ML19322A786
| ML19322A786 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 12/01/1966 |
| From: | DUKE POWER CO. |
| To: | |
| References | |
| NUDOCS 7911250004 | |
| Download: ML19322A786 (39) | |
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I v TABLE OF CONTENTS Section Page, 5 CONTAINMENT SYSTEM 5-1 5.1 REACTOR BUILDING 5-1 5.1.1 DESIGN BASES 5-1 5.1.1.1 Postulated Accident Conditions 5-1 5.1.1.2 Energy and Material Release 5-2 5.1.1.3 Contribution of Engineered Safeguards System 5-2 1
5.1.2 STRUCIURE DESIGN 5-2 )
5.1.2.1 Design Conditions 5-2 1
5.1.2.2 Design Leakage Rate 5-3 '
5.1.2.3 External Loadings 5-3
- g 5.1.2.4 Codes 5-5 5.1.2.5 Drawings 5-6 5.1.2.6 Penetrations 5-6 5.1.2.7 .41ssile Protection Features 5-7 5.1.2.8 Corrosion Protection 5-13 5-/.5 5.1.2.9 Insulation .5 44 5 ts 5.1.2.10 Shielding 5-M 5.2 ISOL'. TION SYSTEM 5-15 5.2.1 DESIGl; BASES 5-15
! 5.2.2 SYSTEM DESIGN 5-15 5.3 VENTIIATION SYSTEM 5-20 5.3.1 DESIGN BASES 5-20 5.3.1.1 Coverning Conditions 5-20 p 5.3.1.2 Sizing 5-20
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O Section Pajte 5.3.2 SYSTEM DESIGN 5-20 5.3.2.1 Isolation valves 5-20 i
5.4 LEAXAGE lONITORING SYSTEM 5-21 l 1
l 5.5 SYSTEM DESIGN EVALUATION 5-23 5.6 TESTS AND INSPECTION 5-24 l 5.6.1 PREOPERATIONAL TESTING AND INSPECTION 5-24 l
S.6.1.1 During Construction 5-24 5.6.1.2 Structural Test 5-25 )
5.6.1.3 Initial Leakage Test 5-25 5.6.2 POSTOPERATIONAL SURVEILIANCE 5-25 5.6.2.1 Leakage Monitoring 5-25 :
5.6.2.2 Surveillance of Structural Integrity 5-25 i
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. LIST OF TABLES !
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5-1 Missile Energies 5-10 1 i
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5-2 . Missile Penetrations 5-12 l t
i 5-3 Reactor Building Isolation 5-17 thru 5-19 ;
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Figure No. Title 5-1 Reactor Building Typical Details 5-2 Typical Electrical and Piping Penetrations
- 5-3 Details of Equipment Hatch and Personnel Lock 1
5-4 Reactor Building Isolation Valve Arrangement 5-5 Reactor Building Isolation Valve Arrangement 5-6 Reactor Building Normal Ventilation System 5-7 Reactor Building Instrumentation O
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4 5 CONTAINMENT SYSTEM l The containment for this station consiste of two systems which are:
! a. The Reactor Building to provide biological and missile shielding and to contain the energy and material that l
could be released by an accident.
I i b. The engineered safeguards systems which limit the maximum value of the energy released by an accident.
l 5.1 REACTOR BUILDING
, The Reactor Building is a fully continuous reinforced concrete structure j
in the shape of a cylinder with a shallow domed roof and a flat foundation slab. The cylindrical portion is prestressed by a post-tensioning system consisting of horizontal and vertical tendons. The dome has a three-way poi,t-tensioning system. The foundation slab is conventionally reinforced with high-strength reinforcing steel. The entire structure is lined with one-quarter inch welded steel plate to provide vapor tightness.
The approximate dimensions of the Reactor Building are: inside diameter, 116 feet; inside height, 206 feet; vartical wall thickness, 3-3/4 feet; dome thickness, 3-1/4 feet; and the foundat!.on slab, 8-1/2 feet. The build-ing encloses the pressurized water reactor, steam generators, reactor coolant loops and portions of the auxiliary and engineered safeguards systems. The Reactor Building is shown in Figures 1-10 through 1-15.
The interior arrangement is designed to meet the requirements for all anticipated conditions of operation and maintenance, includ ag new fuel and spent fuel handling.
l The building is designed to sustain safely all internal and external loading conditions which may reasonably be expected to occur during ;
the life of the station. Due consideration has been given to all site factors and local environment as they relate to the public health and 1 safety. The design of this structure is described in detail in this i
section.
Full advantage is being taken in the design of this Reactor Building of the experience gained in the review of similar designs with the AEC for che Florida Power and Light Company's Turkey Point Plant - (Docket Nos.
< 50-250 and 251), Consumers Power . Company's Palisades Plant (Docket No.
50-255) and Wisconsin-Michigan Power Company's Point Beach Plant (Docket No. 50-266), as well:as containment designs by others which meet the same
! functional requirements.
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5.1.1 DESIGi BASES
- 5.1.1.1 Fostulated Accident Conditions The containment . system is designed to provide protection for. the public
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from the consequences of a worst loss-of-coolant accident as defined in Sec-tion 14, " Safety Analysis . " The accident is based on a sudden break in the reactor coolant system. 1ressure and temperature behavior subsequent to the accident is determined by the combined influence of those engineered safe-guards assumed to operate, and the energy sources and heat sinks. This is discussed in detail in Section 6, " Engineered Safeguards," and Section 14,
" Safety Analysis."
5.1.1.2 Energy and Material Release The source available for the release of energy and materials into the con-tainment system are:
Reactor stored heat Reactor decay heat Reactor coolant system stored heat Metal-water reactions Hydrogen combustion Fission products from the core The energy released by each of these sources during the accident is discussed in Section 14, along with the post-accident time-dependency functions of the released energy. Energy contribution from the secondary steam system is not included in the calculation of Reactor Building design pressure.
5.1.1.3 Contribution of Engineered Safeguards Systems As the design application and contribution of the engineered safeguards systems are discussed fully in Section 6, only their relation to the con-tainment system design is included in this section.
Engineered safeguards systems are provided to minimize the consequences of the accident by: removing heat from the containment system; removing heat from the fuel; and inserting negative reactivity into the reactor.
These systems are:
A high pressure injection system A low pressure injection system Core flooding tanks A Reactor Building emergency cooling system A Reactor Buil': g spray system A Reactor Buili 2g isolation system A penetration rooms ventilation system These systems are designed to limit the maximum value of the energy released in the Reactor Yuilding.
5.1.2 STRUCIURE DESIGN 5.1.2.1 Design Conditions
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The Reactor Building is designed to withstand an accident causing an internal pressure of 59 psf g with a coincident temperature of 286 F.
5-2 (Revised 5-25-67 ) 000 359
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The internal net free volume is 1,900,000 cubic feet. Material and design !
features uf the liner plate and penetrations are covered fully in Appendix SE, " Liner Plate Technical Specification." The pressure transient curves '
for the design basis accident are contained in Section 14 beginning with Figure 14-43. The Reactor Building is Class 1 as defined in Appendix 5A, "Sttuctural Design Bases," and is designed according to these bases. De- j tailed information on structural analysis and a discussion of materials l used in the building are presented in Appendix 5B, " Design Program for ;
Reactor Building." '
The foregoing structural design conditions are essentially the same as the containment structures for the Florida Power and Light Company's Turkey Point Plant, Consumers Power Company's Palisades Plant and Wisconsin-Michigan Power ,
Company's Point Beach Nuclear Plant. The final design values of moment, ;
- shear, deflection, meridional force and hoop tension are being developed but have not yet been completed. For preliminary detail drawing of cylinder wall-foundation slab intersection, cylinder wall-dome intersection and liner plate anchorages, refer to Figure 5-1. Criteria for design of the Reactor Building I are contained in Appendix SC, " Design Criteria for Prestressed Concrete Re- l actor Building. "
5.1.2.2 Design Leakage Rate The Reactor Building will be tested at the conclusion of construction to
("T demonstrate that leakage at design pressure does not exceed 0.5 per cent
( by volume in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Either the reference volume method or the absolute method will be used.
