ML19326C789
| ML19326C789 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 10/01/1968 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| References | |
| NUDOCS 8004280724 | |
| Download: ML19326C789 (72) | |
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October 1,1968 SAFETY EVALUATION BY THE DIVISION OF REACTOR LICENSING U.S. ATOMfC ENERGY COMMISSION IN THE MATTER OF ARKANSAS POWER AND LIGHT COMPANY RUSSELLVILLE NUCLEAR UNIT DOCKET NO. 50-313 4
THIS DOCUMENT CONTAINS POOR QUAUTY PAGES g$
8004 280 72 y/
. TABLE OF COIEENTS Page No.
1.0 INTRODUCTION
1 2.0 SITE AfD PLANT DESCRIPTION 4
2.1 Site Description k
2.2 Plant Descript!.on.
6 30 IMPORTANT SAFEIT CONSIDERATIONS.
11 31 Site Saitability 11 32 Acceptability of the Nuclear Steam Supply System Design.
15 33[EngineeredSafetyFeaturesAdequacy.
18 3.h
' Foundation and Structural Design Adequacy.
22 35 Adequacy of Instrumentation, Control and anergency Power Systems..
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36 Radioactive waste Disposal System Adequacy 27 37 Analysis of Radiological Ccusequences from Potential Accidents.
28 3.8 Design Conformance to AEC General Design Criteria.
30 39 Energency Plans.
31 h.O RESEARC'H AND DEVELOPMENT 32 4.1 Other Matters to be Further Evaluated During Construction.
34 4.2 ' Conclusion 36 50 REPORT OF THE ADVISORY COINITTEE ON REACTOR SAFEDUARDS 37 6.0 TECHNICAL QUALIFICATIONS OF THE APPLICANT.
38 6.1 Technical Competence 38 6.2 - Quality Assurance....
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h2 70 COMON DEFENSE AND SECURITY..
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8.0 CONCLUSION
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hk Appendix A - Report of Ad'risory Committee on Reactor Safeguard Appendix B - Chronology - Regulatory Reviev.
47 50 Appendix C1 - Report of U.S. Fish & Wildlife Service..
54 Appendix C2 - Report of U.S. Fish & Wildlife Service.
55 Report of U.S. Weather Bureau Appendix D Appendi.c E - Report of U.S. Geological Survey............
57 60 Appendix F - Report of U.S. Coast & Geodetic Survey.........
Report of Nathan M. Newmark Consulting Engineering Appendix G 62 Services.
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1.0 INTRODUCTION
On November 29, 1967, the Arkansas Power & Light Company (applicant) submitted an application to construct and operate a sin 61e-unit nuclear power plant, to be known as the Russellville Nuclear Unit. Ten supplements to'that' application have since been filed with the Atomic Energy Commission. The reactor site is located about 6 miles from Russellville on a peninsula in the Dardanelle Reservoir on the Arkansas River, Pope County, Arksnsas.
The facility architect-engineer and construction manager will be the Bechtel Corporation, the nuclear steam supply system will be furnished by the Babcock u Wilcox Company (B&W), and the turbine generator will be supplied by the Westinghouse Corporation.
The plant will use a B&W prescurized water reactor designed to operate at 2452 megawatts thermal (Mwt) and produce 850 megawatts of electrical power (Mwe ).
The' expected ultimate capacity of this plant is 2568 Mwt. The appli-cant has designed the major plant components including the containment and other engineered safety features for a power level of 2568 Mwt, and has used this power level in analyzing postulated accidents in conformance with the siting guidelines of Title 10 - Chapter I, Part 100 of the Code of Federal Regulations (10 CFR 100).
We evaluated the containment and other engineered safety features for 2568 Mwt; however we evaluated the thermal and hydraulic characteristics at 2452 Mwt.
Before operation above s power level of 2452 Mwt
. is authorized, the-Commission's regulatory staff must perform a safety evalu-ation to assure that the facility can be operated safely at that power level.
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, D-The application, including the Preliminary Safety Analysis Report (PSAR) and Supplements 1-10 (hereinafter collectively referred to as the " application")
was the basis on which the Division of Reactor Licensing conducted the technical evaluation of the preliminary design of the proposed plant. The staff used the following approach in its review of this application.
Performed an in-depth evaluation of site-related features.
a.
b.
Identified and compared all of the design and safety features of the
.-3 Russellville Nuclear Unit for similarity to those previously reviewed.
Where justified, we relied upon previous in-depth evaluations of like systems, components, and structures without performing separate, duplicate evaluations.
Determined that the design features and the treatment of safety matters c.
were consistent with current regulatory criteria and policy, and that the applicant adequately addressed concerns which have been identified by the Advisory Committee on Reactor Safeguards (ACRS) in previous reviews.
d.
_ Identified and evaluated those design features and related safety matters that are new or unique, or which, a'+. hough reviewed in the past for O
other applications, continue to require review.
Within the Division of Reactor Licensing, the Reactor Projects group was responsible for the review, and for coordinating parts of the review involving personnel within.the Division representing various special technical disciplines from the Reactor Technology and Reactor Operations groups, as well as consultants
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. and other governmental agencies outside of the Division of Reactor Licensing.
The reports of our consultants are attached as Appendices C through G.
During the review a number of meetings were held with representatives of the applicant to discuss the proposed plant.
As a consequence, additional information was received from the applicant.
The Advisory Committee on Reactor Safeguards has considered the applica-tien, has visited the site, and has met with both the applicant and the staff.
A copy of the ACRS report to the Commission on the Russellville Nuclear Unit is included as Appendix A.
A chronology of the principal actions relating to the processing of the application.is attached as Appendix B to this report.
The review and evaluation of the proposed design and construction plans of the applicant prior to construction constitute the first stage of a continuing AEC review of the proposed facility.
Prior to issuance of an operating license, the Commission's regulatory staff will review the final, as-built, design and operating features to determine that all of the Commission's safety require-ments have been met. The unit would then be operated only in accordance with the terms of the operating license and the Commission's regulations, and under the continued surveillance of the' Commission's regulatory staff.
The issues' to be considered, and on which findings must be made by an Atomic Safety and Licensing Board before the requested construction permit may be issued, are set forth in' the Notice of Hearing published in Federal Register on September 20, 1968, 33 FR lh243
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2.0 SITE AND PLANT DESCRIPTION 2.1 Site Description The Russellville Nuclear Unit will be constructed on an 1100. acre site located on a peninsula in the Dardanelle Reservoir on the Arkansas River in Pope County, Arkansas approximately 6 miles from the town of Russellville (1967 population,11,154) and 2 miles-from the village of London (1967 popu-lation,495).
An exclusion area with a radius of 0.65 miles (3430 feet) from the reactor
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has been. established for this plant.
All land within this radius, except for the bed and banks of the Dardanelle Reservoir is owned by the applicant. The bed and banks of the reservoir are controlled by the U.S. Army Corps of Engineers. The applicant has obtained an easement from the Corps of Engineers for the area which will permit it to exclude all persons from this area in the event conditions at the plant warrant such action. The applicant has specified a low population zone' (LPZ), as defined in 10 CFR 100, of. 4 miles.
The area around the site is largely undeveloped.
In 1964 practically no land was under cultivation out to 4 miles; out to 10 miles less than 0.4 percent was under cultivation.
In 1964 approximately 20% of the land out to -5 miles and 27% of the land out to 10 miles from the site was classed as pasture land.
The nearest population' center with over' 25,000 people is Hot Springs,- Arkansas, 55 miles south of the site. The applicant has estimated a 1967 population of
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3146 within '4' miles (LPZ) and 22,993 within 10 miles of the site.
Projections of the total population within these distances have been made by the applicant for the: year 2012 and are given as 5700 and 34,827, respectively.
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- The meteorology of the site is typical of continental locations, with lighter wind speeds and slower diffusion conditions at night than during the day.
The. site is in an area with appreciable tornado activity with kl tornadoes reported per 1 dedree squareb! over a 45-year period (1916-1961).
With respect to hydrology, the maximum probable flood, as computed by the Corps of Engineers, combined with failure of the upstream dam will flood the reactor site to 361 feet or 8 feet above plant grade level.
An onsite pond which will provide the _ source of emergency cooling water will be available in the unlikely event that there is a loss of such cooling water from the Dardanelle Reservoir.
In terms of geology, the site is near the axis of the Scranton syncline, one of several westward-trending gentle folds that characterize the Arkoma Basin--a major structural and topographic feature of Arkansas and eastern oklahcma.
The site is underlain by a thick sequence of gently-dipping shales and sandstones of Pennsylvaninn age.
overburden consists of alluvial clay and stity clay that ranges in thickness from 13 to 23 feet.
No identifiable active faults or other recent geclogic structures exist that would localize earthquakes in the immediate vicinity of the site.
Although several ancient faults are associated with the folded structures in the area, none appear to have been tectonically active since latest Paleozoic time (about 230 million years ago).
' 1/. A 1 degree square as used here is that earth surface area bounded by 1 degree of lattitude and 1 degree of longitude. -.At the Russellville site a 1 degree square contains approximately 3000 square miles.
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. A sc=ewhat unique site feature is the buried natural gas transmission pipe line which crosses the site approximately 600 feet from the containment struc-ture. The line, which does not supply this facility, will cross 4 feet beneath the bed of the plant's discharge water canal.
A discussion of the acceptability of 'the site is given in Section 31.
2.2 Plant Description The Russellville plant will have a closed-cycle, pressurized-water nuclear steam system housed in a prestressed concrete containment building, a steam and power conversion system housed in an auxiliary building and an outside electrical switchyard.
It will also have those auxiliary systems and struc-tures required to safely operate and maintain the plant under normal and emergency conditions. These auxiliaries include a radioactive waste disposal system, fuel storage and handling facilities, emergency power systems, and
- other engineered safety features.
The principal features and design bases for the Russellville Nuclear steam supply system are essentially identical to those of the Metropolitan Edison Company's Three Mile Island. Nuclear Station, for which a construction permit has been issued by the Commission. The nuclear steam supply system consists of a pressurized water reactor, a reactor coolant system, and associated auxiliaries. The reactor coolant system consis'ts of two parallel recircula-tion circuits, each sending reactor coolant through a steam generator (reactor coolant side) where it splits and flows through two pumps and associated piping, back to the reactor vessel.
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_7 An electrically-heated, spray-cooled pressurizer is connected to one of the two flow circuits. The reactor core uses fuel rods of uranium dioxide pellets clad in Zircaloy-4 tubes. The fuel rods are supported in assemblies by spacing grids and fittings, and a perforated can, all made of 30h stainless
- steel, Reactivity is controlled by movement of control rods (Ag-In-Cd), clad a
.vith 304 stainless steel, and by varying the boric acid concentration in the reactor coolant.
The control-rods are positioned axially in the core by the use of electro-mechanical, rack-and-pinion rod drive mechanisms and tripped (gravity insertion for least reactivity) by deenergizing a magnetic clutch. The clutch design permits the drive motor to apply down-drive force should a rod not fall freely.
A control system monitors reactor system temperatures, pressure, flows, neutron flux and load demand, and adjusts reactor power, steam generator feedvater flow, and turbine throttle within prescribed operating limits.
A reactor protection system monitors reactor coolant system temperatures, flows, and pressure, core neutron flux startup rate, and neutron flux level.
If an operating limit is reached, this system shuts down the reactor by releasing rod drive' clutches and allowing the control rods to drop into the core.
t The principal. engineered safety features are the emergency core cooling.
system (ECCS), the_ containinent ventilation system, and the containment spray -
i system (with chemical additives).. A protection system monitors primary cool-ant and reactor building pressures and vill. automatically initiate operation of the engineered safety feature systems if preestablished safety _ limits are reached.1
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o 8-The containment structure vill be a steel-lined, prestressed, post-tensioned concrete, vertical cylinder with flat bottom and shallow domed roof. The con-tainment is of the same basic design as that used by Bechtel for the Commission-licensed Duke Power Company Oconee Units 1, 2, and 3, the Florida Power & Li ht Company Turkey Point Units 3 and 4, and the Consumers Power Company Palisades Plant. The design details of the Russellville containment differ from this basic design in that the design details provide for a modified prestressing system using three vertical buttresses and 240 -span horizontal tendons rather than m
0 for a prestressing system using six vertical buttresses and 120 -span horizontal tendons.
