ML20002E539
| ML20002E539 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 01/09/1981 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20002E535 | List: |
| References | |
| DD-81-01, DD-81-1, NUDOCS 8101290017 | |
| Download: ML20002E539 (28) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION 0D-81-1 HAROLD R. DENTON, DIRECTOR In the Matter of
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Duke Power Company
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Docket Nos.
50-413 (Catawba Nuclear Station,
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50-414 Units 1 and 2)
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DIRECTOR'S DECISION UNDER 10 CFR 2.206 By petition dated January 28, 1979, Jesse L. Riley, President, Carolina Environmental Study Group (CESG), petitioned the Commission to reopen the safety phases of the licensing proceedings for Duke Power Company's Catawa Nuclear Station and McGuire Nuclear Station.
On March 7, 1979, the Director of the Office of Nuclear Reactor Regulation l
advised CESG that its request to reopen the McGuire proceedings had been referred to the Atomic Safety and Licensing Board since the matter of issuance of an operating license for the McGuire facility was currently pending before that Board.
However, because the Catawba case was not currently pending before any Licensing or Appeal Board, the request to reopen the Catawba proceedings was to be treated as a request under 10 CFR S,2.206 of the 810 12900G
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Commission's regulations.
Notice of the consideration of the Catawba request under 10 CFR S 2.206 was published in the Federal Register. 44 Fed Reg 14654 (March 13, 1979) 1 Before examining the specific issues raised by CESG in this petition, it is appropriate to review the criteria for evaluation of requests for action under 10 CFR 2.206.
A petition shall specify the action requested and set forth the facts that constitute the basis for the request. 8-The factual basis of the petition j
i should identify new information regarding the issues under consideration.
In order to have a hearing reopened on the basis of new information, as CESG l
seeks to do, the Appeal Board has held that the new information must identify a significant unresolved safety issue or a major change in facts material to l
the resolution of major environmental issues.2 Although the Director, in
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considering a request for action under 10 CFR 2.206, is not bound by the Appeal Board's standard for reopening a licensing proceeding on the basis of new 110 CFR 92.206(a) 2 Vermont Yankee Nuclear Power Corporation (Vermont Yankee Nucear Power Station), ALAB-124, 6 AEC 358 (1973); Commonwealth Edison Company (La Salle, Units 1 and 2), ALAB-153, 6 AEC 821 (1973).
The Director of NRR has previously applied this standard in denying another petition under 10 CFR 2.206 which requested suspension of construction permits pending reconsideration of the need for power issue after the proceeding on issuance of construction permits for the facility had been closed.
Georgia Power Company ~(Alvin W. Vogtle Nuclear Plant, Units 1 and 2), 00-79-4, 9 NRC 582 l
(April 13,1979) (Docket Nos. 50-424 and 50-425).
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3 information, this standard is persuasive in considering requests under 10 CFR 2.260 because, as the Commission has indicated on another occasion,
"[P]arties must be prevented from using 10 CFR 2.206 procedures as a vehicle 4
for reconsideration of issues previously decided..."
Consolidated Edison Company (Indian Point Units 1-3), CLI-75-8, 2 NRC 173, 177 (1975).
CESG's petition provides no explanation, by reference to the record or otherwise, why the matters it identifies support reopening of the record under i
this standard.
This failure would justify denial of the petition at the outset because the petitioner has not, as required by 10 CFR 2.205, specified the i
facts that constitute the basis for the request.3 However, the staff has conducted its own review of CESG's petition to reopen and has found no good cause to reopen the record at this time.
Accordingly, the petition to reopen the Catawba safety hearing must be denied.
The staff's analysis follows.
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In its petition, CESG has asserted a number of issues as the basis for its subsequent to the issuance of construction 4
request to reopen the safety hearings l
permits.
These issues are:
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3See also the Director's denial under 10 CFR 2.206 in Duke Power Company (0conee Nuclear Station, Units 1, 2 and 3), DD-79-6, 9 NRC 661 (May 24, 1979)
(Docket Nos. 50-269, 50-270, and 50-287).
4The Director does not have the power to direct a Licensing Board or Appeal Board to conduct further proceedings on the matters which CESG raises.
The Director could recommend to the Commission that the hearings be reopened or the Director could issue an order based on the matters raised by CESG under which interested persons may have a right to request a hearing.
sConstruction permits for Catawba Nuclear Station were issue'd'on August 7, 1975.
The licensee tendered its application for an operating license on March 21, 1979.
The application has not yet been docketed by the NRC staff.
After the i
application is docketed, a notice of opportunity,for hearing will be published in the Federal Register.
See 10 CFR 2.105.
At that time, interested persons-may seek a nearing on the proposed issuance.of an operating license.