5.1.2.3 External Loadings 5.1.2.3.1 Barometric Pressure The Reactor Building is designed for an external atmospheric pressure of 2.5 psi greater than the internal pressure. This corresponds to the differential pressure that could be developed if the building is sealed with an internal temperature of 120 F and with a barometric pressure of 29.0 inches Hg and the building is subsequently cooled to an internal temperature of 80 F with a concurrent rise in barometric pressure to 31.0 inches of Hg. Since the I
building is designed fc,e this pressure differential, vacuum breakers are not requirad.
i 5.1.2.3.2 Wind, Including Tornadoes l
Simultaneous external loadings to be used in the design of tornado resistant j Class 1 structures are: '
- a. Differential pressure developed over 5 seconds - 3 psi .
- b. External wind forces resulting from a tornado funnel having a peripheral tangential velocity of 300 mph and whose center is traveling at 40 mph.
- s. A torsional moment resulting from the horizontal peripheral
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tangential velocity.
(/ Missile equivalent to an 8 in, diameter x 12 ft long piece of wood traveling end-on at 250 mph. -
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The forces due to the wind will be calculated in accordance with methods described in ASCE Paper No. 3269 entitled, " Wind Forces on Structures."
Applicable pressure and shape coefficients will be used. There is no variation with height nor gust factor.
There is no published suggested criteria for tornado resistant design.
A tornado is broadly defined as a whirling vortex of air and most authorities agree that there is an upward wind force of considerable magnitude. The diameter of the vortex may range from a few feet to one mile. The largest known diameter of 1.25 miles was measured in Europe.
It has been estimated that the tangential velocity of a particle of air in the cone may range up to 500 mph. (1,2) A velocity of 200 to 250 mph would represent an average tornado. (2,3) Hoecker (4) suggests that the upward velocity (at least for the Dallas tornado) is of the same order of magnitude as the tangential velocity. There is undoubtedly an inward flow of air at grade elevation with the velocity increasing as it approaches the funnel. (2) Reduced pressures exist in the vicinity of the funnel.
Values of reduced pressure are not well defined or reliable. A measure-ment recorded during the 1966 Topeka tornado 292 ft. from the funnel indicated a pressure drop of 0.7 inches of Hg in 30 seconds. The maximum values of pressure at the funnel are thought to be 100 to 200 mb below j ambient. (5,6,7 and 8) 1 i
Investigations of areas struck by tornadoes indicate that wood frame buildings and non-reinforced masonry buildings suffer heavy damage.
Steel frames of structures stand with relatively little damage but usually a good part of the siding and roofing are lost. Well designed and constructed reinforced concrete structures suffer the least damage. A heavy steel frame structure with well anchored roof and walls would also suffer minimal damage. A concrete water tower, the 8-story Federal Reserve Building and the Washburn University 2-story Science Hall are examples of well designed structures which withstood the Topeka tornado. The water tower (150' high, 92' dia.) that is similar in shape and general structural integrity to.the Reactor Building, suffered no damage other than the loss of doors and hatches in the supporting cylinder.
- 1. On the Nature of Tornadoes, Silberg, P. A. , Raytheon Co., June '61.
- 2. Tornadoes of the United S tates, Flora, S. D. , Univ. of Okla. , '57.
- 3. Personal Communication with Dr. C.C. Chang, Catholic Univ. , Aug. '66.
- 4. Wind Speed and Flow Pattern in the Dallas Tornado of April 2, '57, I
I Hoecker, W. J. , Monthly Weather Review, Wash. , D.C. , May 1960.
S. Three-Dimensional Pressere Pattern of the Dallas Tornado and Some Resultant Implications, Hoecker, W.J. , Monthly Weather Review, Wash., D.C., December 1961.
- 6. The Topeka, Kansas Tornado, ENR, April '66.
7 Tornadoes and Related Phenomena, Brooks, E. N., Compendium of Meterology, Boston Metrological Soc., '51.
- 8. Tornado Studies, Gloser, Arnold, Texas A & M Department of Oceano-graphy and Metrology, Contract Cub 8696 '56.
000 361 5-4
(q ~jh Ihe general land terrain and adjacent structures influence the magnitude of the tornado. The flat midwestern states have experienced the most severe tornadoes while those in hilly or mountainous areas, similar to our station, are generally of lesser magnitude.
5.1.2.3.3 Snow or Ice A live load of 20 psf, distributed uniformly on the dome of the Reactor Building will be used for snow and ice, in accordance with the Southern Standard Building Code.
5.1.2.3.4 Floods or Inundations The plant grade is at elevation 796 feet and is situated on a ridge between two draws so the drainage is away from the site. The Keowee and Little River dams will create a reservoir in which the maximum tmpounded water level design elevation is 800 feet. The site is sepa-rated from the reservoir by natural higher terrain which protects the site. Refer to 2.4, " Hydrology and Groundwater" and 2.5, " Geology."
A site drainage ;em will be provided as equired. Therefore, the plant site will not be subject to inundations.
5.1.2.3.5 Seismic Considerations The design seismic ground acceleration (reference 2.6) is 0.05 g hori-zontally and vertically. Structural design principles will be followed Os in the design for seismic loads, including a dynamic analysis for all Class I structures, components and systems. Plot of the response spectra is shown in Appendix 2B, " Seismology."
The basis for classification of the plant structures and systems as shown in the PSAR are the following references:
- 1. Design of Nuclear Power Reactors Against Earthquakes, Housner, G. W., Proceedings of the Second World Conference on Earthquake Engineering, Vol. I, Japan 1960, pg.133, 134, and 137.
- 2. AEC Publication TID-7024, Nuclear Reactors and Earthouakes, ;
pg. 111. '
1 5.1.2.4 Codes The Reactor Building is designed under the following codes as modified herein by the structural design bases:
- a. Southern Standard Building Code
- b. Building Code Requirements for Reinforced Concrete (ACI 318-63)
- c. AISC Manual of Steel Construction i
- d. ASME Boiler and Pressure Vessel Code,Section III, ;
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Nuclear Vessels,Section VIII, Unfired Pressure Vessels, I Section IX, Welding Qualificatirns (Applicable Portions) !
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5.1.2.5 Drawings ~
Figures 1-9 through 1-14 are plans and elevations showing principal dimensions of the Reactor Building. Plot plan is shown in Figure 2-2.
5.1.2.6 Penetrations 5.1.2.6.1 Piping and Other Mechanical and Electrical Penetrations Penetrations for process piping, instrumentation lines, ventilation ducts and electrical lines are designed to withstand Reactor Building design pressure and temperature as well as anf forces due to differen-tial expansitn between piping systems and the structure.
Where required, bellows are provided between piping and the Reactor Building wall to prevent excessive forces on the piping or on the Reactor Building. Portions of penetrations subject to stress by Reactor Building pressure are made of impact-tested steel, such as ASTM A-201 or A-212 produced to A-300 specifications. Where they are not located in a heated enclosure, penetrations are insulated from outside ambient temperature.