All penetrations will be pressure-resistant, leak-tiSht, velded assemblies.
Personnel hatch openings vill have interlocked double doors, the equipment hatch will have a double-gasketed, bolted door and an isolation system vill be I
provided to close all fluid lines that penetrate containment and are not required for cperation of engineered safety features.
The emergency core cooling system (ECCS) vill be designed to provide core cooling for any location and size primary coolant pipe break, up to and including the double-ended rupture of the largest pipe--the 36-inch reactor outlet pipe between the reactor pressure vessel and the steam generator. The ECCS will consist of two operating and one spare high pressure injection pumps, two core flooding tanks (accumulators) and two low pressure (decay heat) pumps.
A recirculation system using the two low pressure pumps, returns water from the containment su=p to the ECCS. The Russellville ECCS does not differ in concept
- or capacity from the ECCS reviewed and approved for Metropolitan Edison's Three Mile Station-Unit 1.
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An emergency containment spray system vill provide bcrated water contain-ing dissolved sodium thiosulfate and sodium hydroxide to limit containment accident pressure (by heat removal) and to remove iodine (by chemical action) in the event of an accidental energy release from the primary system. A containment ventilation system, consisting of three fin-fan air coolers, is used to maintain containment temperatures at nor=al values during normal plant cperat' ions.
During accident conditions either the coolers alone or the core spray system alone vill be capable of keeping the accident pressure within the decign limit.
The major plant auxiliary systems are the chemical and volume control system, the vaste disposal system and the fuel handling system. The chemical and volume control system is used to adjust the concentration of the chemical neutron absorber (boric acid) in the reactor coolant and to maintain the proper amount of water in the primary system. The vaste disposal system is used to accumulate radioactive gases, liquids and solids from plant operation, process the radioactive vastes, and control and monitor the release of radio-active gases and liquids frem the plant to the air and to the reservoir respectively. The fuel handling system includes equipment and facilities designed to transport spent fuel under water from the reactor to the water-filled spent-fuel storage pool from where the spent fuel vill be shipped to an offsite processing plant.
The closed steam-feedvater cycle of the steam and power conversion system removes heat energy from the reactor coolant in the two cnce-through steam generators in the form of steam, converts steam energy into electrical i
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. energy in passing throu6h the turbine generator, condenses the steam into feedwater which is purified, chemically controlled for opti:m2m pH and minimum oxygen content, preheated, and recycled to the steam generators.
The condenser circulating water system condenses the steam leaving the turbine generator unit in the main condenser. The pumps for this system will withdraw water from the Dardanelle Reservoir by way of an intake canal and pump it through submerged conduits to the main condenser, and thence back to the reservoir through submerged conduits and a discharge canal. Cooling water for vital plant functions, which must remain operable in the event of an accident, vill be supplied by the service water system.
This system will drav vater from an intake structure which is normally supplied through the intake canal from the Dardanelle Reservoir. The service water portion of the intake structure can be isolated-from the intake canal and be gravity-fed by submerged piping from an elevated emergency cooling water pond to be con-structed on the site.
Onsite emergency power to operate post-accident emergency core cooling systems, the containment cooling systems, and other vital systems vill be supplied by two.2750 kW diesel generators. Two separate 125 volt d.c. systems, complete with charged storage batteries, will also be provided to supply vital instrumentation and provide emergency lighting and switching power.
There are two independent offsite sources of power.
Offsite power can be provided automatically upon loss of the main generator, through one of two transformers from a 161 kV transmission system which will be supplied power
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over separate lines from different sources. Offsite power can-also be.provided
-autcmatically in.a similar fashion from a 500 kV transmission system which can.also bring power into the plant over. separate lines from two different sources.
30 IMPORTANT SAFETY CONSIDEPATIONS In our evaluation of this application, we have given special consideration to a number of site and design features which are new, unique, require continu-ing evaluation, or have important safety implications. The more important of these safety considerations are discussed in the following sections.
31 Suitability of the Site In evaluating this reactor site, we have considered the following aspects:
t the characteristics of the proposed reactor; the containment capability; the nature and amount of radioactive vaste products generated; the site character-istics relating to meteorology, hydrology, geology, and seismology; abnormal weather conditions, such as tornadoes and floods; the population distribution in the surrounding area; and the potential radiation exposures at the site bound-l ary and offsite as a consequence of any of the postulated design basis accidents.
The area around the site is sparsely populated; however, the site does present one potential problem related' to evacuation of the few persons on the
~ Bunker Hill section of the peninsula which extends into the Dardanelle Reservoir.
Since the land evacuation route for these people vould be across the applicant's property inside the exclusion area, che' applicant vill provide boats to evacuate these persons by water if the land route is unsafe.
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The applicant states that no water is removed for either industrial or potable purposes downstream between the plant and the Ndssissippi River.
To establish background radiation levels, the applicant has outlined an environmental program which vill be initicted 12 to 18 months prior to opera-tion of the Russellville plant. This program will include onsite monitoring of radiation exposure levels and radionuclide concentrations in soil, vegetation,
' lake bottom, vater, fish, and air.
Offsite monitoring vill include analyses of milk, pasture forage, truck crops, and public water supplies. The applicant has consulted with various state and federal agencies in establishing this program.
The applicant's program has been reviewed by the Fish & Wildlife Service (Appendices C1 and C2).
The Fish & !!ildlife Service has recommended that the applicant's program include pre-and post-operational survey studies regarding specific radionuclides and. their.effect on selected organisms indigenous to the area. On the basis of our review of supplementary information submitted in response to Question 2 9 in Supplement No. 3 of the PSAR, we conclude that the applicant intends to comply with these recommendations of the Fish &
Wildlife Service.
We conclude that with the incorporation of these recommendations, the
. applicant's proposed program is acceptable.
On the basis of available data, we conclude that the site meteorology does not present any unusual problems. However, to supplement and verify the exist-ing data, the applicant has indicated that an ensite meteorological measurement 4
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We find the scope of this program to be acceptable.
The applicant's meteorological assumptions relating to site diffusion factors are considered to be adequately conservative.
This finding is based on inde-
' pendent adalyses performed by the staff and by the Environ = ental Science Services Aiministration, whose ecmments are attached as Appendix D.
To meet our safety criteria, certain aspects of the site required further definition and/or changes to the material originally presented in the applica-tion. These matters are discussed in the following paragraphs.
The applicant added an emergency cooling water pond on the site to ensure, in the unlikely event of failure of the Dardanelle Lock an
- Dam, a continued source of emer6ency cooling water for vital plant functions.
The onsite gas transmission line, described in Section 2.1, has been evaluated for effects on the Russellville plant.
The buried line, at its nearest point, is 600 feet from the reactor containment building and is k feet below the bed of the discharge canal. The applicant has calculated the energy potential for this line due to an explosive rupture and that due to ignition of gas dis-charged from the open line.
Neither the explosive rupture nor the radiant energy from gas ign!. tion at the break are considered capable of damaging this facility.
The applicant has further indicated that, should such events occur, the gas line owner. will close control valves on both sides of any break in the plant vicinity
-within 2' hours of notification. The applicant has stated that priar to plant operation, the existing pipe will be replaced with pipe constructed to Type C
- specification of-ASA Code L 31.8 for a distance of 600 feet on each side of the crossing.
On the basis of our review of this information and analysis, we do not-consider this pipe 'line to be a significant hazard to the safe operation of tLis plant.'
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. The geology of the site was found to be generally favorable by ua and the U.S. Geological Survey, whose report is attached as Appendix E.
In summary, 4
cur review shows that the site is underlain by shales and sandstones of Pennsylvanian age. Overburden consicts of alluvial clay and silty clay that ranges in thickness from 13 to 23 feet. no identifiable active faults or other recent geologic structures exist that would localize earthquake in the immediate vicinity of the site. The limited subsurface data available indicate that the major units of the nuclear facility will be founded on a hard, dense shale which should provide an adequate foundation.
Considering the site geology, soil conditions and earthquake history, the U.S. Coast & Geodetic Survey (USC&GS) and we concluded that an accelera-tion of 0.1 g would adequately represent earthquake disturbances likely to
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occa within the lifetime of the facility and that an acceleration of 0.2 g votad adequately represent the ground motion from the maximum earthquake likely to affect the site. The applicant will use these parameters in the seismic design of all Class I structures and systems. The USC&GS report is attached as Appendix F.
'The applicant'a original design criteria considered tornadoes having a tangential velocity of 300 mph, translational vind velocity of 40 mph, and a
' barometric pressure drop of 3 psi in 5 seconds. Following discussion with the regulatory staff, the applicant agreed to change these criteria to design for a tornado having a translational wind velocity of 60 mph and a barometric pressure drop of 3 psi in 3 seconds.
Design basis missiles equivalent to a h-inch.by 12-foot plank traveling end-on with a velocity of 300 mph at any l
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height and a h000-lb auto traveling through the air with a velocity of 50 mph at a height of 25 feet or less are proposed. These values are consistent with valuss' used by other nuclear plants recently approved for construction in areas having a significant history of tornado activity and, in our judgment, are reasonable design criteria. - We conclude that the tornado design bases including the effects of tornado-generated missiles are acceptable.
In the unlikely occurrence of the maximum probable flood concurrent with the failure of the upstream Ozark Dam, the site would be flooded to a level of 361 feet which is 8 feet above plant site grade level. The applicant has considered this in the facility design and has stated that all vital equipment including service water cocling pumps either will be Incated above maximum probable flood level or will be protected by waterproof Class I structures.
We therefore conclude that the applicant will provide adequate flood protection for this facility.
We conclude that the applicant has adequately considered the important characteristics of the proposed site.
We find the proposed site to be accept-able.
32 Acceptability ' f the Nuclear Steam Supply System Design o
The reactor design characteristics for the -Russellville Plant are essentially-the same as those for the Commission-approved Tpree Mile Island, Crystal River, and Rancho Seco plants. As in those f sn's, 13eration will be at 2452 Mw thermal with a mattimum fuel burnup of 5 ;.,
jgawatt-days per metric ten of uranium (Mwd /FE'U).
During-part of the first' fuel cycle the core is predicted to have a slightly positive moderator temperature coefficient of reactivity.
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~ calculations indicate that, with this coefficient, the core could withstand a loss-of-coolant accident and not exceed 20000 F peak fuel clad temperature.
An ' acceptable. final design value of the positive moderator temperature coef-1 ficient vill be set at the operating license stage.
The applicant has agreed to reduce or eliminate this positive coefficient, if nece'ssary, to bring the consequences of.the applicable a;cident within acceptable limits.
B&W has provided for the evaluati.on of xenon oscillations and in-core neutron detectors in its research and development 2rogram. To date, calcula-tions have been performed which indicate that xenon oscillations are not expected in the azitathal or radial direction, and are not likely in the axial direction at any time during the initial fuel cycle.
Further analyses will be made using final values of core properties. Calculations have also been made to show feasibility of controlling a divergent xenon oscillation using part-
' length control rods.
Since xenon oscillations are relatively slow flux varietions which could be detected by the proposed in-co' e flux instrumenta-r tion, we believe that such a control technique is feasible and could be provided.
The above-mentioned in-core flux inctrumentation consists of 52 fixed-position self-powered flux detectors distributed th.roughout the core. Normal readout is provided by the plant computer. Data obtained from this system vill provide a history of fuel burnup, power distribution, and power disturbances during' operation. -In 'the event that the plant computer fails, there is an alternate realout system for selected in-core detectors.
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With respect to the ther=al-hydraulic parameters and design features, our review revealed nothing new or different from recently authorized pressurized water reactors.
However, as noted in Section 4.0, additional analytical and experimental verification to support the ch'oice of the fuel damage limit, the use of stainless steel shims and the use of part-length rods will be obtained before the Russellville plant receives an operating license.
