ML19309C543

Chapter 5 to TMI-1 PSAR, Containment Sys. Includes Revisions 1-11

ML19309C543
Person / Time
Site:	Three Mile Island
Issue date:	05/01/1967
From:	JERSEY CENTRAL POWER & LIGHT CO., METROPOLITAN EDISON CO.
To:
References
NUDOCS 8004080735
Download: ML19309C543 (68)

Text

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TABLE OF CONTENTS Section M

5 CONTAINMENT SYSTEM 5-1 51 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 Mass Releases 5-1 5 1.1.3 Contributien of Engineered Safeguard System 5-2 5.1.2 STRUCTURE DESIGN 5-2 5.1.2.1 Design Conditions 5-2 5 1.2.2 Design Leakage Rate 5-3 5 1.2.3 External Loadings 5-3 5 1.2.h Codes 5k O 5.1.2 5 piwings 5-5 5 1.2.6 Penetrations 5-5 5 1.2.7 Missile Protection Features 5-6 5 1. 8 Corrosion Protection 5-9 5 1.2 9 Insulation 5-10 5 1.2.10 Shielding 5-10 5.2 ISOLATION SYSTEM 5-10 5 2.1 DE"IGN BASES 5-10 5 2.2 SYSTEM DESIGN 5-10 5.3 YENTILATION SYSTEM 5-12

, 5 3.1 DESIGN BASES 5-12 5 3.1.1 Governing Conditiens 5-12 5.3.1.2 Sizing 5-12 5_1 0001 009

5.3.2 SYSTEM DESIGN 5-13 5 3.2.1 Isolation valves 5-13 5.h LEA % AGE MONITORING SYS'"D4 5-14 55 SYSTEM CISIGN EVALUATION 5-16 5.6 TESTS AND INSPECTION 5-16 5.6.1 PREOPERATIONAL TESTING AND INSPECTION 5-16 5.6.1.1 During Construction 5-16 5.6.1.2 Structural Test 5-17 5.6.1.3 Initial Integrated Leakage Rate Test 5-17 5.6.2 POSTOPERATIONAL LEAK MONITORING 5-17 5.6.2.1 Leakage Monitoring 5-17 0

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! i LIST OF TABLES Table No. Title Page.

i 5-1 Missile Energies 5-18 4

4 i 5-2 Missile Penetrations 5-18 r t

1 5-3 Reactor Building Isolation l j Valve Information (Two Sheets) 5-19 i

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i LIST OF FIGURES Figure No. Title

, 5-1 Reactor Building Typical Details 5-2 Typical Electrical and Piping Penetrations

'3 Details of Equipnent Hatch and Personnel Lock 54 Reactor Building Isolation Valve Arrangement 5-5 Reactor Building Normal Ventilation System 5-6 Reactor Building Instrumentation O

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l 5 COTIAINMENT SYSTEM The containment for this station consists of two systems which are:

a. The Reactor Building to provide biological and missile shielding and to contain the energy and material that could be released by an accident.

b. The engineered safeguards systems which limit the maximum value of the energy released by an accident.

51 REACTOR BUILDING The Reactor Building vill be a reinforced concrete structure cmposed of cylindrical vans with a flat foundation mat and a shanov dome roof. The foundation slab vill be reinforced with conventional mild-steel reinforcing.

The cylindrical vans will be prestressed with a post-tensioning system in the vertical and horizontal directions. The dme roof vill be prestressed utilizing a three-way post tensioning system. The inside surface of the Reactor Building vill be lined with a carbon steel liner to insure a high degree of tightness during operating and accident conditions. Liner plate thickness will be 3/8 in, for the cylinder and dome and 1/h in. for the base.

I The foundation mat vill be bearing on sound rock and vill be approximately 9 ft thick with a 2 ft thick concrete slab above the bottom liner plate.

The cylinder portion vill have an inside diameter of 130 ft, vall thickness of 3 ft 6 in. , and a height of 157 ft fra top of foundation slab to the i

/. spring l'ne. The shallow dome roof vill have a large radius of no ft, a transicion radius of 20 ft 6 in. , and a thickness of 3 ft. The Reactor Building is shown in Figure 5-1.

l The experience and knowledge gained in the design at Rochester Gas and Electric Capany's Brookwood Plant No.1 (AEC Docket Ne~ 50-244), as ven as signs by others that are similar in functional requirements, vin be uti .:ed in the design of this Reactor Building.

5 1.1 DESIGN BASES 5.1.1.1 Postulated Accident Conditions The Reactor Building is to provide biological shielding for normal and accident conditions. It will enclose the reactor and the reactor coolant

, system and will be designed to insure that an acceptable upper limit of leakage of radioactive =aterial vill not be exceeded under the maximum loss-of-coolant accident as described in Section 14, " Safety Analysis."

The accident is based on a double-ended pipe break in the main coolant system and produces pressures and temperatures that are influenced by the safeguard system, heat sinks , and energy sources. This is described in Section 6, " Engineered Safeguards" and Section ik " Safety Analysis."

5.1.1.2 Energy and Mass Releases Additional energy and mass vill be available for release into the Reactor O Building frca the following sources:

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a. Stored heat in the reactor

b. Reactor decay heat llh

c. Stored heat in the reactor coolant system

d. Water-Metal reaction

e. Eydrogen combustion

f. Fission products in the core The energy released from these sources is discussed in Section ik.

The energy contribution from the secondary steam system is not included in the calculaticus for the Reactor Building design pressure and temp-eratures. The supports for the reactor coolant system vill be designed to withstand the forces associated with a break in the reactor coolant pipe so that a rupture in the secondary system vill not be considered to act simultaneously with a reactor coolant pipe break.

5 1.1 3 Contribution of Engineered Safeguard Systems The contribution of the engineered safeguard system is discussed in Section 6. These systems will be actuated to minimize the accident conditions by removing heat fran the Reactor Building and inserting negative reactivity into the reactor.

These safeguard systems will be:

a. A high pressure injection system

b. A low pressure injection system

c. A core flooding system

d. A Reactor Building emergency cooling system ,)

e. A Reactor Building spray system.

f. A Reactor Building isolation system 5.1.2 STRUCTURE DESIGN 5 1.2.1 Design Conditiens The Reactor Building vill be a Class 1 structure as defined in Appendix 5A.

The internal free volume vill be at least 2,000,000 cubic feet. It will be designed for an internal pressure of 55 psig with a coincident temperature of 281 F at accident conditions and an external pressure differential of 2.5 psi at normal working conditions. Due consideration vill be given to the dead load, live load, temperature gradients , and penetrations at accident and working conditions. The functional requirements for the liner are covered in detail in Appendix SE, " Liner Plate Specification".