Penetration assemblies are seal-welded to the Reactor Building liner.
Typical electrical and piping penetrations are shown in Figure 5-2.
Piping penetrations are the rigid all-welded type where tha penetration consists of the process line pipe itself rather than a separate assembly.
Such piping penetrations will be tested in the general leakage test of the Reactor Building. Personnel access locks, the equipment hatch, and penetrations with resilient seals .and expansion bellows can be teste locally at design pressure.
5.1.2.6.2 _ Equipment Access Opening A 19 foot alameter pressure-sealed door is provided to enable passage of large equipment, such as reactor coolant pumps and motors and reactor
, vessel 0-rings, into the Reactor Building during a unit shutdown.
l Analytical solutions for the determination of state of stress in the vicinity of equipment openings are obtained from reference to the article:
l State of Stress in a Circular Cylindrical Shell With a Circular Hole, by A. C. Eringen, A. K. Naghdi and C. C. Thiel - Welding Research Council Bulletin No. 102, January 1965.
The analysis of the Reactor Building as a whole is first carried out without considering the openings in it. This analysis will be done through computer aid using the finite element program.
The Reactor Building with the opening in it is then analyzed in the following steps:
- a. Formulation of aifferential equations for the shell in complex variable form with the center of the hole as the origin.
(See the aforementioned reference.)
- b. Solution of the differential equations. m 000 ..
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l c. Formulation of the boundary conditions based on the l stresses obtained from the vessel analysis above without the hole.
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- d. Evaluation of constants in the solution.
- e. Calculation of membrane forces, moments, and shears around and at the edge of the opening.
- f. The wall thickness around the opening will then be increased and reinforced to carry the higher forces, moment, and shears.
- g. Finally, the design will be checked to insure that the 4
strength of the reinforcement provided replaces the strength removed by the opening. This check is to i maintain a good degree of compatibility between the general vessel shell and the area around the opening.
I Details of the reinforcing and deflected strand pattern around the equipment opening and personnel lock is shown in Figure 5-3.
The pattern of stresses at design accident loading is not expected to be significantly different than the pattern of atresses during
, the acceptance test, since the pressure stresses ar far more significant than other stresses.
5.1.2.6.3 Personnel Locks A personnel airlock is provided for access for maintenance services.
- This lock has two gasketed doors in series which are interlocked so that only one door may be opened at a time. Operation of the lock doors is by mechanical means and without power assist. The open or )
a closed condition of each door will be indicated in the control room. '
A personnel escape lock with similar features is also provided.
5.1.2.7 Missile Protection Features 5.1.2.7.1 Reactor Building l
Missile protection for the building liner will be provided to comply with the following criteria: ,
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- a. The building and liner will be protected from loss of function due to damage by such missiles as might be generated in a loss-of-coolant accident,
- b. The engineered safeguards systems and components required to maintain containment integrity and to meet the site criteria cf 10 CFR 100 shall be protected against loss of function due to damage
(, by the missiles defined below. ,,
, 4-5-7 * ~ ~ *
- During the detailed plant design, the missile protection necessary to meet the above criteria will be developed and implemented using the f.
following considerations:
- a. The reactor coolant system will be surrounded by reinforced concrete and yteel structures designed to withstand the forces 2EY!$2Yf: with double-ended rupture of a main reactor coolant pipe and designed to stop the missiles.
- b. The structural design of the missile shielding will take into account both static and impact loads and will be based upon a barrier cross section with energy absorption capacity at least 25 per cent greater than the impact energy of the considered potential missile.
- c. Misaile velocities will be calculated considering both fluid and mechanical driving forces which can act during missile generation.
- d. Components of the reactor coolant system will be examined to identify and to classify missiles according to size, shape and kinetic energy for purposes of analyzing their ef fects. The components to be examined will include:
All valve stems up to and including the largest size to h be used.
All valve bonnets.
All instrument thimbles.
Various type and sizes of nuts and bolts.
Reactor vessel head bolts.
5.1.2.7.2 Main Steam Turbine Missiles The turbine-generator supplier has made a study of failure of rotating elements of steam turbines and generators. The postulated types of failures are: (a) failure of rotating components operating at or near normal operating speed and (b) failure of components that control admission of steam to the turbine resulting in destructive shaft rotational speed.
(a) Failure at or Near Operating Speed All of the known turbine and generator rotor failures et near rated speed resulted from the combination of severe strain concentrations in relatively brittle materials. New alloys and processes have been developed and adopted to minimize the probability of brittle fracture in rotors, wheels and shafts. Careful control of chemistry and detailed heat treating cycles have greatly improved the mechanical properties of all of these components. Transition temperatures (the temperature at which the character of the fracture in the steel changes from brittle to ductile, often identified as FATT) have been reduced on the low temperature wheel and rotor applications for nuclear units to 5-8
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well below start-up temperatures. Inproved steel mill practices in vacuum pouring and alloy addition have resulted in forgings which are much more uniform and defect free than ever before.
More comprehensive vendor and manufacturer tests involving improved ultrasonic and magnetic particle testing techniques are better able to diccover surface and internal defects than in the past. Laboratory investigation has revealed some of the basic relationships between structure strength, material strength, FATT and defect size and location so that the reliability of the rotor as a structure has been significantly improved over the past few years.
New starting and loading instructions have been developed to reduce the severity of surface and bore thermal stress cycles incurred during service. The new practices include:
- 1. Better temperature sensors
- 2. Better control devices for acceleration and loading
- 3. Better guidance for station operators in the control of speed, acceleration and loading rates to mlnimize rotor stresses.
('~') Progress in design, better materials and quality control, more i _f rigorous acceptance criteria and improved machine operation have substantially reduced the likelihood of burst failures of turbine-generator rotors operating near rated speed.
(b) Failure at Destructive Shaf t Rotational Speeds 1
Improvements of rotor quality discussed above, while reducing the chance of failures at operating speed, tend to increase the hazard l level associated with unlimited overspeed because of higher bursting speed. Therefore, turbine overspeed protection systems have been evaluated as follows:
- 1. Main and secondary steam inlets have the following valves in series:
- a. Control valves - controlled by the speed governor and. tripped closed by emergency governor and backup overspeed trip, thus providing three levels of control redundancy,
- b. Stop valves or trip throttle valve - actuated by the emergency governor and backup overspeed trip, thus providing two levels of control redundancy.
Since 1948 there have been over 650 turbines, of over 10,000 kw each, shipped and put in service by the Oconee
(,-) turbine supplier with no report of main stop valves v
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failing to close when required to protect the turbine.
Impending sticking has been disclosed by means of the fully closed test feature so that a planned shutdown could be made to make the necessary correction. This almost always involves the removal of the oxide layer which builds up in the stem and bushing and which would not occur on a low temperature nuclear application.
- c. Combined stop and intercept valves in cross around systems - these are actuated by the speed governor, emergency and backup overspeed trips. These valves also include the testing features described above.
The speed sensing devices for the governor and emergency governor are separate from each other, thus providing two independent lines of defense.
- 2. Uncontrolled Extraction Lines to Feedwater Heaters If the energy stored in an uncontrolled extraction line is sufficient to cause a dangerous overspeed, two positive closing nonreturn valves are provided, to be actuated by the emergency governor and backup overspeed trip. These are designed for remote manual periodic tests to insure proper operation. The station piping, heater and check valve system are reviewed during the design stages to make sure the entrained steam cannot overspeed the unit beyond safe limits.
