ML19351A187
| ML19351A187 | |
| Person / Time | |
|---|---|
| Issue date: | 02/27/1981 |
| From: | NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| Shared Package | |
| ML19351A177 | List: |
| References | |
| REF-10CFR9.7 ENVS-810227, NUDOCS 8106260211 | |
| Download: ML19351A187 (14) | |
Text
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.s ORAFT ENVIRONMENTAL. IMPACT ASSESSMENT FOR PROPOSED RULEMAKING TO AMEND 10 CFR PART 50 CONCERNING ANTICIPATED TRANSIENTS WITHOUT SCRAM (ATWS) EVENTS 1.
Introduction The NRC staff has recently completed a review and evaluation of information that har been developed over the past 10 years on Anticipated Transients Without Scram (ATWS) events and the manner in which' they should be considered in the design and safety evaluation of nuclear power plants.* The result of these' i
l, efforts indicate the potential extent and probability of serious consequences from such events and the NRC is proposing to amend 10 CFR Part 50, Section 50.49 (SECY-80-409).
Since this amendment to the regulations. governing the licensing of production and utilization facilities is substantive and may have a signif-
.icant impact on the human environment, an environmental assessment has been L
prepared to determine whether an environmental impact statement should be developed fcr the propdsed rulemaking.
2.
Need for the Amendment The significance of ATWS for reactor safety is that some ATWS events could result in melting of the reactor fuel and the release of a large amount of radio-active fission products.
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. The NRC has concluded that the probability of ATWS events occurring over l
the lifetime of nuclear power plants and the potential magnitude of consequences arising from such events, should they occur, are sufficiently great to warrant l
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- Anticipated Transients Without Scram for Light Water Reactors, NUREG-0460, Vol. 1 through 4.
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l 8106260211 1
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the imposition of additional requirements to reduce the probability and mitigate the consequences of ATWS events.
3.
The Proposed Action The proposed rule would require each~ light-water-cooled nuclear power plant to be designed, constructed and operated such that the consequences of postu-lated anticipated transient without scram events calculated in accordance with.
an approved evaluation model conform to specified criteria for primary system pressure, fuel integrity, radiation release, containment and long-term shutdown cooling.
Although the. equipment necessary to comply with this requirement will vary from plant to plant, the equipment will generally consist of additional' separate i
and diverse electrical control circuitry to supplement the reactor trip system and some engineered safety systems.
In addition, some PWR plants may be required to install additional safety valve capacity and BWR plants may be required to install a higher capacity standby liquid control system.
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The NRC staff proposed alternatives representing different levels of safety.
The following table summarizes the requirements of these alternatives for the Pre-1984 and Post-1984 plants:
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TABLE,1 -
SUMMARY
.0F REQUIREMENTS
(* Indicates implicit requirements.)
Post-1984 Plants Pre-l'834 Plants VENDOR SPS SPS,
AMSAC B&W, CE AMSAC Cont Isol Cont Isol Analysis Analysis Instr
- a Instr Safety Valve
- Safety Valve
AMSAC Cont Isol Cont Isol Analy. tis Analysis Instr
GE AMSAC Cont Isol Cont Isol Analysis Analysis SD l
SD l
Logic
- Logic
- Instr
- i Instr
- SLCS-Auto
- Incr. Cap SLCS-Auto
- I Hi-Cap Nomenclature:
Diverse and independent ATWS mitigation acuation circuitry.
l ASMAC:
from the reactor protection system to actuate:
h PWR's - Turbine trip, auxiliary feedwater BWR's - High pressure coolant infection (HPCI),
1.
Standby Liquid Control Systems (SLCS), Recircula-2.
tion Pump Trip (RPT).
Analysis with acceptable evaluation models of performance I
Analysis:
following ATWS events.
Containment isolation initiated by early detection of fuel Cont Isol:
failures.
Instrumentation necessary for shutdown that can withstand Instr:
ATWS conditions.