The design moments , shears , deflections , membrane forces , and stresses are being developed but have not been ecmpleted. For preliminary structural concept of the Reactor Building refer to Figure 5-1. The design criteria for the Reactor Building is covered in Appendix SC, " Design Criteria for Concrete Reactor Building".

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10 V 5.1.2.2 Design Leakage Rate The Reactor Building vill be designed to limit the leakage rate to 0.2 per cent by weight in 2h hoies at the design pressure.

5.1.2 3 External Loadings 5 1.2 3 1 ,

External Pressurd The Reactor Building vill be designed for an external atmospheric pressure of 2 5 psi greater than that of the internal pressure.

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5 1.2.3.2 Tornadoes Structures vill be designed to withstand short tem tornado loadings including i tornada generated missiles where such structures house systems and components whose failure could result in an inability to safely shut devn and isolate the reactor. Structures that vill be so designed include the following: j

a. Reactor Building

b. Control, Relay, and Battery Rooms

<. Designated areas of Auxiliary Building The tornado design requirements are-I

a. l Tangential vind velocity of 300 mph.

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b. An external vacuum of 3 psig.

c. Missile equivalent to a utility pole 35 ft long, 14 inches in diameter, weighing 50 PCF and traveling at 150 mph.

d. Missile equivalent to a one-ten automooile traveling at 150 mph.

A 300 mph wind will be applied in accordance with standard vind design .

practice and utili::ing applicable pressures , shape factors , and drag coefficients l fran ASCE Paper No. 3269, " Wind Forces on Structures." The vacuum of 3 l psig is conservative considering that most measured pressure drops are in the order of magnitude of 1.5 psig.

Recognition is given to the fact that the steel superstrr.cture of the Auxiliary Building and Spent Fuel Building could be damaged. All vital equipment and component in these areas will be located below the operating floor of the Auxiliary Building. The operating floor slab at elevatien 329'-O' 6 and all exposed portic = of the Auxiliary Building belov elevation 329'-O'~

vill be designed for the loading conditions stated above.

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5-3 (Rerised 1-8-68)

The structural concrete valls and base mat of the spent fuel pit vill be designed te resist all tornado loads thereby ensuring no excessive less g of water dus to structural failure.

5.1.2 3.3 Wind, Snow, or Ice The Reactor Building vill be designed to withstand a snow or ice load of 2

35 lbs/ft and a vind velocity corresponding to a 100 year frequency wind.

The vind velocity as a function of height and drag coefficients will be established on the basis of ASCE Paper No. 3269 (Wind Forces on Structures).

5 1.2.3.4 Ground Water and Floods The top of the Reactor Building foundation mat will be approximately at elevation 280 ft. The ground water elevation vill be approximately 260 ft. The foundation slab design vill take into consideration ground water pressure. The plant site vill be protected from any flood conditions.

Fluctuations in the ground water due to flood and normal variation vill be given due consideration in designing the valls and foundation slab.

Refer to Sections 2.h, " Hydrology and Groundwater" and 2 5, " Geology."

5.1.2.3 5 Seismic conditions The site seismology and response spectra are described in the Seismology section of Appendix 23.

The seismic design of the Reactor Building vill be based on the response to a maximum horizontal component of ground acceleration of 0.06 g. In )

addition, the design will be checked to ensure no loss cf function based on the response to a maximum hr;,rizontal component of ground acceleration of .12 g. The vertical component vill be taken as 2/3 of the horizontal etmponent and will be assumed to occur si=ultaneously with the horizontal

component.

l 5.1.2.h codes The Reactor Building vill be designed under the following codes:

a. Regulations for Protection Frcm Fire and Panic - Ccmmonwealth l of Pennsylvania I b. Building Code Requirements for Reinforced Concrete (ACI 318-63)

c. AISC Manual of Steel Construction

d. ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessel, Sections VIII, Unfired Pressure Vessels ,Section IX, Welding Qualifications (Applicable Portions ).

For additional design criteria see Appendix 5A " Structural Design Bases."

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5.1.2.5 Drawings Figures 1-2 through 1-10 and Figure 5-1 are plans and elevations showing principal dimensions of the Reactor Building.

5 1.2.6 Penetrations 5.1.2.6.1 Piping, Duct, and Electrical Penetrations Penetrations for process piping, ventilation ducts , fuel transfer, instrumentation lines, and electrical cables vill be designed to withstand the following leads:

a. Incident pressure and temperature including response of the Reactor Building shell.

b. Pipe reactions based en themal flexibility and seismic loads. -

c. Expansion of containment shell under test conditions.

d. Pipe thrust loads to ensure the vapor barrier is not breeched due to a primary system pipe rupture which might result in subsequent failure of the primary coolant system.

All pipe penetrations will normally be anchored at the Reactor Building shell. Penetrations vill be =ade of =aterial which exhibits by test a transition temperature at least 30 F below the minimum service metal temperature.

'. Typical piping, duct, and electrical peneteations are shown in Figure 5-2. 6 All penetrations vill be of the double bar.aier type. Where temperature changes require, the second barrier of piping penetrations vill include expansion bellows. All penetrations c e designed to provide a captive air space that can be pressurized to the Reactor Building design pressure for leak testing and accident conditions. All piping systems that may be open to reactor building atmosphere vill be designed for adding a fluid block.

5 1.2.6.2 Personnel Access Locks Two personnel access locks are provided, one of which penetrates the dished head of the equipment hatch. Each personnel hatch is a velded steel assembly with double doors equipped with double gaskets to provide l an air space that can be pressurized to the Reactor Building design pressure for leak testing and accident conditionn. The doors are interlocking to ensure that only one door is opened at a time. Remote, indicating lights and annunciators vill be provided in the Control Room to indicate if a door is open.

The personnel locks will be designed and fabricated so as to comply in all respects with the requira.ments of Section III of the ASME Boiler and Pressure Vessel Code for Class 3 vessels.

5.1.2.6.3 Equipment Access Hatch An equipment hatch with an inside diameter of 22 ft k in. is provided to g1 enable, passage of large equipment and components into the Reactor Building 5-5 (Revised 1-8-68) 0001 017

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during a plant shutdevn. 'Ihe iters brought into the vessel include in part the reactor coolant pu=ps and motois , and reactor vessel 0-rings .

5 1.2.7 Missile Protection Features 5.1.2.7.1 Reactor Building l

Missile protection for the Reactor Building liner: I

a. The building and liner vill be protected frem loss of function due to damage by such missiles as might be generated in a loss-of-coolant accident for break sizes up to and including the double-ended severance of a =ain coolant pipe.