Special field tests are made of new components to obtain design information and to confirm proper operation. These include the capability of controls to prevent excessive overspeed on loss of load.
Careful analysis of all past failures has led to design, inspection and testing procedures to substantially eliminate destructive overspeed as a possible cause of failure in modern design units.
The study of postulated ruptures made by the turbine-generator supplier concludes that the missile having the highest combination of weight, size, and energy is the last stage wheel. The properties of this missile are summarized in Table 5-1. Initial velocities and energies shown below are based on 186 per cent rated speed before leaving the casing. Based on the turbine manufacturer's recommendation, the analysis is based on 60 per cent of the initial energy being absorbed in penetrating the casing.
Table 5-1 Weight Impact Area 6600 lbs Side On - 8.89 sq ft End On - 3.91 sq ft Velocity Kinetic Energy Ft-lbs Initial - 700 fps Ini:ial - 50.2 x 106 Impact Impact ,
Cylinder - 20.1 x 106
, Cylinder - 443 fps Dome - 390 fps Dome - 16.48 x 106 000.. 4 5-10 (Revised 5-25-67)
Analysis of the above missile is based on calculations using methods pre-sented in Nav. Docks P-51* to determine the depth to which this missile would penetrate the concrete Reactor Building. Conservatively, no reduction of missile energy was made for penetration of the Turbine Building and/or impact with intervening equipment and structural components after leaving turbine shell. The energy loss from 20.1 x 106 ft-lbs to 16.48 x 106 ft-lbs is caused by air friction. This effect has been calculated by using a drag coefficient of 1.0. Since the offset between the Turbine and Reactor Build-ings is relatively short, about 170 feet, no account has been taken for air friction losses for the case in which the missile is ejected nearly hori-zontally to strike the cylinder wall. Following are results of analysis:
Case I:
" Side on" impact. Missile could penetrate the concrete cylinder wall to a depth of approximately 5-1/4 inches and the dome to a depth of approximately 5 inches. The tendons will not be damaged since they are protected to a depth of 7-3/4 inches in the cylinder wall and 8 inches in the dome.
Case II:
"End on" impact. In this case the missile could penetrate the concrece cylinder wall to a depth of approximately 12 inches and the dome to a depth of approximately 11 inches. The tendon arrangement is such that the missile could strike two adjacent tendons in the dome or a maximum of s ,) three horizontal and one vertical tendons in the cylinder wall. The local effect on the tendons could be one of either partial deflection or possible severance. However, analysis of the structure indicates that the structure can withstand the loss of three horizontal and three vertical tendons in the cylinder wall or five adjacent tendons in the dome without loss of function and a greater number of tendons without building failure.
Case III:
As a final analysis, an extreme case was considered in which none of the initial kinetic energy of the missile is absorbed by its penetration through the turbine casing. The total initial energy of 50.2 x 106 ft-lbs is avail-able for penetration of the cylinder wall and 32.88 x 106 ft-lbs for penetra-tion of dome where the reduction is due to air friction only. The maximum depth of penetration of cylinder wall is 37 inches and the dome is 29 inches.
The missile can strike five tendons in the dome or three horizontal and one vertical tendons in the cylinder wall. The local effect in the impact area would be as described in Case II above even though the depth of penetration is greater.
Depths of penetration of Reactor Building wall are summarized in Table 5-2.
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- Nav. Docks P-51, Design of Protective Structures, Amirikian, A;
( ,/ Bureau of Yards and Docks, Department of the Navy, August 1950.
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5-11 (Revised 5-25-67) l
Table 5-2 Depth of Penetration of Concrete Case I Case II Case III Cylinder Dome Cylinder Dome Cylinder Dome 5-1/4" 5" 12" 11" 37" 29" Since the thicknesses of the cylinder wall and dome are 45 inches and 39 inches respectively, it can be seen that the turbine missile, even under ex-treme assumptions, does not penetrate the Reactor Building.
5.1.2.7.3 Tornado NEssiles For an analysis of missiles created by a tornado having maximum wind speeds of 300 mph, two missiles were considered. One is a missile equivalent to a 12 foot long piece of wood 8 inches in diameter traveling end on at a speed of 250 mph. The second is a 2000 pound automobile with a minimum impact area of 20 square feet traveling at a speed of 100 mph.
For the wood missile, calculations based on energy principle indicate that because the impact pressure exceeds the ultimate compressive strength of wood by a factor of about four, the wood would crush due to impact. However, this could cause a secondary source of missiles if the impact force is sufficiently large to cause spalling of the free (inside) face. The com-pressive shock wave which propagates inward from the impact area generates a tensile stress pulse and a shear stress pulse that are of concern. The tensile pulse, if it is large enough, will cause spalling of concrete as it moves back from the free (inside) surface. This spalled piece moves off with some velocity due to energy trapped in the material. Successive pieces will spall until a plane is reached where the tensile pulse becomes smaller than the tensile strength of concrete. From the effects of impact of the 8 inch diameter by 12 foot long wood missile, this plane in a conventionally reinforced concrete section would be located approximately 3 inches from the free (inside) surfacc. However, since the Reactor Building is prestressed, there will be residual compression in the free face, as the tensile pulse moves out and spalling will not occur. Calculations indicate that in the impact area a 2 inch or 3 inch deep crushing of concrete should be expected due to excessive bearing stress due to impact.
For the automobile missile, using the same methods as in the turbine failure analysis, the calculated depth of penetration is 1/4 inch and for all practi-cal purposes the effect of impact on the Reactor Building is negligible.
From the above, it can be seen that the tornado generated missiles neither penetrate the Reactor Building wall nor endanger the structural integrity, of the Reactor Building or any components of the reactor coolant system.
O'j W
6 ,
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5-12 (Revised 5-25-67 ) i l
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5.1.2.8 Corrosion Protection Galvanic corrosion normally occurs underground, underwater, or in the presence of a corrosive medium. Atmospheric conditions may cause surface attack but there will be no galvanic corrosion unless metals of two different electro-chemical levels exist and the medium between them permits current flow.
Consequently, if the materials used are steel, and precautions are taken to prevent water from providing a conducting path between them, there should be no galvanic corrosion. (1)
Values of electrical resistivity of the soil as measured at the station site are extremely high as shown on the graph on page 5-14.
The mean resistivity at the station site of 80,000 ohm-cm is classified as non-corrosive. (2 )
Because the ground corrosive influences are practically non-existent dta to high soil resistivity, the steel liner and its attachments that extend below the ground level to base slab will not require any protection against galvanic corrosion.
The surface of the exposed liner plate will be protected by an initial surface cleaning and prime coat of paint applied at the fabrication plant and, further, by application of a finish paint. The prestressed tendons will be surrounded
, and protected by a corrosion protective compound.
It is planned to use Dearborn Chemical Company's No-Ox-Id A Special or C-M Casing Filler, or equal, as a protective compound for the tendons. This filler will be applied to the tendons as a non-drying protective coating in the factory after fabrication, and after tendon installation in the casings will be pumped into all voids surrounding the tendons.
No-Ox-Id is essentially a modified refined petroleum oil base (about 98 per cent petroleum jelly) which contains no solvent, although it does contain certain proprietary chemical additives and inhibitors to prevent corrosion of the steel. The chemical composition of this material indicates that it would possess the normal stability of the linear hydrocarbons subjected to ambient temperature levels. This material has a pour point of 110 to 115 F and will be applied at approximately 130 F to drive air and vapor from the voids before solidifying to a soft gel.