It-i should be noted that CESG has participated as a party in the Catawba construc-tion permit proceeding and both the McGuire construction permit and operating license proceedings.
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'(1) need for the Catawba facility and the effect a reduced level of need for that facility would have on the cost / benefit balance, especially the consideration of risk, (2) inadequate treatment at the construction permit proceeding of Class 9 accidents, stud bolts, and the ice-condenser pressure suppression containment, (3) the degree to which the construction permit safety evaluation of Catawba was " infected" by deficiencies which CESG claims are present in the Reactor Safety Study.
I.
NEED FOR POWER i
CESG contends that there is a diminished need for the Catawba facility thus affecting the cost / benefit balance struck at the construction permit stage and requiring a renewed examination of the risk involved in licensing the units.
The linchpin of CESG's contention in this area is the claim of diminished need for the Catawba units.
Given a continuing need for these units, the original cost / benefit balance struck at the construction permit stage remains valid.
Other than its claim of diminished need, CESG offers nothing additional in its petition which could call the original cost / benefit balance into question.
The need for power issue was thoroughly explored in the Catawba construction permit proceeding.
CESG was an active participant.
In the construction permit'
5 proceeding, the Licensing Board,8 as well as the Appeal Board,7 determined that the need for power evaluations warranted the construction of the facility.
An examination of these proceedings clearly exhibits that the need-for power issue had been exhaustively treated at that time.
The question for consideration then is whether CESG has identified such new information as would clearly mandate a change in result.
The staff has analyzed the information presented in CESG's petition and has found that this information does not identify a major change in facts which would alter the need for power determination as originally analyzed in the construction permit proceedings for the Catawba facility.
CESG claims that, given Duke's current high reserves, there is virtually a certainty that the Catawba Units 1 and 2 will not be required at any time in the foreseeable future.
The staff has examined the capacities and demand requirements of the Duke system and continues to be of the view that system reserves justify the addition of the Catawba units in accordance with Duke's latest capacity 6 Initial Decision, Duke Power Company (Catawba Nuclear Station, Units 1 and 2) LBP-75-34, 1 NRC 626, 656-666 (1975).
7 Partial Decision, Duke Power Company (Catawba Nuclear Station, Units 1 and 2) ALAB-355, 4 NRC 397, 404-414 (1976).
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expansion plan.
Current scheduling calls for Catawba Unit 1 to be available i
by the summer of 1984 and Unit 2 to be available by the winter of 1985/86.
l Table 1 presents Duke's latest forecast of peak demand and capacity plans for the winter of 1984/85 through the winter of 1989/90.
Reserve margins as a percentage of peak demand are reported for two cases reflecting capacity 4
estimates with and without the Catawba units added as scheduled.
Based on these projections, reserve margins on the Duke system during winter peak t
demand will not be adequate to insure reliability of service unless the proposed Catawba units are added in this time frame.
i The reserve margins calculated in Table 1 show that even with the scheduled addition of the Catawba units, reserve margins are expected to range between 7.3% and 26.6%.
Without their addition, reserves are estimated to become i
unacceptably low as of the winter of 1984/85 (14.3% reserves) and become progressively worse through the forecast period.
The Duke Power Company has identified reserves ranging from 17% to 25% as necessary to provide minimum l
acceptable reliability.
The Department of Energy has indicated that reserve i
i margins in the 15% to 25% range characterizes systems that are reasonably l
l reliable.
Based on these reserve margin standards, the staff concludes that l
the Catawba units are needed in order for the Duke system to maintain reliable l
service.
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This conclusion that the Catawba units are needed is largely predicated on Duke's expectation that peak deqand will grow at an average annual growth rate of about 4.5% over the forecast period. ' The' staff Views this growth rate as a i
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TABLE 1 Peak Load Demand, Capacity, and Reserve Margins With and Without the Catawba Nuclear Station -- Winter 1984/85 through Winter 1989/90 RESERVE MARGINS %
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5 WITH WITHOUT YEAR PEAK LOAD CAPACITY WITH CATAWBA CAPACITY WITHOUT CATAWBA MWe MWe MWe CATAWBA as % of Peak Load Demand Winter 1984-85 12692 15646 14501 23.3 14.3 Winter 1985 13157 16656 14366 26.6
- 9. 2 Winter-1986-87
~13631 16563 14273 21.5 4.7 Winter.
1987-88 14092 16563 14273 17.5
- 1. 3 Winter 1988-89 14631 16388 14098 12.0 3.6 Winter 1989-90 15179 16280 13990
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- 7. 8 a/
. Peak load estimates based on Duke Power Company's long term forecast of June 1980.
