ML100820279
| ML100820279 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 01/20/1972 |
| From: | Boyd R US Atomic Energy Commission (AEC) |
| To: | Walker R Public Service Co of Colorado |
| Barr J | |
| References | |
| Download: ML100820279 (97) | |
Text
DOCKET 50267-38 FORT ST. VRAIN NUCLEAR GENERATING STATION.
UNCLAS Safety Evaluation. (Division of Reactor Licen sing (AEC),
Washington, D. C.).
20 Jan 1972. 89p.
80-4 UNLTD DIST
UNITED STATES ATOMIC ENERGY C8iv1MISSION JAN 2 '0 '1972 WASHINGTON. 0 C.
]054')
Docket No. 50-267 r
Public Service Company of Colorado ATTN:
Hr. Richard F. Halker Vice President P. O. Box 8110 Denver, Colorado 80201 Gentlemen:
I am enclosing for your information a copy of a Safety Evaluation, d~ted January 20, 1972, by the Division of Reactor Licensing, United. States Atomic Energy Commis-sion in the matter of the Public Servlce Company of Colorado, Fort St. Vrain Nuclear Generating Station.
Sincerely, G~Y:c~~'d Roger S. Boyd, Assistant Director for Boiling Water Reactors Division of Reactor Licensing
Enclosure:
AEC Safety Evaluation cc r w/encl.
Lee, Bryans, Kelly &Stansfield.
ATTN:
Bryant O'Donnell, Esq.
bce:
w/encl.
H. J. McAlduff, ORO H. Mueller, GMR/H J. A. Harris, PI J. R. Buchanan, ORNL tA': W. Laughlin, DTIE N. H. Goodrich, ASLB S. Robinson, SECY J. Saltzman, SLR I'
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Issued:
January 20, 1972 SAFETY EVALUATION BY THE DIVISION OF REACTOR LICENSING U.S. ATOMIC ENERGY COMMISSION IN THE MATTER OF PUBLIC SERVICE COMPANY OF COLORAOO FORT 5T. VRAIN NUCLEAR GENERATING STATION DOCKET NO. 50-267 DISTRIBUTION Of THIS OOtOMJ:NT IS UNUMlltI VI L
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TABLE OF CONTENTS PAGE 11
- 1. ()
INTRODUCTION 2.0 SITE AND ENVIRONMENT 2.1 Site Description 2.2 Geology and Seismology 2.3 Hydrology 2.4 Meteorology 2.5 Environmental Radiological Monitoring 3.0 FACILITY DESIGN 3.1 General 3.2 Reactor Design 1
3 3
4 4
5 7
7 7
6 3.3 Primary Coolant System 3.2.1 3.2.2 3.2.3 3.2.4 3.2.5 3.3.1 3.3.2 3.3.3
- 3. 3.4 Fuel Design Core Physics Reactor Control Core Mechanical Design Thermal and Hydraulic Design Helium Circulators Orificing System Inservice Inspection Steam Generators 9
10 11 12 13 14 14 15 15 16 3.4 Prestressed Concrete Reactor Vessel and Class I Structures 17 3.4.1 3.4.2 3.4.3 3.4.4 3.4.5 Prestressed Concrete Reactor Vessel Class I Structures Design Criteria and Seismic Analysis PCRV Thermal Barriers Structural Acceptance Testing and Surveillance 17 18 18 19 20 3.5 Instrumentation and Control 20 3.5.1 3.5.2 General Protective Systems 20 20
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i 3.5.2.1 Reactor Trip System 3.5.2.2 Steam Generator Dump and Loop Shutdown 3.5.3 Safe S~utdown System 3.5.4 Other Instrumentation 21 21 22 22 i
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3.6 Electrical Power System 3.6.1 Offsite Pow~~
3.6.2 Onsite Power 23 23 24 3.7 Auxiliary Systems 25 4.1 General 32 4.2 Reactivity Transients 33 4.3 Applicant's "Maximum Credible Accident" 34 4.4 Rapid Depressurization of PCRV-DBA #2 35 4.5 Permanent Loss-of-Forced Circulation Cooling - DBA #1 38 4.6 Anticipated Transients with Failure of Protective Action 40 5.0 EMERGENCY PL'~NING 3.8 Radwaste Management Systems 25 25 26 27 28 28 30 32 43 43 45 General Cooling Water Systems Fuel Handling Fuel Storage Liquid Waste Management System Gaseou8 Waste Management System 3.7.1 3.7.2 3.7.3 3.7.4 3.8.1 3.8.2 4.0 ACCIDENT ANALYSIS 6.0 CONDUCT OF OPERATIONS AND TECHNICAL QUALIFICATIONS 7.0 TECHNICAL SPECIFICATIONS j
.1 8.0 REPORT OF ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 46 9.0 COMMON DEFENSE AND SECURITY 46
APPENDIX F -
AEC Regulatory Staff's Evaluation of Financial 82 Qualifications
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J 47 47 54 61 49 51 FINANCIAL QUALIFICATIONS FINANCIAL PROTECTII)N AND INDEMNI'IY REQUIREMENTS iii CONCLUSIONS APPENDIX A -
Report of Advisory C01ll1littee on Reactor Safeguards, dated May 12, 1971.
APPENDIX B2-Report of Nathan M. Newmark, Consulting Engineering Services, dated March 3, 1971.
APPEI'!DIX C -
Report of U. S. Department of Commerce, 67 National Oceanic and Atmospheric Adndnistration,.
dated March 3, 1971.
APPENDIX Bl-Report of Nathan M. Newmark, Consulting Engineering Services, dated Decetrber 8, 1970.
APPENDIX Dl-Report of U. S. Department of the Interior, 69 Fish and Wildlife Service, dated July 14, 1971.
APPENDIX D2-Report of U. S. Department of the Interior, 71 Fish and Wildlife Service, dated April 14, 1971.
APPENDIX E - Report of L. A. Booth, Staff Scientist, :Os Alamos 73 Scientific Laboratory, dated August 3, 1971.
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1.0 INTRODUCTION
This report is the Atomic Energy Commission's Safety Evaluation of the application by the Public Service Company of Colorado (herein referred to as the applicant) for a license to operate the Fort St.
Vrain Nuclear Generating Station.
The facility, which will utilize a high temperature gas-cooled reactor, is located on a site con-sisting of 2,238 acres in Weld County, Colorado, approxim3tely three and one half miles northwest of the center of Platteville II and about 35 miles north of Denver, Colorado.
The facility has been under construction since the U. S. AtoDdc Energy Commission (herein referred to as "the AEC" or "the Commission")
issued a construction pe~it on September 17, 1968.
On November 4, 1969, the Public Service Company of Colorado subDd~~ed the Final Safety Analysis Report (FSAR) as Amendment No. 14 to its application for a construction permit and operating license for the Fort St. Vrain Nuclear Generating Station.
This safety evaulation report summarizes the results of the technical evaluation of the Fort St. Vrain Station performed by the Commission's regulatory staff.
Our evaluation included a technical review of the information submitted by the applicant with regard to the following principal matters:
1.
We reviewed the population density and use characteristics of the site environs, and the physical characteristics of th~ site. including seismology, meteorology, geology and hydrology to determine that these characteristics had been detemined adequately and had been given appropriate consideration in the plant design, and that the site char-acteristics were in accordance with the Commission's siting criteria (10 CFR Part 100) taking into consideration the design of the facility including the engineered safety features provided.
2.
We reviewed the design, fabrication, construction, testing, and expected performance of the plant structures, systems, and components important to safety to determine that they are in accord with the Commission's General Design Criteria, other appropriate codes and standards, and the CoDmdssion's Quality Assurance Criteria, and that any departures from these criteria have been identified and justified.
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3.
We evnl ua t ed the response of the facility to various an-ticipated operating transients and t~ a broad spectrum of posrul.ated accidents, and det ernd.ned that the potential con-sequences of a few highly unlikely postulated accidents (design basis accidents) would exceed those of all other accidents considered.
We performed conservative analyses of these design basis accidents to dete~ine that the calculated potential off-site doses that might result in the very un-likely event of their occurrence would not exceed the Commission's guidelines for site acceptability given in 10 CFR Part 100.
4.
We evaluated the applicant's plans for the conduct of plant operations, the organizational structure, the technical qualifications of operating and technical support personnel, the measures taken for industrial security, and the planning for emergency actions to be taken in the unlikely event of an accident that might affect the general public, to deter-mine that the applicant is technically qualified to operate the plant and has established effective organizations and plans for continuing safe operation of the facility.
s.
We evaluated the design of the systems provided for contrnl of the radiological effluent. from the plant to determine that these sys terns can control the release of radioactive was ces from the station within the limits of the Commission's regulations (10 eFR Part 20) and that the applicant will operate t.he facility in such a manner as to maintain radio-active releases to levels that are as low as practicable.
During our review of the information submitted in the Safety Analysis Report we requested the applicant to provide additional information we needed for our evaluation.
The additional infor-mation was provided in subsequent amendments to the application.
In the course of our review we held numerous meetings with the applicant to discuss and clarify the techn:f.cal information s ub-mitted.
As a result of our review we required a number of changes to be made in the facility design; these changes are described in the applicant's amendments and are discussed in appropriate sections of this report.
Our technical safety review and evaluation of the facility has been based on Amendments 14 through 24, which comprise the FSAR and supporting information.
All of these documents are available for review at the Atomic Energy Commission's Public
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Document Room, 1*711 H Street, N. W., Washington, D. C.
In addition to our review, the Advisory Commdttee on Reactor Safe-
~uards (ACRS) reviewed the application and met with both the applicant and us to discuss the facility.
The report of the ACRS ~ dated May ll, 1911, is attached to this Safety Evaluation as AppendiX A.
The proposed operation of the plant would be at core p~Aer levels up to 842 MW thermal for a
~et electrical output of approximately 330 MW.
The facility has an installed capability for operation up to 819 MW thermal and the safety analysis for operation at this higher power level has been included in the FSAR.
Since the applicant may at some subsequent time request authority to operate the reactor at the higher
~ower rating, our evaluation of all structures and system. which relate to safety is baaed on the higher power level (879 MWt), while the thermal and fluid flow characteriytics of the reactor core were analyzed and evaluated at the initial power l~vel (842 MWt).
Based on our evaluation or the plant 88 summarized in subsequent sections of this report, we have concluded that Fort St. Vrain Nuclear Generating Station can be operated,.. proposed, at power levels up t.o 842 MWt without endangering the health and safety of the public.
2.0 SITE AND ENVIRONMENT 2.1 Site Description The Fort St. Vrain site consists of 2, ~38 acres in Weld County, Colorado, apprOXimately three and one half miles nortbweet of the center of Platteville, Colorado, and approximately 35 miles north of Denver, Colorado.
The reactor building lies slightly to the north and east of the center of a one mile square exclusion area.
The shortest distance from the reactor to the exclusion area boundary is 1,935 feet to the east, and the distance to the nearest site boundary is 3,511 feet to the east-northeast.
A low population zone (LPZ) distance of 10 miles has been selected, and the distance to the nearest edge of Greeley, Colorado, the nearest population center, is about 14 miles.
On the basis of our evaluation of the population data, and of the calculated potential offsite doses that might result from the postulated Design Basis Accidents, we conclude that the exclusions at "
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the low population zone distance, and the distance to the nearest population center meet the guidelines given in 10 CFR Part 100.
2.2 Geology and Seismology During our review of this site prior to the issuance of a con&truc-tion permit, we and our consultant, the U.S. Department of Interior, Geological Survey, concluded that the applicant's geological analyses, field explorations, and laboratory tests provided sufficient infor-mation to determine that the site was suitable from a geological standpoint for the construction of the plant.
No info~tion was developed during excavation or construction that has changed this conclusion.
The applicant has stated that components, equipment, and structures necessary to assure a safe and orderly shutdown (designated as Class I) will be designed such that stress levels resulting from an earth-quake having a maximum horizontal ground acceleration of 0.05g (Operating Basis Earthquake, OBE) will not exceed those allowed by applicable codes.
For a maximum horizontal ground acceleration of 0.10g (Design Basis Earthquake, DBE) there will be no loss of function of critical structures and components necessary to assure a safe and orderly shutdown of the plant.
During our construction permit review, we and our seismic consultant, the U.S. Coast and Geodetic Survey, concluded that the site is in an area of low seismic activity, that there are no known faults within 20 miles of the site, and that these design basis accelerations are acceptable.
The facility design has been reviewed by our consultant, Nathan M. Newmark Consulting Engineer-ing Services of Urbana, Illinois, who has concluded, and we agree, that the facility was designed and constructed in accordance with the above seismic design criteria.
Surveillance for possible seismic ground motion is being provided by two strong-motion accelerographs and three seismoscopes.
In the event of a seismic event at the aite, the data collected with these instruments would be employed in the subsequent evaluation of the effects of the seismic event on the continued safe operation of the facility.
2.3 !!ydrology The flow of ground water on the site is toward the alluvial deposits of both the South Platte River and the St. Vrain Creek with ground water flow from the reactor location probably flowing toward the South Platte River.
There are very few radioactive or potentially radioactive liquid systems in this gas-cooled plant, and special design provisions of such liquid systems, make it extremely unlikely that any significant amount of radioactive liquids could ever reach the ground water.
However, if a spill or leak should occur, it would be detected by virtue of the environmental monitoring program and appropriate control measures could be taken.
Water from the two streams and shallow wells located on the si te will be used to supply water to the cooling towers which will provide water for condenser cooling.
This water is isolated from the reactor primary (helium) and secondary (water) heat transport systems and therefore will not contain radioactive contaminants.
The quantities of radioactive liquid waste expected to be released from the Fort St. Vrain facility are very small, i.e., approximately 0.2 mCi per year, and the releases of radioactive effluents are limited by the Technical Specifications to be well within 10 CFR Part 20 limits for unrestricted areas.
These normal radioactive liquid effluents will flow via Goosequlll Ditch to Goosequill Pond.
Overflow from the pond will be directed to the South Platte River on the extreme north of the plant boundary.
Flooding at the site is unlikely.
In the event of the "probable maximum flood" on the South Platte River (500,000 cfs), the Omaha Dts trict Office of the U.S. Arrtrj Corps of Engineers estimates that the water level would be at least 10 feet below plant grade elevation.
We and our hydrological consultant, the U.S. Geological Survey, Department of the Interior, considered the potential flood levels in the South Platte River and St. Vrain Creek during our review prior to issuance of a construction perndt.
We concluded at that time that the site elevation (plant grade) offered adequate protection against flooding for the reactor and its safety related equipment.
Our operating license review has not altered this conclusion.
2.4 Meteorology The applicant's site meteorological measurements program consisted of wind speed and direction measurements at 202 feet above grade and measurements of temperature differences between 20 and 202 feet.
We, our meteorological consultant, the National Oceanic and Atmospheric Administration (NOAA), and the applicant have used the data collected in this program for the one year period from October 1967 through October 1968 in our evaluAtion of the site meteorology.
The most
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significant feature shown by the data is a high frequency (18%) of annual wind occurrences from the northern 22-1/2° sector.
The wind-speed measurements indicate that the site is not well ventilated when compared with wind speed measurements taken at other locations in the United States.
As a result the atmospheric dispersion factors used in the accident analyses are more conservative than those we used in our calculations prior to issuance of the construction permit review.
A comparison of the atmospheric dispersion factors that we have derived from the site meteorological data with those derived by the applicant are shown in Tables 4.2,4.3 and 4.4.
1be differences between our dispersion factors and the applicant's result from our more conservative assessment of some of the plant features.
For example, the applicant considers that gaseous releases th rough the plant roof-top vent would be elevated because of an effective stack height resulting from the 125 ft/sec exit velocity *.
The plant vent is a pipe extending 10 feet above che roof of the 170 foot tall reactor building.
We concluded, however, that the co~
bination of exit velocity and release height is not sufficient to assure that effluents would be free of reacto~ building wake effects that could pOHsibly draw releases, emitted from the plant vent, back to ground on the lee side of the reactor building.
We have calculated an annual average atmospheric dilution factor for a ground level release at the most critical site boundary (900 meters to the south) of 3.5 x 10-5 sec/m3*
Our meteorological con-sultants (NOAA) have concurred, in their report (Appendix C), with this atmospheric dilution factor for the annual average dispersion conditions.
The applicant, however, assumed that releases through the plant vent represent an elevated source and calculated an annual average dispersion factor of 1.37 x 10-6 sec/m3*
Release limits for gaseous effluents from the plant, based on our more conservative annual average atmospheric dilution factor, have been incorporated into the Technical Specifications.
The applicant will have a continuing onsite meteorological measurements program after Fort St. Vrain commences operation to measure wind direction and wind speed and temperature at ground level and at higher elevations.
We have concluded that the instrumentation provided will provide meaningful atmospheric data when properly calibrated and correlated with data from the instrumentation on the meteorological tower.
Using this meteorological instrumentation to determine atmospheric conditions. the applicant will make controlled.
batchwise releases of radioactive gaseous waste only during those periods when favoTable dispersion conditions exist.
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2.5 Environmental Radiological MOnitoring The principal requirements for the applicant's environmental radiatjon monitoring program are listed in the Technical Specifications.
These requirements state that a minimum program will be maintained to sample forage, soil, water, bottom sediments, aquatic biota, air, and milk and to measure terrestrial gamma radiation.
Crops grown on the plant property will also be sampled.
Recommp.ndations from our con-sultants, the Fish and Wildlife Service of the U.S. Department of the Interior, have been incorporated into the applicant's environmental radiation monitoring program.
We conclude that the applicant's program will be adequate for monitoring the radiological aspects of plant operation on the environs and for assessing the health and safety aspects of the release of
~* q,d i oac t iv i ty to the environment from the operation of the plant.
3.0 FACILITY DESIGN 3.1 General The Fort St. Vrain Nuclear Generating Station consists basically of a high temperature helium-cooled nuclear reactor which supplies super-heated steam to a conventional turbine generator.
The entire reactor primary system, including the steam generators and the helium circula-tors, is located within a steel-lined prestressed concrete reactor vessel (PCRV).
This PCRV not only serves as a pressure vessel for the reactor. but also provides a containment function similar to that provided by the containment building for a water cooled power reactor.