Logic of control circuits to reduce vessel isolation events Logic:
and runback feedwater.
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Add *'.fonal safety valve relief capacity.
Snfe'ty. Valves:.
Supplementary protection system that is d SPS:
protection system --
tm
.B&W' - 8055, a diverse four-channel backup scram sys e tection N?S, a diverse. four-channel supplementary pro CE -
system W
l redundant scram air header exhaust valves 1
GE -
Scram discharge volume for GE control rods that is less susceptible to common mode failure SD:
Automatically initiated, Standby Liquid (neutron poison) i SLCS-Auto:
Control Syste:n--
Capability to simultaneously inject with both pumps Incr-Cap:
Single failure proof system, with approximately Hi-Cap:
400 gpm capacity l
e in stages The proposed rule provides for implementation of the requirem l
time and at in order to gain the greatest increase in safety in the shortest the least cost.
i k sites Some existi,ng nuclear power plants are considered to be at h l
d other factors.
owing to population density, meteorological conditions, an e
tion and Identification of these sites is a subject of another Commission a ld be subsequently con-any additional ATWS requirements for these units wou sidered.
Impact of the Proposed Action ized as 4.
The environmental impact of the ATWS proposed rule can be c f
follows:
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The effect on the natural environment and the people in it due to 1) risk of exposure to radioactivity.
The effect of replacement power.
11) by The environmental impact on the natural environment can be e er plant considering the relative radiological influence that the nuclear pow The latter has with respect to the background radioactivity already present.
c s other than includes natural radiation background and man-made exposure sour e from the nuclear power plant operation.
Some occupational exposure occurs in all operating Occupa'tional Exposure:
Additional exposures plants in the normal course of operation and maintenance.
ifications that can occur in operating plants during any installation of mod Additional occupational exposure would be expected in i might be required.
i The greatest occu-menting the ATWS modifications required by this rulemak ng.
judged to be that pational exposure due to the proposed ATWS requirements is l
the primary l
associated with the installation of new safety relief va ves on I
the staff l
To conservatively estimate the extent of such exposure, t st number of system.
selected an operating plant, that is estimated to need the grea e To establish the plant and additional valves to meet the ATVS requirements.
idered the nuclear l
the maximum number of relief valves needed, the staff cons i
i g of cycle steam supply system power / relieving fiow ratio, utilized beg nn n ture coeffi-(BOC) core conditions, and assumed'a conservative moderator t Based on these considerations, cient (MTC) value for the pre-1984 plants.
additional relief h
the Calvert Cliff Unit-1 plant was selected and two or t ree his i;ype of plant to valves are conservatively estimated to be required for t Technical staff from the Office of meet the proposed ATWS modifications.
t tion Inspection and Enforcement experienced in nuclear power pla 5-
l at the Calvert practice worked with the resident.t 3pector and plant personne k s Cliffs Unit-1 plant to estimate t i amount of radiation exposure that wor i ary system. The would receive in the weldig of three relief valves on the pr m cssumptions used for this s ttmate included:
The installation will be performed several days after cold shutdow (i)
(ii) -The lines will be purged prior to start of installation.
The size of the line will be two to three inches maximum.
(iii) d Manual shielded metal arc or tungsten inert gas welding will be us (iv)
Completed welds will be liquid penetrant examined.
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(v)
Pipe material is austenitic stainless steel.
(vi) The lines are assumed to emit 0.4 to 0.5 R/hr of radiation on c (vii)
(These values were provided by or 0.1 to 0.2 R/hr at a distance of 18 inches.
the Calvert Cliffs radiation protection officer.)