W'e have reviewed the applicant's seismic design bases pertaining to the reactor vessel, reactor internals, and other Class I (seismic) mechanical systems and components. These systems will be designed to withstand normal design loads of mechanical, hydraulic, and thermal origin, plus applicable earthquake loads, as well as concurrent accident-induced blowdown loads.
Our evaluation cf the proposed design criteria for reactor internals and Class I mechanical systems and components indicates that they will provide an adequate margin of safety.
One aspect which we are reviewing in detail is that of thermally-induced stresses in the pressure vessel during actuation of the emergency core cooling system. The initial results of the applicant's analysis of this accident indicate that no loss of vessel integrity would be experienced even if large flaws were presumed to exist in the vessel wall at the beginning of the quenching.
However, in viev 'of the uncertainties associated with the analytical methods used to arrive at these results, the applicant plans to continue his work on this problem.
While there remain uncertainties in the analyses being pursued, it is important to note that there is a significant time available (about 5 years) until material properties will be affected by irradiation to an extent that vill
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be of concern.
Further, it appears that there are means that can be employed, if necessary, to : _ ace the potential for vessel failure resulting from thermal shock and to mitigate the consequences of such a failure should it occur.
As recommended by the ACRS (Section 5 0), we vill continue to reviev information subsequently developed concerning thermal shock on the pressure vessel to ensure that the calculational models used are not in conflict with experimental data.
Provided that the development program substantiates the reactor design 4
characteristics discussed above, we conclude that the design of the nuclear steam supply systen is acceptable.
33 Engineered Safety Features Adequacy Engineered safety features for this plant include the emergency core cool-ing system (with reactor vessel internal vent valves), the containment ventila-tion systems, and the containment spray systems, and associated iodine removal system.
The emergency core cooling system (ECCS) is described in Section 2.2 of this report. The applicant's design basis is the same as that of Crystal River and other Babcock & Wilcox-designed systems recently reviewed. That basis is to prevent fuel clad melting for the entire spectrum of reactor coolant system fail-ures from the smallest leak to complete severance of the largest reactor coolant pipe. To provide assurance that this criterion is met and to prevent any mechan-ical damage that might interfere with core cocling, the applicant has sized the emergency core cooling systems to limit the clad temperature transient to 2300 F or less. The calculated peak clad temperature, about 1950 F, occurs transiently 2
during the postulated hot-leg break (a 14.1 ft break).
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We have. reviewed the applicant's failure mode analysis of the ECCS and have concluded that adequate short-term cooling can be provided at high and
. low vessel pressures even in the event of failure of any single active compo-nent.
In addition, adequate redundancy is provided to accommodate failure of a single active or passive component without jeopardizing the ability for long-term core cooling with the ECCS in the recirculation mode. To achieve this, the applicant revised his originally proposed ECCS design to provide two systems with no sharing of active components and minimum practical sharing of passive components. This applicant's ECCS as revised is the same as those systems previously reviewed. The result is that there are two separable core cooling systems which share only the passive borated water storage tank, core flooding tanks, and containment building sump.
Sharing of the tanks is acceptable since they are in use for. only a short period of time.
Sharing of the reactor building sump is acceptable since the recirculation lines for the two systems take suction from different locations of the sump, the sump is covered with a grating and heavy duty strainers are provided.
As was done in the B&W nuclear steam supply system design provided for the Three Mile Is1hnd, Crystal River, and Rancho Seco plants, the Russellville design incorporates one-way internal vent valves in the reactor core barrel to prevent steam binding above the core.
In the event of a loss-of-coolant accident initiated by a break in a cold leg of a reactor loop, the valves will open to permit steam generated in the core to flow directly to the leak and thus not prevent the emergency core coolant system from keeping the core adequately covered. These valves have been previously authorized for use in the Three Mile Island plant.
B&W has made a preliminary sensitivity analysis using vorst case parameters
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to show how loss of core flow, by shunting reactor coolant through a failed l
(open) valve, affects the DNB ratio (design limit is 13 or greater). The I
preliminary analysis shows the reduced-flow DNB ratio is 1.68 at 100% power, 1 30 at 112% power and 1.2h at 114% power (the highest thermal power calculated in any operational transient). An analysis based on the final design of the core is expected to meet the 13 DNB ratio design requirement at 114% of rated power as well.
W'e also considered the ability to detect, by change in measured reactor coolant loop flow, the failure of more than one vent valve. Based on the preliminary design data supplied by B&W, the total system flow is increased by 1.1% by failure of one valve. The applicant has stated that flow distribu-tion studies will be made using a model of the reactor to simulate failure of the vent valves.
Completion of valve testing, including vibration tests, is expected by January,1969.
At the operating license review on this plant, or earlier, we vill evaluate the results of these tests and verify the ability to identify failure of the vent valves by detection of changes in reactor coolant flow.
We conclude, at this stage of our review, that the vent valve design is satisfactory subject to completion of the final design, design analyses, testing, and verification of ability to use flow change to detect failure.
Two diverse methods are provided for containment heat removal under accident conditions:
(1) two 120 x 10h Btu /hr capacity containment spray syste=s, each of which takes relatively cool water (initially from the borated water storage tank and later from the containment sump) and delivers it to the containment
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atmosphere through a spray header and (2) three 80 x 106 Btu /hr capccity-containment cooling systems, each consisting of a fan and tube cooler, which removes heat from the containment air and transfers it to the low-pressure service water systen.
The containment cooling requirement is that the post-blowdown reactor building pressure be maintained below the containment design pressure. This 6
requires an initial heat removal capacity of about 240 x 10 Btu /hr. This requirement can be satisfied if either all sprays or all containment cooling systems are assumed to be inoperative.
It can also be satisfied if one spray and one cocler are inoperative. On the basis of our review of these systems, we conclude. that adequate capacity has been provided to initially limit and subsequently reduce the containment pressure (and thereby reduce leakage) after the design basis accident, in the event such an improbable accident should occur.
A chemical additive (sodium thiosulfate with sodium hydroxide) vill be mixed with the spray water to remove iodine from the containment at=csphere i
following a loss-of-coolant accident. Two spray systems are provided as dis-cussed above.
Each spray system has the desi5n capability to deliver an adequate amount of the chemically treated spray to the containment atmosphere to prevent exceeding 10 CFR 100 guidelines for potential radiological doses at the site boundary and' at' the low population zone boundary.
Section 3 7 gives the calcu-
. lated doses using a single spray system and also states that, in the event additional chemical iodine spray tests now underway indicate that the spray system is not as effective as anticipated, iodine reducing charcoal adsorber units can be added to remove iodine.
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l The service water system shown on Figure 9-h of the PSAR provides all water required for emergency cooling of the engineered safety feature equip-l ment including the containment building coolers and the emergency diesel 1
generators. Redundant pumps and piping and an emergency reservoir are provided such that no single failure can cause loss of required cooling.
34 Foundation and Structural Design Adequacy In evaluating the foundation and structural design of the plant structures, we and our consultaat,b/ considered the following general aspects: the geology and nature of the subsoils, the seismic design parameters, site finading, tornado vind loadings, and the effects of missiles generated from tornadoes and internal plant sources.
We considered the following specific aspects in our evaluation of the containment and other Class I structures: design criteria, specifications and inspection for concrete reinforcing, selection of loads, load combinations and allevable stresses for the structure, liner and liner anc' orage criteria, tendon and tecdon anchorage criteria, design of penetrations, and containment strength and leak testing.
All structures and equipment required for plant safety and to maintain the integrity of engineered safety feature systems have been designated as Class I.
All other structures and equipment are Class II.
All Class I structures will be designed to behave elastically under normal and accident loads, except that
-limited yielding vill be permitted under a combination of dead load, piping thermal shock or rupture, and design-basis earthquake (0.2 g).
Class II struc-tures, which do not perform vital safety functions, will be designed to Zone 1 requirements of the Uniform Building Code. Class II equipment will be designed for an equivalent horizontal loading of 0.05 g.
_ lj Nathan M. Newmark Consulting Engineering Services.
See report attached as Appendix G.
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. The containment building, as noted earlier, is similar.to other Bechtel designs including the design for the Rancho Seco plant, except that heavier tendcas and three instead of six buttresses are used. The cylinder has 0
staggered 240 -span instead of 120 -span horizontal tendons.
The vertical and the dome tendon systems are similar to those used in previous designs, except for anchorage designs and tendon sizes.
In response to our questions on several aspects of structural design, the applicant provided additional supporting details on methods of analysis, and construction details.
We and our consultants have reviewed the proposed tendon systems tenta-tively selectea by the applicant.
We conclude that use of the tendon systems proposed, with up to 184 wires per tendon, would be acceptable.
The liner anchorage design is similar to that proposed for the Rancho Seco plant. The liner anchorages are designed to fail before the liner itself can fail.
We have expressed concern that, with the liner in compression and tending to buckle locally, anchors may fail rapidly and sequentially.
On the basis of our review, we do not believe the analyses presented in the PSAR are
-conclusive.
We have discussed this with the applicant and (as noted in Section h) prior to construction we vill 6btain confirmatory test specimen data that deal with gross liner failure considerations.
-In the tendon anchor zone, we are concerned that sufficient reinforcing be included in the design to cover all possible tension stresses that may exist in this-cone. The usual design methods ne61ect two potentially significant tensile stresses, those generated by temperature gradients and by concrete
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. shrinkage.
As noted in Secticn 4, prior to construction of these anchorages, we will obtain from the applicant analyses and qpalification test data to i
confirm design adeqyacy.
The problem of tornado-induced loss of water from the fuel pool leading to fuel melting and-fission product release is of continuing concern.
We have examined the analysis provided by the applicant in this regard and find that it contains no-new information or arguments that have not been presented in previous ~ applications.
We are continuing to examine the requirements for spent fuel pool design and. <e conclude that the design of the fuel storage pool should be such that protection of the pool from water 1 removal effects could be added if this is found necessary. The applicant has agreed to provide this capability in the design of ',he Russellville fuel storage pool.
The applicant has proposed e 2% statistical sampling program for strength testing of the Cadweld reinforcing bar splices to be made in the structures.
Since this may result in a small number of welds being tested, we are examining the area further.
In the event that a modified testing program is considered necessary requiring a larger number cf welds being tested or placing more emphasis.cx1 selected weld locations, we conclude that these relatively minor changes can be agreed upon with the applicant prior to the actual placing of these welded splices in the structures.
From our in-depth review, we and our seismic design consultant conclude that the containment, foundation and general structural designs proposed for
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. the Russellville plant are acceptable except for the submission of confirmatory data on liner and tendon anchorages and Cadweld splice tests.
These items have been left for later consideration as discussed in Section 4.0.
35 Adequacy of Instrumentation, Control and Energency Power Systems The ' instrumentation and control systems were evaluated and found to comply with the Commission's General Design Criteria (see Section 3 8) and IEEE 279, Proposed Criteria for Nuclear Power ~ Plant Protection Systems. A comparison was also made with the systems proposed for the Three Mile Island Nuclear Station, and Crystal River Unit No. 3 The applicant has verified, and we concur, that the proposed design of the instrumentation and control systems for the Russellville plant and the above mentioned plants are substantially identical in concept except that the Russellville plant (a) uses not one, but all four of the redundant reactor power level channels in an averaging system as inputa to reactivity control; (b) initiates reactor trip upon loss of any two pumps while the other plants utilize systems which permit continued operation with the loss of one pump in each loop provided p "ar is below a predetermined safe limit; (c ) supplements reactor coolant systems code safety valves with a pilot actuated relief valve which is not provided in the other plants; and (d) varies boiler feed pump speed as the major means of controlling feedwater flow as opposed to reliance in the other plants solely upon feedvater valve centrol. The differences noted' in (b), (c ) and (d) above are considered to be minor and to have no significant effect on reactor safety.
In evaluating item (a) we examined the proposed design and found it to be in compliance with IEEE 279 In particular, the protection system-has four redundant power level channels. The randem
,, failure of any one channel leaves three for protection, only two cf which are required.