For this reason, each of the penetrations in the PCRV is equipped with two separate, independently sealed closures, and lines emerging from the vessel are purposely made small in diameter and are equipped with redundant Deans of isolation.
The PCRV is housed in a conventional steel frame building (the reactor building) which is sufficiently leaktight to assure positive flow of ventilation air through filters and to a vent.
While this building does not provide a containment function as such, it does serve as a confinement structure for fuel and waste handling activities in a manner analogous to that provided by the reactor building for a boiling water reactor.
MOst of the auxiliary systems directly associated with the reactor and the PCRV are housed in the reactor building.
The control room, the turbine generator, and most auxiliaries associated with the steam system and the electric system are located in the adjacent turbine bUilding.
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'lhe reactor core structure consists of a cylindrical stack of hexagonal graphite blocks.
Each block is perforated by an array of vertical coolant holes.
Adjacent to these coolant holes are other sealed vertical holes containing the nuclear fuel.
This structure is supported by a floor about mid-height in the PCRV.
The lower part of the PCRV contains the steam generators and the compressor rotors of the helium circulators.
Coolant flows downward through the reactor core, the steam generators, the circulators, and then back to the top of the core via an annulus between the t*!a11 of the PCRV and the core.
Since the temperatures of the circulating helium are quite high, cooling coils and thermal insulation are provided for the PCRV liner and the support floor.
Control rods enter the top of the core and move within vertical channels in the hexagonal graphite blocks.
Although the use of a prestressed concrete reactor vessel for a reactor is new in the United States, several gas-cooled reactors utilizing PCRV's are in operation or under construction in France and England.
Some of these European reactors utilize a plant design and arrangement quite similar to that used for the Fort St. Vrain plant.
3.2 Reactor Design 71:1! reactor is a high temperature gas-cooled reactor utilizing a t' "a,' i um-th or i um fuel cycle.
The fuel is in the form of ThUCZ and
~ '; C " particles coated with Layers of pyrocarbon and silicon carbide.
i: ': " ;.e I particles are bonded in a graphi te matrix in the form of
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(' '::'>:=':,/Jnal graphite blocks of the core structure.
The blocks of
'_'.lI.:J.clar grade graphite (H-327) are stacked to form the 19.5 feet in diameter by 15.6 feet high active core.
The core is surrounded by about a four-foot thickness of unfueled, solid graphite reflector blocks.
A very small fraction (much less than I percent) of the fission products produced within the fuel is expected to diffuse through the relatively impermeable fuel coatings and be absorbed by the graphi te or pass into the coolant gas-stream.
Fission products in the coolant srre.un will be removed by a bypass purification sys tern and by plateout and decay in the coolant system.
Reactivity control is provided by 37 pairs of B4C-graphite, steel clad, control rods which operate in vertical holes in the graphite.
Each adjacent pair of rods is operated by a winch and cable type drive.
A reactor reserve shutdown system, consisting of small spheres of B4C in graphite stored in hoppers over the core, is also provided.
Remote manual actuation of the reserve shutdown system would allow the granules to fall through guide tubes into separate channels provided in the core.
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1 The hellum coolant enters the top of the core at a temperature of 760°F and a pressure of about 700 psis and flows downward through the core wherp 1t Is heated to about 1430° F.
The flow is then divided b(~tween two identical cool ant loops J each contain! ng two helium cir-culators and a six-modulestE!am generator.
3.2.1 Fuel Design '
The fuel 'f or the Fort St. Vrain reactor consists of fissile and fertile fuel particles.
'The' fissile particles "cont ai n both thorium and uranium~235 (93% enriched), and the fertile particles contain only, thorium.
The fissile, 'lbUC2' and fertile, ThC2, particles are provided with 'a three-layer coating (TRISO-II), consisting of an inner layer of porous pyrolytic carbon, an intermediate layer 'of silicon carbide (SiC) J and an outer layer 0 f highly impervious. high densi ty isotropic pyrolytic carbon.
The coatings on fuel particles serve to contain fission gases released from the fuel kernel.
Results from the applicant's irradiation test program have demonstrated the ability of properly designed ' and fabricated fissile' and fertile particles to experience the maximum burnup (20%), maximum temperatures (about l300°C), and maximum fluence (about 8.0 x 1021 n/cm2) predicted for fuel in Fort St. Vrain with less than 1% coating failures.
Experiments conducted by the Oak Ridge National Laboratory confirm that the particle can operate under these conditions with no significant coating,failures and very low fission product release.
Based on the results of these experimental programs J we conclude that the fiss ion product release from fuel particles coating contaminants (uranium and thorium) and from diffusion of fission products through intact coatings will be small under relatively normal reactor operating conditions.
The fuel particles are bonded together in a hot-injection molding process to form a fuel rod one-half inch in diameter and about two inches long.
In this process, the required -coated particle loadings are blended into a close-packed array (about 65 volume percent) in a mId cavity.
A fluid binder (as powdered graphite filler suspended 'i n coal tar pitch) is then injected into the mold (at about 20QoC) to fill the interstitial space and to surround each particle with matrix material.
The rod is cooled I
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The performance of the fuel particle in bonded rods has been exaadned experi-mentally.
The applicant's prototype fuel has been irradiated in the General Electric Test Reactor.
Hot-injected fuel rods were irr~diated for various combinations of fluence and tempera-ture.
Test results have establiahed the ability of the fuel to operate for the proposed aix-year cycle with less than 1%
failed coatings.
The percent of failed fuel is small.
The reactor and associated plant have been designed for operation with coatin~ failures as high 88 10%.
The primary concern related to fuel particle coating failures is an increase in the fission product inventory in the primary helium coolant.
The Technical Specifications contain a limdting condition of opera-tion th~t states the maximum quantity of fission products that may be present in the primary coolant.
This control on primary coolant activity haa been established to assure that releases of radioactivity in gaseous effluents resulting from normal operation of the facility or as a result of an accident will be kept within acceptable liadts.
Baaed on the results of the experiments to determine the effects of irradiation on fuel particles and on bonded fuel rods and on studies of the mechanical integrity of the fuel particle coatinga for long periods of time in a thermal gradient. we have concluded that the design is satisfactory.
3.2.2 Core Physics During our review of the Fort St. Vrain facility prior to issuance of the construction permit. we evaluated the analytical techniques used by the applicmt for the core physics design. and concluded that they were acceptable.
We have reviewed the core physics inforuation 'given in the Final Safety Analysis Report and have determined that the calculational methods and nuclear cross section information as checked by calculations of critical experiments and the Peach Bottom Unit 1 plant startup and test program (Cor~s I and II) are adequate to calculate safety-related reactor parameters to within required margins.
T~ *;e ind~.vidual and banked control rod worths t temperature reactivity coefficient (negative over temperature ranges of interest) and power reactivity defects, reactivity shutdown margins and gross and local power distributions for normal and abnormal control rod confdgurattons have been determined satisfactorily and are acceptable.
3.2.3 Reactor Control nte reactor is contro lIed by 74 control rods operated in pairs by 37 control rod drive mechanisms that are mounted inside the PCRV top head refueling penetrations.
Each mechanism is an electrically-powered cable winch capable of fulfilling the regulating, shim and rapid shutdown (scram) requirements of the reactor.
Rods are inserted by gravity.
During scram, the speed of the rod is limited by braking the cable drum to prevent mechanical damage to the control rods.
Burnable poison rods, in the form of particles of boron carbide in a rod of graphite, are installed in the fuel ele-ments to reduce the control rod requirements, liDdt control rod movement during full power operation, and to help control the overall core power distribution.
As a result of depletion of the burnable poisons, the Ddnimum reactivity shutdown margins occur during the initial fuel cycle at about mid-cycle.
Wi th all rods :l.nserted the minimum shutdown margin is -0.082 l\\ k at room temperature.
With the highest worth rod pair withdrawn the reactor would s till be shutdown by -0.037 6k at room temperature about mid-cycle of initial core.
The minimum shutdown margin is -0.014
~k during the equilibrium cycle wi th the maximum worth rod pair wi thdrawn at room temperature.
The control rods consls t of a s tack of annular absorber sections (boron carbide) sheathed in concentric tubes, with end caps attached to a Jpline passing through the assembly.
In the 19 central control rod pairs the boron carbide molded compacts contain 30 weight percent boron, and in the 18 outer control rod pairs, the compacts contain 40 weight percent boron.
The outer diameter of the control rods is 3-1/2 inches and the hole in the graphite in which it travels is 4 inches in diameter.
This clearance permi ts the rod to pass freely for any credible combination of bowed and displaced holes in the central stack of fuel elements.
A crushable-type shock absorber, attached to the lower end of each control rod, is capab Le of absorb ing the maximum kinetic energy developed in the 120 pound assembly, should the cable fail, without damaging the lower graphite reflector b locks or caus ing a loss of function of the control rod.
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r, l The control rod drives consist basically of e1ectric-motor-driven drums, or winches, that raise and lower the control rods by means of flexible steel cables.
'nle design of the drive mechanism provides for redundant indication of control rod position th rough the use 0 f potentiometers, redundant "rod-in" and" rod-out" limit switches, and redundant indication of broken support cables by means of slack cable sensing swi tches.
Based on our revlew of the design and testing programs for the control rods and their associated control rod drive mechanisms, we conclude that there is reasonable assurance that they will meet their functional requirements.
A reserve, or backup, reactor shutdown system independent of the normal control rod system has been provided.
It consists of 37 storage hoppers containing charges of graphite-boron carbide balls (7/16 and 9/16-inch spheres).
Sufficient negative reactivity control will be provided by this reserve sys tern to shut dawn the reactor to refueling temperature from any reactor operating condition without movement of the control rods, even if two of the reserve shutdown storage hoppers were inoperative.
'!be applicant has performed tests on a full-scale prototype reserve shutdown system under conditiona of temperature and helium chemical purity that it will experience during normal operation.
Provisions have been made for in situ tes ting of the integrity of each reserve shutdown hopper without actually releasing the neutron absorber.
In addition, periodic functional tes ting will be performed by removing individual control rod drive-reserve shutdown system and orifice control assemblies and placing them in the hot service facility (a
shielded remote handling facility) for testing and inspection.
We have concluded that the design and testing program will give reasonable assurance that the reserve shutdown system will perform its intended function.
3.2.4 Core Mechanical Design Prior to issuance of a construction permit, we reviewed the core mechanical design of the Fort St. Vrain facility and concluded that it was acceptab Le, We have also reviewed the mechanical design information con-tained in the FSAR, and have considered the loadings on the core structure and the other reactor internals which result from dead weight of materials, helium coolant flow, pressure differential forces, and seismic forces.
The seismic analysis of the reactor internals uses the PCRV support ring response as input data, and the vessel internals are modeled as a lumped mass system with modal analysis -
respon~e spectrum procedures being applied. The resulting computations indicate that the design as currently implemented is acceptable.
Dynamic responses to acoustic pressures and flow excitations have been considered in the design of the Fort St. Vrain reactor internal structures.
Extensive test programs have been conducted on fuel assemblies, core support structures, control rods, and circulators to identify forcing functions for vibration analysis and to confirm the calculated responses of critical internal components.
Each component was tested under normal and abnormal conditions which reflected anticipated and unusual operating conditions.
Acoustic pressure level and vibration amplitude measurements will be made during the pre-operational hot flow tests.
After completion of the hot flow tests, the applicant will submdt a report containing test results, visual observations, a compari-son of measured and calculated values, and an evaluation of measurements or observations, if any, that exceed acceptable limits.
Based on our review of the design and the testing program, we conclude that the core mechanical design is acceptable.
3.2.5 Thermal and Hydraulic Design The helium coolant gas flow in the Fort St. Vrain reactor 1s downward through the core.
Helium enters the core inlet plenum at a temperature of about 760°F and exits from the bottom of the core at a mixed-mean temperature of about l440°F.
The total helium flow rate is about 3.4 x 106 lb/hr, and a typical pressure loss of about 8.4 psi is calculated for the core.
The amount of coolant-gas flowing through each refuel-ing region is adjusted during operation by means of remotely-operated flow control orifice mechanisms located at the top of the core.
This adjustment provides means to keep the exit gas temperatures from all refueling regions about the same.
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In addition, the applicant has conducted flow experiments using an approximately one-half scale model of a portion of the reactor core.
The maximum fuel temperature was "det ermined to be about 2300°F with a volume median temperature of 150QoF and a minimum temperature of 800°F.
A maximum heat transfer coefficient of 400 BTU/hr-ft2oF and a maximum surface heat flux of 140,000 BTU/hr-ft2 were calculated.
Based on the analytical method used, the thermal properties, the hot spot evaluation, the assessment of the uncertainties in physical properties of core materials, and possible dimen-sional variances in the fuel elements, we have concluded that the reactor thermal and hydraulic design is acceptable.
3.3 Primary Coolant System The primary coolant.ystem for the Port St. Vrain reactor consists of two coolant loops, each containing two helium circulators, helium shutoff valve., and a steam generator consisting of six parallel DOdule-type exehangers with integral superheaters and reheaters.
The entire primary coolant system i. loeated within the prestressed con-crete reactor vessel (PCRV).
After. passing through the core, the coolant gives up its heat to the 8team generators and is discharged into the circulator inlet plenum.
The circulator compresses the helium and directs it through the helium shutoff valves to the circulator outlet plenum from which it returns to the reactor core inlet plenum.
3.3.1 Helium Circulators The four helium circulators are mounted in penetrations located in the bottom head of the PCRV.
A circulator consists of a single-stage, axial-flow helium compressor driven by either a single-stage, axial flow steam turbine or an auxiliary water turbine when steam supply is unavailable.
The circul~tor steam turbine normally operates on cold reheat steam from the exhaust of the high pressure stage of the main turbine.
The water turbine normally is driven by the steam generator feedwater supply, but is also capable of being driven by condensate from the condensate pumps or by water from the firewater pumps.
Anyone of the four circulators driven by either steam or water is capable of providing adequate gas flow to cool down the shutdown reactor, even if the primary coolant system has been depressurized to atmospheric pressure.
Based on the results of the developmental, operational, and acceptance tests performed on the helium circulators, we con-clude that there is reasonable assurance that the helium circulators will perform as intended.
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Orificing System An inservice inspection program for the primary coolant system 1s required by the Technical Specifications.
Although inspec-tion of the primary coolant system pressure barrier in a large Based on our review of the design, the installation procedures, production, and preliminary operating tests, we conclude that the orificing mechanism is acceptable.
Inservice Inspection In order to compensate for the variations in power generation 1n the various regions of the core, a flow control system has been provided for each of the 37 fuel regions.
The flow con-trol system is composed of an orifice control mechanism and a variable-orifice flow control assembly.
The orificing control mechanism is mounted in the PCRV top head above a radiation shield and immediately below the control rod drive mechanism.
The -I"q1.gn arrangement of the helium circulators is such that the
&..ugle-*stage axial flow helium compressor proj ects up into the circulator inlet plenum region.
A missile barrier has bean provided to protect the equipment located within the PCRV fr;,
missiles that could possibly be generated by the failure of the compre8sor.
This barrier is an annular ring that surrounds the compressor disc and blading.
The applicant has conducted a series of developmental tests on 1/4-scale models of the disc to establish the composition and location of this missile barrier (disc catcher).
'!he applicant also has conducted a series of tests to verify the capability of the circulators and helium shutoff valves to withstand postulated accident conditions without loss of function and to remain operational following a rapid depres-surization of the PCRV.
Tests have been conducted to deter-mine the response of circulator assemblies to thermal shock resulting from transfer from steam to water drive, overspeeds greater t.han 130% of design speed, and depressurization rates corresponding to the design basis depressurization accident.
3.3.2 3.3.3
3.3.4 lITGR using a prestressed concrete reactor vessel is difficult because of liner inaccessibility, the applicant has developed a liner integrity surveillance program designed to detect and monitor any det~rioration in the liner.
This program uses ultrasonic inspection equipment in addition to removal and inspection of assemblies consisting of portioos of the thermal barrier and liner plate.
The applicant will review the inservice inspection program after five years of reactor operation and at that time will propose any necessary modifications, based on the five years 0 f experience, to the AEC for review and approval.
We conclude that these provisions for the inservice LuspectIon program for this reactor are acceptable.
Steam Generators The primary coolant system s t.eam generators and reheaters are the heat sinks for the reactor for all normal, shutdown, and off-normal operating modes.
T'ne steam generators are located in each of the primary coolant system loops and consist of six modules located beneath the core support floor inside the PCRV.
The individual modules consist of economzer-evaporator-superheater and steam reheater sections.
The sections are made up of heH ca.lIy-wound tubes supported by perforated support plates.
The steam generatorR and reheaters are designed suCh that only individual tuLes and small subheaders penetrate the PCRV pri-mary and @econdary closures.
A flow limitin~ orifice is pro-vided in each of these subheaders to limit the rate at which steam and water would be introduced into the primary coolant system in the event of a rupture of a steam generator tube or subheader.
The steam generators have been fabricated in accord with the Class A requirements of Section III of the ASME Pressure Vessel Code and the power piping code USAS B3l.I.0.
A full-size steam generator module has been subjected to air
~low testing to confirm theoretical calculations of shell-side (gas-side) flow distribution, flow resistance and vibration properties outside the steam generator tubes.
In addition, air flow tests were uged to dete~ne the acceptability of the sleeve-wedge tube protectors provided at all potential points of contact of steam generator tuclng.
Based on our review of the design and the design criteria used, we conclude that there is reasonable assurance that the steam generators will be able to perform their intended function.
3.4 Prestressed Concrete Reactor Vessel and Class I Structures 3.4.1 Prestressed Concrete Reactor Vessel The reactor and the entire primary coolant system is contained in and shielded by the prestressed concrete reactor vessel (PCRV), a right circular cylindrical structure having an inside cavity about 31 feet in diameter and 75 feet high with walls 9 to 15-1/2 feet thick.
The concrete walls and head are constructed around a 3/4-inch-thick low-carbon-steel liner and are reinforced with bonded reinforcing steel and prestressed with steel tendons.
The major penetrations are located in the top and bottom heads of the PCRV, with only a few small diameter penetrations in the side walls.
Each penetration is provided with two clos-ures and seals, both of which are designed and tested at the design (reference) pressure.