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b. The engineered safeguard ecmponents required to =aintain con-tainment integrity and to =eet the site criteria of 10 CFR 100 vill be protected against loss of function due to danage by the missiles defined below.

During the detailed plant design, the missile protection necessary to meet the above criteria vill be developed and Lnplemented using the folicwing considerations :

a. The reactor coolant system vill be surrounded by reinforced cencrete and steel structures designed to withstand forces associated with double-ended rupture of a main coolant pipe and any missiles that may be generated.

b. The structural design of the missile shielding vill take into '

account both static and impact loads and vill be based upon a barrier cross section with energy sbsorption capacity at least 25 per cent greater than that required when considering a po-tential =issile,

c. Components of the reactor coolant systen vill be examined to l identify and classify potential missiles according to size, shape, kinetic energy, and driving force for purposes of analyzing

their effects.

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d. The types of missiles for which =issile protection vill be provided are:

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1. All valve stems up to and including the largest sise to be used.

2. All valve bonnets

3. All instru=ent thi=bles

h. Various sizes of nuts and bolts

5. Reactor vessel head bolts e

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I v 5 1.2.7.2 Main Steam Turbine Missiles The turbine-generator supplier has =ade a study of various failures of turbine-generator rotating elements and has concluded that they are of two general types: (a) Failure of rotating emponents operating at or near nomal operating steed. (b) Failuz a of ecmponents that control admission of steam to the turbine resulting in destructive shaft rotating speed.

ta. Turbine Rotor Failure at or Near Operating Speed Turbine and generator rotor failures at or near rated speed have resulted from the cabination of severe strain concentrations in relatively brittle materials. To minimize the probability of brittle fracture in rotors , wheels , and shafts , new alloys , and pro-cesses have been developed and adopted. Careful control of chemistry and detailed heat treating cycles has greatly improved the mechanical properties of all of these cmponents. Transition temperatures (the  ;

temperature at which the character of the fracture in the steel changes l fra brittle to ductile, often referred to as FATT) has been reduced '

on the lov temperature wheel and rotor applications for nuclear units to well below start-up temperatures. Improved steel mill practices in vacuum pouring and alloy addition have resulted in forgings which are more uniform and defect free than ever before. More emprehensive ]

vendor and manufacturer tests involving improved ultrasonic and magnetic l particle testing tec%iques are better able to discover surface and ,

internal defects than in the past. Laboratory investigation has revealed I scue 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 loadir4

3. Bes.ter guidance for station operators in the control speed, acceleration, and loading rates to minimize rotor stresses.

Progress in design, better material and quality control, more rigorous acceptance criteria, and improved nachine operation have substantially reduced the likelihood of burst failures of turbine-generator rotors operatica near rated speed.

b. Turbine-Generator Overspeed Failure The improvements of roter quality discussed above, while reducing the chance of failures at operating speed, tend to increase the hazard level O , associated with unlimited overspeed because of the greater missile energy associated with higher bursting speed. Thus, the turbine overspeed pro-tection systems will be as follows:

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1. Main and secondary steam inlets vill have the following valves in series:

a. Control valves - controlled by the speed governor and tripped closed by emergency governor and back-up overcpeed trip

b. Stop valves or trip throttla valve - actuated by ,he emergency governor and back-up overspeed trip. Energency =ain stop valves of the stem sealed design have been used on General Electric steam turbines 10,000 kw and larger since 1948. There have been over 650 turbines shipped and put in service during this period and there has been no report of the main steam stop valve failing to close when required to protect the turbine. Impending sticking has been disclosed by =eans of the daily full closed test feature so that a planned shutdown could be made to make the necessary l correction. This almost always involves the re= oval of the oxide layer which builds up on the stem and bushing and which would not occur on a lov temperature nuclear application,

c. Cembined stop and intercept valves in cross around systems where required to centrol overspeed to the valves mentioned above.

l These are actuated by the speed governor, emergency and backup 1 overspeed trips. These valves also include the daily testing features described above.

2. Uncontrolled Extraction lines to Feedvater Heaters l

If the energy stored in an uncontrolled extraction line is sufficient to cause dangerous overspeed, the positive closing non-return valves I are provided, to be actuated by the emergency governor and back-up

! overspeed trip. The.se are designed foi remote manual periodic tests l to insure proper operation. The station piping, heater, and check l 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 both to obtain design l information and to confirm proper operation. Such tests include the l capability of centrols to prevent excessive overspeed on loss of load.

l Careful analysis of all past failures have led tn design, inspection, and +,esting procedures to substantially eliminate destructive over-speed as a possible cause of failure in =odern design units.

The turbine-generator supplier has made a study of the major missiles 6 that might escape the turbine-generator exciter housing as a result of a hypothetic failure. The last stage wheel is censidered to have the voret ecebination of ucight , size, and energy. The latest analysis indicates that the last stage wheel could fail at an overspeed of 169 percent of the rated speed. Properties and penetration into the Reactor Building for the last stage wheel missiles are shown in Table 13.2-1. (Sucplement 3) The maximu= penetration vill occur with the 120 degree segment as shown in rigure 13.2-1. (Supplement 3)

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A ccmparison of the penetration depths of the 120 degree segment at a 69 6 L percent overspeed, as stated above, and the 13h deg;ee segment at 'an 86 percent overspeed (as shown in section 5.1.2.7.2 of the PSAR) indicates that the difference in penetration depths is negligible and that the difference in percent of overspeed does not apprecially influence the penetration depth. For example, penetration in concrete is 12.8 in. at 69 eercent overspeed as opposed to 13.2 in. at 86 percent overspeed.

Missile penetrations when equal to or less than one-talf the shield thickness vill not produce a breech of the shield due to' a local material failure. In the case of the Reactor Building this =eans that

• penetrations less than 1.75 feet vill not breech the vapor barrier.

5.1.2.8 Corrosion Peetection 1 The centainment structure is protected against external corrosive influences by the following means:

1. A vater proof tendon access gallc,y at the base of the cylinder, as shown en Figure 5-1.

2. A retaining vall around the vessel which precludes contact af ground water with the shell.

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3 A cover of concrete in excess of that for normal construction as I exemplified by code requirements. (Refer to the answer to l Question 7.12.1). I O' h. Use of a water-tight steel condu'i t for tendons with added precaution of a thicker valled, rigid confuit in the base mat and extending immediately above into the cylinder.

5. An inboard-oriented haunch which results in only nominal tensile stresses of the outer fibers; these stresses are within the caeacity of concrete in tension.