The welder will use some type of protective shielding during (viii) ill be The results of this estimate were that approximately 4 to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> w diation exposure to needed for the installation of each valve and that the ra Thus, for welding, the the welder. is estimated to be about one rem per valve. is conservatively three valves on the primary system, the occupational exposure estimated to be less than 10 man-rems.
difi-All other oc:upational 6xposures necessary to implement the AT l tive to the cations required by the proposed rule are judged to be less re a i
they are of occupational exposure from welding the valves; however, assum ng re is estimated to the same order of magnitude, the total occupational exposu be less than 100 man-rem per plant.
b of To estimate the impact of the occupational exposure, the total r rate wer; plants (25 BWR, 16 CE and B&W, 28 Westinghouse) licensed to op difications for each addressed taking into consideration the proposed A1VS mo 6
Based on this assessment, the estimated type of plants as listed in Table 1.
i occupational exposure for the 25 BWR's at 80 man-rem / plant is 2,000 man-rem; for the 16 CE and B&W plants.at 100 man-rem / plant is 1600 man-rem; and for the The total collec-28' Westinghouse plants at 40 man-res/ plant is 1120 man-rem.
-tive occupational whole-body dose is estimated to be 4720 man-rem.
l It should be recognized that t,1e occupational' exposure for these ATWS modifications are essentially a one-time exposure and that they would be only a small fraction of the total occupational exposure expected from normal opera-tion over the 40 year life time of a plant.
For example, the average exposure per reactor year for the six years (1973 through 1978) shown in Table 2 is I
This could about 500 man-rems or 20,000 man-rems over a plant life of 40 years.
be compared to the one time occupational exposure of 100 man-rems per reactor for the bounding type case of this assessment, which indicates that the ATWS modifications would f r:rease the total occupational exposure expected from the In addition, the 100 man-rr.m currently operating plants by about 1/2 of 1%.
occupational exposure is generally within the year-to year variation of ' t.e i
It is also well within the year-to year average exposure as shown in Tacle 2.
variation among individual plants which for the four years (1975 through 1978) varied from a low of 21 men-rem to a high of 3142 man-rem.*
Most authorities ** are in agreement that a reasonable, and probably con-l l
servative estimate of the statistical relationship between low levels of radi-I ation exposure to a 'arge number of people and the subsequent health effects per million man-rem is about 100 potential premature deaths due to cancer an about 220 genetic changes appearing over a span of five generations.
- NUREG-0692, " Final Environmental Statement Related to Steam Generator Rep at Survey Pewar Station, Unit No.1 Virginia Electric and Power Company,"
- Meetin'g on development of NRC position on health risk estimators - Michael A July 1980.
Parsont; Radiological Health Standards, October 22, 1980.
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1AstE 2*
i SummRY Or menat Exposures f
Rfr0RIED BY NUCLEAR POWER FACitlIIES i
1973 1978.
Average No.
Average Wo.
Man-Rees Per Total No. of Workers Total Average l
Annual Dese of Man-Rees. of Workers Reacter Number of Reacters Number With Measurable Megawatt-Yrs.
(Rees/ Worker)
Per teacter Per Reactor _
Megawatt-Yr.
i Generatc<f _
Deses Yarr Type included
_ et Man-Rees 1.00 783 787 2.5 j
0.85 380.
445 1.3 1973 PWR 12 9,399 9.440 3,770 2
0.94 582 616 1.9 BWR 12 4,064 5.340 3,394 Total 24 13, % 3 14,780 7,164 0.68 331 485 1.0 l
1974 Nt 20 6,627 9,697 6,824 507 626 '
1.7 i
0.81 543
- 1. 3 -
14 7,095 8,769 4,059 0.74 404 BWR 13,722 18,466
- 10,883 0.76 318 419 8.7 Total 34 i
0.86 701 812 2.2 l
'975 PWR 26 8,268 10,684 11,983 0.82 475 579 1.2 BWR 18 12,611 14,607 5,786
)
lotal 44 20,879 25,491 17,769 0.79 460 586 1.0 0.71 549 776 1.5 l
1976 PWR 30 13,807 17,588 13,325 0.75 499 669 1.2 SWR 23 12,626 17,859 8,586 g
Total 53 26,433 35,447 21,911 0.65 396 614-0.8 -
1977 PWR 34 13,469 20,878 17,346 828 930 2.1 9
0.89 742 1.3 l
SWR 23 19.042 21,388 9,103 0.77 570 i
f Total 57 32,511 42,266 25,584 055 428 659 0.8 0.74 604 811 1.3 39 16,708 25,700 19,840 i
0.69 497 718 1.6 I
1978 PWR 15,096 20,278 11,774 l
Total i4 31,804 45,978 31,614 SWR
.t5 been in commercial aperation for at least ene year as of "The figures on this tabi's are based en the number of nuclear power reacters that hadIndian Point 1, although defuele December 31 of each of the years indicated.