5 All capacity _ estimates assume MWe net firm purchases and the following retirements.
135 MWe effective by the winter of. 1985-86 35 MWe effective by the winter of'1988-87 85 MWe effective by the summer of 1988 90 MWe effective by the winter of 1988-89
-108 MWe effective by the winter of 1989-90 E
Only planned additions over this timeframe are McGuire and Catawba Nuclear units:
- McGuire Unit 1 - available for 1981 peak McGuire Unit 2 - available for 1982 peak Catawba 1 - available for 1984 peak Catawba 2 - available for 1985 peak Source: Duke Power Company, submittal of S. B; Nager to S. Feld (NRC), July 29, 1980
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-reasonable estimate of future growth as the staff's own independent forecast of growth in electrical energy demand for the State of North Carolina between 1976 and 1990 is 4.4% per year.
This estimate is based on a state level econometric forecasting model of electricity demand developed at the Oak Ridge National Laboratory.
The staff notes that the Duke Power Company has initiated a load management program aimed at promoting the increased use of interruptible contracts on the part of Duke's larger customers.
Customers interested in this program will be offered more attractive rates with the understanding that service may be interrupted during periods of peak demand.
Because of the nature of this contracted service, Duke will not have to maintain peak reserves to support this interruptible load.
Thus, that portion of the peak load represented by interruptible contracts can be added to reserves in assessing reliability.
During the winter peak, the interruptible load is projected to grow from 27 MWe in the winter of 1980-81 to 389 MWe by the winter of 1989-90.
The effect of this program on reserve margins would be to increase winter reserves.
However, its impact is e>pacted to be minimal throughout this forecast period with maximum impact occurring in the winters of 1987-88 through 1989-90 when reserves would be effectively raised by about 2.5 percentage points.
Correct-ing for this adjustment would not alter the underlying conclusions reached above.
Given a continuing need for the Catawba facility, the cost / benefit balance struck at the construction permit proceeding remains valid. While CESG implies t
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the factors identified by CESG as contributors to additional risk are support-able as is discussed in the next portion of this decision.
Consequently, there is no basis for CESG's claim that the cost / benefit balance originally i
struck is now invalid.
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1 II. ALLEGED INADEQUATE TREATMENT OF CERTAIN CONSTRUCTION PERMIT ISSUES 1
1.
Stud Bolts i
CESG raises the issue of stud bolts, which are used to mechanically
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secure the reactor top closure to the reactor vessel flange, as a factor leading to an increased level of risk at the Catawba facility.
This i
issue has been litigated by CESG in proceedings before this agency with respect to both the Catawba and McGuire facilities.
i CESG raised the stud bolt matter in both the Catawba and McGuire con-l struction permit proceedings and, in each instance, the Licensing Board found against CESG.9 kppealBoardreviewsir each instance supported the findings of the Licensing Board.10 "CESG Petition,. page 2, paragraph 4 9 Duke Power Com an (William B. McGuire Nuclear Station,' Units 1 and 2)
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l LBP-73-7, 6AE
, 106-108 (1973); Duke Power Company (Catawba Nuclear Station, Units 1 and 2) LBP-75-34, 1 NRC 626, 642-646 (1975) 100 uke Power Company (William B. McGuire Nuclear Station, Units 1 and 2 ALAB-128, 6 AEC 399, 401-404 (1973); Duke Power Company (Catawba Nuclear l
Station, Units 1 and 2) ALAB-355, 4 NRC 397, 414 (1976) i N
e 10 At the construction permit proceeding, CESG contended that 18% of the stud bolt stock specimens in an infinite population would fall below the acceptance value of 130 ksi, set by the ASME Code.
During the Licensing Board hearings on Catawba held in Rock Hill, South Carolina, in April 1974, this matter was extensively discussed.
At that time. the discussion pointed out that the data used to establish Code allowables could be considered to have come from an essentially truncated distribution rather than a normal distribution.
However, statistics, applied in the normal sense of data collected from a small sample of an infinite population, really are not applicable in this case for the reasons which follow.
Stresses are not set by statistical means.
Data are collected and normalize 1 by dividing the value of the property for each set at elevated temperatures by the room temperature value of that set.
Then using linear regression techniques, a ratio trend curve is established.
1 In using the ratio trend curves, values higher than 1.0 are reduced to l
- 1. 0.
The ratio trend curve may be further modified by the responsible ASME committee after reviewing and adjusting the data as considered necessary before being accepted as the " trend curve," or " characteristic variations of property with temperature" for the material.
The ASME Code includes among its criteria for establishing allowable stresses not only fractions of yield and tensile strengths at elevated temperatures but also fractions of the specified minimum values of these
11 properties at room temperature.
In effect, allowable stresses become anchored to the requirements of the purchase specification.