The exposed surface of the liner vill be given a protective ccating of Carbonzine #11 Gray, as manufactured by the Carboline Company. The outer surface of the, steel vill be in direct contact with the concrete, which provides adecuate corrosion protection because of the alkaline properties of the concrete.

The tendons vill be inserted in steel conduit embedded in the concrete which will provide an additional water barrier, as well as an electrical shield against stray currents. The inner surface of the steel conduit, as well as the tendons including anchorages, will be protected with NO-0X-ID "CY Casing Filler, as manufactured by the Dearborn Chemical Division of

• J.R. Grace and Company. This material is ecmposed essentially of a selected paraffin-base refined mineral oil, blended with a microcrystalline petroleum-derived base (petrolatum) which has a definite melting point and penetration range. Additives consisting of lanolin, and sodium petroleum sulphonates are incorporated as water-displacing, surface-active agents and corrosion inhibitors.

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The proportion of oil to microcrystalline wax in the formulation is 3 adjusted to give a pour or gelling point within the range of 110 to 120 F. llh The oil and wax are highly refined long-chain, saturated, paraffinic, petroleum derivatives, resistant to oxidation and chemical or physical degradation within the temperature ranges to which they will be exposed in this service. The lanolin is a polar substance which enhances inhibitor performance and wetting of the metal surface by the microwax blend. The petroleum sulphonate is a surf ace-active, water displacing corrosion inhibitor of long tested merit. The physicai properties and tests per-formed on this material are more complately described in the Fourth Supplement to the Preliminary Facility Description and Safety Analysis Re port for Brookwood Nuclear Station Unit No. 1 (Docket No. 50-244) . The end anchorages will be covered with a' metal container and provision made to control the humidity .to a safe limit so that the temperature of the air within the enclosure is maintained above the dew point for all operating conditions.

The retaining wall and drainage system around the Reactor Building provides excellent protection for the liner and tendons against ground corrosion,and therefore no cathodic protection system will be provided.

All metallic components including the liner plate and tendon conduit will be electrically connected to prevent stray current corrosion. The tendons will be enclosed in a metallic tube so as to isolate them from outside electrical influences.

Permanent reference electrode stations will be installed to f acilitate measurements of structure potentials. These stations will consist of -

plastic pipe extending from the ground surface to the point at which the j structure potential measurement is required. The plastic pipe will function as a salt bridge. Standard reference half cells placed into the pipe,down to the ground water levci, will be used to make structure potential measurements.

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5.1.2.9 Insulation The Reactor Building liner place will not be insulated. Insulation and/or cooling coils will be provided when required for hot line penetrations through the Reactor Building so as to limit concrete temperatures during normal operation to no greater than 150 F.

5.1.2.10 Shielding The Reactor Building completely encloses the reactor coolant system and is designed to contain all radioactive material which might be released following a loss of coolant accident. The building provides adequate biological shielding for accident and normal operating conditions. The interior walls provide adequate shielding for limited access during normal operating conditions.

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-conse-quence-limiting systems is to be minimized by a double barrier so that no single, credible failure or malfunction of an active com-ponent can result in loss of isolation or intolerable leakage.

The installed double barriers take the form of closed piping sys-() tems, both inside and outside the Reactor Building, and various types of isolation valves.

Rea; tor Building isolation occurs on a signal of approximately 4 psig in the Reactor Building. Valves that isolate penetrations that are direct-ly 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 op-eration of the engineered safeguards in order to prevent leakage of radioactive materials to the environment. Fluid penetrations serving engineered safeguards also meet this design basis.

All remotely operated Reactor Building isolation valves are provided with position lbsit indicators in the control room.

5.2.2 SYSTEM DESIGN The fluid penetrations that require isolation af ter an accident may be classed as follows:

Type I. Each line connecting directly to the reactor coolant system has two Reactor Building isolation valves. One valve is ex-ternal and the other is internal to the Reactor Building.

pg These valves may be either a check valve and a remotely oper -

! (,) ated valve or two remotely operated valves depending on the I

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Type II. Each line connecting directly to the Reactt"- Building atmo-sphere has two isolation valves. At least one valve is exter-nal and the other may be internal or external to the Reactor Building. These valves may be either a check valve and a re-motely operated valve or two r-motely operated valves depend-ing on the direction of nomal flow.

Type III. Each line not directly connected to the reactor coolant system or not open to the Reactor Buildirg 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 that penetrate the Reactor Building and are connected to either the building or the reactor coolant system, but which are never opened during reactor oceration, have provisions for locking in a closed position.

There are additional subdivisions in each of these =ajor groups. The indi-vidual 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-k 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 nomal position, and its position under Reactor Building isolation conditions. The specific system penetrations to which each of these arrangements is applied are also presented. Each valve vill be tested periodically during nomal operation or during shutdown conditions to insure its operability when needed.

The accident analysis for failure or malfunction of each valve is presented with the respective system evaluation of which that valve is a part, for example, chemical additien and sampling system, etc.

There is sufficient redundancy in the instrumentation circuits of the safe-guards actuation system to minimi::e the possibility of inadvertent tripping of the isolation system. This redundancy and the instrumentation signals that trip the isolation system are discussed further in Section 7.

The system abbreviations in column three of Table 5-3 are defined as follows :

MU Makeup and Purification System DH Decay Heat Removal System RB Reactor Building Cooling System SF Spent Fuel Cooling System WD Waste Disposal System CA Chemical Addition and Sampling System .

RSS Reactor Building Spray System '

IC Intemediate Cooling System SW Nuclear Services Cooling Water System RW River W.r.cer SP Secondary Plant l

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5.3 VE::TILATION SYSTE 5.3.1 DESIGN BASES 5.3.1.1 Governine conditions The Reactor Building ventilaticn system perfoms the following functions:.

a. Removes or adds sensible heat under nor=al conditions of operation to maintain the building at or below a predetemined maxi =u= temp-erature and at or above a predetemined minimum temperature.

b. Removes sensible and latent heat under emergency accident conditions to maintain the building at or below predetemined maximum values of temperature and pressure.

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c. Deleted

d. Filters all air.

e. Filters contaminated air throu6h charcoal filters before discharging it to atmosphere,

f. Purges the Reactor Building with outside air whenever desired.

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5 3.1.2 sizine To provide nomal cooling in the Reactor Building, the ventilation system s will be sized to centrol the interior air temperature to 110 F maximum in accessible areas during operation and 60 F minimum during shutdown.