While these channels are also connected to the plant's reactivity control-system, a single random failure in any one channel is prevented from i
causing a control failure by isolation devices and by the manner in which they are combined. Further the applicant reports that tests have been success-fully performed simulating open circuits, short circuits, grounds, and faults to high voltages with no failures propagating beyond the channel in which the simulated failure was imposed.
As a result of our evaluation of item (a) we conclude that the design provides satisfactory protection against random failures.
We vill continue to work with the applicant to ensure that it takes into account, in completing the desiEn of protection and control instrumentation, the possibilities of common failure modes such that by the suitable use of redundant devices with functional and equipment diversity, the proposed interconnections of protection and control instrumentation vill not adversely affect plant cafety.
The control room contains instrumantatien and controls necessary for safe operation of the nuclear facility.
Safe occupancy of the control room during al ormal conditions is provided for in the design.
In the event the control room becomes uninhabitable, sufficient instrumentation and controls are provided at local stations which permit the operator to maintain the reactor in a hot standby condition. 'iurther, the applican; has stated that the capability to perform an orderly cold shutdown from outside the control room, should this rocm become inaccessible for a long period of time, will be provided.
We conclude that.the control room design bases meet the intent of Criterion 11 of the General Desi6n Criteria.
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. The applicant has established criteria for the selection, protection, and routing of all control, power and instrumentation cables.
We conclude that adequate measures will be taken to prevent and minimize the possibility of fire or other damage in electrical cabling.
We have evaluated the proposed offsite and onsite electric power systems and have concluded that they comply.vith Criterion 39 of the General Design 1 Criteria.
In its letter on the Three Mile Island Nuclear Station, the ACRS recom-mended that consideration be given to the development and utilization of instrumentation for prompt detection of gross failure of a fuel element.
The applicant has indicated that-it vill provide continuous radiation monitors in the reactor coolant makeup and letdown line and in the containment atmosphere sample line with. sufficient sensitivity to promptly detect a gross fuel element failure.
Information on the response time as a function of fuel failure severity will be made available during the detailed 'esign of the plant.
We vill review this matter on other plants scheduled for operation before the Russellville plant, and at the operating license stage review of the Russellville plart.
On'the basis of the foregoing, we have concluded that the reactor intcru-mentation, control, and emergency power systems are acceptable for this construc-tion permit stage of review.
36 Radiesetive Waste Dispesal Adequacy The radioactive liquid vastes generated in normal plant operations will be collected, stored, treated, measured for activity, and discharged on a batch
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basis.vith continuous moni'oring during discharge through a line to the plant's circulating water discharge canal. Gaseous wastes will be collected, monitored, diluted and released to the atmosphere.
If the activity levels exceed precribed limits, the gases will be compressed and stored in vaste gas decay tanks. Follow-ing decay, the stored gases vill again be monitored prior to release to assure that release is within prescribed limits. Solid radioactive wastes accumulated from plant operation will be temporarily stored onsite.
Shipment from the site vill be in containers approved for that purpose.
We reviewed the possibility of activity release due to system failures.
The solid and liquid disposal equipment is located in shielded, controlled-access-areas of a. Class I structure with provision for contamination control in the event of spills-or leakage. Calculations by us and the applicant indicate that failure of a vaste gas tank containing maximum activity would result in whole body doses of less than 2 rem at the site boundary which is well below 10 CFR 100 limits.
On the basis of our review, we conclude that the proposed radioactive vaste disposal system will adequately control the radioactive vastes generated from plant operations.
37 Analysis of Radiological Consequences from Potential Accidents Potential accidents which could result in radioactive releases to the environment have been analyzed by the applicant.
We have evaluated these acci-dents and the engineered safety features provided to mitigate or limit the potential offsite exposures. Accidents which have been considered are: the loss-of-coolent accident, the rod-ejection accident, rupture of a steam pipe,
i
.. rupture of a steam generator tube with loss of offsite power, fuel-handling accident, accidental release of radioactive liquid and gaseous waste, and rupture of a recirculation line in the eme.rgency core cooling system. Of those accidents considered to have a potential for significant releases of radioactivity to the environment, the loss-of-coolant accident would result in the highest potential offsite doses.
For accidents involving loss of coolant from the primary system, the emergency core cooling systems are designed to limit fuel cladding temperatures to well below the melting temperature, to prevent shatter of the fuel cladding, and to limit fission product release from the fuel. However, for conservatism we assume that the containment and its associated engineering safety features must be capable of limiting-potential doses in conformance with 10 CFR Part 100_ guidelines assuming releases of fission products from the fuel based on TID-1484k release fractions. Using these fission product release fractions available ~ for leakage from the containment, and assuming ground release, conservative meteorological diffusion parameters and design data on the con-tainment sprays, we calculated potential doses at the exclusion boundary and the low population zone radius.
Utilizing conservative values for drop size spectrum and deposition velocity and the specific characteristics (e.g., ' droplet size, flow rate, fall distance, terminal velocity of drop) of the Russellville plant's iodine removal system, we have calculated that iodine removal factors of 4.1 for the 2-hour dose and 10 for the 30-day dose are achievable by the sprays. These dose reduction factors assume as much as 10% of the iodine in 1/ _ TID-14644, Calculation of Distance Factors for Fower and Test Reactor Sites,
- din em ), J.'
J., et al, March 23,'1962.
s the containment is in "nonremovable" (organic ) form. Allowing these dose reduction factors for iodine removal, the potential 2-hour doses at the exclu-sion area boundary (0.65 miles ) are 4 rem whole. body and 210 rem to the thyroid and the 30-day doses at the low population zone radius (k miles) are about 2 rem body and 81 rem to the thyroid. The applicant has stated that analytical and experimental work on the efficiency of chemical additive sprays is being conducted by B&W, Oak Ridge National Laboratory and others.
In addition to sodium thiosulfate, other chemical solutions are also being evaluated.
In -
the event that the results of these development programs indicate that the spray systems might not be as effective as anticipated, the applicant has stated that space will be reserved in the plant so that charcoal adsorber units can be'added to further reduce the iodine concentration in the containment.
38 Design Conformance to AEC General Design Criteria The applicant has assessed the Russellville Nuclear Unit. design with respect to conformance with the Commission's General Design Criteria published in the Federal Register on July 11, 1967 We have evaluated the application for conformance with the revised criteria and have ecncluded that the preliminary design of the proposed unit conforms to the intent of these criteria. Recog-nizing that the proposed criteria, as revised, may be further modified as a result of comment by interested parties, and that the final design may differ
-somewhat from the preliminary design, vu intend to review the proposed unit for conformance to the General Design Criteria again at the operating license stage.
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. 3 9' Emergency Plans The scope of emergency planning by the applicant, including proposed preparation of written procedures covering reasonably foreseeable emergency operating conditions, is acceptable. Detailed-emergency plans for the low population zone will be developed by the applicant in cooperation with state and local authorities. We will evaluate these plans at the operating license review stage.
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4.0 RESEARCH AND DEVELOPMENT There are a number of areas related to pressurized water reactors for which additional research and development will be required. These areas are summarized in this section. We will follow the programs listed below by meet-ir.g with the applicant and his contractors and by evaluating reports submitted on these programs.
(Expected completion dates are parenthetically noted).
(1) B&W Development of the Emergency Core Cooling System Design The core cooling research and development being conducted by B6W, must specifically include (a) the coupletion of the analysis of the spectrum of scall break sizes in the loss-of-coolant accident, (b) the development of the analytical techniques for determining blowdown forces on reactor internals, and (c) demonstration that the injection coolant will cool the core including cc sideration of core bypass or formation of a vapor lock. Experimental vibration tests will also be performed to show that induced-vibration will not unseat the core barrel vent valves.
(July 1969).
'(2)
B&W Development of Final Reactor Thermal-Hydraulic, Nuclear and Mechanical Design Parameters Development work to be performed includes the following:
a.
Thermal and Hydraulic Programs The applicant has proposed scaled flow distribution tests on the vessel and internals and rod bundle tests to determine local mixing and flow effects. This further experimental and analytical work must be done to determine the limiting heat fluxes at various positions within the fuel bundle if the design is to be based on the B6W heat transfer data.
(prior to 1969) 4 e
b.
Fuel rod failure mechanisms during loss-of-coolant-accident (LOCA).
Various failure modes of the fuel rods during the LOCA, such as clad melting, eutectic formation, bulging, splitting, or brittle failure, will be examined in an experimental program to assure the continue core cooling capability during a LOCA.
(late 1969).
c.
High burnup fuel tests Fuel specimens will be tested at heat rates ranging up to 21.5 kw/ foot, burnup ranging up to 75,000 MWD /MTU, and with cladding surface temperature of 650 F.
(June 1970).
d.
Xenon oscillations The applicant will further develop analytical techniques to determine the stability margins with respect to xenon oscillations (late 1969).
If the stability margins are found to be insufficient, a system for stabilizing and controlling the oscillations will also have to be developed. Results from physics tests on Duke Power Company's Oconee Unit I will be used to confirm the analytical results. (2nd quarter 1971).
(3) B6W Control Rod Drive Unit Tests The prototype tests are being conducted on the B&W control rod drive units under operating temperature, pressure, flow and water chemistry and should provide design adequacy information on the operability and reliability of the system.
(prior to 1969).
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(h) B6W In-Core Neutron Detectors Tests
)
The self-powered in-core neutron detectors, wh4.ch have been developed by B6W, are currently under life testing at B6W's Lynchburgh facility and at the Big Rock Point Nuclear Power Plant. The status of the tests to date are acceptable s
(5) B&W Once-through Steam Generator Development and Tests Investigations of steady-state conditions and operational transients have been completed. Vibrational tests, including vessel response to primary system blowdown, have also been investigated and the thermal response to both primary and secondary blowdown determined. The remaining work involves the development and verification of analytical models for steam system blowdown analyses.
(1st quarter 1969).
(6) B&W Development of the Design Details of Iodine Removal System (Chemical Additive to Containment Sprays )
The Russellville plant iodine removal system is being developed by B6W. Chemical characteristics, iodire removal characteristics, compatibility, and radiolysis of spray materials are being evaluated. Experimental investi-gation of the relationship of absorption rate of containment atmospheric conditions, the effects of process variables on spray nozzle performance and the extent of radiolysis are being conducted by B&W, Oak Ridge National Laboratory, and Battelle Memorial Institute. (early 1969) 4.1 Other Matters to be Further Evaluated During Construction (1) -Instrumentation There are two areas of instrumentation which will require further infor-l mation and review.
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. Design of the pro ~ pt fuel failure detectors a.
m The applicant has not yet completed the design 6f these detectors.
Upon completion of these detectors, which are to be of two types, one to sample reactor coolant (in the letdown line) and the other to sample containment air, we will review their design capability for adequacy and speed of response as a function of percent of fuel
- failed, b.
Interaction of control and protection systems As discussed in Section 3.5 we and the applicant will continue evaluation of the protection and control instrumentation systems with regard to interaction. In particular, we are reviewing the proposed design as it is finalized, for common failure modes, taking into account the possibility of systematic, nonrandom, con-current failures of redundant devices, not considered in the single-failure criterion.
(2)
Containment Design Details Three containment items have been selected for further evaluation prior to construction of the affected subsyv. ems. This information, which will be developed in the normal course of design, includes the design details and associated analyses for the tendon anchor system and for the liner
- anchorages.
For tendon anchorages, the applicant has agreed to submit a report giving both predictions and results of the tendon anchorage qualification test. This report, will identify analytical methods and material properties used in the predictions, results of actual tests and comparision of predictions
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with test results.
We plan to review this data as it becomes available as well as additional design information prior to construction of the tendon anchorages.
For liner anchorages, the applicant has agreed to perform tests demon-s strating his design vill not result in seqpential anchorage failures.
We plan to review these tests as well as additional design information prior to con-struction of the liner anchorages.
For Cadweld splices, we and the applicant will agree on the relatively minor changes, if any, reqpired in the statistical sampling strength testing program prior to use of such plices in the plant structures.
(3) Quality Assurance Information After the constructor has been selected and prior to starting any major construction at the site, we vill review the additional qpality assurance information, indicated in Section 6.2, which the applicant has agreed to submit.