In turn, the yield and tensile strengths at temperature as characterized by the trend curves may be developed without specific concern for their exact posi-tions relative to the strength scale since they are adjusted to specified minimum strengths.
In contrast, many foreign codes require a statistical assessment of minimum strength at temperature since no room temperature specified minimum value is used to anchor the data.
Another way in which the data could be developed is the use of an unnormalized unanchor' g with the variance of the data to form the basis for a statisticas definition of a minimum trend curve in terms of an appropriate confidence level.
Perhaps this is what CESG may be alluding to.
However, the Code has chosen not to use statistical means to establish allowable stresses but instead establishes these stresses by multiplying the value selected from the ratio trend curve by the specified minimum room temperature property value and then by the appropriate stress basis factors selected from Appendix III of Section_III of the Code.
The specified minimum strength value in the case of the stud bolting is as established. in Table 2 of the material specification, SA-540. A test coupon is removed from each end of bars selected to represent each heat of a given size for each tempering change or each 10,000 pounds whichever is less. Ten thousand pounds represents approximately 15 stud bolts of the size used on McGuire.
The specified
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minimum property values must be met or the lot is rejected.
Therefore, the conclusion that specified minimum values of yield strength at room temperature will not be below 130,000 psi is valid, and the statistical inference implied by CESG that 18% of the specimens would be below that figure is invalid.
i CESG offers nothing in its petition to place into question these prior determinations in agency proceedings.
Therefore, given the consideration this issue has already received and the absence of any new information on this subject in CESG's petition, and in light of the above discussion, there is no justification for CESG's claim that this issue increases the level of risk associated with the Catawba facility.
2.
Ice Condenser Pressure Suppression Contaig ent CESG also raises the issue of ice condenser pressure suppression containment l
as a factor leading to an increased level of risk at the Catawba facility.
CESG has raised the ice condenser pressure suppression containment issue in both the construction permit and operating license proceedings for the McGuire facility.
The containments for the McGuire and Catawba Units are virtually identical.
In the McGuire construction permit proceeding,
s 13 the matter was found against CESG by the Licensing Board.
The Appeal 10 Board confirmed this finding The only information offered by CESG in its petition on this issue refers to certain internal documents of the Comission dated in 1972 alledgedly...." expressing reservations about pressure suppression containments..."II In 1972 Dr. S. H. Hanauer (then Technical Advisor to the NRC's Executive Director for Operations) wrote a memorandum that raised several questions on the viability of pressure suppression containment concepts. As the memo expressed reservation with respect to pressure suppression contain-ments, an evaluation addressing each of the points raised by the memo was undertaken.
9 Duke Power Company (William B. McGuire Nuclear Station, Units 1 and 2)
LBP-73-7, 6AEC 92, 101-104 (1973) 100 uke Power Company (William B. McGuire Nuclear Station, Units 1 and 2)
ALAB-128, 6AEC 399, 401-404 (1973)
II CESG petition, p. 3 4
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. In view of the extensive testing and analyses sponsored by the owners of pressure suppression containments and the in-depth reviews by the Advisory Committee on Reactor Safeguards, the staff issued a report on this matter entitled, "A Technical Update on Pressure Suppression Type Containments in use in U.S. Light Water Reactor Nuclear Power Plants,"
NUREG-0474, dated July 1978. Each of the technical concerns identified by Dr. Hanauer in his 1972 memorandum is discussed in detail in the report.
It was concluded that pressure suppression types of containments were conceptually acceptable. Consequently, the staff's earlier findings were confirmed.
Since the filing of CESG's petition, a new concern has developed with respect to certain types of containment systems, to which the staff has responded.
1 The accident at TMI-2 indicated a need to consider the possibility of hydrogen generation well in excess of the amounts considered in 10 CFR 50.44 of the Commission's regulations. The staff has undertaken a study of the potential of excess hydrcgen generation, the effects such concentrations of hydrogen would have on the various types of plants, and the effectiveness of various mitigation systems in protecting the plarit against such situa-tions. The resu'lts of our studies to date are presented in the SECY 80-107 series of occuments:
1.
SECY 80-107, Proposed Interim Hydrogen Control Requirements for Small Containments, dated February 22, 1980.
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SECY 80-107A, Additional Information Re: Proposed Interim l
Hydrogen Control Requirements, dated April 22, 1980.
3.
SECY 80-1078, Additional Information Re: Proposed Interim Control Requirements, dated June 20, 1980.
When the Comission approved the licensing of the Sequoyah plant for full power operations, certain additional requirements for hydrogen control were i
imposed as license conditions. These include requiring an acceptable interim system for hydrogen control by January 31, 1981, and an acceptable final system by January 31, 1982. Based on Commission guidance, we are proceeding with plans to implement similar requirements for all other i
operating ice condenser plants on a case-by-case basis.