Air frem the system vill be distributed over and around all heat producing equipment. he anticipated total nomal c.,oling load is h.3 x 106 Btu /hr. l6 l To provide for e=ergency accident cooling in the Reactor Building, the ventilation system will be additionally selected and sized to meet the abnomal conditions of te=perature, pressure, and latent heat which vill be be vill i= posed 281 F, upon it isand 55 psis, such anxevent.6 240 10 It is anticipated that these conditions l Stu/hr. 1 Deleted To provide a flcw of outside air through the Reactor Building, the purge ccmponents of the ventilation system will be selected and sized for a purge rate of 50,000 cfm. All air entering the Reactor Building vill be filtered and heated as required. All air leaving vill be passed through HEPA and charcoal filters before being discharged to the atmosphere through 1 the vent.

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The cooling which is required during normal operation vill be obtained 1 frem vater in a closed circuit which is cooled in'an evaporative-type industrial cooler to a temperature of 85 F maxi =um.

"'he cooling which is required during an emergency condition vill be obtained from vater frem the nuclear services cooling water system which vill be available at a maximum temperature of 95 F. l1 The nornal cooling water circuit will be entirely independent of the emergency cooling vater circuit.

5.3.2 SYSTDi DESIGN Figure 5-5 is a flow diagram of the ventilation system for the Reactor Building showing the cooling and purging components.

The cooling components consist of three fan assemblies, each of which will '6 contain a normal coil, emergency coil, filter, direct-driven fan, and suit-able casing. To insure adequate air distribution to the major head loads, auxiliary fans and ducts vill be employed.

DELETED The purge. exhaust air component consists of two fan assemblies each of which vill contain a HEPA filter, a charcoal filter, fan, and casing.

Each assembly vill handle one half the total air quantity (25,000 cfm each).

The purge exhaust air component consists of two fan assemblies each of which vill contain a HEPA filter, a charcoal filter, fan, ar.1 casing.

Each assembly vill handle one half the total air quantity (25,000 cfm each).

All of the purge co=ponents of the ventilation system, except the interior ducts and isolation valves as shown en Tigure 5-5, vill be located outside the Reactor Building.

The purge discharge to the station vent vill be monitored and alarmed to prevent release exceeding acceptable limits.

5.3.2.1 Isolation valves As the cooling component of the ventilation system is contained completely within the Reactor Building, it will not include provisions for any isolation valves other than on cooling water lines.

1

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' I't! 5-13 (Revised 1-8-68) 0%1 DN

I l

The purge ecmponent of the ventilation system vill be provided with double auto =atic isolation valves (or dampers) in both the supply and discharge llh ducts. These valves vill be normally closed and vill be opened only for ',

the purging operation. They vill be electrically actuated inside the Reactor Euilding and pneu=atically actuated outside.

The isolation signal and controls are discussed in section 5.2. The closure ti=es and sequence vill be developed during detailed design and safety analyses.

Operability testing of the isolation valves is acecmplished each time the purge system is put into operation.

5.h LEAKAGE MONITORING SYSTEM Provision for monitoring of leakage throu6h the penetrations and penetration sleeve to liner velds will be provided.

The ner=al barrier to leakage frem the Reactor Building vill be a steel liner three-eighth of an inch thick in the cylinder and dame and one-quarter of an inch thick in the base. All penetrations through the containment shell for pipes , ducts , electrical conductors , and access will be velded to the liner plate before the concrete is placed. Insofar as possible the penetrations and reinforcing plates vill be shop velded assemblies. All penetrations in the Reactor Building vill be of the double barrier type and will be equipped for continuous leakage monitoring with means provided to isolate and locate a leaking penetration. All penetrations as shown typically in Figure 5-2 vill be designed, fabricated, and tested so as to ensure leak tightness. All penetration sleeve to liner plate velds will be covered by a test channel to permit verification of leak tightness.

The liner plate vill be anchored to the concrete shell so as to ensure elastic stability under all leading conditions and ccmposite action between the liner and the concrete. Elastic stability will be ascertained by analy ing the liner as a flat plate between supports subjected to the biaxial stresses. Should liner stresses under the factored load conditions exceed yielding ccmpression, the analysis will assu=e plastic behavior at a stress of 1.2 times the minimum guaranteed yield stress. The Reactor Building vill be tested for leak tightness under conditions more severe for possible liner l'eakage than under the conditions of the hypothetical accident and other simultaneously occurring loads. Surveillance of the quality of all materials and work =anship will be maintained so as to ensure that the containment shell including the liner, penetrations , and reinforced concrete does meet the intent of the design in all respects.

The details of a program to demonstrate integrity of the structure and the leah tightness of the vapor barrier are described in Section 5.6 and su=narized

( as follows :

( a. The test to demonstrate structural integrity vill be at 115 per cent of design pressure. This test is established to duplicate, insofar i

as possible, the incident leads plus the most severe effects of the design earthquake.

O w, .

~

,_1, 0001 027

b. The integrated leak rate test will be at design pressure and will demonstrate that the leak rate is no greater than 0.2 per cent of the contained volume in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the design pressure. The conditions for this test vill impose greater tensile or lower em-pressive membrane stresses on the liner than would exist under accident conditions, thereby ensuring that the test is performed under the most severe condition insofar as leakage is concerned.

The pressure load in cabination with the thermal load results in the lesser probability of leakage. Leak rate measurements will also be made at one or more lower pressures (i.e. below design pressure) to demonstrate leak tightness and to establish a reference for future testing.

Under all normal operating conditions and even including the most severe less-of-coolant accident, there is virtually no possibility that leakage would occur th.t would be a hazard to the public or to the station per-sonnel. The emplete safety analyses are presented in Section 14.

As described in Section 51.2.6, the personnel access hatches will be interlocked so as to ensure one barrier vill always remain intact and that indicators and annunciators will be provided in the Control Rom to indicate when a door is open. The permanent equipment access hatch cannot be opened except by extensive deliberate action which cannot be taken except at plant shut down.

All liner seams will be tested individually to demonstrate their leak tightness . Where the liner vill be inaccessible for future inspection (i.e. where it will be totally encased in concrete), test channels will s' O be provided on all veld sesms. These channels vill ensure leak tight-ness of the plate to plate veld and the channel to plate veld and will be segmented so as to permit pressurization at specified locations. On liner plate veld seams where test channels are not provided, testing vill consist of the use of soap film sad a vacuum box. The penetration sleeve to liner plate velds are also covered with test channels to per-mit testing similar to that for inaccessible veld seams. Each shop assembled unit including single and multiple penetration units will be shop tested to demonstrate leak tightness. More emplete details of testing procedurec on the liner and emponents are included in Appendix SE.

