.these inspection results.
o Reactor Internals'- The' reactor internals'have been visually examined on a regular basis since the start of plant operation.
Bolting problems were experienced during the early years of operation resulting in redesign and replacement of the lower shroud tube package in 1972. Since that time, visual inspections have shown no evidence of degradation.
o Reactor Vessel - The vessel nozzles were ultrasonically examined in 1984. The examination was performed in accordance with ASME Section XI and the near-surface sensitivity requirements of NRC Regulatory Guide 1.150. No indications of any significant
' deterioration were'found during this inspection. 'l The vessel'and head tlange welds have been examined regularly since 1970 with no indication of problems. The head cladding has been monitored since the late 1960's.
The cladding of the reactor vessel is stainless steel sheet spot welded to the vessel interior. These-spot welds have a " quilted" appearance. Some of the quilts have apparent fatigue cracking ecross them. These. cracks have been monitored, however, and have-shown no apparent growth during the last 15 years, o Pressurizer - The pressurizer was included in the initial ISI Program inspections which began in 1970. The pressurizer cladding i has experienced the same type of quilt cracking as the reactor vessel cladding. More recent inspections of the pressurizer j
5784R/4.70 ;
i
L i
cladding has shown that no change has occurred over the inspection intervals. No other evidence of deterioration has been found.
o Steam Generators - The steam generator pressure boundary welds have been examined on the primary side since 1970 and on the secondary side since about 1978. No deterioration of these welds has been observed. The steam generator tubes undergo periodic eddy current exams. Yankee uses Type 304 stainless steel tubes, with a 3/4-inch OD, and average 0.072-inch wall thickness. To date, a total of 309 out of 6,480 tubes have been plugged. Thus, only about 4.8% of the total tubes have been removed from service during 27 years of I operation..
o Piping - Nondestructive Examination ~(NDE) methods have been used to examine piping components throughout the life of the plant. Formal examinations began in 1970 when the ISI program was initiated. The majority of the primary piping is wrought Type 304 stainless steel, although some cast stainlesa steel is used for the fittings. Welds have also undergone liquid penetrant, ultrasonic, or radiographic examinations. None of the inspection results to date have indicated any significant age-related deterioration of the systems.
3.3.3 Plant Modifications A number of major design modifications have been made during the 27 '
years since Yankee went into operation. These changes have been made to upgrade plant equipment or safety systems or to a lesser degree to replace
-equipment which has failed, become obsolete, or reached its end-of-useful f life. A chronology of the more significant additions is provided in Table 5, of which a few are discussed briefly below.
o Three independent emergency diesel generators with three independent trains of emergency power were installed in 1970. !
o In 1972, a new dedicated Emergency Core Cooling System (ECCS) was installed. The system includes an accumulator, three trains of high and low pressure pumps, and associated instruments and l controls.
1 l
u______1______ _ . _ .
]
s
o In.1980, two new motor-driven emergency feedwater pumps were added. These-two pumps were in addition to the existing steam-driven emergency feed pump. The new motor-driven pumps feed through two independent paths to the steam generators.
o In 1981, the main steam nonreturn valves were automated, which enhanced protection in the event of a main steam line break.
o In 1982, a Safety Parameter Display System (SPDS) was installed as a joint demonstration by EPRI and Yankee. This was one of the l first operational systems of its kind in the ind_atry. The system I
provides a display for the plant operators of the plant critical safety functions to aid operational decisions during anticipated 1 transients or postulated accidents. ,
o In 1985, a new Safe Shutdown System was installed. This is a dedicated system designed to enable the plant operators to maintain i the plant in a safe shutdown condition following postulated external events, such as earthquakes, tornados, or fires.
l o During 1980-1985, a portion of the Reactor Protection System and the entire Feedwater Control System were replaced with updated, state-of-the-art, systems. The original pneumatic Feedwater Control System was replaced with an electronic analog system. The instrumentation used for the six process variables associated with the Reactor Protection System were upgraded from the original ,
magnet.ic amplifier type to electronic analog components including sensors, cables, and electronics.
In reviewing the plant modifications during the first 27 years of operation, it can be concluded that:
- 1) Most af the modifications have involved additions rather than f replacements of equipment due to age-related failure. Component f- aging has not had a significant effect on plant operation, mainly
} because the effects are gradual, not precipitous and can be l detected and tracked by routine plant surveillance and maintenance.
5784R/4.70 E-__ _
- 2) The net effect of these changes has been to enhance the safety provided by the plant _ systems.- This conclusion is supported by the-results of the NRCs Systematic Evaluation Program and the Probabilistic Safety Study (see Section 3.2.2 and 3.2.3).
- 3) .Because a number of important plant components have been added or replaced with more modern equipment since the beginning of plant operation, the effective age of the plant equipment is significantly less than 27 years.
3.3.4 Regulatory Performance The Systematic Assessment of Licensee' Performance (SALP) Program was initiated by the NRC in 1980. The purpose of this program is to collect available observations and data on a periodic basis to evaluate licensee performance in selected functional areas important to nuclear safety and the environment. Areas evaluated under the program include: plant operations, radiological controls, maintenance and modifications, surveillance, fire protection and housekeeping, emergency preparedness, security and safeguards, refueling and outage management,. assurance of quality, training and qualification effectiveness, and licensing activities.
Based on the NRC evaluations, the performance in each area is classified Category 1 (highest), 2, or 3 according to the following definitions:
Category 1 - Reduced NRC attention may be appropriate. Licensee management attention and involvement are aggressive and oriented toward nuclear safety; licensee resources are ample and effectively used such that a high level of performance with respect to operational safety is being achieved. j i
Category 2 - NRC attention should be maintained at normal levels.
Licensee management attention and involvement are evident and concerned l with nuclear safety; licensee resources are adequate and reasonably .
1 I effective such that satisfactory performance with respect to l l .
operational safety is being achieved.
i
. , )
5784R/4.70 )
i L.
Category 3 - Both NRC and licensee attention should be increased.
Licensee management attention or involvement is acceptable and considers nuclear safety but weaknesses are evident; licensee resources appear strained or are not effectively used such that minimal satisfactory performance with respect to operational safety is being achieved.
Table 6 provides a summary of recent SALP ratings received by Yankee.
-This track record of safe facility operation, as judged by the NRC, is among the best in the industry and provides a high level of confidence that the plant will continue to be operated'and maintained in a way which will meet, if not routinely exceed, the level of safety performance required by the approved licensing basis.
3.4 Assurances for Continued Functional Capability of Safety-Related Components In order to assure the present level of safety is maintained during future plant operation, it is necessary to assure the continued functional i capability of safety-related components. These are components associated with systems which are designed to prevent or mitigate events that could cause a release of radioactivity to the environment. The systems that provide this protection, and the assurances for continued functional capability of their components, are described in References 7-21 and discussed in Section 3.2.
The following discussion is based on a review of these documents. The components discussed will be classified into three groups as defined in Table 7.
3.4.1 Mechanical Components
' 3.4.1.1 Main Coolant Pressure Boundary The mechanical components associated with the Main Coolant System pressure boundary include: the reactor vessel, piping, valve bodies, pump casings, steam generators and pressurizer. These components were designed in i
accordance with the codes and standards listed in Table 8. The design of these components included consideration of potential effects of age-related 5784R/4.70 l
--a-----------
phenomena such as corrosion, thermal cycling fatigue, and radiation-induced embrittlement. The consideration of these effects was also taken into account when the operating limits and surveillance requirements were established in the Technical Specifications.
In accordance with the latter requirements, the Main Coolant System is included in the In-Service Inspection Program (see Section 3.2.5.1). Al l components are located such that critical areas are reasonably accessible for the required inspections and/or tests. The reactor vessel design enables accessibility to the internal vessel surface near the nozzles connecting to the main coolant piping, the vessel cladding surface, and the external and internal surfaces of the top head. The design of Main Coolant System components is such as to allow inspection of external surfaces.
The potential for corrosion was accounted for by using corrosion resistant materials. All mechanical components that are in contact with reactor coolant, except the fuel, are either made of or clad with stainless ;
steel. The fuel is clad with Zircaloy. The Main Coolant System water l chemistry is selected to minimize corrosion. A periodic analysis of the coolant chemical composition is performed to verify that the coolant quality )
is within specifications.
Components of the Main Coolant System 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 start-up and shutdown operations. During ,
1 startup and shutdown, the heat-up and cooldown rates are limited to less than i 100 F/ hour (typically 25-30 F/ hour), consistent with system design specifications. Also, there is a limit of 200 on the number of heat-up and cooldown cycles that can be experienced during the plant lifetime. These limits are set by the Technical Specifications. As of August 1987, after ;
27 years of plant operation, only 61 heat-up and cooldown cycles have been experienced. On this basis, the number of cycles expected over 40 years of plant operation is 90, which is well below the Technical Specification limit of 200 cycles.