In addition, a two-step rule-making process is currently underway which proposed that an interim rule be put in place expeditiously for the near term, and l
that a final rule be developed for the longer term.
Interim Requirements Related to Hydrogen Control and Certain Degraded Core Considerations,"
45 Fed. Reg. 65466 (October 2, 1980). With respect to ice condenser I
containments, the proposed interim rule will require owners to perform
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certain extensive analyses of accident scenarios involving hydrogen releases and furnish the staff with a proposed approach for mitigating these hydrogen releases. Upon evaluation of these interim measures, a final rule will be developed for longer term requirements. The Catawba facility will be required to meet, the regulatory requirements in this area.
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l In sumary, CESG again offers nothing in its petition to place into question i
prior agency determinations in this area.
Given the consideration this issue has received, and current measures being undertaken to address new issues, there is no justification for CESG's claim that this issue increases 9
the level cf risk associated with the Catawba facility.
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i 8.
Class 9 Accidents J
l CESG also contends that the risk level for the Catawba facility is affected by the failure to consider Class 9 accidents at the construction permit proceeding. The Comission's current policy differs somewhat from the policy 3pplied at the Catawba operating license proceeding. The tem
" Class 9 accident" was first used in a Comission' rulemaking proposed in i
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t 17 December 1971.
" Consideration of Accidents in Implementation of the j
i National Environmental Policy Act of 1969," 36 Fed. Reg. 22851 (1971).
An Annex to Appendix 0 of 10 CFR Part 50 was proposed to establish the manner in which various categories of accidents should be taken into account in the environmental review for a nuclear power plant.
The Commission has since withdrawn the proposed Annex and has replaced it with new interim guidance for the treatment of accident risk considerations in NEPA reviews.12 In following the Commission's interim guidance, considera-tion of serious accidents is now planned as part of the Staff's review of a licensee's application for an operating license.
It is u3eful, however, to briefly review the withdrawn Annex and other events leading to the Commission's new interim policy.
In the proposed Annex, nine classes of accidents were created based on their range of severity.
Each class of accidents, except Classes 1 and 9, was required to be investigated in environmental reports and state-F ments. Class 1 accidents were exempt because of their trivial consequences.
In dealing with Class 9 accidents, the Annex stated:
"the occurrences in Class 9 involve sequences of postulated successive failures more severe than those postulated for the design basis for protective systems and engineered safety features.
Their consequences could be severe.
However, the probability of their occurrence is so small that their environ-mental risk is extremely low.
Defense in depth (multiple T7"kuclear Power Plant Accident Considerations under the National Environmental Policy Act of 1969;" 45 Fed. Reg. 40101 (June 13,1980).
18 physical barriers), quality assurance for design, manufacture, and operation, containued surveillance and testing, and conservative design are all applied to provide and maintain the required high degree of assurance that potential accidents in this class are, and will remain, sufficiently remote in prob-ability that the environmental risk is extremely low."
36 Fed.
Reg. 22862 (1971)
Even though the Annex was never formally adopted by the Commission, the Annex was used as an " interim guidance." The proposed Annex was used consistently from 1971 to 1979, i.e., Class 9 accidents were not considered in environmental statements.
In September 1979, the Commission announced that it would complete the rulemaking started by the Annex and review the policy regarding accident considerations 23 On May 16, 1980, the Commission withdrew the Annex and issued a statement of interim policy.
This new interim policy was pub-lished in the Federal Register, on June 13, 1980.
The new policy requires environmental impact statements for ongoing and future NEPA reviews to consider a broader spectrum of accidents, including severe accidents that 131n Offshore Power Systems (Floating Nuclear Power Plants), CLI-79-9 10 NRC 257 (1979), The Commission determined that consideration of a Class 9 accident in the environmental review for floating nuclear power plants was appropriate.
10 NRC at 260-61.
The Commission did not use the proceeding to resolve the ;eneric issue of consideration of Class 9 accidents at land-based reactors, but noted that "such a generic action is more properly and effectively done through rulemaking proceedings in which all interested persons may participate." Id. at 262.
See also Public Service Co. of Okla. (Black Fox Station, Units 1 and 2) CLI-80-8, 11 NRC 433, 434-435 (1980)
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19 were once designated as " Class 9."
The Commission gave the following guidance for considering environmental risks, or impacts, attributable to accidents at a facility:
"In the anaiysis and discussion of such risks, approximately equal attention shall be given to the probability of occurrence of releases and to the probability of occurrence of the environmental consequences of thnse releases..