Dearborn Company has stability data going back 10 years which indicates that the casing filler will not deteriorate during the 40 year life of the station.
5.1.2.9 Insulation The liner plate will not be insulated. Insulation and/or cooling coils will be provided for hot line penetrations through the Reactor Building. Cooling coils will be provided in the primary shield wall for reduction of gansna heating in the concrete.
b (1) The Corrosion Handbook, Uhlig, H. H. , N. Y. 1948, Pg. 481-496.
(2) Cathodic Protection, Applegate, L. M., McGraw-Hill, 1960, Pg. 66.
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5.1.2.10 Shielding The Reactor Building completely encloses the reactor coolant system to minimize release of radioactive material to the environment should a failure of the system occur. This building provides adequate biological shielding for accident situations. In combination with the interior concrete shielding walls it provides adequate shielding for limited access during normal operation.
5.2 ISOLATION SYSTEM
- 5.2.1 DESIGN BASES The general design basis governing isolation valve requirements is:
Leakage through all fluid penetrations not serving accident-consequence-limiting systems is to be minimized by a double barrier so that no single, credible failure or malfunction of an active component can result in i loss-of-isolation or intolerable leakage. The installed double barriers l take the form of closed piping systems, both inside and outside the Reactor Building, and various types of isolation valves.
Reactor Building isolation occurs on a signal of approximately 4 psig in the Reactor Building. Valves which isolate penetrations that are directly open to the Reactor Building such as the Reactor Building purge valves and sump drain valves, will also be closed on a high radiation signal.
The isolation system closes all fluid penetrations, not required for opera-tion of the engineered safeguards systems, to prevent the leakage of radio-active materials to the environment. Fluid penetrations serving engineered safeguards systems alco meet this design basis.
l All remotely operated Reactor Building isolation valves are provided with ;
position ibnit indicators in the control room. j 5.2.2 SYSTEM DESIGN l l
The fluid penetrations which require isolation after an accident may be l 4 claased as follows: l Type I. Each line connecting directly to the reactor coolant system has two Reactor Building isolation valves. One valve is external, and the other valve is internal to the Reactor Building. These valves may be either a I
check valve and a remotely operated valve, or two remotely operated valves, depending upon the direction of normal
- flow.
Type II. Each line connectiEg directly to the Reactor Building j atmosphere has two isolation valves. At least one valve is external, and the other may be internal or external to the Reactor Building. These valves may be either a check valve and a remotely operated valve or two remotely
_,/ operated valves, depending upon the direction of normal flow.
6 "
5-15 (Revised 6-16-67) .. .
Type III. Each line not directly connected to the reactor coolant system or not open to the Reactor Building atmosphere has one valve, either a check valve or a remotely operated valve. This valve is located external to the Reactor Building.
Type IV. Lines which penetrate the Reactor Building and are connected to either the building er the reactor coolant system, but which are never opened during reactor operation, have provisions for locking in a closed position.
There are, additional subdivisions in each of these major groups. The individual system flow diagrams show the manner in which each Reactor Building isolation valve arrangement fits into its respective system.
For convenience, each different valve arrangement is shown in Table 5-3 and Figures 5-4 and 5-5 of this section. The symbols on these figures are identified on Figure 9-1. This table lists the mode of actuation, the type of valve, its normal position and its position under Reactor Building isolation conditions. The specific system penetrations to which each of these arrangements is applied is also presented. It may be noted that only motor-operated or check valves are used inside the Reactor Building to eliminate the need for additional pneumatic lines and associated isolation valves for the actuating air associated with air-operated valves.
Each valve will be tested periodically during normal operation or during shutdown conditions to insure its operability when needed.
The accident analysis for failure of malfunction of each valve is presented with the respective system evaluation of which that valve is a part, e.g.
chemical addition and sampling system, etc.
There is sufficient redundancy in the instrumentation circuits of the engineered safeguards protective system to minimize the possibility of inadvertent tripping of the isolation system. Further discussion of this redundancy and the instrumentation signals which trip the isolation system is presented in Section 7.
The system abbreviations which are used in column three of Table 5-3 are defined as follows:
HP High Pressure Injection and Purification System LP Low Pressure Injection and Decay Heat Removal Sys;em RB Component Cooling System (Reactor Building)
SF Spent Fuel Cooling System WD Waste Disposal System CA Chemical Addition and Sampling System RBS Reactor Building Spray System LPSW Low Pressure Service Water 1
e 9
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l 5-16 (Revised 6-16-67)
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TABLE 5 REACTOR BUILDING ISOLATION VALVE INFORMATION Valve Location Position Pene- Feferred Method Normal with Post tration Flew Valve to Valve Line of Valve Power Position Accident No. Service System Direction Arrgt. R. B. Type Size Actuation Signal Position Failure Indication Position 1 Pressuriser CA Out 10 Inside Gate 3/h" DIO* ES Closed As is Yes Closed and Reactor Cate 3/h" EMO ES Closed As is Yes Closed Coolant Sample Lines Outside Gate 3/h" Air ES Closed Closed Yes Closed 2 Steam Gen. CA Out 5 Inside Gate 3/h" EMO Remote Closed As is Yes Closed Sec. Water Manual Sample Line Gate 3/h" EMO Remote Closed As is Yes Closed Manual Outside Gate 3/h" Air ES Closed Closed Yes Closed 3 Component RD In 1 Inside Check 6" -- -- -- -- No --
Cooling Water Inlet.Line Outside Check 6" -- -- -- - No --
L Component RB Out h Inside Gate 6" EHO ES Open Closed Yes Closed y Cooling Water y Outlet Line Outside Gate 6" Air ES Open Closed Yes Closed 9 5 Reactor Bldg. WD Out b Inside Gate 2" EMO ES Closed Closed Yes Closed 4 Sup D Mn E Line Outside Gate 2" Closed Closed Air ES Yes Closed 1
7 6 Let Down Line HP Out 5 Inside Gate 1-1/2" EMO ES Open As is Yes Closed 7.f* to Purification Gate 1-1/2" ENO ES Closed As is Yes Closed O Demineralizers M Outside Gate 2-1/2" Air ES Open Closed Yes Closed 7 Reactor HP Out h Inside Gate 3" ENO ES Open As is Yes Closed Coolant Pump Seal Outside Gate 3" Air ES Open Closed Yes Closed Return Line 8 Reactor HP In 18 Inside Stop Ck 3" -- -- Open --
No Closed Coolant M Puse Seal Outside Globe 3" Air dp Throttled Closed Yes Closed d Inlet Line Globe 3" Air dp Closed Closed Yes Closed
'," Check (2) 2-1/2" -- -- - -- No --
- All valves with electric motor operators are also equipped with handwheels, t
I
Valve Location Position Pine- Referred Method Normal with Post tration Flow Valve to Valve Line of Valve Pwer Position Accident No. Service Svatem Direction Arrgt. R. B. Type Size Actuation Signal Position Failure Indication Position 9 Normal Makeup HP In 6 Inside Check 4" -- -- -- --
No Open to the Reactor Outside Cate 4" Air ES Closed Closed Yes Open Coolant System Globe 1-1/2" Air RC Throttled Open Yes Open level Globe 1-1/2" Air RC Closed Open Yes Open level Check (2) 2-1/2" -- -- -- --
No --
10 liigh IIP In 11 Inside Check L" - -- -- --
No Open Pressure Injection Outside Cate h" Air Isi Closed ts is Yes Open Line Check (2) 2-1/2" -- -- -- -
No Open 11 Fuel SF In/Out 6 Inside Special 30" - -- Closed -- - --
12 Transfer Closure Tubec Outside Gste 30" Manual --
Closed -- No Closed 13 Reactor Bldg. RBS In 7 Inside Check 6" -- -- -- --
No Open Spray Inlet Line Outside Gate 6" Air ES Closed Closed Yes Open 5 lh Reactor Bldg. RBS In 7 Inside Check b" -- -- -- -- No Open Q Spray Inlet W Line Outside Gate d" Air ES Closed Closed Yes Open i 15 Low Pressure LF In 7 Inside Check 12" -- -- -- --
No Open
, Injection and f Decay Heat Outside Gate 12" Air ES Closed Closed Yes Open
& Removal Line les Low Pressure LP In 7 Inside Check 12" -- -- -- --
No Open Injection and Decay Heat Outside Gate 12" Air ES Closed Closed Yes Open Removal Lire
$ 17 Decay Heat LP Out 9 Inside Gate 10" D'O Remote Closed As is Yes Closed g Suction Line Manual
. & Fuel Canal Gate 10" EMO Remote Closed As is Yes Closed
? Drain Line Manual m Gate lod Manual -- Closco Closed No Closed g Outside Gate 10" Air Remote Closed Closed Yes Closed g , Manual b .