5784R/4.70 1
l I
During plant startup and shutdown, the Main Coolant System I pressure / temperature combination is also limited by the Technical Specifications. These limits are established to maintain conditions which j i
assure adequate fracture toughness of the reactor vessel, in accordance with the requirements of 10CFR50, Appendix G and Section III of the ASME Boiler and
' Pressure Vessel Code, Appendix G. In determining these limits, consideration is given to the cumulative effects of fast neutron embrittlement on vessel RT NDT*
NDT is based on a best-estimate calculation of fast neutron fluence and a conservative prediction of material )
toughness. Pressure / temperature limits are adjusted at the beginning of each operating cycle to account for the RT shift predicted for. that cycle.
NDT ]
1 l
The fracture toughness of the reactor vessel to thermal shock during a j postulated Loss-of-Coolant Accident (LOCA) has also been determined, in !
accordance with.the requirements of 10CFR50.61. The determination included an analysis of the effect of cumulative neutron fluence on Reference NDT. The i fluence was calculated assuming 40 years of operation at full power with an average plant capacity factor of 80% (32 EFPYs). The analysis results !
indicated the limiting reactor vessel material to be the lower plate. Because of the relatively low copper / nickel content of this material, the calculated maximum RT PTS C rre8Ponding to 32 EFPYs was only 253 F. This value is well below the NRC PTS screening criterion of 270 F for plate material. I Therefore, the Yankee reactor vessel can be operated for at least 32 EFPYs i without significant risk resulting from pressurized thermal shock. Assuming a future capacity factor of 80%, 32 EFPYs of generation would not be accumulated until October 2002. Moreover, the NRC screening criterion of 270 F would not be reached until March 2020. The NRC recently issued a favorable safety evaluation of this analysis (Reference 22).
3.4.1.2 Other Mechanical Components The passive components (tanks, pump casings, and valve bodies) associated with other safety-related systems are designed to the same codes as the components that comprise the Main Coolant System pressure boundary. Also, consideration was given to possible aging effects of corrosion, erosion, and thermal cycling fatigue. Therefore, the expected service life of these passive components is greater than 40 years, as for the Main Coolant System 5784R/4.70
boundary. Nevertheless, these components arc included in the plant In-Service Inspection and Preventive Maintenance Programs, so that unexpected age-related degradation will be identified and corrected if it occurs.
Many of the active (moving or rotating) mechanical components, on the other hand, are expected to wear out and be periodically replaced during the plant's operating lifetime. These components are periodically inspected and maintained under the In-Service Inspection and Preventive Maintenance Programs. Age-related degradation will therefore be identified and corrected, and component functional capability will be maintained.
Thus, assurance is provided that the functional capability of both active and passive mechanical-components will be maintained for 40 or more years. Passive mechanical components are designed such that they are not expected to be replaced during that time. The functional capability of active components will be maintained through maintenance and/or periodic replacement. Accordingly, it can be concluded that safety-related mechanical components are designed to function more than 40 years or will be inspected and maintained such as to assure continued functional capability.
3.4.2 Electrical Components Electrical components, which would be required to function in a harsh (significantly worse that normal) environment during a design basis event, are covered by the Environmental Qualification (EQ) Program. This program was established in accordance with the requirements of 10CFR50.49, as discussed in Section 3.2.5.2. The program provides assurance that the components can perform their safety function in their normal or (if necessary) design basis environments for their qualified lifetime. This assurance is based on analysis and/or tests which take into account: (1) environmental conditions expected during the design basis event and (2) aging due to cumulative !
exposure to the normal service environment.
l The EQ Program is a continuing program. Components covered by the program are subjected to surveillance and maintenance to ensure that they I remain qualified throughout the plant service life. If the qualified lifetime of a component is determined at any time to be less than the expected plant l
5784R/4.70 L__________.
i +
, a 4 n r
,,I..gI /.
a
' service 111fe,thecomponentwill.berequalified'toa'longerNualified 111fetime,-replaced,'or upgraded. l I
Safety-related plectrical components which would not be subjected to af j harsh environment and %re'therefore not' included in the'EQ. Program, are. !
covered by the. Preventive Maintenance Program.- This-program includes' periodic surveillance, maintenance,;replachwat,orrefurbishmentasnecessaryto s
assure their continued funct (n'lui capability. S'ee discussion in. V
~
section 3.2.5.3. M l-3.4.3 Structural Components 3.4.3.1 . Vapor Container- 1 l'
The' Vapor Container is' a ipherical shell which is made of welded steel plate. It was designed and' fabricated in accordance with Section VIII of the a
ASME Boiler ~and Pressure Vessel Code,.1956 Edition,and. applicable codetcases.
- During the SEP,'the Vapor Container design was compared to the requirements of Section III of the 1980 Edition of this code and found acceptable ,
(Reference 16). -
i Corrosion of the shell is limited by a protective paint which is .J applied to both'the interior'and exterior surface. The interior surface l
- coating is inspected at every refueling outage for deterioration due.to high humidity. The exterior surface coating is.also periodically inspected. These 'l inspections, and any required maintenance. are done in accordance with the !
- Technica. Specifications and the Preventive Maintenance Program.
y
]
J i
Prior to the beginning of plant operation, the shell seams were '
completely radiog % 'trever possible. Welds'not radiographer were -
subjected to a magnetic y, cicle inspection. DuYIng plant operation, the l
.iH
_ Vapor Container is maintained at a pressure slightly higher than atmospheric Io facilitate detection of any leakage through welds or penetrations. This leakage is continuously monitored. Also, the Vapor Container is subjected to periodic pressure testing in accordance with Technical Specifications; and the results'are reported to the NRC. ;
i 5784R/4.70 l
E .
/
L d -
Inspections and testing have not indicated any deterioration in the structural integrity of the Vapor Container over the first 27 years of plant
- operation.- Moreover, for the reasons Biven above, it should be possible to maintain the integrity of this structure for well beyond 40 years of plant operation.
3.4.3.2 Other" Structures Other critical plant stEuctures are made of steel and/or reinforced concrete. Structural steel components were designed and fabricated in accordance with the American In'stitute of Steel Construction (AISC), Steel Construction Manual. Concrete structures were designed in accordance with the requirements of the American Concrete Institt.te (ACI), Building Code. During the SEP, critical structures were re-evaluated to current codes and standards. In each case, the design was either found to be in compliance with the newer version of the code or otherwise shown to be acceptable with modifications currently being implemented.
Plant structures are subject to periodic inspections and maintenance.
l Such maintenancelocludes periodic reapplication of protective coatings and concrete surface repair.
)
Experiende in other industries with similarly designed structures indicates that, with an aggressive inspection and maintenance program, a service life well'An excess of 40 years can be anticipated. On this basis, it j s
should also be possible to maintain the integrity of these structures well l beyond 40 years.
3.5 Conclusion The preceeding assessment provides a review of: (1) documents which describe the present licensing bases. (2) plant operating history, and (3) assurances for the continued functional capability of safety-related components thrdugh at least 40 years of plant operation.
l I
5784R/4.70 w____-_______ _.
The documents which describe the present licensing bases include:
o Technical Specifications which set forth the conditions that are acceptable for plant operation.
o A Final Safety Analysis Report (FSAR) which provides a description of the overall plant design and safety evaluation.
o NRC Systematic Evaluation Program Reports, which provide the results of a favorable NRC evaluation of the plant relative to current safety design criteria.
o A Probabilistic Safety Study, which demonstrates that plant operation poses a smaller risk to public health and safety than larger contemporary plants such as were considered in WASH-1400.
o Surveillance and Maintenance Program documents, which describe the plant's In-service Inspection, Environmental Qualification, and Preventive Maintenance Programs.
The plant operating history demonstrates the validity of the present licensing basis. More specifically, it demonstrates that the plant's ;
reliability and safety have been maintained. This is evidenced by:
o A lifetime average plant capacity factor of 74.2 percent. !
o Favorable results under the In-Service Inspection Program.
o A High Regulatory Performance Rating Under the NRC's SALP Program.
1 I
o A history of plant safety improvements.
5784R/4.70
N ,1 ~-
o L/: -
lD Safety-related systems and components are designed, constructed, h3 operated, and naaintained to assine their continued functional capability.
P through 40 years of plant operation. In particular: l
- s. ,
o The fracture toughness of the reactor vessel to thermal shock l during a postulated' loss-of-Coolant Accident (LOCA) has been I determined to be acceptable, in acaordance' with the requirements of 10CFR50.61. This determis tion in 1uded consf.deration of a cumulative neutron fluenceJ ccrrespc$d.ing to LO yants of operation +
l at full ri' er with an average. plant capacity Ezetor of 80% ,
(32 ETPYs ). c
, , x.