The interspace between these closures is filled with purified helium at a pressure slightly greater than the primary coolant helium pressure.
The only lines connected to the primary cooiant' system that pass through the walls of the PCRV are of small diameter.
The largest of these is the regeneration line of the helium purification system with an inside diameter of approximately two inches.
The PCRV is designed for a "reference" pressure of 845 psig, compared with the normal and peak working pressures of 688 and 704 psig,**"I:'espectively-,' -and has been pro'of-t"estetf"at '115% "
of this reference pressure.
The design limit on leakage of primary coolant containing radioactive impurities is 1% of the primary coolant inventory (weight) per year.
The maximum permissible leak rates for the primary and secondary closure seals have been detenDined and are included in the Technical Specifications.
We reviewed the PCRV design and aesign analysis prior to issuance o~ the construction pennit and determined that they were acceptable.
We have reviewed the information con-tained in the applicant's FSAR and determined that the material I
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3.4.3 Desi$n Criteria and Seismic Analysis 3.4.2
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therein is substantially the same as that reported during the construction permit review and does not alter our conclusion.
The FSAR contains a compilation of previously-submitted infor-mation on the design of the PCRV as w~:i.l as new updated material.
The analytical and experimental studies of the PCRV top and bottom head are also described.
The design analysis uses finite element elastic, elastic-creep, and elastic-crack analysis pro-cedures to evaluate the design and to demonstrate conformance with the applicant's criteria.
Based on the manner in which the PCRV has been designed and constructed, we conclude that the PCRV is acceptable.
Class I Str~ctures The reactor-turbine building is a composite steel frame at i uc-ture with concrete shear walls.
The design criteria for this building, as specified by the applicant, include the allowable stress values of the ACI 318 and AISC Specifications (1963 editions) for load combinations involving the operating basis earthquake (OBE).
An increase of one-third of these va.lues for load combinations involving tornado and the design basis earthquake (DBE) have been used.
The design criteria and elastic analysis procedures used are acceptable.
The FSAR c0"tains a detailed list of Class I structures, systems and equipm~nt.
All Class I structures, systems, and equipment, with only a few exceptions, are located in the reactor-turbine building, which itself is designed as a Class I structure.
The classificationa of structures, systems, and equipment are appropriate and acceptable.
The design and analysis of all Class I structures other than the PCRV has followed conven-tional practice and is acceptable.
Design basis loadings arising from seismic, wind, and tornado forces have been considered in the design.
Detailed studies of regional and site geology and seismology and local soil conditions have resulted in selection of peak foundation acceler-ation values for the operating basis (OBE) and design basis (DBE) earthquakes of 0.05 g and 0.10 g, respect ively.
The design wind load is that associated with a wind velocity of 100 mph.
The design tornado loads include consideration of a tornado resulting in a 3 pai differential pressure in 3 seconds, a 300 mph uniform horizontal wind, and tornado-driven missiles.
The missile
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resistant nature of the reactor building siding has been established experimentally as being suitable to withstand the design missile:
a fir plank, 12 ft. by 1 ft. by 4 in.
weighing 105 lbs.
The loadings specified by the applicant are acceptable.
The principal seismic-dynamic structural analyses for the plant were the PCRV-reactor-turbine building analyses and the reactor internals analyses.
The PCRV-reactor-turbine building analysis was a detailed modal-response spectrum analysis considering the torsional characteristics of the reactor-turbine building, the nonsymmetric reactor support flaxibility, the bounding of building foundation conditions, and the dynamic foundation coupling between this building and the PCRV.
The resultant seismic stresses in the reactor vessel are law as shown in Table E.l3-l of the FSAR.
The procedures used and the results reported are acceptable.
3.4.4 PCRV Thermal Barriers The PCRV thermal barriers perform the important function, in conjunction with PCRV cooling systems, of maintain~ng the liner and adjacent concrete at acceptably low temperatures (150-200°F).
The applicant has described the design and the results of the Thermal Barrier Research and Development Program in the FSAR.
Class A insulation, consisting of one or two layers of Kaowool, a ceramic fiber blanket material, is designed for service in areas where the hot side is exposed to helium at 755°F (normal operating temperature at 842 MWt).
Class B insulation, consisting of layers of silica felt, is designed for service in areas where the hot side is exposed to helium at 1470°F.
Class C insula-tion is a composite consisting of one layer of Kaowool and three layers of flat blocks of slip-cast ailica.
The normal operation design of the Class C thermal barrier is based on a full power-steady-,tate hot face operating temperature of 2000°F.
Tests of the thermal barriers have included chemical properties, materials handling, permeability, vibration, conductivity, moisture absorption, depressur~zation, thermal cycling, fric-tion and wear, and fatigue tests for the attachment fixtures.
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Structural Acceptance Testing and Surveillance 3.5.1 3.5.2 20 -
Based on our review of the thermal barriers and the redu_dant rCRV c00ling system (coils), we conclude that the PCRV liner and adjacent concrete will be protected from experiencing excessive temperatures.
The structural acceptance testing program for the PCRV includes extensive instrumentation, controlled pressurization, determina-tion of deflection and cracking under load, evaluation of the results and the filing of a report thereon.
We have evaluated the testing program submitted by the applicant and consider it acceptable.
The in-service surveillance program proposed by the applicant is also acceptable.
The program includes continuous monitoring of 27 selected load cells and the taking 9f data point samples from strain, temperature, and moisture monitors at selected intervals.
At our request the applicant included a require-ment that selected tendons from each major group be monitored.
PCRV deflection and cracking, liner and tendon corrosion, and helium permeation will be monitored at selected yearly intervals.
GENERAL The design criteria for safety related plant instrumentation, as stated in the FSAR, are in agreement with the criteria we evaluated and approved prior to issuance of the construction permit.
In our review of the application for an operating license, we have evaluated the design and the conform<1nce of the installation to the criteria.
The following paragraphs discuss the instrumentation and controls for various systems.
Protection Systems The reactor plant protection system consists of the instrumenta-tion and controls requlred to ir.itiate automatic corr~ctive actions upon onset of a potentially unsafe condition.
The protective systems include (1) the reactor protective circuitry, (2) the instrumentation and controls for the engineered safety features, and (3) circuitry provided to protect various plant components from major damage.
3.5.2.1 Reactor Trip System The reactor trip system instrumentation and logic circuits, with th~ exceptions of the sensors and output devices, are located in the main control room.
The three inde-pendent channels are located in vertical panels separated by ph~sical and thermal barriers.
Output signals from indicating devices located on other control boards are individually isolated at the reactor trip control board to prevent a failure of these remote circuits om affecting any reactor trip signal.
Manual reactor trip can be actuated from the control room or the switchgear room.
The contacts of both manual trips are connected in series with each of the three logic matrices and are independent of the automatic reactor trip circuits.
The final design of the system is in agreement with the preliminary design evaluated during the construction permit review.
Our review of the implementation of the design identified certain items of nonconformance.
In order to assure that the essential circuits were properly designed and installed, the applicant has conducted s detailed audit of the design and installation.
The applicant has certified that all exceptions to the criteria identified in our r~view and in its audit have been redesigned and are n~. installed or will be installed in accordance with the criteria.
On the basis of our review of the design, the installa-tion, and the applicant's audit and rework of the reactor trip system, we conclude that the system meets the require-ments of the Commission's General Design Criteria (GDC),
and the proposed Institute of Electrical and Electronics Engineers Criteria for Nuclear Power Plant Protection System (IEEE--279) dated August 28, 1968, and is acceptable.
3.5.2.2 Steam Generator Dump and Loop Shutdown Steam generator dump and loop shutdown are initiated by signals indicating loop high moisture, reactor high pressure, or steam generator penetration innerspace high pressure.
In addition to the signals that initiate both steam gener~tor dump and loop shutdown, the following signals initiate loop ghutdown only:
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High reheat header activity b.
Low superheat header temperature c.
Steam pipe rupture d.
Trip of both circulators in a loop.
The applicant has included these circuits in his audit and has certified that these circuits are now, or will be upon completion of the circuit installation, in conformance with the approved criteria.
We conclude that the steam generator dump and loop shutdown circuits comply with the IEEE-279 criteria and are acceptable.
Safe Shutdown System The instrumentation and controls required for safe shutdown and removal of decay heat following the occurrence of an earthquake or violent storm, were designed to comply with the IEEE-279 criteria.
In our review we requested that the applicant include the safe shutdown circuits in its audit for compliance with its criteria.
The applicant reviewed the desi~ and installa-tion and has certified that this circuitry is now, or will be upon completion
~f the circuit installation, in compliance with its criteria.
Based on our review of the design, the installation, and the applicant's audit of the safe shutdown instrumentation and controls, we conclude that they are in conformance with the criteria and are acceptable.
3.5.4.
Other Instrumentation We have reviewed the instrumentation for the control rod drives, steam generator dump tanks, and PCRV cooling systems.
We find that the instrumentation will provide the information necessary to determine the status of these components and systems even in the event of a single component failure and conclude that it is acceptable.
~he reactor helium outlet coolant temperature monitoring system is provided for monitoring the outlet temperature from each of the 37 refueling regions in the reactor core.
Together with the overall coolant flow, overall core power, orifice position, and control rod position, the coolant temperature thermocouples are used to assure acceptable fuel temperatur.es during power operation.
As initially proposed, only one of four thermocouples per region was connected to the readout instrumentation.
At our request the applicant has modified this design and will monitor two thermo-couples per region.
One thermocouple from each region is connected to a scanner and the remaining one from each region is connected to a data logger.
On the basis of its redundancy and separation, we conclude that the helium outlet coolant temperature monitoring system, as now designed and installed, is acceptable.
3.6 Electrical Power System 3.6.1 Offsite Power The Fort St. Vrain Nuclear Generating Station will be connected to the Public Service Company of Colorado system through 230 kV circuits.
Power from the main generator feeds the 230 kV switch-gear.
The switchyard has two 230 kV buses arranged in a "breaker-and-a-half" configuration.
There are four transmission lines connected to the switchyard.
There are three rights-of-way with steel towers spaced such that two independent transmission paths are located on the site and three independent paths leave the site boundary.
The towers are spaced on the three rights-of-way such that collapse of a tower would not affect any other tower or transmission path.
A single reserve auxiliary transformer is provided for startup and shutdown plant requirements.
A back-up source of power for loads connected to the reserve auxiliary transformer can be provi ded by backfeeding through the main and unit auxiliary transformers.
The applicant has stated that the connections required for backfeeding can be made in less than two hours.
We have determined that this is acceptable for the Fort St. Vrain reactor since analysis indicates that a loss of all cooling for two hours would not result in fuel temperatures that might cause fuel particle coating failures.
The main turbine generator is provided with net load rejection capability and can supply station power in the event offsite power should be lost.
The ability of the plant to withstand a
loss-of-~1fsite power without reliance on the onsite emergency diesel generator will be demonstrated by tests simulating loss-of-power while the generator is supplying at least 50% of full load.
The system transient stability study made by Public Service Company of Colorado shows that the system will maintain stability on loss of the Fort St. Vrain main generator while operating at full load.
Consequently, 10s8 of the main generator is not expected to result in interruption of offsite power to the safety-relAted plant loads.
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We conc l ude th at aufficlent redundant and independent BoUrCt~B of offHlte power are provided to assure that no single failure could C8UHe the lOBR of offs i te powe r to the engineered H8fcty fenturea or components and systems needed for safe shutdown of the plant, that the offsite pover meets the requirements of General Design Criterion 17, and that it is accep tab Le.
1.6.2 Onsite Power The essential safety-related loads are arranged on two 480 Volt essential buses either of which will supply minimum safety requirements.
TWo 1210 kW diesel generator sets are provided, with one set assigned to each 480 volt essential bus.
A third 480 volt bus (alAo designated an essential bus) is provided and is supplied by either of the other two essential buses.
The applicant's analysis indicates an expected safety load require-III!nt of about 1150 kW with a minimum safe shutdown load of 866 k.W, both of which are well within the continuous rating for a Ringle die~el generator set.
Each diesel generator set il!Ji started automatically upon loss of voltage on the unit auxillaiy transformer, the switchyard bus, or its own 88sociated 480 volt./bus.
The essential safety loads are automatically sequenced on ~ach diesel generator set.
Redundant sequencers are prOVided for 'each 480 volt bus.
The third 480 volt essential bus is automatically connected to the bus supplied by the first diesel generator s~t to become ready for service.
We have reviewed the loading sequence proposed by the applicant and have concluded that the essential loads will be supplied and that required-safety related loads will not overload a single diesel generator set.
The diesel generator sets are located in separate Class I (seismic) rooms.
Edch diesel generator set has its
~n lubrica-tion oil, fuel oil, and cooli~g and air starting system.
The a1 r starting capaci ty for each set will supply three lO-second cranking periods.
The fuel 0::1 supply available on site is more than sufficient for opert:ting one diesel generator set at full load for seven days.
Each Fort St. Vrain diesel genuratnr set consists of a generator rated at 1350 kW (continuous) a~iven by two diesel engines, each rated q,t 605 kW.
The diese1 en.dnes are connected to the generator th rov I an air-operated dfs conm ct device with one engine located
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3.7 AUXILIARY SYSTEMS 3.7.2 Cooling Water System i
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Some additional makeup water (260-350 gpm) is required to replace We conclude that the oneite power system meets the requirements of General Design Criterion 17 and is acceptable.
on either end of the generator drive shaft.
This is the first application of a tandem diesel generator set in a nuclear power plant.
General There are three separate and Lndependene 120 volt instrument power buses.
Two of the three buses are supplied by separate and independent dc-to-ac inverters.
The third bus is supplied by one of two transformers each connected to a separate 480 volt essential bus.
The 125 volt dc system con.lsts of two separate, redundant and independent 125 volt batteries each with its own charger and dis tribution bus.
The batteries are located in two separate, independently ventilated rooms.
Each battery will supply power to minimum shutdCMn dc loads for at least one hour after loss of all ac power.
We conclude that the 120 volt ac and the 125 volt dc systems are acceptabIe.
We have reviewed the auxiliary systeus provided in the Fort St. Vrain Nuclear Generating Station.
The systems reviewed include the systems discussed in the fol~owing paragraphs, and the helium purification system, the helium storage system, the nitrogen system, the instrument and service air system, and the fire protection systems.
We have concluded that these systems will perform their intended functions as required.
Since this is the fi rst nuclear application of this. diesel generator arrangement, we have required that the applicant perform tests to demonstrate that the probability that a diesel generator set will fail to start is less than one chance in a hundred at a 95 percent confidence level.
We have reviewed the applicant's proposed test program and have concluded that successful comple-tion of this program will demonstrate the required reliability.
3.7.1
3.7.3 water continuously drained from the cooling tower basin (blow-down) in order to control the concentration of soluble solids
- 1.0 the cooling tower recirculati.ng water.
Circulation in the service water system is provided by three electric-motor driven pumps.
Each pump is capable of deliver.ing one-half of the required circulating water flow (about 5250 gpm).
Cooling for the main steam condenser, located beneath the main turbine, is supplied by the plant circulating water system.
The circulating water 1s cooled in the ten-cell induced-draft main cooling tower.
Makeup water (about 5500 gpm) for evaporative and blow-down losses from the main cooling tower is supplied f,om two large storage ponds, each having a capacity of 1.1 X 10 gallons.
Water for the storage ponds is pumped from the South Platte River or St. Vrain Creek or both.
Flow in the plant circulating water system is provided by four electric-motor driven pumps.
Each of the two larger pumps circulates about 67,000 gallons per minute and each of the two smaller pumps circulates about 11,000 gallons per minute.
All four pumps are required for normal full power operation.
Interconnections are provided between the circulating water system and the service water system such that the main circula-ting water system can be used to supply cooling to essential plant systems and components in the event the service water system should be unavailable.
During normal operation the valves in the interconnections are closed.
All structures, components, and piping in the service and circulating water systems, that are required for cooling the shut down plant and maintaining it in a safe condition, have been designed in accordance with Class I seislnic and tornado protection design criteria.
In addition, certain essential plant cooling systems have interconnections (valves closed during normal operation) that permit the fire-water pumps to supply cooling water directly to the heat exchangers under emergency conditions (as described in Section 9.8.3 of the FSAR).
On the basis of the above considerations, we conclude that the cooling water systems are acceptable.
- ouel Handling Refueling takes place through the refueling penetrations located in the top head of the PCRV after reactor shutdown, coolant
3.7.4 27 -
depressurization~ and removal of the control rod drive mechanism-orifice control assembly and the control rods for the refueling region.
The removal of spent fuel and reloading of new fuel is accomplished by means of a fuel handling
~chine which is lifted and transported by the reactor building overhead crane.
Shielded gate-type isolation valves are used at the peRV refueling pene-trations and at the lower end or the fuel handling machine to protec~ the operating personnel during transfer of radioactive components and to preserve the helium atmosphere in the reactor and the fuel handling machine.
The fuel handling machine consists of a cask and an extensible fuel transfer mast which has a ruel grapple capable of azimuthal, vertical, and radial positioning.
We have reviewed the design, the planned sequence of operatIo.ia and the extensive developmental, prototype, and preoperational testing that has been and will be conducted on the equipment required during fuel handling and conclude that the equipment will perform its intended functiont Fuel Storage The irradiated fuel storage facility consists of nine fuel storage wells supported in three concrete shielded vaults.
Each storage well consists of a cylindrical tank with a watertight subcompart-mente Cooling tubes along the outer walls of this compartment maintain the temperature of the fuel elements below 750°F.
Each well opening is provided with high-integrity seals and a shield-plug-closure to maintain a dry helium atmosphere in which the uncanned irradiated fuel and reflector elements will be stored.
Each well contains four storage compartments for fuel elements surrounding a central compartment used for storage of irradiated reflector elements.
The nine wells have storage capacity for one-third of the fuel from the core (about 490 elements) plus a total of approximately 1~0 refle~tor elements in the central compartments.
In addition to the redundant water-cooled heat removal system provided for each storage well, 1500 cfm of ventilating air is routed through each vault to serve as a back-up heat removal system in the event all cooling water circulation b ~ ou l d be lost.