O t
O O O
-- . ~ . _ _ _ . . . - . -- . . _ _
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\v) k- L2 Valve Location Position Pene- heferred hathed Nomal with Fost tration Paow Valve to Valve Line of Valve Power Position Accident
!:o. Service System Direction Arret, rt . B. Tyre Size Actuation Signal Pcsttion Failure Indicotton Position 18 f.uench Tank WD Out 16 Outside Gate 2" E Closed Air Closed Yes Closed Vent Line 19 Reactor Hldg. In 12 Inside butterfly 36" D.O ES Closed Closed Inlet Purge Yes Closed Line Outside Putterfly je," Air FS Closed Closed Yes Closed 20 Reactor Bldg. Out 13 ~. side Butterfly 36" Ds0 F3 Closed Closed Yes Closed Outlet Purce Line outside Butterfly 36" Air F3 Closed Closed Yes Closed 21 Heactor LISW In and lh Outside Stm Ck. 10" Manual Open Open 22 Coolant No Closed Cut Pump Outside Gate 10" Air FS Open Closed Yes Closed Motors and Lute oil Coolers 23 Ncactor Bldg. LPSW In an! lh Inside Stop Ck. 8" Manual --
Open Open No Closed m 2h Normal Air Out g Coolers Outside Gate b* Air ES Ol en Closed Yes Closed 9 25 Feedwater In and 15 Outside Swing Ck. 20" -- --
Open Open No Closed
- J 26 and Out Steam Lines. Outside Turb.Stop 30" Hydr. Open Closed ES Yes Closed a- 27 ~ Feedwater In and 15 Outside Swine Ck. 20" -- --
Open Open Closed a 23 and Out No Steam Lines outside Turb.Stop 30" Hydr. 5 Open Closed Yes Closed 29 fteactor Coolant WD Out h Inside Gate 2-1/2" EMO ES Closed Closed Yes Closed System Drain Line outside Cate 2-1/2" Air Closed Closed ES Yes Closed 30 Reactor Bldg. LPSW In 2 Outside Gate 6" Air Remote Open Closed Yes Fhergency Ogen 31 P.nual
% 32 Cooler Inlet g Line
. 8%, 33 Reactor Bldg. LPSW Out 3 Outside Globe 6" EMO ES Throttled Closed Yes Open Y' 3h EmerEency
'd 35 Cooler outlet g .Line
'36 Reactor Bldg. LP Gut 17 Outside Gate 16" Air heete Closed Closed Yes Open S 37 Sump Recire. hanual Line i _ _ _ _ - _ _ _ _ -_______- ___ __ __ - - - - _ - - - - - -
l 5.3 VENTIIATION SYS*EM 5.3.1 DESIGN BASES 5.3.1.1 Governing conditions The Reactor Building normal ventilation system is composed of the normal cool-ing system and the purge system and accomplishes two functions. One function is the removal of normal heat loss from equipment and piping in the Reactor Building, and the other is to purge the Reactor Building with fresh air when-ever desired. .
Rc4 eron Schowc The'demergency cooling units in th: "- ete ;;ild' ; ;;d :: ;: :rt ;;;1in. --
_ _,. ... are completely separate from the Reactor Building's normal ventilation system and are described in Section 6, " Engineered Safeguards."
5.3.1.2 Sizing To provide for access to the Reactor Building, the normal ventilation system will be sized to control the interior air temperature to 104 F in accessible areas during operation and a minimum of 60 F during shutdown.
The purge system equipment will be sized to provide a minimum of one-third air change per hour in the Reactor Building. The normal cooling units will be utilized and sized to distribute adequate air over and around all heat producing or releasing equipment, i
5.3.2 SYSTEM DESIGN A flow diagram of the normal ventilation and purge systems is shown in Figure 5-6.
The normal cooling system will consist of fan-cooler units located throughout the building outside the secondary shielding. These will recirculate and cool the Reactor Building atmosphere. The coolers will use low pressure service water as the heat removal medium. The fan units will discharge the cooled air through ducts to provide adequate distribution for the equipment and areas including the control rod drives. In addition, a circulating fan unit will discharge air into the reactor cavity.
The purge system will consist of a supply fan with a heater and filters and a discharge fan-filter unit. All of the purge system, except interior ducts, and two isolation valves will be located outside the Reactor Building. Ducts will be provided inside the Reactor Building for adequate distribution. To provide spare or standby ability, the fan units will be manifolded between Units 1 and 2.
The purge system discharge to the station vent will be nonitored and alarmed to prevent release exceeding acceptable limits.
5.3.2.1 Isolation valves As the normal cooling system is contained completely within the Reactor Build-ing, it will not include provisions for any isolation valves other than on --
5-20 l4
- + . , - - - - - - - , - - ,
y g -- ---y--rer
cooling water lines. The purge system will be provided with double automatic _
isolation valves (or dampers) in both the supply and discharge ducts. These valves will be normally closed and will be opened only for the purging opera-tion. They will be electrically actuated inside the Reactor Built'.ing and pneumatically actuated outside the Reactor Building. Refer to 6.5 for a discussion of the penetration room ventilation system.
The isolation signal and controls are discussed in 5.2. The closure times and sequence will be developed during detailed design and safety analyses.
Operability testing of the isolation valves is accompliahed each time the purge system is put into operation.
5.4 LEAKAGE MONITORING SYSTEM No continuous leakage monitoring system will be provided.
The barrier to leakage in the Reactor Building is the one-quarter inch steel liner plate. All penetrations of whatever type are continuously welded to the liner plate before the concrete in which they are embedded is placed.
These penetrations, shown on Figures 5-2 and 5-3,become an integral part of the liner and are so designed, installed and tested.
The steel liner plate is securely attached to the prestressed concrete Re-actor Building and is an integral part of this structure. This Reactor Build-ing is conservatively designed and rigorously analyzed for the extreme loading conditions of a highly improbab's hypothetical accident, as well as for all other types of loading conditions which could be experienced. Thorough con-trol will be maintained over the quality of all materials and workmanship dur-ing all stages of fabrication and erection of the liner plate and penetrations and during construction of the entire Reactor Building.
The comprehensive program for preoperational testing, inspection and post-operational surveillance is described in detail in 5 6, " Tests and Inspection,"
in the Appendices referenced therein and is summarized in the following para-graphs.