" Events or accident sequences that lead to releases shall include but not be limited to those that can reasonably be expected to occur.
In plant accident sequences that can lead to a spectrum of releases shall be discussed and shall include sequences that can result in inadequate cooling of reactor fuel and to melting of the reactor core." 45 Fed. Reg. at 40103."
When addressing the new interim policy concerning plants for which Final Environmental Statements have been issued, the Commission stated that:
"It is expected that these revised treatments will lead to conclusions regarding the environmental risks of accidents similar to those that would be reached by a continuation of l
current practices, particularly for cases involving special circumstances where Class 9 risks have been considered by the staff..
Thus, this change in policy is not to be construed as any lack of confidence in conclusions regarding the environ-mental risks of accidents expressed in any previously issued Statements, nor, absent a showing of similar special circumstances, as a basis for opening, reopening, or expanding any previous or ongoing proceeding."5
" Commissioners Gilinsky and Bradford disagree with the inclusion of the preceding two sentences.
They feel that they are absolutely inconsistent with an evenhanded reappraisal of the former, erroneous position on Class 9 accidents." 45 Fed. Reg.
at 40103.
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20 "However, it is also the intent of the Commission that the staff take steps to identify additional cases that might warrant early consideration of either additional features or other actions to prevent or to mitigate the consequences of serious accidents.
Cases for such consideration are those for which a Final Environmental Statement has already been issued at the Construction Permit stage but for which the Operating License review stage has not yet been reached.
In carrying out this directive, the staff should consider relevant site features, including population density, associated with accident risk in comparison to such features at presently operating plants.
Staff should also consider the likelihood that substantive changes in plant design features which may compensate further for adverse site features may be more easily incorporated in plants when construction has not yet progressed very far."
The circumstances identified by the staff as "special" fall into three categories:
(1) high population density around the proposed site; (2) a novel reactor design (a type of power reactor other than a light water reactor); or (3) a combination of a unique design and a unique siting mode.
Another exceptional case noted by the Commission that might warrant consideration is the proximity of a plant to a " man-made or natural hazard."14 ts As discussed in Section 1 of the Catawba Safety Evaluation Report, the nuclear steam supply system for each unit will consist of a four-loop Westinghouse pressurized water reactors.
The principal features of the Catawba plants are similar to those previously approved for other nuclear 1* Black Fox, CLI-80-8, supra.
tsSafety Evaluation Report of the Catawba Nuclear Station Units 1 and 2, dated October 12, 1973, Supplement No. I dated January 21, 1974.
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power plants now under construction or in operation, especially the l
McGuire Nuclear Station Units 1 and 2 (Docket Nos. 50-369 and 50-370) and 1
j the Donald C. Cook Nuclear Plant Units 1 and 2 (Docket Nos. 50-315 and 50-316).
Therefore, Catawba is not a novel reactor design.
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l The staff has developed population density guidelines, which are given in Regulatory Guide 4.7, for determining when the population surrounding a proposed new site is sufficiently high to require that special attention 4
be given to the consideration of alternative sites with lower population J
densities.
1 The following table shows the cumulative population and population density out to a radius of 30 miles around the Catawba site for the years, 1970, 1985, and 2019.
The 1970 population was based on census data, while the.1985 and 2019 projections were developed by the applicant.
POPULATION DISTRIBUTION - CATAWBA SITE RADIUS CUMULATIVE POPULATION POPULATION DENSITY, PERSONS /MI2 l
Miles 1970 1985 2019 1970 1985 2019 l
0-5 5626 13196 19007 72 168 242 0 - 10 65227 85433 130262.
208 272 415 0 - 20 442339 601884 961164 352 479 765 0 - 30 705691 943730 1500059 250 334 531 l
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22 As shown in the table above, the cumulative population density surrounding the Catawba site is estimated to be less than the 500 persons per square mile density guideline of R.G. 4.7 which applies at the startup date (1983) out to a distance of 30 miles.
The projected growth rate for the area surrounding 9
the site indicates that the population density will stay well within the s
1,000 persons per square mile guideline over the lifetime of the plant.
While the population density within 20 miles for the year 1985 is projected to be close to the value of 500 persons per square mile, it approaches this value only at one location, and at a significant distance from the site.
A large fraction of the population at this distance is due to the City of Charlotte, t
N.C., and its environs, located about 17 miles from the Catawba site.
Staff l
studies of accident risk leads the staff to conclude that the risk is higher-i for persons relatively close to the site, and generally decreases with distance.
In particular, the staff has found that the most severe consequences of very l
J large accidents, namely, acute fatalities, would be generally limited to i
distances of about 5 miles or less.
