ML19224B292
| ML19224B292 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/31/1976 |
| From: | Metropolitan Edison Co |
| To: | Mullinix W NRC/IE |
| References | |
| TM-0302, TM-302, ZAR-760531-1, NUDOCS 7906140407 | |
| Download: ML19224B292 (16) | |
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DOCUMENT NO:
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METROPOLITAN EDISON COMPANY.
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Wilda R. Mullinix, NRC 79061404o7 6
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SYSTEM DESCRIPTION, (Index No. 53)
CONTAINMENT SYSTEM JERSEY CENTRAL POWER & LIGHT COMPANY THREE MILE ISLAND NUCLEAR STATION UNIT NO. 2 C
Issue Date May, 1976
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Prepared by:
John Hooper Burns and Roe, Inc.
700 Kinderkamack Road 98 020
- Oradell, N.J.
07649 1
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TABLE OF CONTENTS FOR CONTAIUMENT SYSTEM Section Pace
1.0 INTRODUCTION
1 1.1 System Functions 1
1.2 Summary Description of the System i
1.3 System Design Requirements 3
2.0 DETAILED DESCRIPTION OF SYSTEM 5
2.1 Components 5
2.2 Instruments, Controls, Alarms, and 8
Protective Devices 3.0 PRINCIPAL MODES OF OPERATION _
10 3.1 Startup 10 3.2 Normal Operation 10
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3.3 Shutdown 10 3.4 Special or Infrequent Operation 11 3.5 Emergency 12 4.0 HAZARDS AND PRECAUTIONS 12 G
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CONTAINMENT SYSTEM
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l.0 E TRODUCTIC.N 1.1 S YSTEM FUNCTIONS The primary function of the Containment System is to minimize the consequences of the maximum credible accident to an acceptable limit, thereby protecting the public against potential radiation hazards.
Thus the containment System for Three Mile Island Nuclear Station - Unit 2 comprises
,, _the Containment Structure (or more commonly calle:I the Reactor Building) together with the Engineered Safety Features as controlled by the Safety Features Actuation System _
provided to mitigate the consequences of an accident.
The Reactor Buildirg also functions as a shield for N.o Nucleat Steam Supply System against aircraft-impact and c u.ar xte<7al missiles.
During full power operating conditions, the Reactor Building-functions to attenuate any radiation escaping from the primary-secondary shield complex, so that the radit. tion
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1evel outside the Reactor Building is kept within the limits set forth in 1CCFR 20.
1.2
SUMMARY
DESQRIPTICN OF THE SYSTEM l.2.1 The Reactor Buildinc The Reactor Building is a prestressed (pos t-tensioned ),___ _
reinforced concrete structure composed of a cylindrical wall and L shallow dome roof on a conventionally reinforced flat founda tion mat.
The structure provides the final biological shielding for normal operating and accident conditions.
It encloses the reactor, the reactor coolant pumps, steam generators, pressurizer, and associated piping.and other equipment.
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- Af ter analyzing various credible accident events resulting from component-failures, design deficiencies,and operating errors, the maximum credible accident is found to be one following the " Loss of Coolant Accident" (LOCA), as a resu t of double-ended pipe break in the main coolant system. 19 022 Consequences of this accident have been used as the basis for design of Containment System Components, ad the accident has been termed as " Design Basis Accident (DB A) ".
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l.2.1.1 Liner and Penetrations The inside face of the Reactor Building concrete is lined with carbon steel To' ensure a high degree of lets tightness during operating and post-accident conditions.
Nominal plate thickness for the st' eel' liner is 1/2 inch for dome and ring girder, 3/8 inch for the cylindrical wall and inch for the base.
The liner plate and the concrete shell are penetrated by various openings such as equipment access, personnel access lock and o ther mechanical and electrical
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cetietrations.
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The configuration, basic dimensions and internal ccmponents of the Reactor Building are shown in Figure 1.
1.2.1.2 Inter' - Structtire The Reactor Building interior structure comprises the following:
(see Figure 1) i the intermediate a.
The basement floor at el.
282'-6",
floor at el.
305'-0",
and the operating floor at el.
347'-6".
b.
The primary shield wall,, $ reinforced concrete structure surrounding the reactor vessel and extending upward from the Reactor Building basement floor to form the walls of fuel transfer pool.
The secondary shield wall, a reinforced concr.ete f$ructure c.
surrounding the reactor coolant equipment, including piping, pumps and steam generators.
d.
Equipment supports, pipe supports a'nd pipe whip restraints.
1.2.2
' Safety F'ea tures ~
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The engineered safety features include containment spray systems,, containment air recirculation and cooling syst ems, emergency coolant injection, core flooding and containment t
isolation system.
Containment air recirculation and cooling
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units, backed up by reactor building sprays, are provided for em &cency atmosphere cooling..