Where the liner abuts concrete, it will be in an environment which vill minimiziii any possible corrosion. As described in Section 5.1.2.8 cathodic protection vill be provided to minimize any ground corrosion influsnees . Where the liner is exposed, it will be protected with a coating similar to Carbo Zine as manufactured by the Carboline Corpora-tion. The concrete shell vill protect the liner fra weather iufluences and potential externally generated missiles and insulate the liner plate frm 1cv temperatures. The liner surface on the interior of the vessel vill be protected fra intetnally generated missiles by concrete and in instances a cabination of concrete and steel shields.

Because the vapor barrier is protected in this marmer, and once the ade-quacy of the liner has been initially established, there is no reason to anticipate deterioration of the liner, which would jeopardize its O effgetiveness as a vapor bar::ier. Nevertheless periodic inspections will Ae made of the exterior of the Reactor Building and of interior spaces which are accessible during full power operation.

• l h b 5-15 0001 028

A significant number of steel lined, reinforced concrete containment l vessels are being constructed presently; these vill provide a consider-able background of experience prior to the construction of this vessel.

Full advantage of this knowledge vill be taken in all phases of design, l fabrication, installation, inspection, testing, and operation. Appropriate action vill also te taken to minimiza the possibility of reoccurrence of further leckage, include such redesign as might be necessary.

5.5 SYSTEM DESIGN EVALUATION l

l The Reactor Building, including extensions of the containment boundary l as described in Sections 5 2 and 5.3, vill with the functioning of the l additional engineered safeguards prevent 'an uncontrolled release of radio-l activity to either the plant site or the surrounding areas during normal operation or any accident conditions up to the most severe hypothetical

, accident. Containment integrity is maintained when there is nuclear fuel in the core coincident with either a reactor coolant system pressurized r

above 300 psig or a reactor coolant temperature above 200 F.

5.6 TESTS AND INSPECTION i 5.6.1 PRIOPERATIONAL TESTING AND INSPECTION 5.6.1.1 During construction Test, code, and cleanliness requirements vill acccupany each specification or purchase order for materials and equipnent. 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, vill also be described together with shipping procedures. Standards and tests vill be specified in accordance with applicable regulations, reccgnized technical society codes and current industrial practices.

Inspection vill be perfomed in the shops of vendors and subcontractors as necessary to verify ccupliance with Specifications.

5.6.1.1.1 concrete Testing of concrete materials and concrete as placed is described in Appendix SD, " Quality control." An experienced full-time concrete inspector vill con-tinuously check concrete batching and placing operations.

5.6.1.1.2 Prestressing Testing and inspection of all prestressing materials and special installation equipment is described in Appendix 5D, " Quality control." Full-time super-l vision of the prestressing operations vill be provided by an inspector ex-l perienced in prestressing as well as by the aforementioned concrete inspector.

O s,.

m , e, , 5-16 0001 029

l O 5.6.1.1.3 Reinforcing Steel Testing and inspection of reinforcing steel is described in Appendix SD,

" Quality Control." The concrete inspector vill check the condition and placement of the bars in the foms for compliance with drawings and specifi-cations, including velded splices.

5.6.1.1.h Liner Plate Testing and inspection of the liner plate is described in Appendix 5E,

" Liner Plate Technical Specifications." 1 5.6.1.2 structural Test ,

Following empletion of construction and prior to tha initial fuel loading, the Reactor Building vill be pressureized at 115 per cent of the design pre-sure for one hour to establish the structural integrity of the building.

The response of the building vill be empared with the calculated behavior to confim the design by means of instrumentation described in Appendix 5F, .

" Ins trumente. tion."

5.6.1 3 Initial Integrated Leaktge Rate Test The purpose of the integrated leakage rate test is to measure the percentage by weight of air which can leak out of the reactor building per day at the

, design pressure of 55 psig. The specified design leakage of 0.2 per cent i i. s or less per day will be measured using the Absolute Method. During the test, air pressure, water vapor pressure, and air temperature rill be measured using high accuracy instruments. The reactor building ventilation system vill be used continuously throughout the test to achieve emplete air mixing and control of air temperature. Euration of the test will be a minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The leakage rate test vill be repeated at reduced pressures of approximately 50 per cent and 25 per cent of the design pressure to establish the variation of leakage rate with pressure. The resC.ts of these reduced pressure tests may allow a basis for carrying out future ntegrated leak rate tests at pressures belev the design pressure.

5.6.2 POSTOPERATIONAL LEAK MONITORING 5.6.2.1 Leakage Monitoring Periodic integrated leak rate tests of the Reactor Building vill be carried out to verify its continued leak tight integrity. The postoperations'.

leak rate test vill be conducted at a pressure established from the pr. .;rsa of initial leakage rate *,ests. The acceptable leakage rate at this pressure vill be determined fra the measured variation of leakage rate with pressure.

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,_, 0001 030 i

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hb13 $-3 Reactor Ika1141rq Iso 1Ttico V 'In Infossation m

I%ne- Flow locattua Line tee tbat S nraal Ibst-tration Direc- Valve hferroi valve Site, of baa ve Ibaitten Accident N. Se rv ic e Systm tion Arrat. to R.B. Type to. Actuation Sigral Ibeltion Im11 cation Ibeltion 437(*) Ivessuriser CA Gat S Inside Gate 3/e De0* ES Closed Yet Closed

( t,) ami Reactor Gate 3/4 Deo ES Cloemt Yes Closed Coulant Sample Lines Gatside Gate 3/4 Air ES Closed Yes Closed 309 Intermediate IC In 4 Inalde O.ec k 6 Open W Closed Cooling Water Sapply Line Outside Gate 6 -- -- Open b Closed 320 latermediate IC Ost 3 Inside Gate 6 De0 ES Open Yes Closed Cooltsw Water Oatlet Line Oatside Gate 6 Air I;S Open Yes Closed Gatside cate 4 Air ES Closei Yes closed Jos Imt thewa Llae Hit Ost S Inalde date 8-8/2 D00 ES Open Yes Closed to I%rtrication Gate I-l/2 D40 ES Closet Yes closed Deatneralizers VI a Oataide Gate 2- 8 / 2 Air ES Open 0 Yes Closed

$ 306 Iteactor Coolant 3%ap Semi Re-seJ Oat 3 Inside Gate 3 Deo E3 Open Yes Closed QO tura Line Oatside Gate 3 Air ES Open Yes Closed g Joy N actor Coolant 3%mp Seal Water teJ In Inside Stop Ch. I -- -- Open b Closed g Supply Outa1Je Globe e A1. ES 61 Erottled Yes Closed **

Globe e Air ES Closed Yes closed" 7

CD 30s brual thkeup NJ In 2 Inside Check 4 -- -- -- M Open to tie Beactor h

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Coolant System Dats tJe 04tside Globe Globe B*l/2 B-t/2 Air -- Yhrottled Yes Open w

haual -- Closed Yee Open Gatside Gate 4 Air ES Closed Yes 03,en y, High Pressure MJ In y Inside Geck N 4 -- -- --

Open Injection Line Outside Gate 4 Air ES Closo! Yes Open 384 Puel Transfer SP In/Out e Inside Special 30 -- -- Closed --

Closed

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Closed b Closed 387 k actor 814g RBS In 7 Inside Oseck e

- Spray i det h Opea Line Outside Gate e Air ES Closed Yes Open w7-C -

. All valves with electric motor operators are also equipped with hand wheels.