(4) Reactor Vessel Thermal Shock As' discussed in Section 3 2 ve are continuing our review of the problem of thermal shock as a potential conseqpence of actuation of the core cooling systems.
h.2 Conclusion We have examined each of the above areas and conclude that they can reascnably be-left for later consideration.
Moreover, on the basis of the descriptions supplied by the applieant, we conclude that the proposed research and development programs are reasonably designed to resolve the identified safety questions.
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5.0 REPORT OF THE ADVIS0FY COMMITTEE CW REACTOR SAFEGUARDS The Advisory Committee on Reactor Safeguards, by letter to Chairman Seaborg, dated September 12, 1968, reported on the Russellville Nuclear Unit.
A copy of this letter is attached as Appendix A.
The letter contains comments and recommendations which we are implementing, as noted in appropriate sections of this safety evaluation.
The Committee has reiterated its belief that additional consideration be given to common mode instrumentation failures not considered in the single-failure criterion.
This is discussed in Section 3 5.
The Committee also emphasizes the importance of quality assurance and quality control programs, discussed in Section 6.2; and early training of a sufficient number of personnel for the operating staff, discussed in Section 6.1.
Modification of the containment prestressing system design is also mentioned.
This is discussed in Section 3.4.
The Committee further calls attention to other matters that warrant careful consideration by the manufacturers of all large, water-cooled, power reactors.
These matters, applicable to the Russellville plant involve the following:
effects of blowdown forces on primary system components, effects of fuel clad perforation on e=ergency core cooling performance, and fuel element performance under operational transients, all of which are addressed in Sections 3 2, 3 3 and 4.0 of this report.
Additional matters about which the Committee expressed concern include pressure vessel shock from cold water injection, discussed in Section 3 2; prompt detection of gross failure of a fuel element, discussed in Section 3.5; and primary system
. quality assurance, discussed in Section 6.2.
These items will be resolved to our satisfaction as the. design work progresses and will be reviewed by the
- ACRS prior to issuance of an operating license.
The report of the ACRS concluded,..... The Advisory Committee on Reactor Safeguards believes that, if due consideration is given to the fore-going items, the proposed reactor can be constructed at the Russellville site with reasonable assurance that it can be operated without undue risk to the health and safety of the public."
6.0 TECHNICAL QUALIFICATIONS J THE APPLICANT 6.1 Tbchnical Qualifications We have reviewed the application with respect to the technical qualifica-tions of the Arkansas Power and Light Company (AP&L) and its contractors to design and construct the proposed facility.
AP&L has over 45 years experience covering design, construction, and operation of conventional steam, hydro, and diesel-electric generating plants which, at the end of 1967, had a total capacity of 1,734 megawatts.
Officers and engineering personnel of AP&L have had previous nuclear experience through AP&L's participation, as a member of the Southwest Atomic Ehergy Associates, in the Southwest Experimental Fast Cxide Reactor Facility, SEFOR, and through AP&L's participation in the Peach Bottom Atomic Power Station project.
AP&L vill rely upon its architect-engineer, contractors, and consultants
-for technical support during the design and construction of the plant.
The Eechtel Corporation has been retained as the architect-engineer and will be responsible for procurement and management of construction of the plant.
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Bechtel has vide experience as architect-engineer and engineer-constructor for several pressurized va er reactor power plants as well as other types of nuclear and conventional power plants.
Babcock and Wilcox vill supply the nuclear steam supply system and two fuel. cores.
E&W has extensive back-ground in supplying nuclear steam supply syste=s.
The turbine generator and its auxiliaries will be supplied by the Westinghouse Electric Corporation.
The number of people proposed for operation of the plant totals 61.
Personnel assigned to the plant will have extensive experience in conventional power plants and all supervisory and operating personnel vill be given special nuclear training including operator training at a comparable nuclear power plant.
The applicant has planned for four-man operating shifts con-sisting of a shift supervisor with a Senior Operator's License, a plant operator and an assistant plant operator, each with an Operator's License and an auxiliary operator who may have an Operator's License.
On the basis of our review, we conclude that the applicant and its principal contractors have the technical competence to design and build the Russellville Nuclear Unit.
We believe, however, that h-man operating shifts may prove inadequate. We vill pursue this matter further with the applicant as it develops its emergency and normal operating procedures and will satisfy ourselve s that its training program vill assure timely availability of adequate operating manpower.
6.2 Quality Assurance and Qualit'y Control We have reviewed the quality assurance and control program proposed for the Russellville facility.
At our request, the applicant has supplemented its PSAR with additional information which is provided in Supplement 3 (ansvers to Questions 8.1 through 8.11, 9.5 and 9 7) and in Supplement No. 9
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The applicant's Safety Review Committee reviews all plant designs, specifications, and procedures to ensure compliance with all plant design criteria, codes and standards as set forth in the PSAR with responsibility and authority to reject those which are not in compliance.
The AP&L Manager s
of Safety, who reports directly to the Executive Vice President, is a member of this committee.
AP&L also has established a Quality Assurance Committee (QAC) for the Russellville plent.
A key member of the QAC is the Chief Quality Control Coordinator who will be in residence full time at the plant site during construction.
He will work closely with the Bechtel Quality Assurance Engineer, who will also be onsite during construction.
The Chief QC Coordinator will review all inspection and test procedures prior to in-spection or test, monitor tests and inspections at the site and at vendor facilities on a frequent " spot-check" basis and review the resulta of all quality control programs.
The QC Coordinator will be assisted in his duties by AP&L Engineering or Production Departrient personnel experienced in plant design and construction.
In areas where AP&L does not now have experienced personnel, they will eitt.er hire or obtain the services of such personnel through a consultant firn:.
In addition to the applicant's orgenization, Bechtel will have a Quality Assurance Engineer (QAE) under the Project Engineer and a separate field inspection force under a J6b Engineer, i The QAE will have access to and will review, for compliance with establishef requirements, all Bechtel and vendor i
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. quality control procedures and reports of all tests and inspections p r-formed by others in vendors' plants and at the job site.
The Bechtel field inspection force reports through the Job Engineer and Project Superintendent to the San Francisco Office Construction Manager while the QAE reports through the Project Manager and Nuclear Power Ehgineering Nhnager tc the San Francisco Engineering Manager.
Bechtel vill also have independent checks on quality assurance during the design and pre-fabrication phase by having design bases, designg and procurement documents, which are prepared by the Project Engineer's staff, reviewed by the staffs of Chief Ehgineers in each engineering specialty.
These Chief Engineers independently report directly to the San Francisco Office Manager Engineering.
B&W, the nuclear steam supply system vendor, has recently established in July of 1968 a quality assurance organization which will be responsible for quality assurance of B&W'a nuclear product line from bid proposal to final customer acceptance.
This organization, which is independent from the previously existing B&W design, production, and quality control groups, reports directly to the Vice President of the B&W Nuclear Power Generation Department and is responsible for assuring that the Russellville nuclear steam supply system furnished by B&W conforms to all established requirements.
.Upon selection of the general contractor for construction of the Russellville facility, the applicant has agreed to submit the following information:
(a) a list of all organizations involved in the design and l
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construction of this plant, (b) description of the various responsibilities of all organization including quality assurance and control, (c) a schedule of major construction activities, (d) a listing of responsible persons (plant site and vendor shops) as contacts for Division of Compliance inspectors, (e) location of complete specifications and quality assurance and control documents, and (f) a list of all major vendor shop locations.
Subject to our review of this additional information, we conclude that the applicant, together with its contractors, will have an adequate quality assurance program and that independent checks on quality assurance and quality control can be provided at all stages, from establishing adequate design bases initially, through design, fabrication, testing and final inspection.
7.0 COMMON DEFENSE AND SECURITY The application reflects that the activities to be conducted would be within the jurisdiction of the United States and that all of the directors and principal officers of the applicant are American citizens. We find nothing in the application to suggest that the applicant is owned, controlled or dominated by an alien, a foreign corporation or a foreign government.
The activities to be conducted do not involve any restricted data, but the applicant has agreed to safeguard any such data which might become involved in accordance with the regulations.
The applicant will obtain fuel as it is net.ded from sources of supply available for civilian
. purposes, so that no diversion of special nuclear material from military purposes is involved.
For these reasons and in the absence of.any informa-tion to the contrary, we have found that the activities to be performed will not be inimical to the common defense and security.
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8.0 CONCLUSION
T On the basis of the proposed design of the Arkansas Power and Light Company's Russellville Nuclear Unit; the criteria, principles, and design arrangements for systems and components thus far described, which include all of the important safety items; the calculated potential consequences of routine and accidental releasc of radioactive caterials to the environs; the scope of the development program which will be conducted; and the technical competence of the applicant and the principal contractors; we have concluded that the appiepriate findings as set forth in the notice of hearing of this proceeding, September 20, 1968, can be =ade by the Director of Regulation.
In summary, we conclude that the proposed plant can be built and operated at the proposed location without undue risk to the health and safety of the public.
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APPCDIX A l
l ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION I
WASH 1NGTON.D.C. 20545 l
SEP 141968 Honorable Glenn T. Seaborg Chairman g
U. S. Atomic Energy Coassission Washington, D. C.
Subject:
REPORT ON EDSSELLVILLE NUCLEAR UNIT
Dear Dr. Seaborg:
At its one-hundred-first meeting, Septanber 5-7, 1968, the Advisory Cor:snittee on Reactor Safeguards reviewed the proposal of the Arkansas Power and Light Company to construct the Russellville Nuclear Unit.
h is project had been considered previously during Subconunittee meet-ings on August 23,190 at the site, and on September 4,1968, in Washington, D. C.
In tra course of its review, the Consmittee had the benefit of discussions with representa?.ives and consultants of the Arkansas Power and Light Company, the Bechtel Corporation, the Babcock and Wilcox Company, and the AEC Regulatory Staff. The Conamittee also had available the documents listed.
The plant will be located about six miles from Russellville, Arkansas, on a peninsula formed by the Dardanelle reservoir. The normal eleia-tion of the reservoir is controlled downstream by the Dardanelle Lock and Dam No. 10 on the Arkansas River. An emergency reservoir en the site will provide adequate storage of water in the unlikely event of failure of' Lock and Dam No. 10. De consequences of the maximum prob-able flood have beef studied, and adequate protection has been provid'ad for the critical equipment of the nuclear unit.
W e proposed nuclear unit is a pressurized water reactor, 2452 MWe and 850 MWe, and is similar to previously approved units (e.g., Rancho Seco, Crystal River, and Three Mile Island, ACES Reports of July 19,'1968, May 15, 1968, and January 17, 1968, respectirely). The Committee con-tinues to call attention to matters that warrant careful consideration by the manufacturers of all large, veter-cooled, power reactors.
Se Coannittee reiterates its belief that the instrumentation design
- should be reviewed for common failure modes, taking into account the possibility of systematic, non-random, concurrent faiLares of redimdant devices, not considered in the single-failure criterion. The applicant.
(45)
Honorable Glenn T. Seaborg SEP 121968 should show t at the proposed interconnection of control and safety instrumentatio.. vill not adversely affect plant safety in a signifi-cant menner, considering the possibility of systematic component failure. The Coasmittee believes this matter can be resolved with the Regulatory Staff.
The containment for the reactor is a prestressed concrete vessel similar to previously approved designs (e.g., naneha Seco), but with modification of the prestressing syntes design.
The Comnittee emphasizes the importance of the implementation and management of the quality assurance and quality control programs necessary to achieve the design, construction,and operation objectives.
Inasmuch as a long lead time is required in the training of the operating staff, the Cossaittee emphasizes the need for early training of sufficiant personnel to assure adequats operating manpower.
The Advisory Committee on Reactor Safeguards believes that, if due consideration is given to the foregoing items, the proposed reactor can be constructed at the Russellville site with reasonable assurance that it can be operated without undue risk to the health and safety of the oublic.
Sincerely yours, Original Signed by l
Carroll W. Zabel l
Carroll W. Zabel Chairmen Enferences Attachen.