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Chemical additives contained in the spray systems will reduce the post-accident concentration of iodine and other radioactive materials in the building atmosphere.
Emergency coolant injec-tion and core flooding protect the fuel cladding by cooling the reactor core and covering it with borated water to provide sufficient negative reactivity to maintain shutdown.
These features are provided to mitigate the consequences of the postulated accident by removing heat from the reactor core and the Reactor Building atmosphere, reducing the concentration of radioactivity in the Reactor Building atmosphere, inserting negative reactivity into the reactor, and isolating the Reactor Building.
The energy releases from the LOCA and as modified by the contribution of the safety features could result in a condition not exceeding the select,d design pressure of 60 psig and peak temperature of 286 F.
1.3
. SYSTEM DESIGN REQUIREMENTS 1.3.1 The Reactor Buil ing In order to fulfill its primary function, the Reactor Building, access openings and penetrations, are required to be designed to ac'ecmiEchs'ts in~~e~x$eEof the" design accident ~ internal
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_y pres 5ure o~f 55.7 psig and a :iiaximum temperatiire 'of 286L_ without,
exceeding the design leakage rate of 0.2 percent by weight of contained atmcsphere in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Specifically the Reactor Building and its components are designed to withst ad the following conditions:
A.
Operating Condition 98 024 This conditicn includes the dead loads, the anticipated live loads, the prestress load, load induced by temperature gradient through the walls in summer and winter, in credible combination with loads due to 80 mph wind or due to the maximum expected seismic disturbances at the site, one half the Safe Shutdown Earthquake.(
SSE), having a me h :
horizontal ground acceleration of 0.06g and vertical ground acceleration of 0.04g).
The design of the Reactor Building also includes the loads due to test pressure of 69 psig and due to transient temperatures during start-up and shut-dewn conditions.
B.
Accident Conditions In the design of the Reactor Bu tiling, a pressure of 60 psig and associated transient temperatures have been considered in combination with normal operating loads and the loads developed as a esult of adverse environmental conditions.
The adverse environmental loadings in this connection are those due to a seismic disturbance of twice that assumed during the operating condition (the " Safe Shut-dcwn Earthquake (SSE) ", having a horizontal ground acceleration of 0.12g and vertical ground acceleration of 0.08g).
Moreover, a tornado loading of 360 mph has been considered in ccmbination with the normal operating loads.
In order to e-('
ensure the adequate margin of safety against the failure of the Reactor Building to fulfill its functional require-ments, the Reactor Building and its components are designed to have an ultimate capacity to withstand internal pressure of 90 psig and associated temperatures together with normal opera ':ing loads.
The following codes, standards and recommended practices provided the guidelin..s in designing the Reactor Building.
a.
Regulations for Protection from Fire or Panic-
___ Commonwealth of Pennsylvania.__
b.
Building Code Requirement for Reinforced Concrete -
ACI 318 c.
AIBC Manual of Steel Construction d.
ASME Boiler and Pressure Vessel code : Section III, t
Nuclear Vessels;Section VIII, Unfired Pressure Vessels;Section IX, Welding Qualification (Applicable Portions).
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1.3.2.
Interior Structure The intermediate and operating floors are designed for the ccmbinations of dead load and anticipated live loads - which include the weight due to Peactor Vess 21 Head (in an assigned arca), the core flooding tanks, piping and other equipment.
The primary shield wall is designed for maximum pressure differentia' of 139.5 psi (with an ultimate capacity of 210 psig) and the secondary shield wall is designed for the pressure differential of 35.25 psi (with an ultimate capacity of 45 psig) and pipe rupture load.
The refueling pool walls are designed for hydrostatic pressure.
The supports for equipment are
. designed for normal opera. ting loads, combined with rupture of
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any pipe plus-design. basi.s earthquake.- - -
2.0 DETAILED DESCRIPTION OF SYSTEM 2.1 COMPONENTS The follcwing are the major components of the Reactor o"11 ding.
2.1.1 Concrete Shell The concrete shell is adequately prestressed to eliminate tensile stresses due to membrane forces from the design loads.
Moreover, the strain levels in the materials of construction are kept within such limits, that the integrity of the liner is maintained under all credible load combinations (Section 1.3.1).
A tendon access gallery below the foundation mat is provided to facilitate 'the installation, tensioning and grouting of the tendons.
198 026 The concrete for the Reactor Building shell and foundation mat has a minimum ultimate compressive strength of 5000 psi.
The reinforcement in the Reactor Bull' ding conforms to requirements of ASTM AG15, Grade 60.
Production and delivery of the concrete was in accordance with B&R Specification 2555-20, Production and Delivery of Concrete.
The placement and testing of the concrete and the reinforcing steel, and the method of cadwelding of the reinforcing bars were in accordance with B&R Specification 2555-54, Reactor Building Concrete.