The applicant states that the design of the fuel storage wells precludes fuel arrangements that could achieve cri t Lca.lLty even if the wells were completely filled with fuel and flooded with
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3.8 RADWASTE MANAG~MF.NT SYSTEMS J.8.1 28 -
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Any fission gases released from the fuel, while in Htoragp. are vented directly to the waste gas system for handling and storage.
We have determined that the design of the spent fuel stor.age
~~cility provides redundant methods of cooling and satis-factory protectj ~n against oxidation of the fuel, against uncontrolled release of fission product gases to the environ-ment. against damage that might result from earthquakes and tornadoes, and against
~ n adve r t en t criticality
~f stored fuel.
We conclude that the fuel storage facility is acceptable.
Liq~!d Waste Management Szstem In Amendment 17, the applicant stated that the design obJe~tive of the waste treatment system for the Fort St. Vrain Nuclear Generating Station was to minimize the release of I ~I :;ioactive effluents and that the anticipated releases of radioactive materials in the plant efflu?nts will be small frac-tions of the limits specified in 10 CFR Part 20.
Since this is a gas-cooled reactor, there will be no source of routinely generated liquid waste.
The applicant has estimated that the total volume of liquid was te effluent will be approximately 3,000 gallons per year (containing approximately 0.1
~Ci).
The one source of routine liquid radwaste will be from the solutions used for decontamination of control rod drives and helium circulators.
In the unlikely event of the rupture of a steam generator tube, the contents of one steam generator would be discharged to the steam-water dump tank, which contains approximately 2,280 gallons of potentially radioactive water that will be transferred to the
]~quid radioactive waste system for storage and disposal.
TW0 other potential sources of small quantities of radioactive liquid wastes, that will be transferred to the liquid radioactive waste sys tern, are the main helium ci rculator bearings md the gas-liquid phase separators in the helium purifi cation sys tern.
The reactor building sumps are another potential source of contaminated liquid waste; however, the waste water from the sumps normally will not be contaminated.
Dtschr.rge from the
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29 -
reactor building sump pumps flows past the redundant liquid radwaste line radiation monitors and, in the event of high radioactivity, the8e monitors will terminate the discharge from the pumps and open valves to divert the flow into the liquid waste receiver tanks.
In addition to the protection provided by the radiation monitors, the Technical Specifications require that the applicant take and analyze two samples of the liquid in the reactor building sump prior to th~ initiation of discharge from this sump.
In the event that this lIquid is shown to be radioactive, the Technical Specifications require that it be processed through the liquid waste demineralizers.
The principal sources of potentially contaminated liquid waste flow to the liquid waste sump, and are pumped through liquio waste filters to the 3,000 gallon liquid waste receiver tank.
The applicant will discharge waste from the receiver tanks to the cooling tower blm~down line.
The Technical Specifications require that the ac~t',ity in the blowdown line, with 1,100 gallons per minute (gpm) dilution flow, must be within 10 CPR Part 20 limits.
If this criterion cannot be met, the waste will be demineralized through the liquid waste demineralizer until 10 CFR Part 20 limits can be achieved.
The Technical Specifica-tions require that no discharge be made unless the instantaneous concentration in the minimum cooling tower blowdown flow (1,100 gpm) will be less than 10 CFR Part 20 values.
The applicant states in Amendment 18 that the concentrations in the discharge in normal operation will be reduced to as near background as obtainable prior to discharge of the liquid waste.
Prior to discharging any liquid radioactive waste the applicant will analyze isotopically a minimum of two samples taken from the liquid waste monitor tank.
A final safeguard against the unplanned discharge of waste exceeding 10 CFR Part 20 concentra-tions is provided by the liquid waste discharge valves which are interlocked with the cooling tower blowdown flow and radiation monitors such that if the minimum b1owdown flow of 1,100 gallons per minute is not available and/or if the con-centration of radioactivity is too high, the discharge of liquid waste will be terminated automatically.
In Amendment 18, the applicant provided a change in the design of the waste discharge system so that the flow now enters a flow diversion box from whence it can be directed to either a concrete-lined canal, which leads to a farm pond (Goosequill Pond) or to a slough which ultimately flows into St. Vrain Creek.
According to the applicant, the normal discharge path will be via the Goosequill Ditch.
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3.8.2 This change in the discharge path provides the advantage of controlling the release of radioactive liquids so that any sedimentation and reconcentration would most likely occur on property under the cont rol 0 f the Publie Service Company rather than 1n a public stream (St. Vrain Creek).
We will review the applicant's operational reports covering waste releases.
If the environmental monitoring program detects significant reconcentration of radioactivity in the Goosequill Pond, the applicant will be required to take appropr Late action to restrict releases of radioactivity and/or to restr:i ct access to the ditches and Goosequill Pond.
Gaseous Was te Management System
'The principal source of radioactive gaseous waste is from regeneration of the low-temperature filter absorber in the helium purification system.
Each train of the helium purifica-tion system is housed in a special PCRV penetration, and four wells in the PCRV head.
A train consists of a high temperature filter-absorber (a charcoal bed that will remove the particulates and halogens from the helium coolant), a helium purification cooler, a puriftcation dryer, a low temperature gas-to-gas heat exchanger, and a low temperature absorber (for removal of the noble gases and other gaseoea impurities).
After leaving the low temperature absorb er, the purified gas flows to a hydrogen-getter unit (a titanium sponge) which is provided to remove gaseous hydrogen and tritium.
One of the two purification trains will nonmally operate for six months after which time it will be isolated, and the spare unit placed into operation.
!he isolated unit will be held for a minimum decay period of 60 days before regeneration.
During regeneration of the low temperature filter absorber, the desorbed gases are released to the waste gas system vacuum tank.
The high temperature filter absorber will never be regenerated and will be disposed of when it is no longer capable of performing its intended function.
The applicant will remove the titanium sponge periodically, without regeneration, and dispose of it as solid waste.
Both the low temperature filter absorber and the titanium sponge are regenerated by passing hot helium through these units, driving off the radioactive gases accumulated and carrying them to the offgas system vacuum tank.
These gases would then be released to the atmosphere under con-trolled conditions following sampling and analysis.
The Technical Specifications require that the applicant provide a minimum of 60 days holdup for the low temperature filter absorber unit radioactive gases.
Gaseous waste containing relatively high levels of radioactivity may occasionally originate in the instrument lines, sample lin~s, the steam wat~r dump tank, the helium purification system regeneration section, cooling water surge tanks, the full storage wells, and the PCRV core support floor vent.
Gases from these sources flow to the waste vacuum tank and are trans-ferred into the gasecus waste surge tanks by the gaseous waste compressors.
Potential sources of low-level gaseous radioactive wastes are collected in ventilation system headers and are sent to the radioactive gaseous waste management system.
Potentially contaminated wastes, originating from the purge system of the fuel handling machine, from the helium circulator handling cask purge system, and from the liquid waste tanks vent headers, flow through a waste gas filter composed of a pre-filter, a charcoal filter, and an absolute filter.
In the event of high level activity, this continuous flow would be terminated by the reactor building plant vent radiation monitors and valves would be repositioned automati~ally sending these gases to the waste gas vacuum tank in the event of high level activity.
Under controlled conaitions, the gaseous waste from the surge tanks is passed through the vent exhaust filters which are composed of a prefilter, an absolute particulate filter, and a charcoal absorber to the atmosphere.
The reactor plant vent radiation monitors will terminate discharge from the surge tanks automatically in the event of high level radioactivity.
Additional potential sources of gaseous waste effluents are the secondary system steam jet air ejector and the reheater relief valves.
In the unlikely event that a tube should fail in the reheater section of the steam generator, radiation monitors on the hot reheat line will actuate isola~ion valves to contain the leak.
The reheat line relief valves may be actuated to protect the isolated line against overpressuriz it i on, thus discharging gases into the plant exhaust system.
The discharged gases will be filtered through the plant exhaust filters prior to release through the plant vent.
The effluent I
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from the stearn jet air ejector, which is normally expected to be nonradioactive, is monitored by the plant vent exhaust uon Ltors prior to release from the plant vent.
The plant vent gaseous monitors measure and record the noble gas activity, and an isokinetic sampler provides meAsurement of the particulate aud iodine activity being released through the plant vent f zom all sources.
The applicant will maintain a frequent gas sampling routine based on the noble gas release rates.
The sampling schedule for determining the relationship of radioiodinee to the noble gases will provide reasonable estimates of iodine releases and reason-able assurance that the halogen release rate is well within the annual average release limits.
The only expected source of potentially contaminated gases containing radiohalogens would be from minor PCRV leakage to the reactor building.
The reactor building air is filtered through the reactor building ventilation system which includes particulate and charcoal filters.
Based on our evaluation of the design and installation of the waste management systems, we conclude that the waste management systems, operating within the Technical Specifications limits, will result in annual effluent releases that are small fractions of the limits specified in 10 CFR Part 20 and, thus, will be as low as practicable.
4.0 ACCIDENT ANALYSIS 4.1 General We have evaluated the response of the facility to various anticipated operating transients.
The events that characterize abnormal operating transients are described in various sections of the FSAR and are tabulated in Section 14.3.
Included among these transients are events such as a turbine trip, loss of offsite power, control rod malfunctions, loss of feedwater flow, and various o~~er incidents resulting from variations in operating parameters.
We have reviewed results from the applicant's analyses of these incidents and conclude that the facility as designed and cO~Rtructed can accommodate the occurrence of these transients without damage to either fuel or primary coolant boundary.
Consequently, these transients will not result in the release rf fission products to the environment.
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We also have evaluated a broad spectrum of accidents that might result from postulated failures of systems or components to perform their intended functions.
We have selected the loss-of-forced-circulation of the coolant and the rapid depressurization of the PCRV as being representative of acci-dents that, although they are highly i~robable, have potentially significant consequences.
The accidents have been designated Design Basis Accidents No. 1 and No.2 (DBA #1 and DBA #2, respectively).
In our conservative analyses of these design basis accidents, we have assessed the capability of the plant and its engineered safety features to control the possible escape of fission products from the facility.
The calculated potential consequences of the design basis accidents exceed those of all other accidents consicered.
The conclusions reached in our review and evaluation of the design basis accidents, as well as for a number of other representative transients used in assessing the safety characteristics of the facility, are summarized in the following sections.
4.2 Reactivity Transients The applicant has considered a number of postulated causes of reactivity transients and has concluded that the greatest potential reactivity insertion rate and magnitude would
~esult from accidental control rod withdrawal.
We have concluded that provisions incorporated in the PCRV design to prevent control rod ejection (i.e., redundant shear anchors for the refueling penetrations and separately anchored back-up cover plates for each refueling penetration) make the proba-bility of a rod ejection accident very low.
Accordingly, we conclude that the more likely rod withdrawal accident is representative of the moat severe reactivity transient that mdght be experienced for this reactor.
The reactivity transients were analyzed with the BLOOST-VI code, which is similar to the CHIC-KIN code used for many pressurized water reactor (PWR) analyses, i.e., 2-dimensional hEat transfer and peint kinetics.
This method is capable of giving conservative results for this reactor, which is neutronical1y small and sluggish compared to a large PWR.From our review of the calculations, we have concluded that the applicant used appropriate ccnaervatIams in selecting input paraw.eters (e.g., the use of weighting factors of unity for reactivity feedback coefficients and peaking factors. without feedback flattening, for use in peak temperature calculations) and that the results f0r average and peak temperatures are acceptable.
Furthermore, various I
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reactor parameters such as delayed neutron fraction and reactivity coefficients used in the calculation will give results that are acceptably conservative.
The results calculated for the rod withdrawal accidents indicate that the maximum temperatures reached in the fuel particles, in the case of withdrawal of either a maximum worth rod or all 37 rods during startup, assuming a scram occurs at 140% power, is less than l500°C.
This temperature would not result in fuel particle failure.
In addi-tion, withdrawal of a maximum wort'l. rod at full power would result in no fuel failures if scram occurs prior to 140% power.
Assuming no overpower scram and continuing the transient until a scram occurs on hot reheat steam temperature (1075°F, at over 100 seconds into the transient), the analysis predicts a failure of less than 2% of the fuel particles and predicts gas temperatures (about 2l00°F) which would not cause loss of system integrity.
From our review of the reactivity transients, we conclude that appro-priate initiating events have been exmnined and conservatively analyzed and that the results demonstrate the capability to limit the consequences to acceptably low values.
- 4. 3 AppIicant 's "Maximum CredibIe Accident II The applicant defines a "maximum credible accident" for the Fort St. Vrain reactor as that due to failures in the seismic Class I helium purification system connection to the regeneration system concurrently with the inability or failure to close the iso-lation valve at the discharge from the high temperature filter/
absorber.
The resulting accident releases primary coolant-helium and the associated radioactivity into the reactor building over a period of about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The initial leak rate is calculated to be 3.4 Ib/sec decreasing to about 1.2 lb/sec in 24 minutes.
In analyzing this accident, the applicant calculates that during the first 5 mdnutes of the transient about 0.13% of the primary coolant system inventory will be released unfiltered from the reactor building.
In order to account for uncertainties in the operation of the reactor building pressure relief sys tem (louvers), we have assumed, for our analysis, that 10% of the primary coolant system inventory constitutes the unfiltered release.
Table 4.1 presents the assumptions used by the staff and the appli-cant in calculating the consequences of this accident.
Also shown
35 -
are the doses at the exclusion area boundary which have been calcu-lated by the staff and by the applicant.
The difference in these values results from our more conservative assumptions on unfiltered release fraction; (10%
VB 0.13%); strontium dose conversion factor (3.6 x 107 rem/Ci vs 1.07 x 107 rem/Ci); and the meterological dilution factors (2.6 x 10-3 sec/m3 va 8 x 10-5 sec/m3).
Using our more conservative assumptions the exclusion area boundary (590m) doses (i.e., Thyroid 4.6 rem, Whole Body 8.6 rem, Bone 36 mrem) are well below 10 CFR part 100 gUidelines.
Low population zone doses for this short term accident are negligible.
4.4 Rapid Depressurization of PCRV_- DBA #2 In the course of our review at the construction permit stage, we concluded that because of the PCRV design and associated engineered safety features, gross structural failure of the PCRV need not be considered as a design basis accident.
At that time we also
~oncluded that a sudden total failure of both the inner and outer closures of anyone of the PC.V penetrations would represent a conservative upper limit to the spectrum of hypothetical failures that might conceivably occur in the primary coolant system envelope.
In addition, flow restrictors and penetration internals are secured against ejection by mechanisms that are independent of the two penetration closures.
Consequently, we concluded that in the event both closures failed, the flow restrictor and internals would remain in place to obstruct and limit the out-flow of the primary coolant gas.
The applicant has investigated the following aspects of the hy-pothesized concurrent failure of the primary and secondary penetration closures:
(a) The abLl.Lty of the reactor internals to withstand the differ-ential pressures experienced during depressurization without loss of function.
(b) The ability to conti.nue primary coolant circulation following rapid depressurization.
(c) The potential for oxygen ingress into the PCRV after equaliza-tion of PCRV internal pressure with the reactor building atmosphere.
(d) The effect of vertical thrust on the PCRV due to depressurization through the bottom head access penetration.
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TABLE 4.1 ACCIDENT ASS~TIONS APPLICANT'S "MAXIMUM CREDIBLE ACCIDENT" -
REGENERATION LINE FAILURE*
10 20 5 m/sec 80 mrem 160 mrem Design*
0.13%
0.7 mrem F
-5 3
8 x 10 sec/m elevated wi th downwash 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Semi-Infinite Cloud Gamma 3.47 x 10-4m3/sec 1.07 x 107 rem/Ci 1.48 x l06rem/Ci APPLICANT 100%
10 10%
3.6 mrem 8.6 rem 4.6 rem STAFF 100%
Bone Whole Body Thyroid Exclusion Radius Doses:
Meteorology:
Dilution Factor (X/Q) 2.6 x 10-3 sec/m3 initial unfiltered release Volume source 8.4 x 10-4 sec/m3 Windspeed 0.3 m/sec Release Point ground level wi th building wake Pasquill Condition F
Durat Lon of Accident Release 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Whole Body Dose - Model Semi-Infinite Cloud Gamma-Beta Breathing Rate 3.47 x 10-4m3/sec Strontium Dose Conversion Factor 3.6 x 107 rem/Ci Iodine Dose Conversion Factor 1.48 x 106rem/Ci ITEM Fraction Escaping Unfiltered Halogen Filter.Reduct i on Factor Source Term Design**
Coolant Inventory Lost Particulate Filter Reduction Factor 20
- FSAR, Section 14.8
- The applicant designates the fission product source term in the primary cool Lng as "expec t ed"; less than 1% coating failures; and "Design" 10%
particle coating failures over the 6 year fuel cycle.
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(e) Pressure effects on the reactor building.
(f) The radiological consequences of the sudden depressurization.
We have reviewed the applicant's analysis rel~ted to each of the above aspects of the rapid depressurization accident and have de-termined that, with the exception of the radiological consequences, the assumptions are conservative and that the PCRV, the core structure, the internal seals and baffles, the circulators, the helium valves and the steam generators have been designed to with-stand a depressurization at
~ 250 lb/sec without loss of function.
In the case of the radiological consequences we believe that the some of the applicants' assumptions are not sufficiently conservative.
The assumptions used by the staff and the applicant in the dose calculations are shown in Table 4.2.
For that fraction of the dose resulting from coolant-borne activity, we and the applicant differ only in the meteorological fodel used in dete~ning dispersion parameters (X/Q
- 8.4 x 10-sec/m3 vs 4 x 10-sec/m3).
Because of uncertainities in the accumulation and the release of plated-out fission products, we use a more conservative approach than the applicant in assessing the potential doses from this source.
In our determination of the
/~oses related to the removal of the plated-out bone seeking fission products and halogens from the surface of the steam generator tubes, we have not considered a 30 year nccummulation of plated-out activity.
Instead, we have calculated a plate-out inventory per primary coolant system loop that, if permitted to accumulate, would give riBe to doses during a rapid depressurization well below 10 CFR Part 100 gUideline values.
Using the source term associated with these plate-out inventories, we calculate d08es at the exclusion area boundary of 150 rem thyroid, 75 rem bone, and 7 rem whole body.
Accordingly, we have included these limdts (1-131 equIvalent
- 5000 Ci/loop and Sr-90 equivalent
- 140 Ci/loop) in the Fort St. Vrain Technical Specifications.