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Honorable Glenn T. Seaborg SEP 121968 References - Russellville Nuclear Unit 1.
Application for Licenses, Arkansas Power and Light Company Russellville Nuclear Unit, dated November 24, 1967.
2.
Volume I - Preliminary Safety Analysis Raport, Arkansas Power and Light Company Russellville Nuclear Unit, dated November 24, 1967.
3.
Volume II - Preliminary Safety Analysis Report, Arkansas Power and Light Company Russellville Nuciaar Unit, dated November 24, 1967.
4.
Supplement No. 1 to Application for Licenses, Arkansas Power and Light Company Russellville Nuclear Unit, dated January 22, 1968.
5.
Supplement No.'2 to Application for Licenses, Arkansas Power and Light company Russellville Nuclear Unit, dated February 14, 1968.
6.
Supplement No. 3 to Application for Licenses, Arkansas Power and Light Company Russellville Nuclear Unit, dated May 3,1968.
7.
Supplement No. 4 to Application for Licenses, Arkansas Power and Light Company Russellville Nuclear Unit, dated June 5, 1968.
8.
Supplement No. 5 to the Arkansas Power and Light Company Preliminary Safety Analysis Report, dated July 3, 1968.
9.
Corrections to Supplement No. 5 to the Arkansas Power and Light Company Preliminary Safety Analysis Report, dated July 10, 1968.
- 10. Supple =cnt No. 6 to Application for Licenses. Arkansas Power and Light Company Russellville Nuclear Unit, dated July 11, 1968.
- 11. Correction to Supplement No. 6 to Application for Licenses, Arkansas Power and Light Company Russellville Nuclear Unit, dated July 15, 1968.
- 12. Supplement No. 7 to Application for Licenses, Arkansas Power and Light Company Russellville Nuclear Unit, dated August 15, 1968.
- 13. Supplement No. 8 to Application for Licenses, Arkansas Power and Light Company Russ=11v111a Nuclear Unit, dated August 26, 1968.
- 14. Supplement No. 9 to Application for Licenses Arkansas Power and Light Company Russellville Nuolaar Unit, dated August 30, 1968.
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APPENDIX 3 CHRONOLOGY REGULATORY REVIEW OF THE ARKANSAS POWER AND LIGHT COMPANY RUSSELLVILLE NUCLEAR UNIT 1.
November 29, 1967 Submittal of Preliminary Safety Analysis Report and License Application.
2.
January 22, 1968 Submittal of Supplemental No. 1, response to AEC General Design Criteria.
3.
January 24, 1968 Meeting with applicant to discuss plans and scheduling of regulatory review.
4.
February 14, 1968 Submittal of Supplement No. 2, design changes in electrical systems and emergency core cooling systems, and data on Dardenell Lock and Dam.
5.
February 28, 1968 Meeting with applicant to discuss areas of the Preliminary Safety Analysis Report that require additional information.
6.
April 3, 1968 Request to applicant for additional information on i
l site, safety analysis, reactor, instrumentation and I
control, emergency power, engineered safety features quality assurance, training schedules, emergency plants, and Linitial tests and operations.
(48) 7.
May 6, 1968 Request to applicant for additional information on Foundation and Structural Design and miscellaneous other items.
8.
May 3, 1968 Submittal of Supplemental No. 3 in response to April 13, 1968 request for additional information.
9.
May 17, 1968 Meeting with applicant to discuss training schedules and operating staff.
10.
June 5, 1968 Submittal of, Supplement No. 4 in response to May 6, 1968 request for additional information.
11.
June 20, 1968 Meeting with the applicant to discuss modified contain-ment design proposed by applicant, site matters and other areas.
12.
July 3, 1968 Submittal of Supplement No. 5, changes in containment design.
13.
July 11, 1968 Submittal of Supplement No. 6, supplemental informa-tion in clarification of areas discussed at June 20, 1968 meeting.
14.
August 6, 1968 Meeting with applicant to discusa containment design
- matters, 9
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(49) 15 August 15, 1968 Submittal of Supple =ent No. 7, supplemental informa-tion in clarification of areas discussed at August 16, 1968 meeting.
16.
August 23, 1968 ACRS Subeccmittee meeting and Russellville site visit.
17 August 26, 1968 Submittal of Supplement No. 8, additional supple-
=entary information in clarification of containment design.
18.
August 27, 1968 Meeting with applicant to discuss quality assurance and quality control plans and organizations.
19 August 30,1%8 Submittal of Supplement No. 9, de nenting informa-tion and oral ccmmitments given et August 27 =eeting and miscellaneous other items.
20.
September 3,1%8 Meeting with applicant to discuss containment liner anchorage design.
21.
September h,1968 ACRS Subecmmittee meeting.
'. 22.
September 5, 1 % 8
' ACRS meeting.
r 23 September 6,1968 Submittal of supplement No.10, updating financial' and perscnnel infomation and correcting minor errors.
24.
September 12, 1968:
-ACRS Report issued.
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- ,FPr oIx C1 m am.v ama m o
a UNITED STATES
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DEPARTMENT OF THE INTERIOR
%,ge FISH AND WILDLIFE SERVICE T-" D' WASHINGTON. D.C.
20240 gg 2 91968 Mr. Harold L. Price Director of Regulations U. S. Atomic Energy Cc mission Washington, D. C. 20545
Dear Mr. Price:
This is in reply to Mr. Boyd's letter of December 11, 1967, requesting our comments on the application by the Arkansas Power and Light Company.
for ecnstruction permit for the proposed Russellville Nuclear Unit, Pope County, Arkansas, AEC Docket No. 50-313 The project would be located on a 1,100-acre site on a peninsula at Dardanelle Reservoir, Pope County, Arkansas. A pressurized water reactor would be used as a power source and the plant is designed for.an ultimate output of 2,568 thermal (880 gross electrical) Mwt. Cooling and dilution water vill be withdrawn from a small inlet embayment vest of the plant at a rate of approximately 1,700 c.f.s. and be discharged into the large Illinois Bayou embayment east of the plant, after receiving radioactive and heat vastes. As currently designed, the te=perature of the cooling water would be raised approximately 150 at the condenser when the plant is operating at full capacity. The applicant is cooperating with the Fish and Wildlife Service and the Arkancas Game and Fish Comission in the development of an environmental surveillance program.
Dardanelle Reservoir, especially the Illinois Bayou e::foayment, supports valuable fish and wildlife resources. The large embayment is a productive nursery and harvest-area for fish. Waterfowl make extensive use of the e
reservoir for resting during the migration pariod. Public and private use facilities on Federal and private land around the embayment are highly developed. Indications are that future development around the embayment will probably result in higher rec. eational use there than any comparable area of the reservoir. Sport fishing is presently, and will continue to be, one of the chief recreational use attracticms in the embayment.
Coccercial fishing is limited but moderately valuable.
The application indicates that the release of radloactive vastes would not exceed maximum permissible limits prescribed under the Code of Federal Regulations. Although these limits refer to maximum levels of radio-activity that can occur in drinking water for man without resulting in any known harmful effects, operations within these limits may not always guarantee chat fish and vildlife vill be protected fra adverse effects.
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If concentrations in receiving water were the only consideration, maximum per=issible limits would be adequate criteria for determining the safe rate of discharge. However, ra !.oisotopes of many elements are concentrat'ed ana stored by organisms that require these elements for their normal meta-bolic cetivities.
Some organisms concentrate and store radioisotopes of elements not normally required, but which are chemically similar to elements essential for ractabolism.
In both cases, the radionuclides are transferred from one organisn to another through various levels of the food chain just as are the nonradioactive.elecents. These transfers may result in fttrther concentration of radionuclides.
In view of the above, we believe that the environmental monitoring program planned by the applicant should include pre-and post-operational radio-logical monitoring of selected organisms which require the waste elements or similar elements for their meta.311c activities. These surveys should be planned in cooperation with the Fish and Wildlife Service and the P.ppropriate Federal and State agencies.
In view of the extensive sport fishery and the potential value of the commercial fishery in the project area, it is imperative that every possible effort is to be made to protect the valuable resources frca radio-active contamination. Therefore, it is recommended that the Arkansas Power and Light Company be required to:
1.
Include in their pre-operational environmental surveillance
~
program radiological monitoring of water and sediment samples and of organisms indigenous to the project area that concentrate and store radioactive isotopes. Water and sediment samples should be collected within 500 feet of the
/
reactor affluent outfall site and be measured for gamma radioactivity. Aquatic plants, mollusks, crustaceans and fish should be collected as near as possible to the reactor effluent outfall site and be analyzed for both beta and gamma radioactivity.
2.
Prepare a report of pre-operational radiological monitoring and provi. five copies to the Secretary of the Interior for evaluation prior to project operatita.
I 3
Centinue a radiological monitoring program similar to that specified in recommendation 1 above, analyze the data, and prepare and submit reports every six months during reactor operation or until it is conclusively demonstrated that no significant adverse conditions exist. Five copies of these reports should be submitted to the Fish and Wildlife Service for distribution to the appropriate State and Federal agencies for evaluation.
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L.
Make modifications in project structures and operations to reduce the discharge of radioactive wastes to acceptable levels if it is determined by the monitoring program that the release of radioactive effluent might result in harmful con-centrations of radioactivity in fish and wildlife.
We urderstand that the Comission's regulatory authority over nuclear power plants involves only those hazards associated with radioactive materials.
However, we recomend and urge that before a construction permit is issued, the possibility of thermal and other detrimental effects on fish and wild-life which may result fron plant construction and operation be called to the applicant's attention.
We are concerned particularly with the possibility of damages to aquatic life from the heated effluent.
Large volumes of heated water discharged into an aquatic environment may not only be detrimental to fish directly, but may also affect these resources indirectly through changes in the environment. The proposed heat load may adversely affect fish habitat ard productivity in the Illinois Bayou embayment during the periods (spring and summer) when fish reproduce and have a m nimum growth rate.
It is likely that the use of the area for spawning will be greatly reduced.
It is likely that fish will disperse and avoid the heat-affected area during the mad. mum temperature months of June through September.
Con-versely, it is expected that fish will be attracted to the discharge channel ard heat-affected area during winter months, resulting in high fisherman-use there.
A General Plan for use of project lands and waters for wildlife conservation f
ard management has been approved for Dardanelle Reservoir by the Secretary of the Amy, the Secretary of the Interior, and the Director of the Arkansas Game and Fish Commission. The Russellville Nuclear Unit would occupy land and water covered, in part, by the General Plan. The General Plan provides for a subsequent management agreement between the Department of the Amy and the Arkansas Game and Fun Commission.
It further pro-vides that the subsequent agreement may make adjustments in the boundaries of the areas shown in the General Plan by the addition or deletion of tracts mutually agreed upon by the parties making the agreement. We understand that the Department of the Amy and the Arkansas Game and Fish Comission are now negotiating an agreement pursuant to the General Plan.
The Company should be made aware of these documents ard plan its operations so that they are in accordance with the Arkansas Game and Fish Commission's fis5 and wildlife management plan for the reservoir.
The applicant has given assurance that additional studies will be carried out, ard has to date cooperated fully with the Fish and Wildlife Service and the Arkansas Game and Fish Com:aission in discussing and developing.
plans for the protection of fish and wildlife in the area. This study l
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(53) program should complement the radiological monitoring program reccmmended above, should be designed to measure habitat changes in the affected area of Dardanelle Reservoir, and should be carried out prior to and during plant operation, so that comparative data will be available fo.r analysis.
In view of the above,we recommend that the Ator.d.c Ene$gy pommission urge the Arkansas Power and Light Company to:
1.
Continue to cooperate with the Fish and Wildlife Service, Arkansas Game and Fish Commission and other interested F;deral and State agencies in developing plans for ecological surveys, initiate these studies at least two years before reactor operation, and continue them during project operation on a regular basis or until it has been conclusively demonstrated
~
that no significant adverse conditions exist.
2.
Meet with the abovevmentioned Federal and State agencies at frequent intervals to discuss new plans and to evaluate results of the ecological surveys.
3.