+ o Main coolant pressure boundary components aro' designed to include '
cor. sideration of potential efiects of age-related phenomena such as torrosion, thermal cycling fatigue, and rtsintica-induced i
en.b ri tt lenen t . Components are also designens to withat md the J ,
fattigue ef fects of cyclic loads due to system temperature and s pressure changes.
I '
o Passive mechanical components associated with safety-related '
systems are designed to include consideration of' t'do potential
- l effectsofage-relatedphenomenasuchascorrosion', erosion,bnd thermal cycling fatigue. Nevertheless, these'cowonents are 'l y ( subject to periodic inspections and maintenance;,
4 o flany active 'mechsnical components associated Qith safety-related systems are expected to wear out. Therefore, these components are
,, periodica1Iy inspected and maintained or replaced under the J
In-Service Inspection and Preventive Maintenance Programs.
f o Electrical components whid would be required to function in a harsh environment during a iesign basis event are monitored and maintained under the Environmental Qualification (EQ) Program.
Other safety-related electrical components, not included in the EQ Program, are monitored and maintaited under the Preventive Maintenance Program. !
i
_33 5784R/4.70 i
o . Plant structures were designed in accordance with applicable codes and standards at the time of construction and are subject to periodic inspections and maintenance.
o Corrosion of the vapor container shell is limited by a protective paint which is applied to both the interior and exterior surface.
Inspection and maintenance are done in accordance with the plant's Technical Specifications and the Preventive Maintenance Program.
The vapor container is also continuously monitored and tested for leakage.
On this basis, it can be concluded that the proposed license amendment
-would not adversely affect the public health and safety.
5784R/4.70
I 4.0 ENVIRONMENTAL ASSESSMENT 4.1 Introduction The purpose of this section is to provide an assessment that demonstrates that the environment will not be adversely affected by the proposed amendment to the plant's operating license. Radiological and nonradiological effects for both on-site and off-site environments during the amendment period are assessed against the four criteria shown below. The assessment was performed under the condition that the plant's mode of operation remains essentialy unchanged through the year 2000.
o Environmental control / monitoring systems and programs meet applicable regulatory criteria and show evitence of: continual appropriate enhancement as well as effectiveness through present and future years of operation.
o The rate of discharge or generation of radiological and nonradiological effluents, solid wastes, and occupational exposures are projected to remain well within the bounds of:
applicable regulatory criteria and permits, or where applicable, the upper limits established through typical plant operation over recent years. .
o The increase in cumulative effects of applicable parameters are projected to be inconsequential.
o The off-site exposures that result from a postulated accident continue to meet the criteria of 10CFR100 through the proposed amendment term.
In effect, these four criteria are the basis for the determinations made by other licensees that have already applied for and subsequently been granted a similar amendment.
In addition to this introduction, this chapter is divided into four additional sections. In Section 4.2, the following systems and programs for 5784R/4.70
environmental control and monitoring are described: Waste Disposal System and Spent Fuel Pool, ALARA Program, Radiation Monitoring Systems, Radiological Surveillance Program, and Nonradiological Surveillance Program. Throughout each description, any refurbishment or upgrading, as well as the findings of environmental studies performed since the plant's startup, are highlighted.
Finally, Yankee's commitment to these programs and systems is verified by a SALP rating of 1.0 in the area of radiological controls. I Section 4.3 presents an assessment of the environmental effects of plant operation during the proposed amendment term based on the applicable criteria that were defined earlier. In addition, the plant's historical data i in each area of assessment is compared with the performance of the industry's average PWR (as defined by the NRC) and, as well, the recent trends of such data are discussed. For the purpose of comparisons and establishing long-term trends, moving average data involving either'a three- or five-year period is used in order to levelize actions such as mandatory retrofitting, major repairs, or unanticipated outages which are beyond the scope of normal operations at any power plant.
In Section 4.4, the off-site exposures from releases during postulated accidents are considered. This environmental effect was previously evaluated in the plant FSAR where the results were found to be within the limits set forth in 10CFR100. Through references to other sections of this assessment and a projection of future populations within 50 miles of the plant, it is shown that the FSAR evaluation will remain valid and 10CFR100 criteria met in future years, including the term of the proposed amendment.
Finally, Section 4.4 presents the summary and conclusions of this assessment.
4.2 Systems and Programs for Environmental Control and Monitoring 4.2.1 Environmental Control Systems and Programs 4.2.1.1 Waste Disposal System I
1 l The Waste Disposal System is described in Section 209 of the FSAR.
I This system receives, contains, adequately treats, and safely disposes of all ,
J l
i 5784R/4.70 '
l u_________.-
i radioactive wastes.. The basic processes used are: natural decay of )
radioactive isotopes, fil'cration to remove particulate matter, evaporation to l concentrate radioactive constituents into a small volume of liquid waste to be solidified in cement, filtration of gases by charcoal and HEPA filters, dilution of low activity liquid and gaseous discharges, and compaction of dry active waste. The system consists of liquid and gas storage tanks, an evaporator, pumps, compressors, heat exchangers, filters, instruments, piping, and valves, as described in the FSAR.
4.2.1.2 As Low as Reasonably Achieveable (ALARA) Program In accordance with 10CFR20.1(c) and the plant Technical Specifications, the Yankee has established an ALARA Program. The purpose of this program is to maintain occupational radiation exposures "As Low As Reasonably Achievable." The program assures that ALARA is considered in all aspects of plant design, operation, maintenance, and inspection.
All radiation workers at the YNPS receive training in how proper work practices can help maintain their exposures ALARA. In addition, a training program for engineers on ALARA design has been developed by the Yankee Nuclear Services Division (YNSD). The training will be provided to all YNPS and YNSD engineers who prepare design changes for the YNPS. The training ensures that design changes reflect appropriate ALARA considerations in the design, installation, operation, and maintenance of the change.
Design changes prepared or ve.rified by YAEC engineering and operations personnel, are implemented in accordance with Engineering Instructions (EI) contained in the YAEC Engineering Manual. These instructions require the consideration of ALARA during both preparation and review of the change.
Design change packages are routed to the Environmental Engineering Department (EED) for review. The EED ensures that ALARA has been considered and informs the plant Radiation Protection Manager (RPM) of significant vork plans. This enables the installation work to be preplanned and allows the plant RPM to have special ALARA concerns factored into the design in a timely manner.
Installation, operation, and subsequent maintenance of design changes are done in accordance with operating procedures which provide a systematic program of f 5784R/4.70 w____________
planning, monitoring, and~ reviewing work activities to ensure that ALARA exposure controls are implemented.
At the end of each refueling outage, work which required significant radiation exposure is reviewed by the plant Radiation Protection Department.
The purpose of these reviews is to identify ALARA-related inadequacies in designs or procedures used for equipment instal 1ation, operation, surveillance, and maintenance. The results of these post-job reviews provide knowledge which can be used to improve future designs and reduce exposures on the same or similar jobs in the future.
An annual plant exposure goal is set based on input from each plant department. Current exposure is reviewed periodically against the goal to identify adverse trends and to implement timely, corrective action when necessary. In 1986, the YNPS purchased and installed a state-of-the-art Radiation Protection Records Management System. This system signi'icantly enhanced the plant's capability to track doses associated with specific jobs, components, and systems and improved the ALARA planning process. j Finally, in future years of operation, Yankee will continue to comply h with these requirements and also apply advanced technology when available and j appropriate.
4.2.2 Environmental Monitoring Systems and Programs i
i 4.2.2.1 Radiation Monitoring System l
The Radiation Monitoring System is described in Section 215 of the l
FSAR. This system monitors radiation levels associated with process systems {
and areas at various locations inside and outside of the Vapor Container. It j is designed for use during normal operation or postulated accident I l
situations. The system includes equipment for detecting, computing, l indicating, and alarming. The present equipment was installed during l
1978-1979, completely replacing and upgrading the original equipment and increasing the number of monitored systems and locations.
l 5784R/4.70 1
Monitoring of the Waste Disposal System is provided as necessary to assure safe waste collection, processing, storage, drumming, and either to:
(1) control any release to the environment or (2) ship waste to established disposal sites.' Setpoints for both liquid and gaseous effluent radiation monitoring instrumentation are based on the methodology of the Off-Site Dose Calculation Manual (ODCM). This methodology ensures that members of the public at or beyond the site boundary are not exposed to annual average ;
radionuclides concentrations exceeding the Maximum Permissible Concentrations (MPCs) after consideration of the maximum accumulated off-site dose allowed by tha Technical Specifications.