During construction, the entire length of every seam weld in the liner plate is leak tested. Individual penetration assemblies are shop tested. Welded connections between penetration assemblies and the liner plate are indivicual-ly leak tested after installation. Following completion of construction, the entire Reactor Building, the liner and all its penetrations are tested at l
115 per cent of the design pressure to establish structural integrity. The initial leak rate test of the entire Reactor Building is conducted at 100 per cent of the design pressure and at successively lower pressures to demon-strate vapor tightness and to establish a reference for periodic leak testing for the life of the station. Multiple and redundant systems based on dif ferent engineering principles are provided as described in 6, " Engineered Safeguards," to provide a very high degree of assurance that the accident con-ditions will never be exceeded and that the vapor barrier of the containment will never be jeopardized.
Under all normal operating conditions and under accidental conditions short of the worst loss-of-coolant accident, virtually no possibility exists that eny leakage could occur or that the integrity of the vapor barrier could be m 5-21 'I 5
1 O
violated in any way that would be significant to the public health and safety or to that of the station personnel. ~ Adequate admir.lstrative controls will be enforced to minimize the possibility of human error. Station operators will be trained and licensed in accordance with regulations. Safety analyses are presented in Section 14.
Penetrations such as the permanent equipment and personnel access hatches cannot be opened except by deliberate action and are interlocked and alarmed by fail-safe devices such that the Reactor Building will not be breached un-intentionally. The liner plate over the foundation slab is protected by cover concrete. Wherever access to the liner plate is blocked by interior concrete, means will be provided so that weld seams can be tested for leakage. The liner plate will be protected against corrosion by suitable coatings and-by c d.d ic p ete: Men. Walls and floors for biological and missile shielding, and for access and operating purposes, also provide compartmentation which constitutes protection for the liner during operating as well as accident conditions .
l Once the adequacy of the liner has been established initially, there is no i reason to anticipate progressive deterioration during the life of the station '
which would reduce the effectiveness of the liner as a vapor barrier. Inside the Reactor Building, the atmosphere is subject to a high de ree of tempera-ture control. The outside of the liner is protected by 3p1 feet of pre-stressed concrete which is exceptionally resistant to all weather conditions. l Inspection on a periodic basis, as necessary, will be conducted in all spaces b' accessible under full power operation. Biological shielding is provided to reduce radiation to limits which make occupancy of spaces adjacent to the liner !
permissible with a frequency exceeding that of most previous stations. An in-tensive visual inspection of the Reactor Building inside and cutside will be conducted at every regularly scheduled shutdown for fuel replenishment.
i All penetrations except the following are grouped and a penetration room is
. located at each group. Any leakage that might occur from these penetra ions will be collected and exhausted through the station vent, as described in 6.5.
In this manner, leakage which might occur from these groups of penetrations will be isolated from leakage which might occur through the containment vessel itself.
l
- 1. Permanent Equipment Hatch
- 2. Personnel Access Hatch
- 3. Emergency Personnel Exit Hatch
- 4. Main Steam Lines mpT A r W m =a n < re m I , ,, o Provisions will be leakage during madeoperation.
normal so that these penetrations TNSF8 M E #""# mayMFbe pressure
- f M**._ tested #/ Mfor f# N"#SS
< cPF.Noneg n -n. ko Tes sin =a BtreatN Tnest D os'a
- d. S M S IS (*N " E'ffD /* Ing /%rMc7R4TMM Within the penetration rooms, provisions will be made such that individual penetration assemblies with resilient seals and expansion bellows may be pressure tested for leakage during normal operation. This degree of control j over leakage through penetrations greatly reduces the probability of unde-l tected leakage for the Reactor Building as a whole.
5-22
A large exclusion radius of one mile has been established and property is owned by the applicant except as noted in 2.2.2.
~
Should there be any indications of abnormal leakage, individual major pene-trations or groups of penetrations will be tested by means of permanently installed pressure connections or temporarily installed pressure or vacuum boxes. If necessary, liner plate weld seams will be tested by the vacuum box soap bubble method where accessible, or by means of the permanently in-stalled backup channels where covered by concrete.
In any event, the source of cxcessive leakage will be located and such corrective action as necessary will be trken. This will consist of repair or replacement. Appropriate action will also be taken to minimize the possibility of reoccurrence of excessive leakage, including such redesign as might prove to be necessary to protect public health and safety. Leak testing will be continued until a satisfactory leak rate has again been demonstrated.
A considerable background of operating experienc e is being accumulated on containments and penetrations. Full advantage of this knowledge will be taken in all phases of design, fabrication, installation, inspection, testing and operation. Three stations with similar contrinment designs will immedi-ately precede this station. Practical improvenants in design and details will be incorporated as they are developed, where applicable.
For the foregoing reasons, it has been concluded that a continuous monitoring system is unnecessary. Since there is no such system provided, there can be no misoperation or malfunction which in itself might constitute a hazard. The steel-lined Reactor Building is self-sufficient, and other than valves and hatch doors, there are no operating parts. The containment boundary is ex-tended only by listed penetrations and further described and tabulated in 5.2,
" Isolation System" and '5.3, " Ventilation Sys tem. "
5.5 SYSTEM DESIGN EVALUATION This containment system is not dependent upon the operation of a system such as a continuous leakage monitoring system for the entire containment, or a continuous leakage surveillance system for containment penetrations and seals, since neither of these systems is being furnished. Therefore, no analyses of the capability of these systems are necessary.
The penetration room ventilation system described in 6.5 provides a partial double containment system and is an additional engineered safeguard.
A full evaluation of the containment system which is provided is included in 5.4, " Leakage Monitoring System," in justification of the omission of such a monitoring system. The Reactor Building with the appurtenant engineered safeguards systems will prevent uncontrolled release of radioactivity to plant and surrounding areas during normal operating and accident conditions, as well as for lesser accidental conditions. Containment integrity is main-tained whenever, simultancously, the reactor coolant system is pressurized above 300 psig, when the reactor coolant temperature is 200 F or above and when there is nuclear fuel in the core.
J -
5-23 ,....
f I
s 5.6 TESTS AND INSPECTION 5.6.1 PREOPERATIONAL TESTING AND INSPECIION 5.6.1.1 During Construction i
Test, code and cleanliness requirements will accompany each specification or purchase order for materials and equipment. Hydrostatic, leak, metallurgical, electrical and other tests to be performed by the supplying manufacturers will be enumerated in the specifications together with the requirements, if any, for test witnessing by an inspector. Fabrication and cleanliness standards, including final cleaning and sealing, will also be described together with shipping procedures. Standards and tests will be specified in accordance with applicable regulations, recognized technical society codes and current industrial practices.
Inspection will be performed in the shops of vendors and subcontractors as necessary to verify compliance with specifications.
5.6.1.1.1 Concrete Testing of concrete materials and concrete as placed is described in Appendix 5D, " Quality Control." An experienced full-time concrete inspector will con-tinuously check concrete batching and placing operations.
4
( An integral part of the Company's operations since 1924, Duke's Construction Department has built 4,500,000 kw of steam-electric and hydroelectric capacity
, in the intervening 42 years. The well-trained and highly competent concrete force has had years of successful experience in mixing, inspecting, placing and testing concrete in a wide variety of complex concrete structures, includ-ing foundations, turbine-generator pedestals, tunnel linings, hydro-scroll cases, dams, bridgec and prestressed walls. Included in the Construction De-partment's nucleus of 400 long-term employees are engineers, technicians, i
foremen and skilled tradesmen whose capabilities are well known and estab-lished, having been in Duke's continuous employ doing concrete work. An in-dependent national company has given surveillance to Duke's concrete 2xperience by analyzing concrete test and quality control data from a number of Duke jobs, and it has reported the Duke record of quality control and test variability to be better than acceptable standards.