'Yostaccident seactor ciutsat system pressume causes valve ta close.

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% N t d Tmt.le 3-3 (comt'd) h e. Flow location Lane nestam.: Ermal N t- b tration entroc - valve fieferred talve Size, of Walv e h itian Acettent W. Srwace $reten tion Arrgt. to II . B. type la. Actamation Signal hitim Inlicatica h iston ice henctor B1J4 likS Ir. p laside O.ec k g -- .- CheJ Ib Open Spray Irdet La ne htsade Gate 4 Air ES Closed Tea Opea 332 Decay Bent De la 6 Ins 14e O eck 02 .- ... Closed b upea Crolmat Saturn htside OeLe 42 Air ES Qosed Yes Open 333 Decay akat Edi la 6 Inside O.ec k 42 -- --

Closed Ib Open Cus.laat hetura Laeta 14e Gate 32 Air ES Closes Tea Olea 380 becay Heat Del Ot 9 laside wate 80 M -- Closed to C losed Coule**. lat. Oste (d) 80 8b80 kamute Closes Tee closed (kna W1 Oatalee Gate 30 Lau genote Closed yes Closaa am i 352 alcactor 214 Out to mis 1Je Gate tMu n .m a Cloemt DAB l4 Yes >

bay flecircula- seam tion Lines 353 Reactor Rida Del Ost 30 Outa14e Gate 94 43 0 Samute Closed Tee Open S ap Acetrcala. w tion Lines 1

FO 337 Isent:t Cec 84ht WD Out IS htaide Cete 2 Air a3 Cle, sed Tea CheJ O 9, ,,, t,3 g 2M 3,,,, ,,, g gj$ , g, 38 lastas btterfly as Es0 Ei Closed Tea Closed 4 Inlet Purge CloseJ Lano uuts1Je Atterfly 4e Air D C L2aed les

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@ LE OF CCNTENTS Section Page 6 E'IGI_NEERED SAFE _ GUARDS 6-1 6.1 24ERGENCY INJECTION 6-1 6.1.1 DESIGN BASES 6-1 6.

1.2 DESCRIPTION

6-2 6.1.3 DESIGN EVALUATION 6-3 6.1.3.1 Failure Analysis 6-5 6.1.3.2 Emergency Injection Response 6-5 6.1.3.3 Soecial Features 6-6 6.1.3.h Check valve Leakage - Core Flooding 1 System 6-6 6.1.h TESTS AND INSPECTIONS 6-6a -

6.2 REACTOR BUILDING ATMOSPHERE CCOLING AND WASHING 6-13 e

6.2.1 DESIGN BASES 6-13

O 6.

2.2 DESCRIPTION

6-13 6.2.3 DESIGN EVALUATIOR 6-lh 6.2.3.1 Failure Analysis 6-15 6.2.3.2 Reactor Building Cooling Response 6-19 6.2 3.3 Special Features 6-19 6.2.h TESTS AND INSPECTIONS 6-19 6.3 ENCINEERED SAFEGUARDS LEAKAGE AND RADIATION C65SIDERATIONS 6-20 6.

3.1 INTRODUCTION

6-20 6.32 SUto!ARY OF POSTACCIDENT RECIRCULATION AND LEAKAGE CONSIDERATIONS 6-20 6.3 3 LEAKAGE ~ ASSUMPTIONS 6-21 6.3.h DESIGN SASIS LEAKAGE 6-22 6.3.5 LEAKAGE ANALYSIS CCNCLUSIONS 6-22 6.3.6 SYSTE! INTEGRITY 6-22a

, 6-1 (Revised 12-22-67)

LIST OF TABLES Table No. Title age 6-1 Core Flooding System Performance and Equipmen:

Data 6-4 6-2 Single Failure Analysis-Emergency Injection 6-7 6-3 E=ergency Injection Equipment Performance Testing 6-12 l

6k Reactor Building Cooling Unit Performulce and 1 Equipment Data 6-13 6-5 Reactor Building Spray System Performance and Equipment Data 6-lb 6-6 Single Failure Analysis-Reactor Building At:nosphere Cooling and Washing 6-16 6-7 Leakage quantities to Auxiliary Building i Atmosphere 6-22 l o

l O

. +

6-11 (Revised T-21-67) t

O LIST OF FIGURES (At rear of Section)

Figure No. Title 6-1 Emergency Injection Safeguards 6-2 Makeup Pump Characteristics 6-3 Decay Heat Removal Pump Characteristics 6-4 Decay Heat Removal Cooler Characteristics 6-3 Reactor Building At=osphere Cooling and Washin6 Safeguards 6-6 Reactor Buildic6 E=ergency Cooler Characteristics 6-7 Reactor Buildin6 Spray Pu=p Characteristics O

i l

O

. 0001 043 6-111

( 6 ENGINEERED SAFEGUARDS Engineered safeguards are provided to fulfill t'ree functions in the un-likely event of a serious less-of-ccolant accid at:

a. Protect the fuel cladding.

b. Insure reactor building integrity.

c. Recove fissica products frcm the reactor building at=csphere and reduce the driving force for building leakage.

E:::ergency injection of coolant to the reacter ecolant system satisfies the first function above, while building atmosphere cooling and washing satisfy the latter two functions. Each of these operations is performed by two cr =cre systems which, in addition, e= ploy =ultiple ecmponents to insure operability. All equipment requiring electrical power for opera-tion is supplied by the emergency electrical p,ser sources as described in 8.2 3 Se engineered safeguards include a core flooding system, high pressure injection equipment, the decay heat removal system, the reactor building cooling system, and the reactor building spray system. Figurea 6-1 and 6-5 shov the operatien of these systems in the engineered safeguards mode, together with associated instru=entation and piping.

, ( Applieble codes and standards for design, fabrication, and testing of

'. components used as safeguards are listed in the introduction to Section 9, and seismic requirements are given in Section 2. The safety analysis presented in Section 14 demonstrates the perfc =ance of installed equip-

=ent in relation to functional objectives with assumed failures.