4.2.2.2 Environmental Radiological Surveillance Program The Environmental Radiological Surveillance Program is described in Section 305 of the FSAR. This program was established in 1958, approximately 2 years prior to the start of plant operation. The purpose of the program is to: (1) verify the effectiveness of the systems and procedures for control of releases of radioactivity from the plant and (2) measure environmental levels of such releases for impact assessment. This is accomplished by periodically measuring radiation levels and amounts of radioactivity in samples at various locations surrounding the plant.
The types of sample media used correspond to the possible exposure ,
pathways. These are direct radiation, airborne, waterborne, and ingestion. 1 j
Lirect radiation is measured by Thermoluminescent Dosimeters (TLDs). Airborne !
radioactivity is collected by passing air first through a glass fibre filter i and then through a charcoal canister. The filter collects particulate I activity and the charcoal collects Iodine-131. Waterborne radioactivity is collected by taking river water samples, samples of ground water, and cores of shoreline sediment. Ingested radioactivity is collected by obtaining samples of milk, commercially or recreationally important fish species, and representative food crops.
The measurements are made within an area which is divided into two zones. Zone 1 is an area that is considered to be within the potential influence of the plant. Zone 2 is an area not considered to be influenced by j the plant. The measurements in Zone 1 are compared to the measurements in 5784R/4.70 l
Zone'2 in order to differentiate between the effects of plant operation and the effects of. natural background or other causes. To ensure that the program continues to include those. locations whose environmental samples are most likely to show plant related radioactivity, a land use census is conducted annually. Changes in sampling locations may be required following the census based on relative potential doses or dose commitments and the availability of samples.
4.2.2.3 Nonradiological Surveillance Program Surveillance of the biological effects of plant operation began in the mid-1970's following enactment by Congress of the Federal Water Pollution Control Act. During 1974-1977, several studies were done to determine the effects of the Circulating Water System discharge on the fish, wildlife, and other organisms living in or on Sherman Pond and adjacent waters, (see discussion in Section 4.3.5). These studies provided the basis for a National
, Pollution Discharge Elimination System (NPDES) Permit originally issued in 1974 from the Environmental Protection Agency (EPA) and the Massachusetts Division of Water Pollution Control (MDWPC), as discussed in Section 4.3.2.2, l
l l
Subsequent to these initial studies, a program was initiated to periodically monitor species indigenous to the pond and adjacent waters. The purpose of the program is to provide continued assurance that the environmental effects of the plant's Circulating Water System remain within the findings of the 1974-1977 studies. j l
1 4.3 Assessment of Environmental Impact During Normal Operations i
4.3.1 Radiological
]
4.3.1.1 Occupational Radiation Exposure Yankee's occupational exposure trend and comparative magnitude with the industry average PWR based on average annual exposures in terms of person-rem per five-year period is summarized in Table 9. The values in Table 9 indicate that Yankee has always had very stable occupational exposures although the values from the periods during 1974 - 1980 are somewhat less than those during 5784R/4.70 w___________ . _ _ _ _ _
I i
1977 - 1984 because of post-TMI mandatory backfitting and increased inspection !
requirements. In comparison with the industry average PWR, Yankee's occupational exposures have always been at least 50 percent less. Likewise, 4
when such occupational exposures are reported on an " average-worker" basis, 1 Yankee is still favorable in comparison to the industry average PWR by a margin of at least 0.1 worker rem. (Yankee's average worker exposure is expected to remain at or below 0.49 worker-rem in future years of operation.)
-Table 10 presents by year the annual occupational exposure and plant outage status. As expected, those years with no outage period had the lowest occupational exposure with an average of only 72 person-rem / year. Conversely, those years with typical refueling outage periods had higher exposures, with J an average of 327 person-rem / year (for conservatism, the lowest value of 116 person-rem in 1975 is not included). The highest annual occupational exposure of 474 person-rem occurred in 1982 as a result of refueling, inspections, and modification of the incore instrumentation.
Given Yankee's continued implementation of its ALARA Program, the plant's historically stable occupational exposure and high SALP ratings in Radiological Controls and Refueling and Outage Management, it is expected that the most recent occupational exposures stated in Table 10 will serve as an upper limit in future years of normal operation. During the proposed amendment term, it is anticipated that Yankee will operate with an approximately 19.5-month fuel cycle mode (see section 4.3.1.4) which means that there will be two refueling outages during the proposed amendment term.
Using the previously-defined average annual exposures of 327 and 72 person-rem for years with and without typical refueling outages, respectively, it is estimated that the total occupational exposure during the proposed amendment term will be about 732 person-rem which averages to 275 person-rem / year. This projection is consistent with the plant's history of stable occupational exposure levels and in accordance with 10CFR20 as well as Regulatory Guide 8.8.
l 4.3.1.2 Off-Site Radiati.pn Exposure Consistent with the 10CFR20 ALARA requirement, the NRC in 1975 issued 10CFR50, Appendix I, which established radioactive design dose objectives for liquid and gaseous (including iodine / particulate radionuclides) effluents.
l 5764R/4.70
- _ _ _ _ _ = _ _ _ _ . ._ _ __ . _ _ _ . _ .__ . _ - . .__
i Table 11 summarizee for the three types of rad live effluents the j Appendix I bases for showing compliance with .tLARA and the plant's actual average releases since Appendix I was incorporated in the Technical Specifications in 1983. I s
A review of the values in Table 11 indicates that the actual performance of the plant to control and limit liquid and gaseous radioactive effluents has been well within the Appendix I objectives. For instance, for i liquid effluent, the maximum individual organ dose averaged no more than 0.053 mrem / year to any organ or 0.04 mrem / year to the whole body, as compared to Appendix I criteria of 10 and 3 mrem / years, respectively. Based on recorded noble gas effluents over the last three years, the calculated maximum gamma and beta doses were 0.33 mrad per year and 0.97 mrad per year, respectively, l or approximately three and five percent of the Appendix I ALARA exposure objectives of ten and twenty mrad per year, respectively. Finally, dose calculations based on the actual iodine and particulate releases over the last >
three years indicate that the average maximum annual organ dose is 0.60 mrem per year which represents only four percent of the Appendix I ALARA dose criteria.
l The three year average ef fluent doses cited in Table 11 are significantly less than the 10CFR50 Appendix I limits. In addition, Yankee also performed an evaluation (in 1975) of the Waste Disposal System that demonstrated that the as-built design was capable of meeting Appendix I objectives during both normal operation and anticipated transients. This information was submitted to the NRC under the title, " Supplemental Information for the Purposes of Evaluation of 10CFR, Part 50, Appendix I," on f June 2, 1976, with two amendments submitted on August 31, 1976, and l October 20, 1976 (Reference 25). Based on the continued operation of Yankee using existing liquid and gaseous radwaste treatment systems coupled with the Radiation Monitoring System and Radiological Surveillance Program, the anticipated liquid and gaseous effluent doses during the period covered by the i
requested amendment remain a fraction of 10 CFR, Part 50, Appendix I limits. l L
4.3.1.3 Solid Waste Generation ;
1 The volume of solid waste generated at the Yankee plant has historically been among the lowest in the nuclear industry. Table 12 presents 5784R/4.70
a summary of the three-year moving average of annual solid waste generation for. Yankee and the average industry PWR. These values show that compared to the average PWR Yankee generates less than half the annual volume of solid waste and has also followed the industry's long-term trend of reducing volume. Yankee's. annual generation of solid wastes from 1980 through 1986 are l
l summarized in Table 13 which shows that in 1986, a minimum quantity of solid waste was generated due to the absence of a refueling outage, the installation of new compaction equipment in late 1985 and improved administrative controls.
During the future years of plant operation - including the proposed amendment term - continued emphasis on lower solid waste generation is expected to at least maintain if not improve the most recent values stated in Tables 12 and 13. Thus, the maximum solid waste volume over the entire ae adment term should be on the order of about 16,100 cubic feet which averages to only 6,000 cubic feet per year, which is consistent with the plant's recent performance.
4.3.1.4 Uranium Fuel Cycle The Yankee reactor contains 76 fuel assemblies. Following Cycle 1, 74 assemblies were removed and replaced with fresh fuel. Following all subsequent cycles, however, only about one-half of the 76 assemblies have been removed because beginning with Cycle 2 a two-region reload core design was implemented. The reload scheme varied somewhat during Cycles 2 through 9, but has been unchanged since Cycle 12 (1976). At each refueling since that cycle the number of assemblies removed f rom the core has alternated between 40 and 36, with the cycle length currently averaging over 19 months duration.
As discussed subsequently in Section 4.3.1.5, Yankee will operate in a similar fuel cycle mode through the end of the proposed amendment term.
Therefore, the additional years of reactor operation will almost I proportionally increase the total fissile uranium required. Thic small impact, however, is justified in light of the continued benefit received from Yankee's operation as discussed in Section 2.
l 5784R/4.70 1 i
4~.3.1.5 .