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,a This-indicates that the total occupational dose estimate of-4720 man-rems for the ATWS modifications corresponds to a statistical risk of less than one premature fatal cancer, and of one genetic effect to the ensuing five generations. These risks are based on risk estimates derived in the BEIR report" from data for the population as a whole.
For a selected population such as is likely for the exposed workers involved in the ATWS modifications
' consisting of males in age range from 20 to 40, these risks would tend to be
.somewhat less.
These risks are incremental risks, risks in' addition to the normal risks of cancer and. genetic' effects we all face continuously.
For a population of 2,MO, these normal risks would statistically indicate 300-400 cancer deaths and 100-150 genecic effects (genetic effects are genetic diseases or malformations).
No early fatalities or sematic effects would occur for this type of exposure from ATWS modifications. The individual occupational' exposures involved in the ATVS modifications will be controlled and limited so as not to exceed the limits set forth in 10 CFR Part 20.
For the foregoing reasons, the staff concludes that the environmental effect due to occupational exposures from the proposedcATWS modificatiens represent an insignificant and societa11y acceptable impact.
Population Exposure:
There are two types of potential impacts on the population exposure due to the ATWS rulemaking requirements.
The first is the radiological impact due to the installation of ATVS modification on ncemal plant operations and the second is the impact related to ATVS events. ATWS modift-cations do not affect normal operations other than improving the reliability
- The Effects on Populations of Exposure to Low-Levels oh Ionizing Radiation (BEIR Report), National Academy of Science, November 1972, Reprinted July 1974.
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i of th scram system and thus the modifications have essentially no impact on
-populat' ion exposure for those conditions.
Since ATWS modifications reduce the overall risk from ATWS events, the modifications'are insignificant relative to further impacting population exposure.
Replacement Power:
Effects from replacement power may be an environmental impact where delay in startup or shutdown of operation is required to make ATWS modifications.
The staff's judgment is that ATWS modifications could likely be Jnade with little or no additional construction time or shutdown time, since immediate retrofit is not proposed.
For operating plants or near term operating plants, the staff believes that most of the ATWS modifications that require
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plant shutdown could be installed at the time of scheduled plant outages. This
~judgeinent is based on the following historical data.
During 1978, the 25 operating BWRs experienced an average of 12.89 weeks of outage time and the 40 operating PWKsan average of 13.01 weeks of outage of which more than 8 weeks were scheduled.
Refueling was the primary cause of scheduled outages at both CWRs and PWRs and the primary cause of forced outages at both BWRs and PWRs was equipment failure. Maintenance or testing also accounted for a large percentage of the scheduled outags time at both types of p~1 ants. Tables 3, 4, and 5 sum-marize the proximate cause of outages durir.g 1976, 1977, and 1978, respectively.
The NRC staff estimates that the ATVS modifications that require plant shutdown can be made in four to six weeks. A utility has estimated from four to eight weeks depending upon what kinds of unexpected problems might arise.