Also, the Commission's recently revised regulations in regard to Emergency Planning, 10 CFR Part 50, Appendix E, a
require emergency planning zones for the plume exposure pathway out to about 10 miles from the reactor site.
Beyond these distances the consequences are t
j expected to diminish significantly.
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Based upon the foregoing findings and considerations, the staff concludes that-4 the population data for the Catawba site do not reflect a sufficiently unique circumstance to warrant considerations of Class 9 accident consequences at this time.
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23 With regard to the third category of special circumstances, the staff has identified a potential special circumstance involving the dewatering system used at Catawaba.
In the event of a serious accident the dewatering system, as presently designed,could potentially result in undesireable releases to the surrounding environment.
The staff will examine this issue during its operat-ing license review to determine its seriousness and potential impact.
In the staff's view, this issue can be resolved prior to issuance of an operat-ing license, and therefore would not form a bases upon which to reopen the construction permit proceedings.
This finding is based on the fact that the range of solutions, including additional structural design considerations or storage capabilities, are of such nature that they do not preclude continued construction, and could result in modifications which can be made in later stages of construction.
In addition to the staff's review, measures have been taken or are under consideration by the Commission and the staff to prepare to meet the possible consequences of a serious accident at a reactor site including:
A rule was issued, 45 Fed. Reg. 162 (Aug. 19, 1980), which signifi-l cantly revises requirements in 10 CFR Parts 50 and 70 for emergency i
planning at nuclear power plants.
24 Recommendations of the Siting Policy Task Force (see NUREG-0625, Aug. 1979) with respect to possible changes in the reactor siting policy and criteria set forth in 10 CFR Part 100.
One goal of the recommendations is to consider in siting the risk associated with accidents beyond the design basis (i.e., Class 9) by establishing population density and distribution criteria.
Proposed " Action Plans" (see NUREG-0660, Vol. 1, May 1980) for implementing recommendations made by bodies that have investigated the Three Mile Island accident.
Among other matters these plans incorporate recommendations for rulemaking related to degraded core cooling and core melt accidents.
In addition, certain of these recommendations have been adopted as requirements-as described in.
NUREG-0737 " Clarification of TMI Action Plan Requirements" (October 1980).
Imposition of additional requirements on operating reactors, e.g.,
the short-term " lessons-learned" recommendations.
See "TMI-2 Lessons Learned Task Force Status Report and Short-term Recommen-dations," NUREG-0578 (1979), and Orders published in 45 Fed. Reg.
2427-2455 (Jan.
11, 1980).
Given the staff's forthcoming review of serious accidents, there is no basis to support CESG's claim that failure to consider Class 9 accidents at the construction permit = stage elevates the level of risk associated with the Catawba facility.
This conclusion is further assured by the additional
25 measures noted by the Commission in its new statement of interim policy on accident considerations.
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26 III.
THE REACTOR SAFETY STUDY The final issue raised by CESG deals with the Reactor Safety Study (RSS).is J
CESG asserts that the McGuire and Catawba " safety evaluations were infected with the same tendencies that the Commission has found in the Rasmussen Report."
H~ sever, aside from averrino that the same personnel were involved, CESG provides no specifics in support of such claim.
Indeed, CESG acknowledges that the RSS was not relied upon in either the Catawba or McGuire safety evaluation.17 This question was extensively dealt with by the Atomic Safety and Licensing Board in the McGuire operating license proceeding.ta The Licensing Board there correctly concluded that "...there is no nexus between the Rasmussen Report and CESG's claim of special circumstances warranting reopening the 1
record in this proceeding."28 The same conclusion applies to the Catawba l
facility.
The Licensing Board in McGuire also reviewed the results of an NRC staff survey of the uses which the Staff had made of the RSS.
The Board concluded l
l 16 Reactor Safety Study, WASH-1400, October 1975.
17CESG Petition, page 3, paragraph 5.
f 18Me" _Jrandum and Order Ruling on Motions to Reopen Record, Duke Power Co.
(McGuire Nuclear Station, Units 1 and 2), unpublished (April 10, 1979).
A copy of this Order is attached.
191d., p. 3.
27 that the Comission's withdrawal of its approval of the Executive Sumary of the RSS was not a basis for reopening the record of that proceeding.2o That conclusion applies with like force to the Catawba facility.
Accordingly, the review of the Rasmussen Report cannot serve as a basis for reopening this proceeding.
Conclusion For the reasons set forth above, the petition of the Carolina Environmental Study Group to reopen the safety phases of licensing proceedings for Duke Power's Catawba Nuclear Station is hereby denied.
. A copy of this Decision will be placed in the Commission's Public Document Room, 1717 H Street, AW, Washington, DC 20555, and at the local public document room for the Catawba Nuclear Station at the York County Library, 325 South Oak Avenue, Rock Hill, South Carolina 29730.