Spent Fuel The spent fuel pool'or pit is located outside the Vapor Container with the irradiated fuel assemblies stored in double-tiered racks. This' double' tier arrangement'was approved.by the NRC in 1982 (Reference 23). The pool capacity,: including both tiers, is 735 fuel assemblies; although, the . licensed'-
maximum inventory is-721 assemblies. The pit is also used for storage of
~
irradiated' control rods or followers prior to shipment off-site for disposal.
A Cooling and Purification System is provided to remove. decay heat from the
' stored fuel and prevent the buildup of radioactive corrosion products in.the.
water.
The fuel discharged from Cycles 1'through 8 was reprocessed by Nuclear Fuel Services, Inc. at West Valley, New York. Fuel discharged from subsequent cycles has remained in the spent fuel pool because of the moratorium placed on spent fuel reprocessing in the early 1970's. The spent fuel inventory in the pool'following the Cycle 18/19 refueling in 1987 is 377, which means that 344 licensed locations remain.
Future cycles lwill be designed to assure that the total number of j assemblies discharged between 1987 and the proposed license expiration in July 2000 will not exceed 344, including full discharge of,the final core.
For the existing two-region alternating 40/36 refueling design, this can be accomplished with eight more cycles averaging at least 19.5 months in length
~
I (starting with Cycle 19 in July 1987). This would result in four discharges of 40 assemblies, three discharges of 36 assemblies, and a final discharge of all 76 assemblies, for a total of 344. Therefore, even if no additional spent l l
fuel assemblies are shipped from the plant before July 2000, the total number of. assemblies to be stored on-site will not exceed the spent fuel pool license limit of 721.
l 4.3.2 Nonradiological 4.3.2.1 Thermal and Ecological Effects of the Circulating ;
7 1
i Water System l
The plant's cooling water is drawn and subsequently discharged to I' Sherman Pond, which is one of a serie= of reservoirs in the controlled 1
5784R/4.70 l \
l Deerfield River Hydroelectric System. The pond hes a surface area of 218 acres and is contained on its southern end by an earth dam with a 7 MW hydrogenerating station. The water flow and thermal conditions of the Deerfield River are determined by natural and artificial factors. Natural conditions of precipitation and groundwater seepage determine the amount of runoff, however, the operation of storage and hydrogenerating reservoirs determines the day-to-day flow of the river. One effect of these reservoirs I i
is that a more constant river flow is maintained than would be the case if no reservoirs were available. The plant, therefore, has a stable source of cooling water even during low and high runoff periods.
The plant uses a once-through, open-cycle system for condenser cooling. Suction is taken from 80 feet below the surface of Sherman Pond and then passes through the remaining portion of the Circulating Water System (CWS). In passing through the condenser, the ci2culating water is heated about 20 F above its ambient temperature. Flow is then conveyed through discharge pipes to the discharge weir, where the heated circulating water is returned to the surface of Sherman Pond adjacent to Sherman Hydrostation's intake. A more detailed description of the CWS is available in Section 221 of the FSAR. 'l During periods of routine operation by the Sherman Dam Hydrostation, l the plant's thermal discharge is immediately drawn into the hydrostation intake and cooled as it mixes with the deeper pond water. Consequently, the discharge surface plume is restricted to the eastern side of Sherman deservoir and occupies about 1 2 acres, or less than one percent of the reservoir's surface area. When there is no hydroelectric operation, the discharge' plume gradually mixes and spreads across Sherman Reservoir. The surface area of the plume depends on the number of factors, including time, meteorology, river flow, season, etc. It is, however, normally restricted to the southern half of the reservoir as a buoyant surface plume.
The potential environmental effects of the CWS include: (1) those of the thermal plume created by the heated water discharged to the pond, (2) '
impingement of fish at the cooling water intake screens, and (3) entrainment of organisms (phytoplankton and zooplankton) as the cooling water passes through the condenser. These effects have been assessed by Yankee in 1974 and 5784R/4.70 4
L L ,
[ 1975, the Massachusetts Division of Water Pollution Control.(MDWPC) in 1975,
~
and:the Massachusetts Division of~ Fisheries and Wildlife (MDFW)~in 1977'(see; 1 i
References 26 through 30).
.]
p :j l
.Res'ults of'these-assessments were used to support Yankee's application I for'a' National Pollution Discharge Elimination System (NPDES) Permit and for a )
. variance under Section 316 (a) and (b) of the Clean Water Act. Tne ',
! Environmental Protection Agency (EPA) and'the MDWPC. jointly. issued a discharge
. permit in 1974 which has'been modified and reissued several-times since.
including the Section 316 (a) and (b) variance.. The permit was most recently b reissued.in 1983 (Reference,31). The permit requires that the cooling' water.
temperature be monitored and controlled to assutJ that the temperature at no- l time exceeds a 23.5 F rise over that'of the receiving water and does not
- l. exceed a maximum of 88 F. The EPA and MDWPC based'their decision to issue
- the permit on the following conclusions:
4 t -
- 1. The as-built' design of the cooling water intake and discharge structures reflect the best technology available for minimizing:
(a) fish impingement on the intake screens and (b) entrainment of organisms in the cooling water passing through the condenser.
Also, the locat3on'and physical characteristics of'the thermal plume are such that it should not interfere with the normal migratory pathways of the indigenous population of organisms
. inhabiting the pond (see discussion in Section 4.2.1.4). !
- 2. The plant has operated since 1960 without any observable impact to i
fish due to thermal effects. Therefore, the existing thermal' limits should continue to ensure the protection and propagation of a balanced indigenous community of fish,' wildlife, and other organisms living within the pond and adjacent waters.
- 3. Accordingly, the combination of the as-built design of the Circulating Water System and thermal limits imposed by the NPDES i permit assure satisfaction of: (a) the technology requirements of.
the Clean Water Act, including the best available ty knology (BAT) economically achievable requirements for toxic pollutents and the
'best conventional pollution control technology (BCT) requirements 5784R/4.70 )
I
i 1
1 y
p a far conventional pollutants, and (b) the Massachusetts Water.
Ouality. Standards.
It is expect'ed-that the basis for the EPA /MDWPC decision will remain
' valid throughout the' present 11cen:,e . term as well as the proposed. amendment
' term.. In short, the environmental effect of the CWS is, and will remain, stable given the controlled Deerfield River and the fixed design of the CWS. -
Yankee will continue to monitor the effects of>the CWS under the Nonradiological Surveillance Program discussed in Section 4.2.2. Likewise,-
'the EPA and MDWPC will periodically re-evaluate their conclusions, since the l 4
NPDES permit requires renewal every five years. 1
)
4.4 Exposure From Releases During Postulated Accidents i s -
The off-site exposure from releases due to postulated accidents has
_+
been previously evaluated in the plants FHSR and more recently in the FSAR with' acceptable results when compared to the criteria defined in 10CFR100.
This type of evaluation is'a function of four parameters: (1) the types of l accidents postulated, (2) the radioactivity release calculated for each accident, (3) the assumed meteorological conditions, and (4) and population distribution versus distance from the-plant. On the basis of the safety assessment.in Section 3 it can be concluded that neither the types of accidents or the calculated radioactivity releases will change through the proposed amendment term.- Furthermore, the site's meteorology.as defined in the FSAR, is essentially a constant and consideration herein is therefore unwarranted. Thus, the one parameter which is dependent on the proposed license amendment is the population size and distribution.
The population size and distribution in the vicinity of the plant has been reviewed several times since the construction permit was issued: FHSR in 1959, FSAR in 1985, and in 1986, a special projection study for the years 1980 through the proposed amendment term (See Reference 32). Table 14 presents a summary of the population. size and distributions stated in these studies with
[ l specific delineation of the' exclusion area, low population zone (LPZ) and
- h. nearest population center. These three parameters are defined by 10CFR100 and ;
remain the same as those defined in the FSAR. For perspective, the exclusion
~
area and LPZ boundaries are outlined below:
5784R/4470
o Exclusion Area: The exclusion area boundary is defined by a
.3,100-foot radius centered on the plant, with the exception of a small segment in the southern sector where the minimum distance is approximately 2,700 feet. All the land in the exclusion area is owned by either Yankee Atomic Electric Company or New England !
Pcwer Company, with the exception of a small parcel, which is owned by a specialty paper company. Written permission has been obtained from the paper mill to have that portion of their land (which is used as a disposal area) which is within the 3,100-foot l radius, under Yankee control in the event of an emergency.
o LPZ: An S-shaped area, approximately two miles wide centered on the Deerfield River, extending two miles upstream, and six miles downstream from the plant.
As presented in Table 14, the population within an aggregate ten-mile radius of the plant is projected through the year 2000 to remain below the level cited in the plant's FHSR, although the LPZ and other isolated areas would incur very minor growth relative to 1cvels cited in the FSAR. A projection for populations between 10 and 50 miles from the plant indicates an increase by about five percent when using the FSAR population as a benchmark.