2.1.2 Prestressina System The prestressing system used for the Reactor Building is the S/H 54-5 Tendon-Anchcrage System as developed by Stressteel.
Corporation.
A tendon-cancists of 54 1/2 inch diameter strands.
Strands conform to ASTM A416, Grade 270, low relaxatien. strands as manufactured by CF&I Reebling Corporation.
The component arts of anchorage hardware consisted of:,
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Bearing plru e ASTM A516, Grade 70 2.
Splay plate AISI 1040 3.
Wedge plate ASTM A517, Grade F 4.~ Wedges.AISI-lll7 (Case hardened a fter machining)
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u Arrangement End spacing of tendons ensures the following minimum effective forces at 40 years:
Cylinder-Hoop 742 k/ft. (156 tendons)
-vertical 500k/ft. (167 tendons)
Dome
- 42 tendons in each of the three bands developing 700 K/ft. to 750 K/ft. in the meridional direction and 300 K/ft. to 750 K/ft. in the hoop direction.
The tendon is held in designated tension by means of wedges and
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wedge plates.
The prestress load is transferred to the concrete through the bearing plate.
The tendon duct is Sh inch ins ide diameter, galvanized steel, ribbed, which is semi-rigid tube except in areas of high curvature where flexible tube has been used.
The ducts are fabricated and installed to conform to the requirements of B&R specification 2555-54.
After installation and tensioning of tendons, the tendon ducts are filled with Portland cement grout. Basicity, which is c.ie inherent property of this kind of grout, prov ides a passivating environment around the prestressed steel and excludes aggressive agents thus precluding corrosion of the prestressing steel.
2.1.3 Liner and Penetrations The liner is designed to ensure that containment integrity is not violated and leakage is limited to a maximum of 0.2 percent by weight in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the design basis accident pressure of 55.7 ps ig.
The liner is anchored to the concrete shell by means of tee-shapped members (ST 3I) running vertically and meridionally along the cylinder and deme, respectively.
Liner plate material conforms to ASME SA516, Grade 55, and penetration material (including the reinforcing and face olates) conforms to ASME SA516, Grade 70 and meets the nil-ductility requirements of the ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels, Paragraph N-1211.
The liner and penetrations were fabricated and constructed according to B&R specification 2555-44, Feactor Building Liner.
The Reactor Building Penetrations Forced Air Cooling System provides cooling which limits the bwo feedwater and four main steam penetration concrete to 1SOF.
Additional details are discussed in System Description No. 35B.
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e 2.1.4 Interior Structure The primary shield wall in conjunction with secondary shield wall, will a ttenuate the rad ia t ion level from sources within the reactor vessel and the reactor coolant system, in order to allow limited access to the Reactor Building dur ing normal full power operation.
The neutron and gamma-ray heating of the pr imary shield will be dissipated to the R' actor Building cooling system.
Reactor Building cooling fans circulate a ir through ducts in and around the primary cavity, to restrict concrete temperatures in the primary shield below 150F.
Audible and visual alarms and a multipoint temperature recorder to monitor primary shield concrete temperatures are located in the Control Room.
Additional details are discussed in the system description for Reactor Building Ventilation and Purge Systems, Index No. 35A.
The material that is used for primary and secondary shield walls and the intermediate and operating floor slabs is reinforced concrete having ar effective density (including reinforcing) of at least 145 lbs/cu.ft.
The primary and secondary shield walls serve as the refueling structure and as restraints for the reactor coolant components under pipe rupture conditions.
Primary and secondary shield walls, toge ther w ith 3 '-0" precas t top slabs, over the Reactor Cavity Area, also serve as a barrier to prevent potential internal missiles from reaching
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the containment liner.
Precast slabs are removable for refueling and repair, maintenance, or removal of equipment.
2.2 INSTRUMENTS, CONTROLS, ALARMS AND PROTECTIVE DEVICES Control ins trumen ta t ion, alarms and protective devices.
e associated with tne atmospheric radiation monitoring in the Containment are described in the system description S r Radiation Monitocing System, Index No. 52.
Instrumentation, provided to detect temperatures, pressures and humidity in the containment atmosphere and including temperature monitoring of the primary shield concrete is described in Table 11 of the System Description for Reactor Building Ventilation and Purge Systems, Index No. 35A.
Monitoring of penetration pressurization is provided and is described in System Description, Index No. 26, Reactor Building Penetrations, Air and Nitrogen Pressurization.
The protective devices used to permit access to the containment during normal operation are discussed below.
Operation of tua Safety Features Actuation System as it controls the Engineered Safety Features is described in System Description No. 50.
Two personnel access openings are provided in the Containment Structure.
The openings are designed with double doors to form airlocks such that access to the interior of the Reactor Building can be made during operation without breach of containment.
Both doors for each lock open toward the interior of the Reactor Building.