The applicant will determine the plate-cut inventory by periodic measurements with a plate-out probe insLalled in each steam generator, by measurement of the gasborne precursors and by calculation of release based on the ratio of release rate to birth rate of Kr and Xel The allowable plsteout inventory limit is included in the Technical Specifications.
We have concluded, based on our review of the applicant's analysis of the rapid depressurizaticn accident and the specification of an fi t'
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allowable plate-out inventory limit t that the calculated doses at the exclusion radius and law-population zone are well within the guideline values in 10 CFR Part 100.
4.5 Permanent Loss-ai-Forced Circulation Cooling, DBA #1 Even though there are many design provisions to preclude the loss-of-forced circulation cooling for the reactor core, the applicant has conducted an extensive analysis of this accident and the resulting consequences.
The accident has be~n discussed in Section 14.10 and Appendix D of the FSAR.
Many variations of the accident have been considered to assure that t even if the multiple failures necessary to produce this accident should occur, the calculated doses would not exceed 10 CFR Part 100 guideline values.
The analysis of this accident is very complex and sophie ticated models and computer codes have been used by the applicant.
In our review of this accident at the construction permit stage, we con-sidered the asaumptions, calculational techniques, and models and concluded that the analysis was reasonable and conservative.
We have again reviewed these aspects of the accident analysis as presented in the FSAR with the assistance of our consultant, Dr. L. A. Booth..
We and our consultant determined that:
(a) The simplifying assumptions, used for approximate solutions, where exact analytical solutions are not possible, were valid and conservative.
(b) The calculational techniques l~ed for the physical processes being analyzed were valid.
(c) The properties of materials used in the analysis were appropriate and represent the bes t currently available.
(d) The experiments performed were relevant to the physical processes being simulated.
We concur with our consultant in his conclusion "... that the presentation of the nuclear, thermal, and fission product release analysis is valid t complete and t in the main conservative."
- Dr. Lawrence A. Booth, Staff Scientist for UHTREX, Los Alamos Scientific Laboratory - Dr. Booth's Report is included as Appendix E.
TABLE 4.2 APPLICANT Design 100%
-4 3 3.47 x 10 m Isec Area Source "G"
1 m/sec 4 x 10-4sec/m3 4 x 10-5sec/m3 5 rem 2.5 rem
< 0.1 rem 5500 Ci/loop 160 Ci/loop 12.4 rem 4.7 rem 17.4 rem 4.8 rem 2.5 rem 0.0057 0.01 STAFF Design 100%
-4 3 3.47 x 10 m /sec Volume Source "F"
0.3m/sec
-4 3
8.4 x 10 sec/m 4 x 10-5sec/m3
< 0.2 rem Calculated Allowable Design 5000 Ci/loop 140 Ci/loop 0.06 O.O~
< 7 rem RAPID DEPRESSURIZATION ACCIDENT:
DBA-~
Thyroid Whole Body Bone 1-131 equivalent Sr-90 equivalent Iodine Strontium Thyroid Bone Thyroid Bon£.!
Whole Body ITEM Meterological Model Pasquill ConditIon Wind Speed Dilution Factor:
X/~ (Exclusion Radius)
X/Q (LPZ)
C~olant Fission Product Inventory Coolant Lost Unfiltered Inhalation Rate ACCIDENT ASSUMPTIONS Doses -
(2 Hr. at exclusion radius) due to Coolant-Borne Activity Plateout Inventories Sour~~
Fraction of Plateout Inv~ntory Released Doses (2 Hr. at exclusion radius) due to removed Plateout Acti".;ity Total Accident Doses at Exclusion Radius
40 -
'l'ab l.e 4.3 presents the assumptions used by the staff and applicant In cal cu l.at Ing the radiological consequences of the permanent loss-of-farced-circulation cooling.
Using our more conservative aBsumptions, we calculate for the course of the accident doscs, at the low population zone of less than 8 rem thyroid, less than 5 rem bone, and less than 0.33 rem whole body.
These doses ar e well below 10 CFR Part 100 guideline values.
Two hour doses at the exclusion area boundary are not appropriate because of the delayed release of flssion products (approximately two hours after loss-of-cooling) that o~cc~s during this accident.
We conclude that the radiological consequences of a pe~ent loss-of-all-forced circula~ion gas cooling of the reactor core will not exceed 10 CFR Part 100 guideline values.
4.6 Anticipated Transient with Failure of Protective Action Although the reactor protective systems for the Fort St. Vrain reactor are highly reliable, the occur=ence rate of anticipated transients leads to the conclusion that there exists a possibility of an anticipated transient with failure of protective action.
We therefore requested the applicant to discuss transients having a probability of occurring once or more during reactor life and the potential conse-quences if protective devices were to fail to function properly.
The applicant has identified two such transients.
The first is a failure of the moisture monitor system to detect and isolate steam leakage, and the second 18 a failu'!'e of the plant protective sys tem to initiate scram.
'!b.e applicant has evaluated the consequences of steam and/or water leakage into the PCRV taking into consideration many combinations of postulated failures of systems
- The evaluation deecne t rarea that backup actions or operator actions (e.g., initiation of the reserve shutdown syste~ would preclude damage to core structures or over-pr~ssure in the PCRV.
Even with the postulated failures of protective systems, none of these transients would result in actuation of the PCRV pressure relief system.
The applicant also addressed the release of fission products from failed fuel particles4by hydrolysis in the event vf steam leakage.
An iucrcase of 3 x 10 Ci of noble g88 activity is calculated during the first 20 minutes in the worst case analyzed; however, all of this actjvity would be contained inside the PCRV and could be removed by the helium purification system.
TABLE 4.3 ACCIDENT ASSUMPTIONS PERMANENT LOSS-OF-FORCED-CIRCULATION COOLING -
DBA III Course of Accident dose at Low Population Zone I
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4.6%
90%
95%
0.2%/day 0.026%
99%
66 mrem 1.9 mrem O 56
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STAFF 879 MWT 180 days 90%
95%
0.2%/day 3.47 x 10-4m3/ae c 0-12 brs 2.9 x 10-4m3/sec
-4 3 1.75xlO m/sec
-4 3
-4 3 2.32 x 10 m /sec 2.32 x 10 m Isec
-6 3
7.8 x 10 sec/m 3.9 x 10-6sec/m3 1-12 hrs 4.7 x 10-7sec/m3 2.4 x 10-6sec/m3 7.8 x 10-7sec/m3 All others 9.5 x 10-9sec/m3 4.6x lO-7sec/m3 3.6 x 10-7rem/Ci 1.07 x 107 rem/Ci
< 8 rem
< 5 rem
< JJO mrem Next most unfavorable 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> For the 1-4 day period For the 4-30 day period For the 30-180 day period Thyroid Bone Whole Body ITEM Power Level Assumed duration of acci~ent emissions uaed In dose calculations H<JlogenH Released to PCRV Halogen Plateout Factor 50% of Core inventory) 2 Halogens Available for
'~ 'o!le88e from PCRV 25% (of Core inventory)
Strontium Available for Release from PCRV 1% (of Core inventory) 100% (of Core inventory)
Noble Gases Available for Release from PCRV Reactor Building Filter Efficiencies Halogens Particulates rCRV Leak Rate Inhalation Rates Most unfavorable 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period Next most unfavorable 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> Remainder of lBO day period Meteorological Dilution Factors Most unfavorable 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period Strontium Dose Conversion Factor We have concluded from our review of the applicant's analyses that secondary system leakage into the PCRV can be accommodated acceptably by the engineered safety features of the plant even in the event of 8 concurrent failure of protective action.
ThC! applicant has considered the following anticipated transients and the effect of a postulated failure to scram on their course and consequences:
(a) partial 10S8 of feedwater supply (b) 1088 of circulator steam turbine drives (c) primary system depressurization (d) turbine trip (e) control rod withdrawal acr.ident The analysis of these accidents demonstrates that the nuclear and thermal characteristic of the HTGR, plus the engineered safety features of the plant, provide considerable margin for accommodating the transients considered.
In cases (a), (b), (c), and (d) above, the principal consequences would be an increase in the fuel and moderator teq>eratures plus changes in main and reheat steam tempera-lure.
The increased fuel temperature is still well below the tempera-ture at which kernel migration
(~ 2900°F) or fuel particle failure
(~ 3600°F) occurs.
For cases (a), (b), and (d) the reactor system conditions would reach equilibrium after a short time (15-30 min.)
without any operator action.
In case (c) the operator would have well over 30 minutes to assess the situation and to activate a manual scram or the reserve shutdown system.
For the rod withdrawal accidents -
Case (e) - some fuel particles would be expected to fail (tV 2%) as a result (*f the postulated multiple failures of protective action.
In the worst case (maximum worth control rod pair withdrawal at full power) average outlet gas temperatures could exceed
~ 2000°F with local hot gas streams in the outlet plemnn of ~ 2600°F as a result of protective action failure at two levels with scram finally occuring on the high reheat temperature trip.
Although additional radioactivity would be released to the primary coolant and some thermal barrier damage could occur at the inlets to the steam generators, the PCRV integrity would not be impaired and safe shutdown of the plant could be accomplished without endangering the public.
We have concluded that there is reasonable assurance that the plant can accommodate these multiple fat.Lures without hazard to the public.
5.0 EMERGENCY PLANNING The applicant has described a plan for coping with the consequences of an accident that might affect the general pubLf.c, Arrangements to deal with radiological emergencies have been made
~ith t~e Colorado Department of Health, the S.tate, Civil Defense Division of the Department of Military Affairs, and the Weld County Sherif:'s Department.
In addition. technical assistance is available through the Radiological Emergency Assistance Team program of the AEC.
Based on data from the activity monitors and meterological instruments at the plant, members of the applicant's onsite staff will provide an initial assessment of the situation to state and local officials concerning the release of fission products from the facility.
The plant is equipped with a decontamdnation area and a first-aid room.
Further medical aid is available at PSC'B Medical Department or Saint Luke's Hospital in Denver.
We have concluded that the arrangements made by the applicant (des-cribed in Section 12.3, FSAR) to cope with the possible consequences of accidents at the site are in conformance with the requirements for Emergency Plans as presented in Appendix E to 10 CFR Part 50.
6.0 CONDUCT OF OPERATIONS AND TECHNICAL QUALIFICATIONS The applicant, Public Service Company (PSC), will be the sole owner and operator of the Fort St.
Vrain Nuclear Generating Station.
PSC has gained exper~~nce in the nuclear field as a member of the Rocky Mountain Nuclear Power Study Group and as one of the member organi-zations of the High Temperature Reactor Development Associates (HTRDA) which supported the research and development work leading to the con-struction of the Peach Bottom Unit I plant.
As a member of the Rocky Mountain Nuclear Power Study Group, several PSC personnel were assigned" '
for about two years to' the National Reactor Testing Station in Idaho to participate in the activities of the group.
As a member of HTRDA, PSC personnel were assigned to GGA in San Diego to participate in the design of the Peach Bottom plant.
Similarly, PSC personnel were stationed at the Peach Bottom plant to participate in its preoperational testing and initial operation.
A nuclear training program'fbr PSC Electric Department personnel, con-ducted by Colorado State University, was initiated in the fall of 1966.
The program consisted of training in basic atomic physics, nuclear physics, and reactor physics ' conducted at the graduate engineer level.
The program has been completed by 75 employees (38 supervisory and 37 nonsupervisory employees).
I1
. All candidates for license prior to fuel loading were sent to the PeAch Bottom Nuclear Power Plant for operating experience.
The final step of the offsite training program was a l7-week course conducted by Gulf General Atomic in San Diego.
All personnel who attended the Peach Bottom training program attended the training program at Gulf General Atomic.
The normal plant operation shift crew will consist of a shift supervisor, a reactor operator, two equipment operators, and an auxiliary tender.
Each shift crew will include at least one person with a Senior Reactor Operator License and t 'h 9 0 persons with Reactor Operator Licenses.
The startup organization will be under the direction of the plant Superintendent and Assistant Superintendent with advisory personnel supplied by Gulf General Atomic.
The operating organization used during startup will be the same organization used for commercial operation, and the number of operators used to start the plant will be the same as used during normal operation.
In order to be properly prepared for all phases of fuel loading, low power physics testing, and initial startup to power, licensed supervisory personnel within the plant organization will be used to supplement the shift crew to meet the licensed personnel :equirements until additional station personnel are licensed after fuel loading.
On the basis of the above considerations, we have concluded that the combination of reactor operating experience and formal training obtained by the staff should adequately preparp. them to perform their intended duties during all phases of startup and operation of the Fort St. Vrain Nuclear Generating Station.
As a means for the continuing review and evaluation of plant opera-tional safety, the applicant has established a Plant Operations Review Committee and a Nuclear Facility Safety Committee.
The Plant Operations Review Committee is composed of senior plant operating personnel and will meet at least once a month.
Its functions are to review the following areas:
all proposed changes for normal and emergency operating procedureR, and other changes or procedures that are determined by the Plant Superintendent to be related to safety, all proposed changes to Technical Specifications, and all violations of Technical Specifications.
Written minutes of all meetings will be kept and transmitted to the Superintendent of Outside Steam Plants and to the Chairrr~n of the Nuclear Facility Safety Committee.
'lh e Nuclear Facility Safety Conunittee is compoaed of five technically qualified persons who are not meni>ers of the pl ant staff and at least one technical consultant.
Its functions are to review proposed changes to the Technical Specifications, to review minutes of meetings of the Plant Operations Review Committee to determine if matters considered by that committee involve unreviewed or unresolved safety questions, to review matters including proposed changes or modifications to plant syateDfi or equipment having safety significance, conduct periodic reviewl'~ of plant operations, and investigate all Technical Specification violations and unusual occurrences reportable under 10 CFR Part 20 and 10 CPR Part 50.
The Nuclear Facility Safety Committee will meet at least aemiannuel Iy, and will be on call by the C1airman.
Written minutes of all meetings will be kept and will be transmitted to the Vice Pr(!8fdent of Electric Operations.
BOHcd on the above considerations, we have concluded that the general s cope of the safety review and audit functions are satisfactory.
7.0 TECHNICAL SPECIFICATIONS The applicant's proposed Technical Specifications were presented in Amendment 15 58 Appendix F to the application.
We have reviewed these proposed Technical Specifications In detail and have held numerous meetings with the applicant to discuss their contents.
Modification. to the proposed Technical Specifications submitted by the applicant are being made to describe more clearly the allowed condf.tLcna for plant operation.
The Technical Specifications, as approved by the regulatory staff, will be available for examination, well in advance of issuance of an operating license in the Commission 'a Puh l Ic Document Room and at the Greeley Public Library. City Complex Bullding, Greeley, Colorado, 80631.
The Technical Specifications will alRo be appended to the proposed operating license.
Included are sections covering Bafety l:f."lit8 and limiting safety system settings, limiting conditions for operation surveillance requirements, design features and adminiRtrative controls. Based on our review, we conclude that normal plant operation within the limits of the Technical Specf f LcatIons will not result in potential offeite exposures in excess of 10 CFR Part 20 limits and that means are provided for keeping the release of radioactivity from the plant within ranges that we consider as law as practicable.
Furthermore, the limiting conditions of operatIor, and survei l l ance requirements will assure that necessary engineered safety features to mitigate the consequences of unlikely accidents will be available.
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- --- H.O REPORT OF ADVISORY COMMITTEE ON REACTOR SAFEGUARDS The Advisory Committee on Reactor Safeguards (ACRS) reported on the application for construction of the Fort St. Vrain Nuclear Generating station at the proposed site in a letter dated May 15, 1968.
We consider that the applicant has been responsive to the recommendations of the ACRS as indicated in their letter, and we conclude that the matters raised have been relolved satisfactorily during the design and construction of the plant.
The ACRS completed its review of the application for an operating license for the Fort St. Vrain Station at its one hundred thirty-third meeting on May 6-8, 1971.
A copy of the ACRS letter, dated May 12, 1971, i8 at tached as Appendix A.
In ita letter. the ACRS made several recommendations and noted several items to be reeolved by the applicant and staff either before plant operation or on an acceptable time scale subsequent to initial operation.
These items have been consi~.red in our evaluation and include:
the reliability of the emergency onsite power system (Section 3.6.2) test procedures for all valves and interlock circuits not used in no~l operation, an audit of all circuit installations for the reactor trip, engineered safety feature, and safe shutdown systems (Section 3.5.2) the results of tests of performance of essential equipment under accident environment conditione, the feasibility of using neutron noise analysis for the detection of anomalous core behavior, diagnostic instrumentation for postulated accidents, and construction and preoperational testing.
I\\s indJ.cated in the appropriate f.lections of this report the applicant haa resolved several of the ACRS items and will implement those recommendations requiring further advances in the technology as suitable approaches are developed.
The ACRS concluded in its letter that if due regard is given to the Items mentioned above. the Fort St. Vrain Nuclear Generating Station can be operated at power levels up to 842 MWt without undue risk to the health and safety of the public.
- 9. ~
COMMON DEFENSE AND SECURITY The application reflects that the activities to be conducted will be wlthin the jurisdiction of the United States and all of the directors and principal officers of the applicant are United States citizens.
l'ub l l c Service Company of Colorado is not owned, dominated or controlled by lin a l i.en, a foreign corporation, or a forei.gn government.
The lie: t Ivi t LPA to he conducted do not involve any res tri cted data, but the up p l.l can t has agreed to safeguard any such data which might become Involved In a c co rdance with the requirements of 10 CFR Part SO.
The nppllcnnt wlll rely upon obtaining fuel as It is needed from sources of nupp ly available for civilian purposes, 80 that no diversion of special nuclear mat er t a I from military purposes, is involved.
For these re aaona and 1n the absence of any information to the contrary, we have found that the activities to be performed will not be inimical to the common
~fense and security.
10.0.f} NANCIAL QUALIFICATIONS We have reviewed the financial LnformatIon presented 1n the application.
'lh e funds necessary to meet operating costs of the facility will come from operating revenues of the applicant as more fully set forth in ItH application.
Information contained in the applic.ation indicates that surh revenues will be ample to cover the estimated cost of operAting this reactor as well as the safe decomndssionlng of the un l t If such should become necessary.