The " nearest population center", which is identified in the FSAR as approximately 20 miles away, is projected to have a relatively stable population through the year 2000, but will still remain the population center as defined in 10CFR100. More importantly, projected populations during the proposed amendment term are shown as undergoing only negligible change for all areas within 50 miles of the plant. Finally, there is no expectation of significant land use changes during the amendment term that would affect off-site dose calculations. l For comparative purposes, Table 15 presents information from a 1979 NRC J study (Reference 33) which compares the pro.jected year 2000 population distributions of Yankee with other nuclear power plant sites that have received approval by the NRC to change their license expiration date to recover their construction period. The comparison indicates that the Yankee site represents 64, 45, and 55 percent of the average population distribution j l within 1, 10, and 50 miles, respectively for the proposed license expiration l
5784R/4.70 L_____________.__
date. Similarly, in comparison to the single highest population density site (Indian Point), Table 15 also shows that the Yankee site population distribution represents only 15, 12, and 10 percent of the total projected 2000 population within 1, 10, and 50 miles, respectively.
It is clear that none of the projected changes in population distribution between 1980 through 1997, and 1997 through 2000 will significantly impact any accident analysis previously calculated.
Furthermore, the current exclusion area boundary, low population zone and nearest population center distance are likely to be unchanged through the amendment term from those originally and currently used by the Yankee plant.
A comparison with other plants that have already been granted a similar amendment shows that Yankee will continue to be representative of a low population distribution through the year 2000. Accordingly, the proposed license amendment will not significantly impact previous conclusions on the potential environmental effect of off-site releases from postulated accidents.
4.5 Summary and Conclusions The environmental effects of Yankee's continued operation through the proposed amendment term have been assessed against four criteria which were established (See Section 4.1) based on the applications for a similar amendment that other licensee's have submitted and subsequently gained NRC approval. The assessments have shown that the environmental effects of Yankee's operation through the proposed amendment term are expected to remain well within the limits set forth by the four criteria. Also, as expected, the plant's environmental effects appear to be independent of chronological age.
Yankee's systems and programs for environmental monitoring and control show that Yankee has established, maintained, and when appropriate upgraded comprehensive environmental monitoring and control programs and systems and meets applicable regulatory criteria. In addition, the plant and other regulatory agencies have performed several surveillance type studies subsequent to the plant start-up, which support the absence of environmental impact as a result of plant operation. )
l 1
l 5784R/4.70
1 1
The' assessment of environmental impacts during normal plant operation was divided between radiological and nonradiological areas, with the following specific conclusions by topic:
o Occupational radiation exposure levels have been stable and at least 50 percent less than those from an industry average PWR.
The ALARA program and continued excellence in refueling and outage management are expected to at least maintain these low occupational exposure levels through the proposed amendment term.
o Off-site radiation exposure from liquid and gaseous, effluents have been less than 5 percent of the applicable Appendix I criteria and are expected to remain at similar levels during future plant operation.
o Solid waste generation has historically been among the lowest in the nuclear industry and has followed the industry-wide trend of reduced volume. Solid weste generated during the amendment period l
is projected to be no more than that generated in recent years of
. operation.
o Uranium fuel cycle impact will be trivial as a result of l operations during the proposed amendment term.
l I
l o Spent fuel storage capacity in the existing facilities is adequate through the proposed amendment term, to include full core l discharge.
l
- o Thermal and ecological effects of the CWS are expected to continue to meet all NPDES permit requirements through future years of I
operation. In cooperation with the EPA and MDWPC, Yankee will continue to monitor and evaluate the impact of the CWS on Sherman Pond.
o Off-site exposures from releases during postulated accidents were previously evaluated in the plant's FSAR. The only parameter used in these analyses which could change during future plant operation 5784R/4.70
g *.I
!is the population' distribution. The population.near the. plant is 1
'relatively low >in comparison with the1 requirements.of 10CFR100 and i j
.is projected:to remain virtually unchanged during. future plant j
- operation. Moreover, the effect of-the projected changes from {
1
, 1997-to 2000 would'beinegligible and also, have no effect on the *
-relationship between' the plant and the population center distance
}L as'definedLin the FSAR.
Based upon these' analyses, it is Yankee Atomic El,ectric Company's-conclusion-that..there are no significant radiological or nonradiological impacts associated with-the proposed action. Issuance.by the NRC of the .
-4 proposed license amendment will have no significant impact on the quality;of
- the' human environment.-
i: .i s
u i
l s
5784R/4.70 ]
4 REFERENCES-i i
- 1. Interim Facility License No. DPR-3, Yankee Atomic Electric Company Docket No. 50-29, July 1960.
'2. "New England Power Pool 1986 Annual Report," NEP00L Executive Committee, February 20, 1987..
- 3. " National Energy Board Reasons for Decision in the Matter of an Application Under the National Energy Board Act of Hydro-Quebec for i Exports to the New England Utilities," EH-1-87, May 1987.
- 4. " Engineering, Cost, and Performance Data for Electric Generating Technologies," Prepared for New England Power Planning (NEPLAN) Staff by Burns and Roe Company, March 1987.
- 5. "New Stationary Source Performance Standards for Electric Utility Steam Generating Units," Federal Register, June 1979.
- 6. " Nuclear Power Issues and Choices," Report of the Nuclear Energy Policy Study Group, Sponsored by Ford Foundation, Administered by MITRE Corporation, 1977. l
- 7. " Final Safety Analysis Report," Yankee Nuclear Power Station, Rowe, Massachusetts, 1985.
- 8. " Integrated Plant Safety Assessment," Systematic Evaluation Program, Yankee Atomic Electric Company Docket No. 50-29, USNRC Final Report NUREG-0825, June 1983.
- 9. NRC Safety Evaluation Reports on " Integrated Plant Safety Assessment Report (IPSAR) Section 4.9, Effects of Pipe Breaks on Structures, Components, and Systems Inside Containment - Yankee Nuclear Power Station,".May 3, 1984 and July 16, 1987.
- 10. NRC Safety Evaluation Report on " Integrated Plant bafety Assessment Report'(IPSAR) Section 4.21.1, Sump Water Chemistry - Yankee Nuclear Power Station," November 19, 1984.
- 11. NRC Safety Evaluation Report on "IPSAR Sections 4.28.1 Battery Current Charge / Discharge and Fuse Open Alarm and 4.29.2 Low Voltage Penetrations - Yankee Nuclear Power Station " September 19, 1985.
- 12. NRC Safety Evaluation Report on "NUREG-0825. Section 4.22, Containment !
Isolation - Yankee Nuclear Power Station," August 28, 1986. I L 13. NRC Safety Evaluation Report on " Yankee Nuclear Power Station -
Ultimate Load Bearing Capacity," June 4, 1987.
- 14. NRC Safety Evaluation Report on "NUREG-0825, Section 4.11, Seismic Design Considerations - Yankee Nuclear Power Station," July 16, 1987.
5784R/4.70 l
1 J
I 15. NRC Safety Evaluation Report on " Yankee' Nuclear Power ~ Station -
, Integrated Plant' Safety Assessment Report-(IPSAR) NUREG-0825,1 .
lSection.4.5', Wind and Tornado. Loading'and'Section 4.8, Tornado Missiles,":May 13, 1987.-
- 16. ;NRC' Safety Evaluation Report on "NUREG'-0825 Section 4.12. Design Codesi Design Criteria,1and Load Combinations - Yankee Nuclear Power ~;
Station," July 16. 1987.
- 17. " Yankee Nuclear Power Station Probabilistic Safety Study," Yankee.
Atomic Electric Company, December 1982.
- 18. " Change to the Administrative Control Section of the Technical-
' Specifications to Conform with the USNRC's Standardized Technical
. Specifications," Amendment:No. 27.to License DPR-3, Yankee Atomic Electric Company Docket No. 50-29, July 1976. ;
- 19. "In-Service Inspection Program Plan for Yankee Nuclear Power Station,"
Yankee Atomic. Electric Company, October 1981.
20.' " Environmental-Qualification Program'for Yankee Nuclear. Power Station,"
Yankee Atomic Electric Company Report.YAEC-1227, Revision 2, July 1986.
- 21. " Plant Procedures Manual," Yankee Nuclear Fower Station,' Yankee Atomic Electric Company Controlled Document,1Rowe, Massachusetts.
- 22. 'NRC Safety. Evaluation Report on " Fracture Toughness ~ Requirements for Protection Against Pressurized Thermal Shock Events," March 10, 1987.
- 23. NRC.Eafety Evaluation Reports on " Yankee Nuclear Power Station - Spent Fuel Pool Storage Capacity," dated November 23, 1982 and April 5, 1984.