Make such modifications in plant structures and operations, including but not limited to facilities for cooling discharge waters, as may be determined necessary to protect the ' fish and wildlife resources of the area.
The opportunity to present our views is appreciated.
Sincerely yours, i
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APPE:rDIX C2
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UNITED STATES
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DEPARTMENT OF THE INTERIOlk 9-"[
FISH AND WILDLIFE SERVICE
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20240
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AUG 2 91968 Mr. Harold L. Price Director of Regulations U. S. Atomic Energy Co::=ission Washington, D. C.
20545
Dear Mr. Price:
ThisisinresponsetoMr.boyd'sletterofJuly16 transmitting Amendment No. 6, dated July 11,1968. to the application by 3
Arkansas Power and Light Company for a construction permit for the proposed Russellville Nuclear Unit, Pope County, Arkansas,.
Docket No. 50-313.,
Modifict', ion of project plans to reverse the direction of cooling water flow through the project would not alter overall effects of the project on fish and vildlife significantly. The recocnen-dations contained in our letter of May 29 are still applicable.
Thank you for the,cpportiin'ty for coments on Amendment No. 6.
Sincerely yours, O
isforis?
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APPENDIX D us mu m u_
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1333 MiltonShaw,Directotj,hII'f rno.u Division of 2cactor Davalopmer.t & Technology scaji:CT:. SAFETY ANALYSIS R22CRTS RDT:NS:S349 Reference is cade to the letters of November 22, 1967, December 11, 1967, and Decc=bar 25, 1967, from the Division of Reactor Licensing, to the Environmental Science Services Administration recuesting cc= cents on the following safety analysis reports respectively:
Rancho Seco Nuclear Generating Station Unit No.1 Sacrc:anto Municipal Utility District Preliminary Safety Analysis Report Volu=cs' I, II, III and IV dated November 1967 t/
Russellville Nuclear Unit '
t.rkansas Powar and Light Preliminary Safety Analysis Report Voluces'I and II dated Nove=ber 29, 1967 Donald C. Cook Nuclear Plant Indiana and Michigan Electric Cc=pany
? aliminary Safety Analysis Report.
Volu=as I, II and III dated December 18, 1957 2cview by tha Environ =antal Meteorology 3rar.ck, Air Resources Laboratory, ESSA,.has'new been completed and their con =ents are attached.
Attach =cnts:
Three Sets of Comments (Orig. & 1 copy)
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(56)
APPENDIX D Comments on Russellville Nuclear Unit Arkansas Power and Light Preliminary Safety Analysis Report Volumes I and II dated November 29, 1967 Prepared by Air Resources Environmental Laboratory Environmental Science Services Administration January 10, 1968 The analysis of the Fort Smith and Little Rock meteorological data indicates.that a continental diffusion climate can be expected at the Russellville site. This means a pronounc'ed difference between daytime and nighttime atmospheric diffusion rates, with the lower wind speeds and slower diffusion occurring at night. The predominant daytime wind direction for the general area would be from the southwest as shown by the Little Rock wind rose. Nighttime wind directions with inversion conditions will most likely be towards the Dardanelle Reservoir of the Arkansas River.
The analysis of the Little Rock hourly weather reports with regard to diffusion types shows an average frequency of about 35% for Pasquill F f
condition during the four months considered (see Table 2A.15). The annual nighttime wind speeds were less than 3 knots about 20% of the time at Little Rock (see Table 2A.6). On this basis, it would seem appropriately conservative to use inversion diffusion conditions (Type F) and a 1 m/see wind speed to compute the initial two-hour average concentration. This would result in a concentration of 6.4 x 10-4 see m.-3 at the site boundary assuming a ground source with no credit for building-induced dilution. Taking ' credit for the building effectasdetenninedempiricallyintestsattheNationalReacgorTesting Station weald result in a concentration value of ebout 2 x 10
, which agrees with the applicant's value.
The analysis of the persistence of a diffusion condition in a unidirectional flow (Tables 2A.17 and 18) shows that no cases persisted longer than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.
Consequently, for the 24-hour average concentration it would be conservative to assume inversion conditions, a 2 m/see wind
'At the site with concentrations averaged over a 221/2 degree arc.
boundary this would result in anaverage concentration of 7 x 10-5 see m-3, which is in reasonable agreement with the applicant's computation.
In summary, a reasonabie, conservative analysis has been made of the '
atmospheric diffusion conditions of the Russellville site which provides a sound basis for a preliminary safety evaluation of the proposed nuclear plant.
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APPENDIX E 84-3/ 4
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GEOLOGICAL SURVEY WASHWGTON. D.C. s0sA2 g
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AUG 161968
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e Mr. Harold L. Price Director of Regulation U. S. Atomic Energy Commission 4915 St. Elmo Avenue Bethesda, Maryland 20545
Dear Mr. Price:
Transmitted herewith in response to a request by Mr. Roger S. Boyd is a review of 8eologic and hydrologic aspects of the site for the Russelville Nuclear Station proposed by the Arkansas Power and Light Company.
The review was prepared by H. H. Waldron and E1L. Meyer and has been discussed with members of your staff. We have no objection to your making this review a part of the public record.
Sincerely yours,'
ActiwD w or Enclosure
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AE0 Docket 50-313 Hydrolony' The site is located on the left bank of the Arkansas River 6 miles upstream from Dardanelle Lock and Dam No.10.
The plant site grade at 353 feet mal (above mean sea level) is 15 feet above the normal operating
' level of Dardanelle Reservoir.
l Flood stages in the pool of Dardanelle Reservoir for a computed maximum
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probable flood of 1,500,000 cfs (cubic feet per second) have been given j
by the Corps of Engineers as 353 feet asl at Dardanelle Dam and 389.5 feet mal at the upstream end of the reservoir. The applicant's estimate of 358 feet mal for the stage of such a flood at the site appears reason-able. The failure of Ozark Dam about 46 miles upstream from the site during such a flood could cause an additional rise in stage. The head differential across Ozark Dam during a maximum probable flood as computed l
by the Corps of Engineers would be 11.5 feet, and on that basis the applicant has estimated an additional 3 feet rise at the site resulting in a stage of 361 feet. This appears to be reasonable.
At a stage of 361 feet the site grade would be overt.opped by 8 feet and the reactor structures would be surrounded by water.. A certain amount of wave action may then be expected and should be reflected in the level of flood protection chosen for essential equipment.
The cooling water requirements of the reactor are given as 1,700 efs (cubic feet per second).
Flow of the Arkansas River has been measured at a gage at Dardanelle 6 miles downstream from the site. Average flow e
during 1937-66 was 34,920 cfs; minimum flow was 416 cfs, and the lowest mean monthly flow was 592 efs in October 1956. Low flow occe : generally in late summer and fall.
- Geoloe, The analysis of the geology of the Russellville Nuclear Generating Plant in Arkansas, as presented in AEC Docket No. 50-313 and supplements, was reviewed and compared with the available literature. The analysis appears to be carefully derived and to present an adequate appraisal of those aspects of the geology ~ that would be pertinent to an engineering evaluation of the safety of the site.
There are no identifiable active faults or other recent geologic structures that could be expected to localize earthquakes in the immediate vicinity of the site.
Tectonically the site is located near the axis of the Scranton syncline, one of several westward-trending, gentle folds that characterize the
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(59)
Arkoma Basin--a major structural and topographic fescure of Arkansas and eastern Oklahoma that developed in late Paleozoic tiue. Although several ancient faults are associated wit.h the Arkoma Basin folded structures in the area, none of these appears,t'o have, been tectonically active since latest Paleozoic time.
The limited subsurface data available indicate that the major units of the nuclear facility will be founded on a hard, dense shale (the McAlester Formation), which should provide an adequate foundation for the proposed structures.
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m nam.v ama m C23 Mr. Harold L.
Price Director of Regulation
'a. S. Atomic Energy Commission Washington, D. C.
20545 De a r ?*.r.
Price:
In cccordance with your request, we are forwarding 10 coplea of our report on the seismicity' of Russellville,
. Arkansas, and vicinity.
The Coast and Geodetic Survey has reviewed and evaluated the information on the sels-mic activity of the area as presented by the Arkansas Pc,wer and Li ht Company in the yeis Report,g' and we are now submitting our conclusions" Prs 11minary Safe on the seismicity factors.
If we may be of further assistance to you, please do not hesitate to contact us.
Sincerely your D
.g ames C. T son Jr. -
Rear dmira,
SESSA Director Enc'.osure
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REPORT OU THE SITE SEISMICITY FOR THE i
i RUSSELLVILLE NUCLEAR UNIT, ARKANSAS At the request of the Division of Reactor Licensing of
. 6 i
the Atomic Energy Commission, the Seismology Division of the Coast and Geodetic Survey has examined the seismicity of the area around the proposed site near Russellville, Arkansas, and has examined a' similar analysis made by the applicant, the Arkansas Power and Light Company in the " Preliminary 1
Safety Analysis Report."
The applicant's report is satis-factory for an evaluation of the seismic factor of the site.
~
Based upon the review of the seismic history of the site anti the surrounding area and the related geologic. con-ditions, the Coast and Geodetic Survey agrees with the app 1i-cant that an acceleration of 0.10 g on good foundation would be adequate for representing earthquake disturbances likely to occur within the lifetime of the facility.
In addition, the Survey agrees with the applicant that the acceleration of 0.20 g would represent the ground motion from the' maximum earthquake likely to affect this site.
We believe this value would provide an adequate basis for designing protection against the loss of function of components important to safety.
'U..S.
Coast and Geodetic Survey Rockville, Maryland 20852 August 14,_1968 P
(62)
APPE:iDIX G NATHAN M.
NEWMARK CONSULTING ENGINEERING SERVK:ES 1814 CIVIL ENGINEERING BUILDING
'u RsAN A.
ILUNOIS elect 19 August 1968 s
Dr. Peter A. Morris, Director Division of Reactor Licensing U. S. Atomic Energy Commission Washington, D.C.
20545 Re:
Cont ract No. AT(49-5)-2667 The, Russellville Nuclear Unit, Arkansas Power and Light Conpany (AEC Docket No. 50-313)
/
Dear Dr. Forris:
We are transmitting herewith two copies of our report entitled
" Adequacy of the Structural Criteria for the Russellville Nuclear Unit,"
p repa red b y Ors. W. J. Ha l l, W. H. Wa l ke r.and mys e l f.
Sincerely yours,
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N. M. Nowmark mlw cc:
W.
J. Hall W. H. Walker Enclosure e
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NATHAN M.
N E W M /. M K CONSUQNG ENGINEERING SERVICES, 1114 CIVIL ENGINEERING BUILDING URsANA. ILUNOIS 61801 REPORT TO AEC REGULATORY STAFF ADEQUACY OF THE STRUCTURAL CRITERIA FOR THE RUSSELLVILLE NUCLEAR UNIT ARKANSAS POWER AND LIGHT COMPANY (AEC Docket No. 50-313) by N.
M.
Newma r k, W. J. Hall 'and W. H. Walker August 1968 O
(64)
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ADEQUACY OF THE STRUCTURAL CRITERIA FOR THE RUSSELLVILLE NUCLEAR UNIT
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Arkansas Power and light Company by N.. M.,Newmark, W. J. Hall and W. H. Walker INTRODUCTION This report is concerned with the adequacy of the containment structures and components for the Russellville Nuclear Unit for which application for a cor,.ruction permit has been made to the U. S. Atomic Energy Commiss ion by the Arkansas Power and Light Company.
The facility is located on a peninsula in the Dardanelle Reservoir, Arkansas River, Pope County, Arkansas, about 6 miles WhV of Russellville, and 2 miles SE of London, Arkansas.
Specifically this report is concerned with the design criteria that determine the ability of the containment system and CIass I equipnent and piping as well as Class II structures and equipment, to withs tand an Operat ing Bas is Earthquake of 0.10g maximum horizontal ground acceleration.s imultaneously 5
with the other loads forming the basis of the design.
The f acility also is to be des igned to withstand a Des ign Basis Earthquake of 0.20g maximum horizontal ground acceleration to the extent of ensuring safe shutdewn and containment.