Wp conclude that the applicant is financially qualified to engage in the ac t lvt t Ies authorized by the provisional operating license.
OUT eval uatIon of the app1ican ts' financial qualifications is preAented In the attached Appendix F.
- 11. 0.~LNANCIAL PROTECTION AND INDEMNI'IY REQUIREMENT~
Purxuan t to the financial protection and indemnification provisions of the AtomLc Energy Act of 1954, as amended (Section 170 and related Rect1onA), the Commission has issued regulations in 10 CFR Part 140.
Th cs e regulations set forth the CotnJllission's requirements with regar-to proof of financial protection by, and indemnification of, licenses for fac11itieR such as power reactors licensed under 10 CFR Part 50.
11.J
~r~~~er~~onal Storage of Nuclear Fuel Th e Commission 'a regulations in Part 140 require that each holder of n c ons t r uc t i on permit under 10 CFR Part 50, who is also to b~ the holder of n llcense under 10 CFR Part 70 authorizing the ownership and pORHCRH Ion for storage only of special nuclear material at the reactor construction site for. future use as fuel in the reactor (after issuance of an operating license under 10 CFR Part 50), shall during the interim storage period prior to licensed operation, nave I
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'9 at the Ditc
~nd at vcriou3 rn~uruet\\,rln3 ft~c:.lit1cs; July 1", 1970, lmd l!arch 10, 1971, at thta oite; 8:'od Uay 4, 1971 in Bethae.1n, Ucrylend.
Durinz itl I
rev1c~I, tho C(\\:.~itt':!a h4d th~ ccnciit 0%
cli~cua8iona tiith repr.::scn-tativ33
~!nc1 cor.s'Jltants of the l'ublic Sorvice CO:l:,any of Colorndo, Culf C~ner31 Ato~lc, Inc., and tha A~C Rczul~tory Staff, end of the docu~::ent8 listod belG17.
Th'! Cc:mdttee re~orted the reou1ts of its conGtruction permit revlew to you in a letter dnt3d 1%3Y 15, 1963.
The Fert St. Vrain St~tion is loeat~d about 35 uulea north of Denver.
The Id.G1t*te~perQtllre sao-cooled re~etcr ie model*Qted with tt'sphitc and cooled "'1th heliu:3.
The prestresEJad concrete reactor ve:JDel, in
~h1ch ths entire pric.ary oY,a tem is cOlltcii"teU, is the fi.rs t
'of lts kln~ in tho U~lted States.
Emergancy on-31te pO~IGr for the pldnt to Bupplied b~ two diesel-gener-ator sats, e!1cb ca,oblo of 8upplying the p~*7.ar needed for the ensineered
.afoty fe3c",re3 and the 8ai"",
Cllutdo~m cool1nr; sy3~em.
E3Ch gC!nerD.tor is dri",an by t~:o d1~ct.!1 cn3in~:J ecaneeeed in tand~~ to the cenerator drive ohaft.
Since this arran~ecent has not been u3ed previously in connection '11th a Ilu:leor plant, the C0t0m11ttee bel:l.cvea that tbe applicant should pr~lid. data to deconstreto that th~ reliability of the O)"D t~iJ 1s
~o:n:':.tr:1blQ to that of en-at t!!
pO~'7~r By9 te~3 providl!d 1n other nuelc~~ p12uts.
this reliability
&r~'Jld be demonstrated to the snt1Dfaetion of the Re3ulatory Staff 'be f or e a signific:lnt inveu~ory of f1~.10n products has been acc~lated from operat~o~ of the reactor.
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REPORT TO THE AEC REGULATORY STAFF STRUCTURAL ADEQUACY OF FORT ST.
VRAIN NUCLEAR GENERATING STATION INTRODUCTION This report is concerned with the structural adequacy of the contn lnment structures, piping, equipment and other critical components for the For t St. Vrain Nuclear Generating Station for which appl ication for a construction permit and an operating license has been made to the U.
S. Atomic Energy Commission by the Public Service Company of Colorado.
The facility is located about 3 1/2 miles NNW of Platteville, Colorado and about 35 miles N of Denver, r
~rado.
T~i~
(t is based on a review of the Final Safety Analysis Report (Ref.
- 1) and
~
supplementary material which has been made available to us (Ref. 2).
The report also is-based In part on the discussion and Inspection associated with the visit to the site on 15 July 1970 by W. J. Hall.
A discussion of the adequacy of the structural criteria presented in the Preliminary Safety Analysis Report is contaJned In our report of February 21, 1968 (Ref. 3), and unless otherwise noted no comment will be made in this report concerning points covered therein.
The design criteria for the containment system and Class I components for this plant were developed for a Design Basis Earthquake of O. 109
~ximum horizontal ground acceleration coupled with other appropriate loadings to provide for containment and safe shutdown.
The plant was also designed for an Operating Basis Earthquake of 0.05g maximum horizontal ground acceleration act ing s imul U'Inf.'OUS ly wi th the other appropr iate loadings forming the bas is of the des ign.
52
~",,~~~.", r.antin(J 8Vlltfml utilize8~ for the most pArt.
equip~ellt tb~t i8 used a100 in the normal operetion of the plnnt. suitably modified or ".uS:iie&1te.:i to provide the redundancy and reli~,L11ity appro-priate to en3ine~rcd s~fcty fQ3turc3.
Tb~
Co~~Jtt~~ believdG thnt proccuures
~Loul~ be d~vc1cped to t~Gt p~~iofl1cQ11yall valve. and tn!:erlock circuits not used in nor~41 uparatlon.
~he8e tests GhoulJ be maao at l£3st tuiCQ a year
- 1'4Y 12, 1971
. Btmorab1e Cl~nn t. Seaborg The
.~pllcant ia roAking' tl c~?lete audit of !lll e1.rcult in9talletlonl for th~ re::etor trip, enr,ltl~tl}rcd ~af#Jty fC!atl\\~eg. end oafe shutd~-m
.yst£rJ~ to detet'rd.ne eonform3nce wJ.th the criteria r~latl1\\a to 6cp~ration and' to 81n~lc f~ilure~.
Tho cpplie~nt has ctete1 that any d~ficiencios found h~vo b~~n or ~~ll be corrected prior to initial o~erQtion.
The 8!,plf.cQllt beta identified e1ectrf.ct1 equ1p~nt uhic:h 10 required to operat~ ~uri~6 ~~ follodlng po'tul~ted c~cidp.~to, nnd h~~
dp.to~i~ed tbe enviror.;..:~ntol cO:.1Jiti01\\:) to which th10 equipo.:~nt '\\1111 be o:tt'0:)(,;d.
Test proceC1ur~a nava &Jeen Q~v.lupts"'.
tLL&~
&;a~~cI tu,....
~.:~.'o...-A.:.
~w provIde t'.:OC~4ln:3 th~t this e!loent1~1 eCX'J1p::tent will perform its functions.
'iho r~~ultl of th-e** t09tl9 abo-lld be rc'V1e~ed by the Rez*
ulator1 Stnff prior to oporatioD of the plant.
I II It to rceo~~nded that the applicant study the applicability of neutron noise ansly3i:; to the c1etection of anom.loua core beh:1v1o: in the Fort St v Vrain rea~tor.
!he Q,plie~nt should review the instrumentation available to the 0PQratora to u~der8t3nd and detercine the COU~S4 Qf events oceurring in the unlikely event of
- savero t.ecldeut.
He should
~ssure himself and the Regulntory Staff that appro~rlate diagnostie in3trumentatioa will rll!1':1aln iu 1&:8
(;;;li:r.t1n~ r=s=
dt:=i~; p=: t~1~t~1 l!eeit1~!\\t~.
The A~lcory Co~ttee on
.~eac tor S4fezua~ds believes th~t. if due regard 18 given to the ite~ centicncd above. and cub~ect to.atia-factory co~?letioD of construction and preoperational te3tlng, there I, re3sonBble a3sur~c~ tha: the FOlt St. Vrain Uuclear Cenerating Station C!U be c,erntad at power level. up to 842 }~;t without undua
~l.k to the health and **flty of' the public.
81Dccr.ly your**
Or1g1n~.1 Slgl113d bY'.
Sp.~ce;r II*.llush.,i
.._. '.0*.
- Spencer B. Bush Ch.~l~~!\\
Refe:enees attached.
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1.
An~nd~ent. Not. 14
- 19 to tho Licenee Applic4tion of the Public Sflrvice C~..:1ny of Colorado for the Fort !ft. Vrain ttuclea\\"
Ceneratlns Statlo4
~,.
lonorable Glenn If. Seabora -3'*
~
'lI4y 12, 1971 I',
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54 -
APPEND:X Bl DRAFT RE PORT TO THE AEC R£:GULATORY STAFF STRUCTURAL ADE QUACY OF FORT ST. VRAIN NUCLEAR GENERATING STATION PUBLIC SERVICE COMPANY OF COLORADO AEC DOCKET NO. 50-267 by N.
M. Newmark and W. J.
Ha 11 Urbana, l l l Inols 8 December 1970 REPORT TO THE AEC REGULATORY STAFF STRUCTURAL ADEQUACY OF FORT ST.
VRAIN NUCLEAR GENERATING STATION INTRODUCTION This report is concerned with the structurai adequacy of the containment structures, piping, equipment and other critical components for the Fort St.
Vrain Nucle~r Cenerating Station for which appl ication for a construction permit and an operating license has been made to the U. S. Atomic Energy Commission by the Public Service Company of Color~do.
The facility is located about 3 1/2 mi les NNW of Platteville, Colorado and about 35 mi les N of Denver, Colorado.
This report is based on a revIew of the FInal Safety AnalysIs Report (Ref.
- 1) and other supplementary materIal whIch has been made avaIlable to us (Ref. 2).
The repcrt also is based in part on the dIscussIon and Inspection assocIated wIth the vIsit to t~e site on 15 July 1970 by W. J. Hall.
A discussion of the adequacy of the structural criteria presented irl the Preliminary Safety Analysis Report is contained In our report of February 21, 1968 (Ref. 3), and unless ctherwise noted no comment will be made in this report concerning pOints covered therein.
The design criterIa for the containment system and Class I components for this plant were developed for a DesIgn Basis Earthquake of O.IOg maximum horizontal ground acceleration coupled with other appropriate loadings to provide for cont~inment and safe shutdown.
The plant was also designen for an Operating Basis Earthquake of 0.05g maximum horizontal ground acceleration acting simultaneously with the other appropriate loadings fo~ming the basis of the design.
I
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56 -
2 COMMENTS ON ADEQUACY OF DESIGN Genere:tl Stiltcrnent During the course of the review of the Final Safety Analysis Report and in conjunction with answers to our questions and those arising from discussions wl th DRS-DRL-AEC personnel, various draft materials were
- revlewed, It appears that all of the important material relating to the seismic des ign has been drawn together and presented in Amendment No. 16.
We wish to cormrend t hc oppl icant on the approach employed, and the att Itude cilsplayed, in answe r l nq questions.
As a result of the efficient approach adopted, it seems unn~cessary to repeat most of the questions, answers, and summary of the supporting studies which were presented to provide the basis for the design evalu~tion.
We have singled out for co~nt, however, certain principal points of concern in the seismic analysis and design on which we wish to make edd l t lonal comre nts ; the seismic design approach Is generally well documented In Refs. I and 2.
SeismIc Analysis (a)
Prestressed Concrete Reactor Vessel (PCRV)
The general approach to the seismic analysis of the PCRV is described In Section E. 13 of the FSAR In some detail.
Additional backup information is presented in Ref. 2(c), 2(f), and 2(g).
The approach employed here is satisfactory with one excaption,namely that the vertical ground excitation was taken to be a constant value of two-thirds of the vertical acceleration at the high frequency e~d of the spectrum.
The applicant advises in the answer to Question 5.12 that all Class I structures are now being analyzed with the verticel response spectra
.j 3
equal to two-thi rds of the horizontal ground spectra, Le,, appropr iate amplification over the entire frequency range.
This approach is acceptable to us.
(b)
Pip i119.
A revised description of the piping design and analysis procedures, document OC-91-1,is made a part of the answer to Quest ion 5.13 and the approach outlin~d in there, which calls for use of a value of 0.5 percent of critical damping for piping with the design basis earthquake, appears acceptable to us.
The ~~tter of accounting for support motiwns of pipe, as discussed in the answo r to Question 5.20, is covered in the rev tsed document DC-91-1 under Section 4.18; it is indicated there that where such movements can occur, they will be con~idered in the stress calculation.
(c)
Equipment The design procedures to be followed for Class I equipment is described at various points in the FSAR and amendments.
A su~ry of the Class I equipment and a 1ist of those responsible for evaluating its adequacy is given in the answer to Question 5.11.
It is indicated there that
~ review is currently under way of all such equipment to insure its adequacy under seismic excitation.
Additional elaboration on the approa h followed in the design of Class I equipment is contained in the answer to Question 5.23.
The approach adopted is acceptable to us.
(d)
Cable Trays The cable trays in this plant, which were particularly noticeable because of the compact arrangement of the plant and the manner in which they were stacked, receive detailed attention in the answer to Question 5.22.
The design approach summarized there is adequate.
4 (e)
Tie-Do\\'/n of Equipment Racks It was noted at the time of the plant inspection that the equipment racks located in the relay room (immediately below the control room) were not fastened to the floor at that time.
We have been advised by DRL personnel that rhl s equ lprrcnt iii being checked and wIll be securely fastened by the time the p lant is approved for operation; this approach is acceptable to us.
(f)
Core iJisurray The analysis of the effect of
!~Ismic excitation on the reactor internals, and particularly upon factors which might affect core disarray and malfunction, is summarized in the answer to Question 5.6 In Amendment 16.
The discussion and analysis presented there indicates that the reactor core wIll function satisfactorily during the Design Basis Earthquake.
Oeslgr. Stress Criteria The design stress criteria are presented at varIous places in the FSAR and a sunvnary is presented in t ne answer to Question 5.21 of Amendment No. 16.
It is indicated rhc rc that the allowable stresses for the combined loadings, including the Design Basis Earthquake and tornadoes,
~pproach the ultimate stress for both the concrete and the steel.
From examination of this table alone one might question the margin of safety Inherent in the design under the maximum loading conditions.
It would be our recommendation that the applicant provide additional inform~ation as to the margin of safety that exists under such loading conditions, especially In the containment structure and other structures and equip~nt which are necessary for safe shutdown and contalnrrent,
It is our belief that the high stress limits cited probably wi 11 be reached in local ized regions, and may have I ittle bearing
~ ~ ~ ~
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5 on the over-all margin of st re nqt h that is inherent in the design.
- However, to reinforce this opinion, or to provide a basis for evaluating what the s l t uo t ion oc tua l lv may be, the appl icant should provide additional discus. ion of the ') i t u.u ions under which these stresses control and corment on the margin of sa f e t y I hat actually exists under the loading conditions cited tn CONCLUDING REMARKS On the basis of the information made available to us concerning the Cl~ss I structures, piping, reactor internals structure, and other Claf3s I items, it is our bel ief that the plant possesses a reasonable margin of safety to ime t the imposed Design Basis Earthquake loadings In conjunction with other app l lcub le loadings. to the extent of Insuring safe shutdown and conta lnnr-nt,
- However, in arriving at this conclusion we should like to have the appl icant prescnt additional Information on the design stress criteria (or the max imum comb i ncd load i n9 cond It Ions e Ited in TabIe 5.21-1 (accompany Ing the answer to Question 5.21).
REFERENCES 1.
"Finul Safety Analysis Report -- Vols.
I -
IV including Supplements 14-16,'1 Fort St.
Vrain Nuclear Generating Station, Public Service Company of Colorado, AEC Docket No. 50-267, 1970.
2.
Supplementary reports:
(a)
Chu, S.
L., J.
M.
l"1cLaughlin, R. Ryan, G.
Patwari, "Fort St. Vrain Nuclear Power Station Unit 1, Turbine-Raactor Building Seismic Analysis, Sargent and Lundy Engineers, A & CD report No.
16, December 1969 (revised June 1970).
60 -
6 (b)
Chu, S. L., J.
M.
McLaughl in, T. Hongladromp, "Fort St.
Vrain Nuclear Power Station Unit I, Combined Turbine-Reactor Building and PCRV Seismic Analysis," Sargent and Lundy Engineers, A &. CD Report No. 22, May 1968 (revised June II, 1970).
(c)
~mill1, R. J., E. C. Rossow, "The Stress Arlalys Is of the PCRV Support St ruc t ure ;!' Sargent and Lundy, January 31, 1968.
(d)
I\\min, M., "A Study of Coupling Between PCRVand Turbine-Reactor Bui ld lnq, Fort St.
Vrain, Unit 1," Saryent and Lundy F.ngineers, 1\\
F',
CO Report No. 11, August 1967.
(c!)
MC'lenrwn, G. A.,
W.
P.
Henschel, and E. P. Esztergar, "Dynamic Response of Inde tcrml notc Space Structures," Gulf General Atomic Inc. Report GA-8692, M~y 1, 1968.
(f)
Amin, M., "A Further Study of the Coupling Between PCRV and Reactor-Turbine Ruilding, Fort St.
Vrain Unit No.1, Sargent and Lundy Engineers, A &. CD Report No. 12, November 1967.
(9)
Hobush, A. L., "Dynamic Soi 1 Property Invest Iqat Ion, Proposed Fort St.
Vrain Nuclear Generating Station, Platteville, Colorado,"
prupored for General Atomic Division, General Dynamics Corporation, 23 June 1967.
3.
- NeWfMrk, N.
M. and W. J. Hall, "Report to AEC Regulatory Staff, Adequacy of the StrucluriJ1 Crltcri~ for the Fort St. Vrain Nuclear Generating Station, Public Service Company of Colorado," February 1968.
APPENDIX B2 REPORT TO THE AEC REGULATORY STAFF STRUCTURAL ADEQUACY OF FORT ST. VRAIN NUCLEAR GENERATING STATION PUBll C SERV ICE COMPANY OF COLORADO AEC DOCKET "'0. 50-267 by N. H. Newmark and W. J. Hall Urbana,Il11nof5 3 March 1971 REPORT TO THE AEC REGULATORY STAFF STRUCTURAL ADEQUACY OF FORT ST.