1
- 24. " Occupational Radiation Exposure of Commercial Nuclear Power Reactors and Other Facilities," USNRC Report, Published Annually, NU. REG-0713, 1969-1984.
25.- " Supplemental Information for the Purposes of Evaluation of 10CFR, Part 50, Appendix I," Yankee Atomic Electric Company, submitted to the NRC on June 2, 1976, followed by two amendments submitted on August 31, 1976'and October 1976. I
- 26. ' " Biological and Thermal Conditions of the Deerfield River," Yankee Atomic Electric Company, YAEC-1069, 1974.
- 27. " Environmental' Data for Harriman Reservoir, Sherman Reservoir, and the Deerfield River," Report by Aquatec, Inc., Prepared for Yankee Atomic Electric Company, 1974.
28.? " Ecological Survey of Sherman Reservoir and Adjacent Waters,"
January-December 1975." Report by Aquatec, Inc., prepared for Yankee Atomic Electric Company, 1975.
- 29. " Study of the Effect of Heated Discharge on the Ecology of the Deerfield River," House No. 5369, Massachusetts Division of Water Pollution Control, 1974.
5784R/4.70
]
l
- 30. " Bear Swamp Pumped Storage Hydroelectric Project Fishery Study," Report by J. Forst and W..Este to Massachusetts Division of Fisheries and
. Wildlife, 1977.
31.< National Pollutant ~ Discharge Elimination Permit No. MA0004367, U.S.
Environmental Protection Agency, Massachusetts Division of Water Pollution Control, 1983.
- 32. " Demographic Update of the Area Surrounding the Yankee Nuclear' Power Station." Report by HMM Associates,. Incorporated, Prepared for YAEC, December 1986.
- 33. - " Demographic Statistics Pertaining to Nuclear Power Reactor Sites,"
NUREG-0348, USNRC, October 1979.
t 5784R/4.70
'l TABLE 1
{
Characteristics of Assumed Fluidized Bed Coal Plant Parameter Value Plant Size (no units x MWe per unit) 2 x 200 ,
Capital Cost ($/kW) 1,830 Fixed O&M Cost ($/kW-yr) 27.5 Variable O&M Cost ($/MW-hr) 6.4 Fuel Cost ($/million Btu) 1.6 Net Heat Rate (Btu /kWh) 9,450 Average Capacity Factor (percent) 74 NOTES
- 1. All parameter values, except fuel cost and capacity factor, are from Reference 4. Fuel cost is based on actual 1987 costs to a New England area utility. Capacity factor is assumed equal to Yankee's lifetime average.
- 2. All dollar values are in 1987 dollars. Data from Reference 4 (which was in 1986 dollars) is' escalated by six percent.
- 3. Capital cost includes 20 percent ($305/kK} for AFUDC. This percentage is based on an assumed interest rate of ten percent per year and an average full cost construction lead t-ime of two years. The latter time is derived from estimates provided in Reference 4 for capital cost cash flow during construction.
- 4. Fixed O&M includes payroll, G&A, insurance, and some M&S.
- 5. Variable O&M includes fuel conditioning and waste disposal M&S.
l 1
l 4
1 i
i l
)
5784R/4.70 1
J
_____..__m _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - -- -
TABLE 2 Comparison of 1998-2000 Power Costs for Yankee Versus Assumed Coal-Fired Replacement Capacity Cost (d/kWh)
Cost Component Yankee AFB Coal Plant O&M 3.1 1.0 Fuel 1.0 1.5 ;
1 Capital J 0 5.5 Total 4.9 8.0 NOTES
.l. All costs are in 1987 dollars.
- 2. An annual generation of 1.13 x 109 kWh (175,000 kW x 0.74 x 8,760 hours0.0088 days <br />0.211 hours <br />0.00126 weeks <br />2.8918e-4 months <br />) is' assumed for both plants.
- 3. Yankee.0&M includes payroll, engineering, G&A, and M&S. AFBC plant O&M includes payroll, insurance, G&A, M&S, and fuel conditioning.
- 4. Yankee capital cost includes depreciation, interest, taxes, and return on equity. It does not include decommissioning, since no collections will be required after 1997.
- 5. AFBC plant capital cost includes depreciation, interest, taxes, and J return on equity.
]
1 i
l l
i l
l l
l I
I l
5784R/4.70 I l
i t __--
TABLE 3 Licensing Basis Documents Final Safety Analysis Report Provides a description of:
(1) The design of plant safety-related systems and components, and (2) Analyses which show that these !
systems and components will function to prevent release of harmful amounts of radioactivity to the environment during normal operation, anticipated transients, and postulated accidents.
Systematic Evaluation Program (SEP) Provides the results of NRC asses: ment Documents under the Systematic Evaluation Program j (SEP). It compares as-built design of plant systems and components to current regulatory requirements.
Probabilistic Safety Study Provides an assessment of public safety-risk due to anticipated transients and postulated accidents. Submitted to the NRC as a partial basis for evaluations made under the SEP.
Technical Specifications Provides specification of:
(1) Plant safety limits and limiting safety system settings, (2) Limiting conditions for operation, and (3) Equipment surveillance requirements.
Surveillance and Maintenance Provides a description of programs Program Documents established to assure continued functional capability of safety-related components throughout the plant's ,
service life. Includes the: i I
(1) In-Service Inspection Program, (2) Environmental Qualification Program, and (
(3) Preventive Maintenance Program.
]
l l
5784R/4.70
__.__.m____
e- . ,
. q. s t
l t- t TABLE'4 / O 4c ,
J
' > gEvolution of ISI Program ~
- 1971 Beganfirst-(accelerated)inspecti$nLinterval. Inspections ]H
, doneJin accordance with the 1970" Edition'of'ASME' 't'i Section XI. Included only.Classji components.
1975 Completed first inspection interval'. . Began lsecond
. inspection interval. . ' Inspections done in accordance with>-
~
1972 Edition of ASME Section XI. ~' $till only included ..
Class.1 components. "y, 0
- a. '1978- .AddedLClass12 and:3 components to the program. ] 1 1
- 1981'- . Completed second period of second inspection interval. '
. Submitted'10-year ISI Program. Plan. This plan covers third period of-second. inspection interval and all of.the third:
inspection / interval. Inspections to be done'in accordance' J.g
-withL1977 Edition, Summer 1978' Addenda of ASME'Section XI. L' r
- f 1984- Completed pressure vessel inspections.
.v ,
lf 4
f
'1 4 w
('
I l
- L / i tl 1
. 5784R/4.70 -i i
I
)
~
T
};
- A TABLE 5 Major Safety Improvements L , Since Initial Construction i-'
Modification Year Added Steam-Driven Emergency Feedwater Pump and 1963
-Third Safety Injection Pump Added Three Emergency Diesel Generators With Three Train 1970 Emergency Power l
Ih Added Three Train High and Low Pressure Pumps and 1972
- g Accumulator to Enhance Safety Injection System Ils ) Added' Post-LOCA Recirculation System 1972 Added Filtration System for Contaminated Areas Exhaust 1975 Added Primary Vent Stack Monitoring and Redundant 1979 Two-Train Containment Isolation Actuation System Added Two Electric Motor-Driven Feedwater Pumps 1980 Automated Main Steam Nonreturn Valves 1981 Added Diesel Fire Pump and Water Storage Tank 1981 Added a Second Vital Instrumentation Power Source 19F2 Added a Custom Designed Safety Parameter Display 1982 System (SPDS) to the Main Control Room Added Post-Incident Cooling System (PICS) to Perform 1982 Fully Remote Plant Cooldown From the Main Control Room Added Dedicated Shutdown System for Fire Protection and 1985 Severe External Events Replaced Reactor Protection and Feedwater Control 1980-Systems With Updated Systems 1985 1
-)
5784R/4.70 1
L_ _. _ _ . _ _ . _ _ . I
TABLE 6
-NRC SALP Ratings Ratings 5/1/82 9/1/83 2/1/85 Functional to to to Area 8/31/83 1/31/85 10/6/86 Plant Operations 1 1 1 Radiological Controls 2 2 1 Maintenance and Modifications 1 1
'{
1 Surveillance 1 1 1 Fire Protection and Housekeeping 1 1 1 Emergency Preparedness 1 1 2 Security and Safeguards 2 2 2
. Refueling and Outage Management 1 1 1 l Assurance of Quality 2 2 1
-l Training and Qualification Effectiveness - -
2 !
Licensing Activities 1 1 1 l I
)
i 1
5784R/4.70
g- , .
=
+ , ; -: ,, ,
4 t j c ;
.C , C '
,,.. *t ..,
, ;j,; 6'
, ' BBLC ~
6(i 1 .
fi, , ,;,j 'r ,..