A copy of this Decision will also be filed with the Secretary of the Commission for review by the Commission in accordance with 10 CFR 2.206(c) of the Commission's regulations.
As provided in 10 CFR 2.206(c), this Decision will constitute the final action of the Commission twenty (20) days after the date of issuance, unless the
" Id., pp. 2-9.
3 1
1 28 Consission on its own motion institutes a review of this Decision within that i
time.
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i Harold R. Denton, Director l
Office of Nuclear Reactor Regulation Dated at Bethesda, Maryland this 9 day of Jan., 1980 i
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UttITED STATES OF A". ERICA NUCLEAR REGULATORY CCF;ilSSION BEFORE-THE' ATOMIC SAFETY Ai;D LICEf1 SING E0ARD
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,In the Matter of DUKE FO'.ER COMPAtlY
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Docket ffes. 50-359 4'
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50-370 (k*illiam 3. McGuire fiuclerr Station, Units 1 and 2)
April 10, 1979 MEMORAfl0UM AND ORDER RULING ON MOTIONS TO REOPEN RECOP.0 1
1.
CESG's Petition to Reopen the Safety Phases of Licensing Proceedings for Duke Power's McGuire and Catawba *:uclear Stations 1
On January 28, 1979, the Carolina Environmental Study Group (CESG), an intervenor in the above-identified proceeding, petitioned the Co:Taission to reopen the safety phases of the licensing proceedings for Duke Power Company's Catawba Nuclear Station and McGuire Nuclear Station. On March 7,1979, the Director of the Office of' Nuclear Reacter I
Regulation advised CESG that its request to reopen the Catawba proceed-l i
ing is being treated as a request under 10 CFR 52.205 of the Cxciission's l
l regulations.
Ho.cavar, sir.ca tha matter of i:s=.nce of operatig liter.sss for the McGuire facility is currently pending before this Board, CESG's I
request to reopen the safety phase of the McGuire proceeding has been referred to us. On the same day, we issued an order advising the parties h
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In the Matter of
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Do:ket I:os. $0-369
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said it "does not regard as reliabic the Reactor Ssichy Study's mmrical estir. ate of the overall risk of a reactor accid rnt."
Under this spe.cial circ =: stance CDS3, intervenor in the referenced docket nrl>:rs, petitions 4
this Cc. nission to reopen th:: sa.fet.y phiscs in ~.,ha lic+nri:.a:.erc, : cedings.
Ch:re is good cause to reopen the.,3 r.t- >rs:
1.
So erronacus s.as Duke Foar Coc.ary's 1971 csbi-ito of I
future caic electrical dest.nd that it is a virtual certair.ty that McGuire unit 2 rad Catau'oa unihs 1 and 2 will not bc required at any tino in the foreser:?olo future. The d=.tc of initini con:r.ercial cycration of I Dire unit I h.s baon sah ba h saven ti:nas. DM:e precently ajoys a 33.% resarr2 Th3 Erecutive Dirastor of the Public 5;aff of the I: orth Carolina Public Utilitics Co.. mission, Mr. Eu;h W211s, stated before the Ato.mic Safety and Licensin; Board h:aring the safety natters in the operating lic;asa pro: ceding, Au,;nst 22, 1973, that "Sased solely upon satir!;;' g the rasarve cc.iterie. found s
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11 arch 1,1979
- 0RN:0U:i FOR:
Harold R. Dan ton, Di rec ti.-
Office of fuclear Reacic;.1:gulaticn F'O.:
Jctr.es F. l:urray Office of Exacutive Lega: Direc:or
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By petition dated January 28, 1979, the Cc. olina Environn:ntal Study Group (CESG) requested that the Cor ission reopen the safety phases of licensing proceedings for Duke Potter Conpany's Cata'.;ba fluclear Station.
The basis for ti request is th2 contention that ncw infor.r.ation - a revis:! esticate of peak requircnants - trould alter the cost /benefi: balance for the racility.
CESG further contends that if t.ere is no nced for these units, the a.cceptance of the level c' risk project?d during orevious proceedings is not justii' ble.
orcorar, since, allegedly, certain safety r.atters have not baan adequatcly v: alt with, if the risks of these factors were c;rrectly assessed, the level of risk would be even higher and evcn less justifiable.
Since there is no ongoing licensing proce fing to considar CESG's request as a motion to rcopen, it is being treated as a icquest for action under 5 2.206.
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This office will work with your staff to d velop an apprc.priate response to this request. A copy of all c.crespondence related to this reatter should be sent to this offi:9.
j Enclosed for your use are drafts of:
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Letter of acknowledgment to the Cacolina Environmental Study Group; l
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