This report is based on information and criteria set forth in the Preliminary Safety Analys is Repor ts (PSAR) and supplements thereto listed at the end of this report.
Also,, we have participated in discuss ions with the' applicant and the AEC Aegulatory Staf f concerning the design of this unit.
CESCRIPTION OF FACILITY The Russellville Nuclear Unit is d+. scribed in the PSAR as consis t ing of a pressurized-water type reactor amploying two closed cooling loops connected in parallel to the reactor vessel.
The system is arranged as two heat transport
(65) 2 loops, each with two circulating pumps and one steam genera' tor; one of the loops contains an electrically heated pressurizer.
The nuclear steam s'upply system will be furnished by the Babcock and Wilcox Ccmpany, and the ' turbine
'o 9enerator is to be supplied by the Westinghouse Electric Corporation.
The plant is to be de's igned for a power level of 2452 MWt (850 Mwe).
The reactor containment structure is a fully continuous reinforced concrete structure in the shape of a cylinder with a shallow domed roof.and a flat foundation slab.
The cylindrical portion, is prestressed by a post-tensioning system of horizontal and vertical tendons.
The dome is pos t-tens ioned us ing a 3-way s ys tem.
The hoop tendons are to be placed in three 240 s ys tems us t.mq three buttresses as anchorages, with the tendons staggered so that half of the tendons at each buttress terminate at that buttress.
The foundation slab is conventionally reinforced with high-strength reinforcing steel.
The cylinder has an internal diameter of 116 ft. and an inside height of 206 ft.
The distance from the top of the foundation slab to the springline of the domed roof is approxinetely 166 f t.-
The vertical wall thickness is noted to be ft. - 9 in, and the dome thickness, 3 ft. - 3 in.
The foundation slab thickness is about 9 ft.
i For_prestressing, the applicant proposes to use 90 to 184 wire tendans, unbonded.
The qiscussion presented in the PSAR suggests that the BBRV type anchorage system will be employed, although the PSAR notes that other l
pres t ress,ing sys tems will cont inue to be s tudied.
The prestress ing tendons will be protected against corros ion by a pressure-injected cas i ng f iller.
The liner plate will conform to specification AS7M-A442, Grade 60, and will be 1/4 in. In thickness.
The reinforcing steel in the base slab of the containment structure will conform to ASTM designation A432-65; this steel
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(66) 3 possesses a minimum yield strength of 60,000 ps!.
Spilces in bars larger than No. 11 will be made by the Cadweld (nethod.
The design of the containment structure for this facility is essentially similar to that employed for the Rancho Seco Nuclear Generating Statibn Unit No.
1.
The geolog real descrIpt ion of the s ite Indicates a stif f clay and silty clay of 13 to 23 foot thickness overlying hard ad dense horizontally bedded shale of the Pennsylvanian McAlester formation.
All major structures of the facility will be founded on the underl'ying McAlester formation shale bedrock.
No active or recent faulting has been mapped in the area of the proposed site.
The closest known f aults are the Lnndon and Perry View faults located 5 or 6 miles from the site.
SOURCES OF STRESSES IN CONTAINMENT STRUCTURES IN' CLASS I COMPONENTS The reactor containment structure is to be designed for the following loadings and conditions:
dead load; live load (including snow and equipment i
loads); prestressed loadings; design accident temperature of about 285 F and pressure of 59 psig; an air test pressure of 115 percent of the design pressure; an external pressure loading with a dif ferential of approximately 2i psi f rom outs ide to ins ide; wind loading co: responding to 80 mph bas ic wind at 30 ft. above grade; buoyancy loadir.gs; tornado loading' associated with a 300 mph tangentiel wind veloc,ity and a 40 mph forward progression velocity, including a dif ferential pressure of 3 psi from inside to outside with associated missi.les; and earthquake loading as described next.
The seismic design is to be made for an Operating Basis Earthquake based upon a maximum horizontal ground acceleration of 0.10g and a Design Bas is Earthquake based upon a maximum horizontal ground acceleration of 0.20g.
(67) 4 The containment walls and liner are shielded by various types of barriers from impact f' rom missiles which possibly could have enough energy to strike or penetrate them.
The high-pressure reactor cooling sys' tem equipment is - *.
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which could be the source of missiles is screened either by the containment shield well ecciosing the reactor cooling loops, by the concrete operating floor, or by a special missile shield to block any passage of missile to the containment walls.
The general criteria controlling the design of piping and reactor internals for seismic loadings are presented in various places in the PSAR.
COMMENTS ON A0EQUACY OF DESIGN Foundations and Daq3 The major f acility structures are to be founded directly on competent bedrock, and on the basis of the information presented in the PSAR and amendments, the foundat ion conditions appear acceptable to us.
The Dardanelle Reservoir f rom which the plant will draw its cooling waters is discus.ted in several places in the PSAR and particularly in Appendix 2F and in the answers to Ques tions 2.7 and 2.8 of Supplement No. 3.
The anal ys is of the Dardanelle Lock and Dam as reported in Appendix 2F suggests that some damage to the L ick and Dam facility might be expected.
Thus, the applicant notes in the answer to Question 2.7 that emergency shutdown cooling water will be supplied f rom an emergency reservoir to be located northwest of the piant
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site.
The emergency reservoir will be excavated in impervlous clay and wij l have an ef fective storage capacity of about 35 acre feet.
We concur in this approach for an assured source of cooling water in view of the possible effects of an earthquake on the Dardanelle Lock and Dam.
p (68)
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The effect of a flood on the structure is discussed in the answer to Question 2.8 of Supplenent No. 3.
It is noted there that the plant grade
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level is elevation 353 ft. and the maximum elevation of a flood is estimated to be 361 ft.
The applic'nt indicates that the early forecast of a 'evere a
s flood of this t,ype would provide ample time for precautionary measures in terms of plant shutdown.
All Class ! equipnent is either located above maximum probable flood level or protected by waterproof Class I structures which are designed for buoyancy effects.
ras Ploeline In the answer to. Ques t ion 2.11 of Supplement No. 3, there appears a discussion of the natural gas transmission pipeline which crosses the discharge water channel.
It is indicated in the answer to that quc5 tion that the existing pipeline cross ing will be re-layed beneath the water' channel with '4 f t. of earth cover.
We understand that it will be pos:ible to valve off this section of line in the event of difficulty.
It is noted that the pipeline will be at its' closest about 400 ft. from the intake s tructure and 600 f t.
from the containnent s tructure.
These dis tances are suf ficient, we believe, to preclude any serious consequer.ces with regard to plant safety in the event of a pipe rupture.
Seismic Des ion and Criteria We are in agreement with the earthquake loading criteria selected for the seismic des ign, namely that associated with an Operating Bas is Earthquake of 0.10g maximum horizontal ground accelerat ion and a Des ign Basis Earthquake of 0.20g maximum horizontal ground acceleration.
These earthquake design criteria are in agreement with those given by the U. S. Coast and Geodet ic Survey (Ref. 2).
g (69) i 6
i The response spectra for the Operating Basis Earthquake and Design Basis Earthquake to be. employed in the' dynamic analys is are presented as i
Fig. SA-l and SA-2 of Appendix SA of the PSAR.
These spectra are. scaled after those presented in p'ub'I'ications by Dr. G. W. Hous ner, and we concur in their use.
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The earthquake analysis will include the ef fects of vertical earthquake exc,itation which will be taken as 2/3 of the horizontal component as noted on page 5-3 of the PSAR.
It is noted in the answer to Quest ion 12.3.6 that the ef fects of vertical and horizontal earthquake motions will be combined linearly and directly with each other and with the other applicable stresses.
We are in agreement with these design criteria.
The perceistage of critical damping to be employed in the analysis is listed on page 5-A-5 of the PSAR, and we are in agreement with the values given there.
The method of dynamic analysis is described in Section 5.1.5.6 of the PSAR.
The method of analys is is not described in enough detall to evaluate it completely; however, it would be our, recommendation that a standard nodal analysis procedure be employed to take account of. structural rocking, lateral
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translation, and the shearing and flexural distortion of the structure.
With proper attention to damping and coupling of the various modes, it should be possible to arrive at reasonable and cons istent values of di rect stress, shear, moment, etc.
The loading' combinat ions to be employed for the des ign of the containment s tructure are given in Section. 5.1.4 of the PSAR.
The loading
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combinat ion express ions given appear acceptable to us, 'and it is noted'that i
for these load factor combinations the resistance will be lass than the yield strength of the structure.
We concur in this approach.
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.F (70) t 7
i The des ign of Class II struc'tures is discussed in the answer to Question 12.3.2 where it is noted that the design of such items will be for Zone I of the Uniform Building Code.
It wuld be our recommendation that for critical Class II Items tha't are of special significance in terms of plant safety, the design be made on the basis of about 2/3 of Zone 3 of the Uniform Building Code.
The design approach as outlined for handling principal concrete tension and combined tension and membrane shear appear acceptable to us.
t.
The design of the 1Iner and anchors is discussed in vartous sections of the PSAR.
We are advised that the liner design is still under study and that further information will be forthcoming during the design phases.
It is our bellef that the 1Iner design can be carrled out satisfactor11y and adequately, and we can see no particular difficulty here which will preclude going forward with the construction oermit.
The general approach outlined for the prestressed design receives attention in various parts of the PSAR and Supplements acd other material made available to us.
The design for this plant employing three buttresses and 90 to 184 wire tendons is relatively new. The applicant indicates that many factors associated with this post-tensioning are receiving added study, as for example the problems associated with the friction arising from the large put iIng arc (240 ).
On the basis of the information avallable to us, and realizing that additional studies are underway and will be carried forward during the design phases, we can see no reason why the proposed system will
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not be acceptable.
The design procedure for handling the statleal design, namely use of the finite element technique, coupled with special study and procedures
.g, (71) 8 for handling the primary and secondary
- loading around penetrations, appears satisfactory to us.
Class I Ploina. Ecu' loment. Vessels and Reactor Internals Only general statements are made in the PSAR concerning the design of Class I piping, equipment, vessels and reactor internals.
However, the t
answer to question 10.4.3 of Supplement No. 3 suggests that the criteria employed for Crystal River Unit 3 of the Florida Power and Light Corporation will be applicable to this plant, and reference is made to the answer to Question 9.11 of the Crystal River application.
It is noted that the calcula-tions and design will not b'e completed until mid 1969.
On the assumption that the approach outlined in the Flor!4 Power and Light Corporation application for Crystal River Unit 3 will be followed, we concur in the proposed approach.
I Controls. Instrumentation. Batteries, etc.
f Only general information is noted in the PSAR concerning the seismic design criteria for critical elements of control, instrumentation, batteries, etc.
It would be our recommendation that criteria for these items be examined in detail during the design phases, to insure that the items can withstand the
. forces, motions and tilt that might be associated with an earthquake.
Quality Control and Inspection The matter of quality control, inspection and acceptance is discussed throughout the PSAR and amendments.
Theproceduresoutlined.Npear acceptable to us.
CONCLUDING COMMENTS l
On the bas is of the information presented in the PSAR and supplements, and in keeping with the design goal of providing serviceable structures and
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components with a reserve of strength and ductility, we believe that the design outilne for the containment and other Class I structures and equipment and for Class II stVuc'tures and components can provide an adequate Aargin of.
safety for sel.smic resistance.
However, in the body of the report we have of fered comments concerning the method of dynamic analysis, and the' design criteria for Class II structures.
It is understood that studies will continue during the design phases on the design of the liner anchorage and the pre-s tress ing tendon system, and it is suggested that the seismic design criteria for eritical instrumentation be developed and implemented durIng the design phases.
REFERENCES 1.
" Preliminary Safety Analysis Report - Voiumes I, II (and Supplements No. 1, 3, 4, 5' and 6)," Russellville Nuclear Unit, Arkansas Power and Light Compa ny, 1968.
2.
" Report on the Seismicity of the Russellville Nuclear Unit Site," U.S.
Coast and Geodetic Survey, Rockville, Maryland 4
e e
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