- 5.6.1.1.2 Pres tressing Testing and inspection of all prestressing materials and special installation equipment is described in Appendix SD, " Quality Control." Full-time super-vision of the prestressing operations will be provided by an inspector ex-perienced in prestressing as well as by the aforementioned concrete inspector.
?
5.6.1.1.3 Reinforcing Steel Testing and inspection of reinforcing steel is described in Appendix SD,
" Quality Control." The concrete inspector will check the condition and O. placement of the bars in the forms for compliance with drawings and specifi-cutions, including welded splicas. --
- - 18 5-24 -----
5.6.1.1.4 Liner Plate s
Testing and inspection of the liner plate is described in, Appendix SE,
'ilner Plate Technical Specifications." Qualification of welding inspectors is described in Appendix SD, " Quality Control."
5.6.1.2 Structural Test Following completion of construction and prior to the initial fuel loading, the Reactor Building will be pressurized at 115 per cent of the design pre-I sure for one hour to establish the structural integrity of the building.
The response of the building will be compared with the calculated behavior to confirm the design by means of instrumentation described in Appendix 5F,
" Ins trumenta tion. "
5.6.1.3 Initial Leakage Test An initial leak rate test of the Reactor Building and its penetrations will be conducted at pressures of 100 per cent, 50 per cent and 25 per cent of design pressure, maintaining each pressure for a minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Values i of Reactor Building ambient dry bulb temperature and relative humidity will be recorded during the test period for correction of data as required. The leak rate test method used will be either the absolute volume or reference volume method. This test will establish the capability of the Reactor Build-ing to contain the pressure for which it was designed at a leak rate not ex-ceeding that specified in the license application. These data will be plotted to establish initial relationships between internal pressure, leak rate, external pres sure and temperature, relative humidity, etc.
5.6.2 POSTOPERATIONAL SURVEILIANCE 5.6.2.1 Leakage Monitoring Periodic leak rate tests of the Reactor Building will be conducted to verify its continued leak-tight integrity. The postoperational leak rate tests will be conducted at an internal pressure of about 15 psig, which is about 25 per cent of the design pressure. The acceptable leakage rate at this pressure will be determined from data plotted from the initial leak rate test. The frequency of the leak rate tests will be consistent with the leak rate approved for this station.
5.6.2.2 Surveillance of Structural Integrigg Duke is convinced that this prestressed, post-tensioned ccacrete structure offers continued structural integrity equal to or greater than alternative-type structures. It utilizies concrete and steel materials stressed in the mos t effective manner. To demonstrate this continued structural integrity and the consequent assurance of long term safety of the Oconee design, as well as to provide data useful in future designs, Duke proposes to give surveillance to the two elements of this design that assure its continued success:
- 1. Freedom of tendons from harmful corrosion.
2 Freedom of tendons from excessive relaxation. .
5-25
(y Q lt is not now feasible to design a lifetime surveillance program, as it can only be intelligently determined after analysis of results of the initial surveillance steps what the subsequent program should be. Since the design and field quality control of the two Reactor Buildings will insure that they are essential duplicates, successful surveillance test results in either build-ing will be considered applicable to its twin. Test provisions will be included in both buildings to permit tests on each should tests on either. indicate such a need.
The test methods outlined below are, at this time, conceptual and may be modified as necessary during final detailed design. Three vertical tendons 120 degrees apart, three dome tendons 120 degrees apart and the three horizontal tendons of one hoop below grade will be specially equipped as necessary for surveillance testing. Additional wires will be provided initially in each of these tendons so that when wires are removed for inspec-tion, they need not be replaced. Initial tests will be made during the fifth year after operation of Unit 1.
Conceptual test methods are as follows:
- 1. Determination of Effective Prestress Hydraulic jacks will be used to take lift-off readings for all nine test tendons and to provide controlled relaxation for the
- s inspection described in Paragraph 2.
- 2. Corrosion Resistance One tendon of each vertical, horizontal and dome group of three will be relaxed and three wires 120 degrees apart will be removed i for inspection. Each wire that is removed will be carefully exam-ined by metallurgical techniques to determine the extent of corro-sion, if any, and its significance to the load carrying capacity of the structure. After the inspection, the tendons will be retensioned to the stress level measured at the lift-off reading.
Should the inspection reveal any significant corrosion ( '.tCng or loss of area), further inspection of the other two tendons in the same g roup will be made. Samples of corroded wire will be tested to failure to evaluate the effecte of any corrosion.
If excessive loss of prestress or corrosion is detected, investigations will be performed to determine their cause, and necessary corrective action will be taken to restore the functional ;ntegrity of the Reactor Building.
The frequency and scope of subsequent surveillance testing will be decided after analysis of these initial tests.
(nv) 20 -
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LEGEN D . TWO ELEMENT STRAIN GAGE ROSETTE, BUDD C09ANY DESIGNATION: C6-121-R2TC, OR EQUALo WATER-Yh I' ROOFED AND ATTAQlED TO REINFORCING BARS TO BE INSTALLED IN WET- C03 CRETE. ORI ENTATION y 0F Tile CAGE IS INDICATED BY T}iE DIRECTION OF TliE SYMBOL. SEE SUPPLEMENT NO. 1, QUESTION 10.1 FOR DETAILS. c,--o ELECTRIC RESISTANCE STRAIN GAGE, BLli DESIGNATION: AS 9-I (VALORE TYPE), OR EQUAL, ENCAPSULATED IN A BRASS ENVELOPE TO BE ATTACliED TO Tile SURFACE OF WET CONCRETE.
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ORIENTATION OF Tile CAGE IS INDICATED BY IllE SYMBOL. g CALIBRATED BARS ATTAQlED TO THE OLTISIDE SURFACE OF CONCRETE AND EXTENDING OUTWARD. A V TRANSIT WILL BE USED TO MEASURE Tile DEFLECTION. ELECTRIC RESISTANCE STRAIN CAGE, BUDD COMPANY DESIGNATION: CP-1101 EX, OR EQUAL, ATTACliED TO Tile OITISIDE SURFACE OF CONCRETE FOR MEASURING CRACK PROPAGATION, TilREE ELEMENT STRAIN ROSETTE, BUDD COMPANi DESIGNATION: C6-141D-R3Y, BLil DESIGNATION: FAR-25-12-(60)S6, OR EQUAL, ATTACilED TO EITilER Tile INSIDE OR OUTSIDE SURFACE OF THE STEEL LINER PLATE.
'NO ELEMENT STRAIN RU3ETTE, 6UDD COMPANY DESIGNATION: C6-141-R2T, BLil DESIGNATION
_l_ FAT-25-12-S6, OR EQUAL, ATTACliED TO EITiiER Tile INSIDE OR OUTSIDE SURFACE OF Tile STEEL LINER PLATE.
'^ CELLS LOCATED AT EACil FND OF 110RIZONTAL AND DOME TENDONS AND ONE END OF A VERTICAL O TENDON.
EP0XY COATING APPLIED TO Tile OUTSIDE SURFACE OF CONCRETE TO OBSERVE CRACKS. H H'N o o e e 55 BLbG. Y Vf+1 2to 15 0, 120, 90 60 30 0 39 I ! I I Y v f MI;0 0
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