For either l
l-estimate, the historical record as summarized in the tables, indicates that the scheduled refueling outages (During 1978: 5.81 weeks / refueling for BWRs and 7.8 weeks / refueling for PWRs) provide ample time to implement these modifications 5
10-
4 TA8tf 3 t
P90XIMATE CAUSE of 00fAGES 00 RING 1976 Scheduled Outages forced Outages Totals Events I
Equipment Maintenance Esgulatory operattual Maintenance Regulatory Training &.
failure or Test Restrictions Error Other or Test Refueling Restrictions Licensing Ahinistrative Oti.or 6
1 328 I
No. of 195 3
2 38 6
58 14 5'
i events DWR 686 100 64174 4
4 llours of 12175 9869 2369 923 491 10098 18263 9280 -
outa0e No. of 354 1
1 51 10 99 24 1
7-1 1
550 -
events 183 w9 26 824 83968 i
PWR llours of 31711 1
120 2599 197 16911 '
31221 outage i
events 549 4
3 89 16 157 38 6
7 7.
2 878 i
Ho. of All Plants
% of 63
<1
<1 10 2
18 4
<1
<1
<1
<1 108 1
Total K0e hours 43886 9876 2489 3522 688 27009 49434 9463 169 632 924 148142 Total out-All Plants 3
E of 30 7
2 2
<1 18 33 6
<1
<1
<1 108 Total i
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TA8tE 4 i
PROMIM41E CAUSE OF OUIAGES DURING 1977 1
i forced Dutages.
Scheduled Outages Totals Events Equipment Maintenance Regulatory operational Maintenance Regulatory Training &
Failure or Test Restrictions Error Administrative other or Test Refueling Restrictions Licensing Administrative other 1
l No. of 205 3
4 31 3
15' 56 18 2
2 9
3 351 events j
BWas llours of 11,388 461 372 799 433 614 11,163 32.940 8,844 111 1,419 91 69,635-l outage 556 25 '
103 26 2
2 1
Ho. of 346 1
50 cvents 1
PWR5 72,540 j
llours of 18.157 378 714 422 17,455-33,717 1,622 31 44 eutage No. of l
svents 551 3
5 81 3
40 159 44 4
4
-10 3
907 All
]
Plants j
Percent of 61
<1
<1 9
<1 4
18 5,
<1
<1 1
<1 100 Total age hours 30,545 461 750 1.513 413 1,036 28,618 66,657 10.466 142 1,463 91 342,175 Total out-l All Parcent 21
<1
<1 1
<1 1
20 47 7
<1 1-
<1 108
. Plants lotal i
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TA8tE 5 PROMINATECAUSEhOUTAGESDURIMG1978 4
Forced Outages Scheduled Outages-Totals Events Equipment Maintenance Regulatory Operational Halntenance Regulatory Training &
Fallure or Test Restrictions Error Administrative Other or Test Refueling Restrictions Licensing Aeslaistrative 2
290 Ho. of 183 2
1 33 1
9
- 35 22 2
twents SWRs 229 54,174 llours of 16,959 160 540 637 27 231 4,499 21.479 9.413 outage 3
1 534 l
No. of 327 3
1 55 2~
13 99 30 avents j
PWRs 39 6
87,894 liours of th,082 270 5,137 1,084 28 258 15,694 39,495 autage No. of avents 510 5
2 88 3
22 134 52 2
3 3
824 All Plants Percent of 62
<1
<1 11
<1 3
16 6
<1
<1
<1 100 Total I
Total out-l age hours 42,841 430 5,677 1,721 55 489 20,193 60,975 9,413 39 235 142,068 All Plants
<1 14 43 7
<1' 41.
100 Percent of 30
<1 4
1
<1 Total 4
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and very little (if any) additional shutdown time would be required. Thus, the impact of needing alternate power sources because of shutdown due to ATWS modi-
'fications is an insignificant environmental consideration.
S.
summary The prcposed regulaton/ action has the effectiof further reducing reactor
, risk which will improve the' human health and safety environment relative to ATWS events.
The environmental impact assessment focused on the occupational exposures and the effects of replacement power requirements that would result from implementing the proposed ATWS modifications.
Based on this assessment, the proposed ATWS regulatory action do'es not require an environmental impact statement since its impact on the human envircnment is judged to be insignifi-cant and nonsubstantive.
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