Component I Clansificatbans 1- 1 Mec'tanical Components Includes p$ essure vessels, ; tanks, heat
, .cxc'mngers , piping, pumps ' valves , and 1
other similar components.
Electrical Components , Includes both electric power ano
'^( , ,
, instrumentation components. Electric f , power components include: power i source, brerker, control circuit, cable, relaying, and operating device (e.g., motor, solenoid, heater, i; relay). Instrumentation components
, i include: power supply, sensors,
, 'l relays, wiring, and final operating
, device (e.g., solenoid, relay).
Structural Components' Includes the' Vapor Container, component l j
~
support structures, buildings, and foundations.
t
/1 r
a .,
i
)
T ,
i e'
I r
,L-9 1
, i 1
.i ,
.t t
6 i
g 5784R/4.70 >'
s y f d' e v',
f t'
i k
TABLE 8 Codes and Standards Used fer Design of Main Coolant Pressure Boundary Reactor Vessel, Pressurizer, Section VIII of the ASME Boiler and Steam Generator _a' Pump Casings Pressure Vessel Codo, 1956 Edition.
Piping ASA B31.1, 1955 Edition, Code for Pressure Piping.
Valve Bodies ASA B16.5, 1957 Edition, Standard for Steel Pipe Flanges and Flanged Fittings.
l l
l l
1 1
1 i
)
l 5784R/4.70 l I l - .- - - - - - - - - - - _ . . _ _ _ _ _
TABLE 9 Yankee and Industry Average Annual Occupational Exposure Per Five-Year Period (As Reported in NUREG-0713)
Average Annual Occupational Exposure Five-Year Period (Person-Rem)
Yankee Average PWR 1974 - 1978 203 493 1975 - 1979 188 452 1976 - 1980 207 527 1977 - 1981 256 557 1978 - 1982 279 561 1979 - 1983 240* Not Published in NUREG-0713 1980 - 1984 281 569 l
- Calculated by Yankee Atomic Electric Company i
i 5784R/4.70 i
l l
l u________ l
TABLE 10 Yankee Historical Annual Occupational Exposures (as published in NUREG-0713)
Cumulative Exposure Year Person-Rem Outage Status 1969 215 Refueling
- 1970 255 Refueling 1971 90 None 1972 255 Refueling and Maintenance **
1973 99 Maintenance 1974 205 Refueling 1975 116 Refueling 1976- 59 None 1977 356 Refueling 1978 282 Refueling 1979 127 Inspection 1980 213 Turbine Failure 1981 302 Refueling 1982 474 Refueling and Maintenance 1983 68 None 1984 348 Refueling
- " Refueling" means normal refueling.and maintenance activities.
-** " Refueling and Maintenance" means normal refueling activities with extraordinary maintenance (e.g., reactor internal maintenance).
5784R/4.70
L TABLE 11 Summary of Off-Site Radiation Exposure Appendix I Criteria Versus Actual' Yankee Performance Actual Performance Appendix I from Yankee Parameter Criteria Operation *
' Liquid Max. Organ _
110.0 mrem /yr 0.053 mrem /yr Max. Whole Body 130 mrem /yr 0.04 mrem /yr Noble Gases
-Beta 120 mrad /yr 0.97 mrad /yr Gamma .1 10 mrad /yr 0.33 mrad /yr Iodine and (15 mrem /yr 0.60 mrem /yr Particulate
- 1
- Based on a three-year average from 1984 to 1986. ,
l 5784R/4.70 j
i i-TABLE 12 J J
Annual Three-Year Average Volume of Solid Waste Shipments '
From Yankee and Industry Average PWR Yankee Industry Average PWR 1 Period (cubic feet) (cubic feet) 1980-1982 8,180 19,325 1981-1983 7,958 17,832 1982-1984 6,770 15,855 1983-1985 6,664 14,125 i
1984-1986 6,028 11,205
- Source: Based upon industry average reported by INPO from " Nuclear Power Plant Performance Indicators."
5784R/4.70 l
1
__ - _ _ - _ _ i
LTABLE 13 Summary'of Annual Volume of Solid Waste Shipments From Yankee 1980 Through 1986 Volume Year (cubic feet) Outage Status 1980 6,265 Turbine Failure
'1981 10,881 Refueling 1982 '7,396 Refueling _and Maintenance
.1983 5,596 None 1984 7,048 Refueling 1985 7,078 Refueling 1986 3,959 None 1
l l
l 5784R/4.70
l' i
l l
TABLE 14 Summary of Population Distributions for the Yankee Plant
)
I I
Projections From Special 1986 Study Change Change
' Distance From FHSR(1 ) FSAR(1) Since Since .j Yankee (miles) (1950 Census) (1980 Census) 1997 1980 2000 1997 Exclusion Area (2) N/A(3) 1 1 0% 1 0%
Low Population N/A 260 277(4) +6.5% 280 +1% ,
Zone
{
1 Nearest Population N/A- 51,974 48,162 Center 0-1 174 61 Not Reported 0-5 2,036 1,499 1,688 +11.2% 1,724 +2.1%
5-10 26,946 22,191 21,432 -4.0% 21,254 -0.8%
0-10 28,982 23,690 23,120 -2.5% 22,978 -0.6%
10-20 15,311 90,300 - - - - - Not Reported - - - - -
0-20 104,293 113,990 - - - - - Not Reported - - - - -
10-50 Not Reported 1,434,470 1,509,791 +5.0% 1,524,442 +1.0%
0-50 Not Reported 1,458,160 1,532,911 +4.9% 1,547,420 +0.9%
1 Actual population data, not an estimate 2 The one occupant in the exclusion zone is a utility employee living in a utility-owned home.
3 N/A = Not applicable at the time that the FHSR was published.
4- Estimated by YAEC.
5784R/4.70 L__-___-__.___. __
_ - _ . _ = _ _ _ _ _ _ _ _ - -
7 __.
1 TABLE 15 l Comparison of-Yankee Population Distribution for the Year 2000 With Other United States Nuclear Power Plant Sites *
(
Population Distribution '
Site 0-1 mile 0-10 miles 0-50 miles Calvert Cliffs 1 and 2 320 29,000 3,900,000 Point Beach 1 and 2 72 33,000 880,000 Beaver Valley 1 800 130,000 5,000,000 McGuire 1 and 2 38 60,000 2,100,000 Indian Point 2 1,000 290,000 24,000,000
-Oconee'1, 2, and 3 0. 53,000 1,000,000 Crystal River 3 0 20,000 290,000 Prairie Island 1 and 2 130 29,000 3,000,000 ,
North Anna 1 and 2 0 14,000 1,500,000 !
Surry 1 and 2 5 98,000 2,300,000 Average 236 75,600 4.397,000 Yankee 150 34,000 2,400,000 Percent of Average Population Distribution. 64% 45% 55%
Percent of Highest Density Site (Indian Point) 15% 12% 10%
I, I
l l
- Data taken from 1979 NRC Study reported in Reference 27 l.
l L_________. !
p
.)
l l
l l
l FIGURE 1
,1661-1986 Annual Capacity Factors 100 , , , , , , , , , , , . , , , , , , , ,,,,,
90 -
1-
'80 -- -
70 -
. CAPACITY 60 - -
FACTOP. 1
(%) 50 -
Average Capacity k -
Factor - 74.2 % l 40 - (1) - l 30 - -
)
20 - - !
(2) ;
10 - - H i
0 ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' '* I 1961 1966- 1971 1976 1981 1986 1 1
1
- )
i NOTES: 1 1
(1) Redesign and replacement of the lower shroud tube package.
(2) Low pressure turbine rotor refurbishrnent. ;
l l
( 70 i I
m__ _ _ . . _ _ _ _ _ _ . _
j
1
. )
FIGilRE 2 Cycles 1-17 Refuelino Outaae Lenaths .l l
i 20 -
18,- _
16 Average Cycle _
Length - 10.1 weeks -
-OUTAGE '4 -
LFNGTH 12 -
IN -
10 w - - - -- w WEEKS -
8 -
6 -
4 -
2 _
0 ' ' ' ' ' ' ' ' ' ' ' ' ' ' '
1 2 3 4 5 6 7 8 9 10 11 12 13- 14 15 16 17 CYCLE 71
- x I
l 1
FIGURE 3 1961-1986 Annual Number of PlantTrips s
13 . . . .. . . . , , . . .
,...., , , , . 4
.l 12 - - i 11 - -
J l
10 . _
NUMBER OF 9 -
TRIPS 8 s
r 7- -
6- Average 4.3 _
5- -
4 '-
3-2-
1 - -
0 ' ' ' ' ' ' ' ' ' ' ' ' ' 'i ' ' ' i 1961 1966 1971 1976 1981 1986 YEAR l
l 72 l
l I
I C_:_____.___...____. ._ _ .h