Each lcck has internal and external operating wheels for the interior and exterior doors as well as means to indicate the position of the opposite door.
Doors are mechanically interlocked so that one cannot be opened unless the other is completely closed, latched and sealed.
Design of air locks also incorporates a locking device on each inside door which.is actuated when the pressure differential across the door exceeds 2 psi.
This feature is intended to arevent sudden increases or decreases in pressure in the lock, when containment pressure is high due to testing or as a result of e 198 Of0 reactor incident.
Each lock has a lighting system and a communication system.
This communication system is connected to the total plant communication system for surveillance and emergency use. ;.
3.0 PRINCIPAL MODES OF OPERATION
',3.1 i y - - STARTUP - - --- - --
'The " containment integrity" (as defined in-Section 16 of the FSAR) has to be established before startup of the plant. - The equipment access hatch cover must
- be in place and as a minimum, one of the doors of the personnel access locks must be closed.
The gate valves and/or blind flanges for each fuel transfer tube must be in closed positions.
The flanges of penetration for Incore Instrumentation Cable Replacement must be in place.
Any penetration dismounted for repairs or testing must be replaced.
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3.2 NORMAL OPERATION During normal operation, at least one door in each of the personnel locks must be in the locked position.
All access to the Reactor Building is to be on a limited basis; i.e.,
the occupancy time for an individual entering the Reactor Building is to be based on health physics survey of the Reactor Building atmosphere and the specific areas to be occupied.
(NOTE : Section 12' of the FSAR describes the Reactor Building shield design criteria and conditions of occupancy, when the plant is in full operation.)
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3.3 SHUTDOWN The equipment access ha tch cover can be removed and/or both doors of the personnel locks opened only during cold sht 1 down and when a health physics survey of the Reactor Building atmosphere has shown that it is safe to breach containment.
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However, during the handling of fuel in the Reactor Building with irradiated fuel in the reactor building, at least one door on each personnel access lock shall be closed and the equip-ment hatch cover shall be in place and secure. 3.4 Soecial or Infrecuent Ooeration 3.4.1 Post Ooerational Leakace Rate Tests Post-operational leakage rate tests will be conducted periodically to ensure that the integrity of containment as defined in paragraph 3.1 is maintained and to determine if any leakage problems have developed since the previous integrated leakage rate test. These tests will consist of integrated leakage rate tests of the containment system and individual leakage rate test of certain components which penetrate or seal the boundary of the containment ~' system. The tests shall be performed at regular intervals, after major repairs or whenever there is any indication of gross leakage. These tests are discussed in Section 16 of the FSAR and in the System Description for Leak Rate Testing System, Index No. 54. 3.4.2 Structural Proof Testine and Prestressing System The prestressing tendons in theReactor Building are grouted cnd their anchorages are protected against the influence of external corrosive agents. Individual tendons therefore, are not accessible for inspection, even at ancherages, without some degree of destruction of the protection provided. Therefore, the continual structural integrity of the Reactor Building will be ascertained by periodic pressure proof tests rather than by direct inspection. These tests are to be performed during the 2nd,[ 10th, and 20th years (approxima tely) after start of commercial 198 032 ~ - - - - - --v-
operation of tne unit. They shall be conducted at routine shutdown periods and it is acceptable tha t they be sequenced with leak rate tests. Procedure for performing _these tests including pressure level, measurements, and visual inspection shall be similar to that of initial proof test. Observa tions recorded during these tests shall be compared with those recorded during initial proof test and evaluated to assesr any change in the behavior of the structure. Details of these tests are discussed in Appendix 3D of the Final Safety Analysit Report. 3.5 Emercency The containmen system is an engineered safety feature which functions to min aize the potential hazardous consequences o f a LOCA. The stract na, along with other eng.ineered safety feature systems, provides this protection. 4.O HAZARDS AND PRECAUTIONS The following conditions are considered to be potentially hazardous and are to be reported so that an evaluation of tbie condition can be made and corrective action, if required, can be instituted. A record of the condition is to be maintained for future reference in determining further degradations. } Deterioration or cracking of shell concrete. a. ~= = , z..== w g=== =.,. _.--.m. b_:_d - ' ' k}.=.&l.. -.'? ' Y~ ~. ~. E 198 033 _ 9 s e
~ ~ b. Rus t or rust streaxs on the Reactor Building shell due to rusting of ferrous material imbedded in the Reactor Building shell concrete. c. Corrosion or peeling of paint on the Reactor Building liner and penetrations. The follcwing precautions will be observed before pressure testing the Reactor Building: a. All light bulbs sh:ll be removed. b. All smoke detectors shall be disconnected electrically. c. For the polar crane; all control panels, switch box doors, gear care inspection covers, test box lid, pendant housing, master switch covers, and similar light duty enclosures shall be opened. 198 034 _ [
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