VRAIN NUCLEAR GENERATING STATION INTRODUCTION This report is concerned with the structural adequacy of the cont a lnme nt structures" piping" equipment and ot hcr critical components for the Fort St.
Vrain Nuclear Generating Station for which application for a construction permit and an operating license has been made to the U. S. Atomic Energy Convniss ion by the Publ ic Service Company of Colorado.
The fac l l ity i!>> located obout 3 1/2 miles NNW of Platteville, Colorado and about 35 miles N of Denver" r
"'rado.
T"'h.
r t is based on a review of the Fi nal Safety Analys is Report (Ref.
- 1) and
~
supplementary material which has been made available to us (Ref. 2).
The report also ls based in part on the discussion and inspection associated with the visit to the site on 15 July 1970 by W.
J.
HCll1.
A discussion of the adequacy of the structural criteria presented in the Prol irninary Safety Analysis Report ts contained in our report of February 21, 1968 (Ref. 3), and unless otherwise noted no comment will be made in this report concerning points covered therein.
The design criteria for the containment system and Class I components for this plant were developed for a Design Basis Earthquake of O.lOg maximum horizontal ground acceleration coupled with other appropriate loadings to provide for containment and safe shutdown.
The plant was also designed for an Ope rat i n9 BuS is Earthquake of 0.05g maximum horizontal ground accelerat ion acting s lmul tanr ous l v with tho other appropriate loadings forming the basis of t he des i 9n,
_..:-'~
2 CO~1ME NTS ON "D EQUP.CY OF DE S!Jili During the course of the review of the Final Safety Analysis Report and in conj unc t.Ion with answers to our questions and those arising f -rom d l scus s lon-with DRS-ORL-AEC personnel, various draft materials were reviewed.
It appears tho t all of the important material relating to the seismic des ign has been drawn together and presented in Amendment No. 16.
We wish to il connmnd the! app l icant on the approach employed, and the attitude displayed, in answering questions.
As i'I result of the efficient approach adopted, it seems unnecessary to repeat most of the questions, answers, and sunrnary of the suppcr t lnq studies which w~re presented to provide the basis for the design evnlutltion.
We have singled out for comment, however, certain principal points of concern in the seismic analysis and design on which we wish to make addit ronal comments; the selsmlc des ign approach is generally well documented in Refs.
1 and 2.
(o)
Pn~stresscd Concrete Reactor Vessel (PCRV)
The general approach to the sersmic analysis of the PCRV is described In S~clion E.13 of the FSAR in some detail.
Additional backup information i s pre 5 Cnted i n Ref.
2 (c), 2 (f), and 2 (g)
- The epprooch employed here is satisfactory with one exception,namely thilt the vcrtic~l ground excitation was taken to be a constant value of two-thirds of the vertical acceleration at the high frequency end of the spectrum.
The app l icant advises in the answer to Quest ion 5.12 that all Class I structures ar~ now being analyzed with the vertical response spectra
- - ---- - - - - ---- 3 equal to t'f,o-thfrds of the hor l zont al ground spectra, i.e., appropriate iltnplifictll:ion over the entire frequency range.
This approach is acceptable to us.
A revised description of the piping design and analysis procedures, document DC-91-1, is made a part of the answer to Question 5.13 and the appronch outl ined in there, which calls for use of
- value of 0.5 percent of critical damping for piping with the Design Basis Earthquake, appears accep t able' to us.
The matter of accounting for support motions of pipe, as discussed In the answer to Question 5.20, Is covered In the revised document DC-9l-l under Section 4.1B; it Is indicated there that wh~re such movements can occur, they will be considered in the stress calculation.
(c)
Equ i prnen t The design procedures to be followed for Class equipment is described at various points in the FSAR and amendments.
A sum~ary of the Class I equipment and a list of those responsible for evaluating its adequacy Is given In the answer to Question 5.11.
It is Indicated there that a review Is cur rent l y underway of all such equipment to Insure I ts adequacy under seismic excitation.
Additional elaboration on the approach followed In the design of Class I equipment Is contained In the answers to Questions 5.23 and Question 7.2 of hnendment 17.
The approach adopted is acceptable to us.
(d)
Cable Trays The cable trays In this plant, which were particularly noticeable because of the compec t arrangement of the plant and the manner in which they were stacked, receive detailed attention in the answer to Question 5.22.
The design approach summarized there is adequate.
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(e)
T ic-DO\\'IIl of Eg,lJ i pm0.n tRacks It was no ted at the time of the plant inspection that the equipment racks loc o ted i n the relay room (ir,lmediately be low the control room) were not I as tcncd to the floor at that time.
We have been advised by DRL personnel that this equipment is being checked and will be securely fastened by the t imc the plant is approved for operation; this approach is acceptable to us.
(f)
Core 0 i!i.~~
The analysis of the effect of seismic excitation on the reactor internals, and particularly upon factors which might affect core dl sarrey and malfunction, is summarized in the answer to Question 5.6 in Amendment 16.
The discussion and analysis presented there indicates that the reactor core will function satisfactorily during the Design Basis Earthquake.
Design Str~ss Criteria The design stress criteria are presented at various places in the FSAR and a summary is presented in the ans\\~r to Question 5.21 of Amendment No. 16.
I t is indicated there that the allowable stresses for the combined loadl nqs,
including the Design Basis Earthquake and tornadoes, approach the ul t lmatc stress for both the concrete and the steel.
From examination of this table alone, one might question the margin of safety inherent in the design under the maxl num loading conditions.
111 regard to this point, the applicant has provided additional information as to the margin of safety that exists under such loading conditions, especially in the containment s t ruc ture and other structures and equipment which are necessary for safe shutdown and containment.
From the data presented it is our bel ief that the s t rcs s l lm I t s c l tcd will bc reached only in localized regions, and will have 1i t II c henr i on the overall margin of strength inherent in the design.
60 -
J CONCLUD ING REMARKS On the bas Is of the i nfo rma.t i cn made ava i l ab l e to us concerning The Class I structures, piping, reactor l n tern a l s structure, and other Class I items, it is our belief that the plant possesses a reasonabJe margin of 5~f~ty to meet the impo5cd Design Ba~is Earthquake loadings in conjunction with other oppl I cabl e loadings to the extent of insuring safe shutdown and con til i nmen t,
REFERENCES 1.
"Flnnl Safety Anal ys ls Report -- Vols.
I-IV including Supplements 14-1.9,"
Fort St. Vrain Nuclear Generating Station, Publ ic Service Company of Colorado, AEC Docke t No. 50-267, 1970.
2.
Supplementary reports:
(a)
(b)
(c)
(d)
(e)
(f)
(g)
Chu, S. L., J. M. McLaughlin, R. Ryan, G. Patwari, "Fort St. Vrain Nuclear Power Station Unit 1, Turbine-Reactor Building Seismic Analysis, Sargent and Lundy Engineers, P~CD Report No. 16, December 1969 (revised June 1970).
Chu, S. L., J. M. McLaughlin, T. Hongladromp, "Fort St. Vrain Nuclear Powe r Station Unit 1, Combined Turbine-Reactor Building and PCRV Seismic Analysis," Sargent and Lundy Engineers, A&CD Report No. 22, May 196B (revi sed June 11, 1970).
Small, R. J., E. C. Rossow, "The Stress Analysis of the PCRV Support Structure," Sargent and Lundy, January 31, 1968.
JVnin, M., "A Study of Coup) ing between PCRV and Turbine-Reactor Building, Fort St. Vrain, Unit 1," Sargent and Lundy Engineers, A&CD Report No. 11, August 1967.
McLennan, G. A., W. P. Henschel, and E.
D. Esztergar, "Dynamic Response of Indeterminate Space Structures,II Gul f General Atomic Inc. Report GA-8692, i~ay 1,1968.
Amin, M.,
IIA Further Study of the Coupl ing between PCRV and Reactor-Turbine Building, Fort St. Vrain Unit No.1, Sargent and Lun~y Engineers, A&CD Report No. 12, November 1967.
Habu sh, A. L., "Dynanl c Soil Property Investigation, Proposed Fort St. Vruin Nuclear Gcnt!rating Station, Platteville, Colorado,11 pr~pur~d for General Atomic Division, Gen~ral Dynamics Corporation, 23 June 1967.
3.
Nm...nark, N. M. and W. J. Hall, "Rcpor r to AEC Regulatory Stuff, Adequacy of l:le Struc.turlll Criteria for the Fort St. Vruin Nuc lear Generating Stillion, Pllbl ic Service Company of Colorado," Fcbruarv 1'360.
67 -
us, OfPAMTMENT OF COMMERCr;
- NN_-\\
En\\:ironmentar Science Sf~rvices Administration RESEARCH l.ABOHATORIES Silver Spring, Maryland 20910 APPENDIX C Comments on Fort St. Vrain Nuclear Generating Station Public Service Company of Colorado Final Safety Analysis Report Volumes I through IV dated Novemher 4, 1969 and Amendment 17 dated December 21, 1970.
Prepared by Air Resources Environmental Laboratory National Oceanic and Atmospheric Administration March 3, 1971 The onsite meteorological statistics show that at a height of 200 feet above the ground wind speeds less than 0.1, 1.0) and 3.0 meters/sec, respectively, occur 3, 28, and SO percent of the time.
Compared to similar Rtatib~~cs for the entire United States which are usually measured at a height of 30 feet above the ground, this site is poorly ventilated.
Inversion statistics, as presented in Table XII of Attachment B, Amendment 17, show thot a lapse rate greater than + l.SoC per 100 meters occurs 64 percent of the time.
When limited to wind speeds less than 2, 1, and 0.5 meters/sec, respectively, this inversion frequency is 30, 16, nnd 8 percent.
We would interpret the latter case to be an atmospheric diffus Ion ra te equivalent to or worse than Pllsquil1 Type F and 0.5 meters/sec.
Extrapolating these statistics to the 5 percent probability level we estimate that this would be equivalent to Type F and 0.3 meters/sec.
The applicant maintains that the elevat~d release of radioactivity by means of a roof-top ' *"mt at a height of 50 meters will not encounter downwash effects with wind speeds equal to or less than S meters/sec.
One of the references cited is that of Sherlock and Lesher [1] whereby they simulated a l25-ft tall building with stacks of from 250 to 400 feet above ground in a wind tunnel.
It should be noted that the point of emission was from 125 to 275 feet above the building roof.
Consequently, until more dnfinitive data are available, it is our view that all releases, whether through the roof-top vent or from the reactor building through roof-edge louvres, should be treated as a ground source with an appropriate factor f.or building wake effect~
_.~--~----~-------'----------- Our estimate of the maximum, short-term (0-2 hours) relative concentra-tion n t LIn: site boundary of 590 meters is 5.3 x 10-3 sec m-3,
assuming TYl'(l 1'" and 0.3 meters/sec.
This agrees with the applicant IS initial cs t Lma t e in F~.gure 17 of Attachment B.
We do not agree with the rr-v i s Ion (unshed line, fig.
- 17) since the effect of the so-called
" r.: a t i ve l y sluggish Bc l for t anemometer" is offset by the effect of the i
I ght (200 Eee t ) of the anemometer in measuring the conditions of a
. i' l:'ound r c l casc,
Using the cross-sectional area of the reactor bui l~ing (52 meters x 24 meters) a building wake factor of cA
- 1/2 x 1233 m was applied as an additional dilution effect resulting in a relativ !
conccntra tion o.f. 2.7 x 10-3 sec m-3*
o In calculating the aver.age annual concentration, a 22-1/2 sector frequ~ncy of 18 percent for winds from the north was used, as observed in the October 1967 to October 1968 site data.
Furthermore, it was observed that at least 2/3 of these cases were with positive temperature lapse rn ces (stable) which we characterized as Pasquill 'Xype F at a mean speed of 3 m/sec and 1/3 of the cases were neutral or unstable which were characteri:.(:!d as Type C at 6 m/sec~
The resulting average concontra t Ion for a ground release was 4 x 10-5 sec m-3 at the 900 m s Ltu boundary to the south.
At a distance of 590 m the average concentration is 8 x 10.5 sec m-3 which compares to the applicant's value of 1.4 x 10-6 as shown in fig.
15 of Attachment B.
Reference
[1]
- Sherlock, R. H. and E. J.
L~sher, "Design of Chimneys to Control Downwash of Gases".
trans. American Society of Mechanical Engineers, 77, 1, p. 1-9, Jan. 1955.
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69 -
APPENDIX Dl UNITED STATES DEPARTMENT OF THE INTERIOR FISH AND WILDLIFE SERVICE WASHINGTON, D. C.
20240 IN Rl!rLY RErER TO ~
JJJL 14 1!J70 Mr. Harold L. Price Director at' Regulation U.8. Atomic Energy Commission Washington, D.C.
20545 Dcar Mr. Price:
'l'hcsc arc our commcnts on the application by the Public Service Company ol' Colorado for an operating licensc for the Fort St. Vrain Nuclear GenczatLru: Station, Weld County, Colorado, AEC Docket No. 50-267.
In rnnponse to Mr. Boyd's letter of November 14, 1969) the Fish and Wild-life Service has reviewed the Final Safety Analysi" Report which pro-vides information on project operations.
We understand that the applicant has implemented only a small part of the environmental monitoring programs recommended in our letter of June 9, 1967.
While the radiological monitoring program inc1udes the collection of water and silt samples, no provisions were made for the collection 01' aquatic biological samples.
In addition, the Fish and Wildlife Service, the Federal Water Quality Administration, and other interested Federal as well as state agencies were not consulted during the formulation of this monitoring program.
The preservation and enhancement of the quality of oUr environment is a primary concern of the Fish and Wildlife Service.
A proper concern for the envirorunent requires that particular attention be e;iven to fish and wildlife resources during the construction and operational phases of a nuclear power fa.cility.
We under-atund that the ulJVlicl.Ult is cOl~izunt 01' the fact that radio-isotopes of many elements are concentra.ted and stored by organisms which require these or chemically similar elements for their normal metabolic activities.
Since acceptable radiation dose rates and body burdens for fish and wildlife have not been established, the amount of radiation exposure to various components of the aquatic environt'lt:nt, including rO(~ organisms of various fish and wildlife species must be known.
There-fore, the applicant should after consultation with the above-named agencies incorporate aquatic biological samples into their on-going radiological monitorin t-~ pr-ogram, We are also concerned with the possibility of damages to aquatic life
- ccsultin{3 from the release of heated cooling tower blowdown waters and r
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0 1' 1,11(' re co lvln-: nr-eas,
Gille(.' Wp nrc or the opl.ni.on
~ho.t the PubL' c Service Oompany of Colorad.o Itan riot Lakon 1.Flcquate stepe to debernd.ne t ! ~:. pous.lb.Lo detrimental effects (J I' pl ant operation on the ecoloc;y of St. Vrain Creek and the South Platte k.l.vor,
'lIe oppose the issuance of an operatine license for this project untIl the applicant provides the Atomic Energy Corrmission with a detailed cnv 'l.r onmcrrtn.L stntement which includes a description of methods by which tlt r~ applicant will provide anaur-ance satisfactory to the Fish and Wild-Ll. i'o Service thut:
J. *
'l'hermul effluents will be discharged in such a manner u.c to preclude measurable adverse ef1'ects on fish, wilcUife, or other aquatic orf"anisms.
'l'he coolln l~ wuter intake structure will not cause Incasur-Ilble physical damo.ge to fish and other aquatic organisms.
~~.
'l'ho concert Lrution of plant w8st(~S to be discharged to Gt. Vl'ain Creek and the: South Platte River will not lie toxic to the aquatic environment.
Sincerely yours, 0~/I:/~~,-~ ~
Commissioner t
APPENDIX D2 United States Department of the Interior FISH AND WILDLIFE SERVICE BUREAU OF SPORT FISHERIES AND WILDLII
ASHI~GT()~, D.C, 20:l-lO Aoonl!o~ mI LY l ilt I'llllC10'1.
OURlAU VI" !.I'lJIH f1SItERn s A~D WILllLII[
Mr. Harold L. Price Director of Regulation U.S. Atomic EllerEY C'ommission Washineton, D.c.
20545 I. I':';*: 'j...:
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Dear Mr. Price:
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Thin r'eoponds to Mr. Boyd's letter 'of
.Febru~~ 25 which requeste~~~\\'~iTi'/':'~"'\\~/" o comments on amendment No. 18 to the applica'~~on for an operating
:~,.-.~
liconso for the :[4"ort St. Vrain Nuclear Generating Station, Weld County, Colorado, AEC Docket No. 50-267.
We have reviewed the amendment and the Seventh Quarterly Progress Report on the Environmental Radiation Surve:i.llanco Program for the site and surrounding area which was in-cluded as a*~tachment C to the amendment.
As a part of this review, the comments of the Center for Estuarine and Menhaden Research, National Marine Fisheries Sorvice" Department of Commerce" were obtained and Are included heredn, The Seventh Quarterly Progress Report summarizes the establishment of six stations at which preoperatio~al and postoperational monitoring of aquatic biota will be accomplished and states that water temperatures will be monitored above and below the outfall structure.
The mvnitoring program is being conducted by the Colorado state Uni-veraity under a contraotual arrangement with the applicant.
The applicant is cooperating with this Bureau, the University" and the Colorado Division of Game, Fish, and Parks in the preparation of amend-ment No. 19 which, we understand, will contain additional details desoribing the results of preliminary sampling and tentative plans for tho future Dampling program.
We understand that the sampling program will continue for one year following ihitial operation of the plant and that prior to conclusion of the initial phase of the program, re-sults will be discussed with the above concerned agencies to determine the ext-ent and frequency of any further sampling required to adequately dotermine *the effects of plant operations of the fish and wildlife resources and tho environment.
The present sampling program is generally satisfactory.
Sinoe the applicant has provided adequate assurance that the monitoring program will be expanded as necessary and continued in cooperation with the concerned agencies and that appropriate studies will be undertaken to assure that the fish and wildlife resources and the environment will be protected adequately, we hereby withdraw our objections to the issuance of t he oporat.Lng license contained in former Conunissioner Meacham's letter dated July 14,1970.
Tn view of tho importance of fish and wildlife resources and the cnvi.rcnmont, in the pro.i oct area, it in imperative that every effort bo made to protect bhoao valuable rcoources from possible damage from hoabnd wnt( l r', losses into the int,ake sbruct.ure, chemicals, and radio-actd vo controntnatdon,
Therefore, tie recommend that the Public Service Comp:-!J1Y of Colorado be requi.red to:
1.
Continue to cooperate with the Bureau of Sport Fisheries and Wildlife, other concerned Federal agencies, and the appropriate State agencies in developing pl.-ms for radio-logic:ll and environmental surveys.
2.
Continue to conduct such surveys to determine the effects of the plant on the envirorunent and prepare a report of the preoperational surveys and provide copies of them to the Director, Bureau of Sport Flsheries and Wildlife for evaluation prior to reactor operation.
3, Conduct postoperational ecological and radiological surveys following plans developed in cooperation with the Bureau of sport Fisheries and Wildlife and other Federal and Sta.te agencies, analyze the datil" a."ld prepare and submit reports annually until it has been conclusively demonstrated that no significant adverse conditions exist.
Copies of these reports should be submitted to the Director ~ Bureau of Sport Fisheries and Wildlife for evaluation.
4.
Make modifications in project. structures and operations as may be determined necessary to protect the fish and wildlife resources and the environment as a result of the radiological and environmental surveys.
The opportunity for providing our conunent*s is appreciated',
Sincerely yours, 2
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August 3,.1971 i 3 -
APPEI'DlX E A REVIJ:W OF THE "PERMANENT WSS OF FORCED CIRCULATION" ACCIDENT FOR THE FORT ST. VRAIN mJCLEAR GENERATING BTJ.TION by L. A. Booth I.
ImRODUCTION This review is ot the interdependence ot the thermal, nuclear, and tilsion product release analysis ot the "Permenen1i LoSI ot Forced C1rcuJ.ation" accident (Design Basis Accident Ho. 1) as described in the Fort St. Vrain Nuclear Generat1ng Station Final Satety ADalys1a Report.
The basis tor Judgement ot the validity ot these analyses is the rel.evancy ot ana]ytical and experimental techniqu.es in predicting the pertinent results ot the described phenomenology.
It is assumed that the phenomenology is acceptable, and therefore no "What 11"8?~r or other possible deviations trom the described conditions are cons1dered.
Furthermore, the veracity at numerical. resu1.ts 1s assumed, therefore none ot the calcu1ations were repeated.
However, some ot the values ot experimentaJ.ly correlated parame1iers normal Jy available in the open literature were verified.
Certainly, the thermal analysis is the most
~rtant, since it must be assured the.t the structural integrity of the PCRV 1s maintained to prevent the uncontrolled release of fission products.
The transport of fission Products from within the fuel particles to locations available for leakage through the peRV is a strong runctton of the temperature levels within the
74 reactor core.
The changes 1n core ree.etiv1ty tram transport am geometric ch8n3es ot shutdown poisons, tiesion product poisons, and tuel migration are alBO dependent upon temperature levels.
The temperature levels are, in turn, depeDdent upon the location or the heat seneration sites that result trom the dec8¥ heat ot the fission produc'te.
In general, these IUBJyses are conservative, i.e., b1aeed toward pre-dieting events more detrimental than they wouJ.d aetna] '1' be.
The condit ions aD1 assumptions tor the thermal analysis resuJ.t 1D maximum core tempera'tures, thOle for tiss10n product lealtaie AMl1'sis resul"t in maximum CODCeDtrations ot nuclides available for leakage, and those tor the nuclear 8J1IL4r8i8 result in maximum positive reactivity changes.
Each ot these anaJyse. ie discussed in more detail in the 8ubsequent sections.
II.
mERMAL ANALYSIS Calculational Tachn1~
The thermal analysis in 'the COROON code 18 based on a clas.ical solution of the Fourier-Poisson conduction equation with volumetric heat generation using a finite dif':rerence method.
This equation describes exactly the physics of heat conduction through matter.
The alternating gradient scheme ot solving s1multaneously the d1tf'erence equations fer each nodal point provides an explicit-implicit solution which is inherently more accurate and taster than previously used relaxation techniques.
This feature permits finer geometric zoning for a given computer memory am. more computing time for iterative determination of temperature dependent conciuctivities.
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... Because the difference equations must be linear for simultaneoua solution, calculation of convection and radiation is done by linear approximations of convection and radiation coefficients, which are included in the conductivities.
The o.pproximations, being temperature dependent, are verified along with the temperature dependent thermP.1 conductivities by iteration during each computation cycle.
Because the computational. method provides an exact solution to the heat conduction equation, the validity ot the thermal analysis depends upon the completeness of the geometric mock-up of the system and the accuracy ot the conductivities of the materials.
Certainly, the PCRV internal components depicted in Fig. D.l-2 comprise the geometric limitations of the system for which the accident phenomenology apply.
Although the Eulerian grids within the material boundaries were not described 1n the FSAR, the detail presented in Figs. D.l-18, 24, and 27 appear to result trom a sufficiently small grid to assure that the thermal gradients are not high enough across individual cells to cause significant errors in the caJ.culation of the cell-to-cell temperature-dependent conductivities.
Material ThermaJ.
Propertie~
The specific heats and thermal conductivities of the hanogeneous materials, graphite, concrete, helium, insulations, and metallic components, are taken from experimentaJ. data within the appropriate temperature ranges.
Where experimenta.1 data are not available, the lower limit values at the ma.ximwn temperature data are used.
Proper evaluation ot the properties of anisotropic gre.phite aceordLng to the geometric orientation he.s been employed.
Although the composite properties for the reactor core and thermal barrier are not verified experimentally, the method of calculation by volwnetric
76 - fractions for density and specific heat and by equivalent resistive networks for thermal conductivity is based on the fundamental physics of these properties.
Heat Generation Heat generation rates are based on fission product decay chain rates, which are well established through the many years of nuclear experience.
Preshutdown operating power levels have been chosen to provide the maximum power peaking, thus maximum core temperatures.
The 100a1;ion ot heat generation sites, initially within the tyel beads, 1s variable because ot release aDd transport of fission product nuclides.
In general, the release and transport analysis tor
~e determination of heat generation is biased toward fission product retention to maximize the production of energy.
The transport calculation was done with the use of the subroutine CHART which was included in the CORCON code.
This subroutine is similar to the SORS code series, which is discussed in Section III below.
The major difference between CHART and the SOR series is that the diffusion process through the graphite fuel element matrix was included in the CHART routine whereas this process was neglF;cted. in the SORS series.
This teature describes more exactly the transport phenomena and should be included in CHART to predict heat generation within the core.
The properties (vapor pressures and diffusion coefficients) used to characterize the grouping of fission products in the calculation (Tables D.3-19 and D.3-20) are conservative with respect to maximum retention of nuclides within the core.
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III.
FISSION PRODUCT TRANSPORT ANALYSIS Because of the complexity of the physics ot the mass transport of all the fission product nuclides, an exact solution of the concentration distribution of each nuclide Within the PeRV system is virtually
~ssible.
Therefore, the calculations are made with several G1mplify1ng assumptions, and the analysis is based primarily on data from experimental. models.
The significance of the calcula.ted and experimentaJ. results can be seen in Table D.I-IO.
The "Fraction of Core Inventory AvaUable tor Leakage" (Leakase Fr&Cticn),
which is used to calculate the dose rates externaJ. to the PCRV, is the product ot the calculated "Fraction Released from Active Core" (Release Fraction) and the experimental "PeRV Escape Fraction" (Escape Fraction).
For the noble gases these fractions are ell 1.0, so the analysie is triviaJ..
For the halogens and lr.t *';.-1.2.*' in volatility groups I through III, the Relea.Li: Fractions are 0.5 to 1.0 a' :'.:, :".,:: Esce.pe Fractions are at least a factor of 20 less.
For the metals in V(*j.*'::. tllity groups IV through VII, the Release Fractions become significant, ranging from 0.02 to 0.2, but the Escape It"'ractions are orders of 1D88nitude less, the highest being 10.5*
Therefore, the experimentally determined Escape Fractions are the most significant am determining factors of the LeakS8e Fractions.
Calculation of Release Fre.ctions The calculation of the Relea.se Fractions (by the SORS code series) is based on a solution of the classical rate equations, describing time-dependent concentrations of nuclides in the fuel particles, graphite fuel element matrix, and coolant channel for each significant nuclide dec~ chain.
The diffusion process through the fuel element matrix is eliminated in this calcuJ.ation by ansum1ng a uniform concentration of each nuclide as i1; is released from the fuel.
This assumption is conservative' because It neglect~ the diffusion delay through the matrix.
Therefore, the time-dependent concentrations are a function of the release rates from the fuel partIcles and the mass transport rates from the fuel element.graphIte surface to the coolant channel gas phase.
The fission product release rates from the fuel particles are determined experimentally.
The nuclides are grouped within nine groups characterized by volatility for faUed and intact particles.
The release rates (Fig. D.l-12) 8J1d the particle failure criteria (Fi. 1.D-l3) have been obtained by GGA from annealing tests ot irradiated fuel partIcles.
These experiments, designed to obtain data on fuel particle behavior tor normal operating conditions" are described in various GOA R&D ::;>rogress reports.
Although the maximum tempera-ture in these experiments was 2500oK, the linear extrapolation of release rate as a function of reciprocal temperature tor the accident conditions is consistent with theory.
The rna.ss transport rates trom the fuel element graphite surface to the coolant channel gas-phase are calculated by a semi-empirical method.
The mass transfer coefficients are calculated from the well-known Sherwood-*Pigford relationship, based on the heat-mass transfer analogy.
Partial pressures tor adsorbed speciea are calculated (by FREVAP) from the empirical Freundlich Adsorption Isotherm constants and are applicable over the limited range of pressures.
Part1aJ. pressures for condensed and carbide species are calculated
. (by FRELlM) from experimental vapor pressure d~ta.
Therefore" the validity of these calcula.tions is dependent on applicable experimental data.
The constants presented in Table D.l-l are consistent with those derived from other sources ot litera.ture.l 1.
D. H. S'tull et aJ.,
JANAF Thermochemical ToulefJ PB l()8370 distributed by Cleo.r1nchouGc for
},(;d.c.!"~,ScienWlc ancLTc£lmic:oJ Informa.tion, August 1965.
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79 - ~ermination of Escape Fraction The determination ot the Escape Fractions is based entire~ on a series of qualitative experi~nts (Table D.1-4).
These experiments were designed to obtain conservative data for estimation of the maximum transport of fission products from the coolant channel to outside the active core rather than to obtain data using appropriate principles of similitude for the derivation of relationships based on theoretical principles.
Although this conservative approach is "somewhat unscientific" and in some cases does not simulate the physical process at all, the eGA experiments are considered useful in demonstrating the alleged plateout phenomena and the simulation is adequate tor the phenomenological temperatures and materials transport properties.
Although the results trom these eXPeriments are quaJ.itative, these results are considered to be valid tor the prediction at Escape Fractions based on the conservative use of high purge flow rates in the experiments.
IV.
NUCLEAR ANALYSIS The assurance of subcrit1cal1ty during the progress of 'the accident is dependent upon results from the the~ and tlssion product transport an~\\ysis.
The principle ecncerv, is positive reactivity changes which result tram geometric changes in the shutdown poiSODS in the active core.
The transport of shutdown poisons, fission product poisons, and fissionable material is celculated by the sane technique as described in Section III.
The data for diffusion and vaporization of U, Th, B, and their car"bides are obtained from appropriate experiments where the transport conditions are properly s Imul.ated,
80 The analyses indica.te that the largest reactiv1ty change 1s the geometric change in the control rod. shutdown poison.
The magn1tude ot th1s change has been determined experimentaJ.ly.
Likewise, exper1mental determination of the geometric changes in the reserve shutdown poison has been made.
The conditions ot these experiments are considered to simulate the accident conditions appropriately.
V.
OOHCWSIOlfS In the matter ot the permanent 10s8 ot torced circulation accident aWyses 1t can be concluded that the calcula1i1oD&1. techn1ques, pbylica.l parameters, aDd assumptions used 111 determ1D1ns 'the temperatl.U'e distr1butions in the core and other internals ot the PCRV have 1ncorporatad conservatis1Il8 which assure that the 53OO-F temperature, calculated tor 3.7j ot the active core volume, represents an upper limit for temperatures that could occur 111 the core during the accident.
It ~ be cODcluded that the release tractions of radioactive nuclides calculated for fuel particles and core graphite during the course ot the accident represent the maximum release which could occur as a result of the temperatures that the core mat~rials are calculated to be during the accident.
The artificially imposed pressure gradient (5 psi) plus the conservative simulation of phenomenological temperatures and material transport properties give assurance that the quantity of fission products escaping trom the PeRV during the accident will not exceed the quantities calculated by the GGA analysis.
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Baeed on the...elnent ot the accuracy ot 'the calculatloDal techniques, the modeling ot the pbf.1cal processes, the relevanc7 of the experj: ~nt&1 data, and the val1d1ty of the phyeical paTemeters ueed in the &nuyses; an ut temperature 41ltr1butloll, _ter1al tl'8D8pol't, IUld tieeioD product escape
- 1. conlidered UIIbIce**U7 tor t. ron st. Yra1D plaDt.
IAvreace A. Booth statt Member, N-DO'l' lAs Alamol Scientific Laboratory
APPENDIX F FINANCIAL QUALIFICATIONS
~ 82 -
The Commission's regulations which relate to the financial data and information required to establish financial qualifications for an applicant for operating licenses are 10 CFR 50.33(f) and 10 CPR 50 Appendix C.
The application of PUblic Service Company of Colorado (PSC),
as amended, and the accompanying certified annual f1.nancial statements provide the financial information required by the Commission's regulations.
These submittals contain the estimated annual operating and maintenance cost of the Fort St. Vrain Nuclear Generating Statio~ plus the estimated cost of permanently shutting down the facility and maintaininl it in a safe shutdown condition.
The estimated annual operating costs of about
$5.4 million over a five-year period will aggregate $26.9 million.
Such costs include nuclear fuel expense, operating and maintenance labor, supervision and engineering labor, and materials and suppues necessary to operate and maintain the facility.
The applicant's estimate of the cosl of purmRuently shutting down the facility is $5.6 million and
$40,000 annually thereafter for maintaining it in a safe condition.
w~ have examined the certified financial statements of Public Service Company of Colorado to determine whether it is financially qualified to meet these estimated costs.
The information contained in PSC's calendar year 1970 financial report indicates that operating revenues for 1970 totaled $220.3 million; operating expenses were $174.5 million, of which $22.7 million represented II I
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depreciation.
The interest on long-term debt was earned 3.1 times; and the nct income for the year was $30.4 million, of which
$2~.0 million was distributed as dividends to stockholders and the remainder of $9.4 million was retained for use in the business.
As of December 31, 1970, the Company's aSS(!tN totaled $788.2 million, most of which was invested in utility plant
($737.1 million); retained earr-ings amounted to $73.2 million.
Financial ratios computed from the 1970 statements indicate a sound financial condi-tion, e.g., long-term debt to total capitalization -.55, and to net utility plant -.51; net plant to capitalization - 1.07; the operating ratio -.79; and the rates of return on common -
11.6~, on stockholders' investment - 9.8%, and on total investment -
6.2~.
The record of PSC's operations over the past 5 years reflects that operating revenue. increased from $165.4 million in 1966 to $220,3 million in 1970; net income increased from $25.6 million to $30.4 million; and net investment in plant from $553.7 million to $737.1 mi,llion.
Moody's In!estors Service rates the Company's first mortgage bonds as Aa (high quality).
The Company's current Dun and Bradstreet Credit Rating is SAl (the highest category).
Our evaluation of the financial data submitted by the applicant, summari~ed abovl~, provides reasonab~e assurance that the applicant possesses or can obtain the necessary funds to meet the requirements of 10 CFR 50.33(f) with respect to the operation of the Fort St. Vrain Nuclear Generating Station.
Coples of the staff's financial analysis of the Company are attached as an Appendix.
PUBLIC SERVICE COMPANY OF COLORADO "
FORT ST. VRAIN FINANCIAL ANALYSIS (dollars in millions)
Calendar Year Ended Deccmber 31 1970 1969 1968
$ 376.1
$ 342.3
$ 320.9 737.1 670.2 637.0.
.51
.51
.50 737.1 670.2 637.0 687.0 623.2 592.2 1.07 1.07 1.0i 310.9 280.9 271.3 188.2 717.6 681.5
.39
.39
.40 26.7 24.9 24.6 230.9 200.9 191.3 11.6%
12.41.
12.8%
30.4 28.6 28.3 310.9 280.9 271.3 9.8%
lO.2~
10.4%
48.6 43.9 42.4 788.2 717.6 681.5 6.2'.
6.1~
6.2%
Long-term debt
~tility plant (net)
Ratio - dr:bt to fixed plant Utility plant (net)
Capitalization Ratio of net plant to capitalization Stockholders' equity Total assets Proprietary ratio Earnings available to common equity Common equity Rat~ of earnings on common equity Net income Stockholders' e~uity Rnt(~ of caru tngs on stockholders' equity Net income before interest Li.nbLll t Lca and capital Rate or enrn Lngs on total inv~8tment Nl ~t income before interest Intcrest on long-term dcbt No. of timcs inter~st (long-term) earned Net income Total revenues Net income ratio Total utility opc~ating expenses Total utility operating revenues Opera ::i ng ratio u~ility plant (gross)
Utility operating revenues Ratio of plant investment to revenues Earnings per share of common 1970 48.6 15.6 3.1 30.4 223.1
.14 174.5 220.3
.79 936.0 220.3 4.25
$1.82 43.9 14.5 3.0 28.6 205.6
.14 161.7 204.7
.79 850.2 204.7 4.15
$1.73 1969 42.4 13.6 3.1 28.3 193.3
.15 150.9 190.0
.79 798.4 190.0 4.20
$1.71
/". "
Amount
%of Total Amount
%of Total Capitalization:
Long-term debt Prcfurr cd stock Common stock & surplus Total
$376.1 80.0 230.9
$687.0 54.8%
11.6 33.6 100.0%
$342.3 80.0 200.9
$623.2 54.9%
12, 32.3 100.0'0 MOOdy's Bond Ratings:
First Mortgage:
Aa Dun and Bradstreet Credit Rating:
5A1
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