The engineered safe 6uards functions noted above are accc=plished with the postaccident use of equip =ent serving nor=al functions. The design ap-preach is based on the belief that regular use of equip =ent provides the best possible means for =cnitoring equipment availability and conditions.

Because sc=e of the equipment used serves a comal function, the need for periodic testing is =inimiced. In cases where the equipment is used for emergencies only, the systems have been designed to permit meaningful periodic tests. Additional descriptive information and design details on equipd:ent used for normal operation are presented in Section 9 This See-tion 6 vill present design bases for safeguards protection, equignent operational descriptions, design evaluations of equip =ent, failure anal-

[ ysis, and a preliminary operational testing program for systems used as engineered safeguards.

1 6.1 EMERGENCY INJECTION 6.1.1 DESIGN BASES D e principal design basis for the e=ergency injection is as follows:

i (37 o &!ii;i 6'

.t 0001 044

Emergency core injection is trovided to trevent clad melting for the entire stectrum of reacter coolant system failures ll ranzing frc= the smallest leak to the cc=clete severance of the lar est reacter coolant cite.

High pressure injection is prcvided to prevent uncovering of the core for 1 small coolant piping leaks at high pressure and to delay uncovering of the core for intermediate-si:ed leaks. The core ficoding syste= and the decay heat removal system (which provides lcw pressure injection) are provided to recover the core at intermediate-to-low pressures so as to maintain core integrity during leaks ranging frc= inter =s'iate to the largest size.

This equipment has been conservatively si:ed to Ir=it the te=perature transient to a clad te=perature of 2,300 F or less.

6.1.2 DESCRIPTICH Figure 6-1 is the sche =atic flev diagram for the emergency injecticn and associated instru=entation.

Escrgency injection fluid, pu= ped to the reactor coolant system during safeguards operations , is supplied in each case frc= the borated water storage tank. This tank contains the volume of borated water necessary to fill the fuel transfer, canal during refueling operations and is cen-nected to the injection pu=p suction headers by two lines. Additional l6 coolant for emergency injection supply is centained in core flooding tanks which inject without fluid pu= ping as described later in this section.

h E=ergency injection into the reactor coolant aystem vill be initiated in the event of an abnor= ally low reactor coolant syste= pressure of 1,800 psi during power operation. The low pressure signals vill autcmatically increase high pressure injection flow to the reactor coolant system with the following changes in the operating mode of the =akeup and purification system described in Section 9: (a) the standby =akeup pu=ps vill start 6 and ec=e en the line, (b) the stop-check valves in each injection supply line to the makeup and decay heat pu=ps will open, and (c) the injecticn valve in each of four injection lines will open. E=ergency high pressure injectica vill centinue until reactor coolant system pressure has dropped to the point where core flooding tanks begin emergency injection. The flow characteristic e m es for each makeup pu=p are given in Figure 6-2 The core flooding syste= is ce= posed of two flooding tanks, each directly connected to a reactor vessel no::le by a line containing two check valves and one isolation valve; the system provides for autc=atic ficoding in-jecticn with initiation of flev vhen the reacter coolant system pressure reaches approxi=ately 585 psi. This injection provision does not require any electrical pcVer, aute=atic switching, or operator action to insure supply of emergency ccclant to the reactor vessel. Operator action is required caly during reacter cooldcyn, at which time the isolation valves in the core flooding lines are c1csed to contain the contents of the core ficcding tanks. The ec=bined ecolant centent of the two flooding tanks is sufficient to recover the ccre het spot aasu=ing no illl, s . .,

6-2 (Revised 1-8-68)

I hk e

liquid is contained in the reactor vessel, while the gas everpressure and flocd- 1 ing line sizes are sufficient to insure core reflooding within approximately 25 ON see after the largest pipe rupture has occurred.

The decay heat removal system (described in Section 9) is nor= ally maintained on standby during power operation and provides supplemental core floodin6 flow through the two core flooding lines after the reactor coolant system pressure reaches 135 psi. Emergency operation of this system will be initi-ated by a reactor coolant system pressure of 200 psi during any accident.

The flow characteristics of each decay heat pump for injection are shown in Figure 6-3; these pumps are designed to deliver 3,000 gpm flow into the reac-tor vessel at a vessel pressure of 100 pai.

Low pressure injection, with supply from the borated water storage tank, using the decay heat pumps will continue until a lov level signal is received from the tank (39 min at a combined low pressure injection and reactor buildir.6 spray flow of 9,000 gpm). At this time, the operator will open the valv>ss controlling suction from the reactor building sump, and recirculation of cool- l6 ant from the sump to the reactor vessel will begin. The decay heat coolers will cool the recirculated flow, thus removing heat from the reactor building fluid and preventing further building accumulation of decay heat generated by the core.

1 The decay heat re= oval pumps are located at an elevation below the reactor building su=p with dual suction lines routed outside the reactor building to the pumps. In the event one suction line is unavailable for recirculation, the lines have been sized so that one line vill be capable of handling the total potential recirculation flow of one 3,000 gpm decay heat removal pump, 6 and one 1,500-gpm reactor building spray pu=p. The NPSH available has been conservatively calculated to be greater than the N?SH requirement of the decay heat recoval pu=ps , and the reactor huilding si, ray pumps.

The heat transfer capability of each decay heat cooler as a function of re-circulated water te=perature is illustrated in Figure 6 k. The heat transfer capability at the saturation te=perature corresponding to reactor building pressure is in excess of the heat generation rate of the core following stor-age tank injection.

1 Design data for core flooding system components are given in Table 6-1. De- {

sign data for other emergency injection components are given in Section 9 l except for those shown in Figures 6-2, 6-3 and 6 k.

6.1 3 DESIGN EVALUATION In establishing the required components for the emergency injection the fol-lowing factors were considered:

a. The probability of a major reactor coolant system failure is very low. *

b. The fraction of a given component lifetime for which the component e is unavailable because of maintenance is estimated to be a small part of lifetime. On this basis, it is estimated that the

\ l ( ,

0001 046 l

'd **9 !I' 6-3 (Revised 1-8-68)

probability of a major reactor coolant system accident occurring '

while a protective ecmponent is out for maintenance is two orders of magnitude below the lov basic accident probability,

c. The equipment downtime for maintenance is a well-operated station often can be scheduled during reactor shutd an periods. When main-tenance of an engineered safeguard component is required during oper-ation, the periodic test frequency of the remaining equipment can be increased to insure availability.

d. Where the systems are designed to operate normally or where meaning-ful periodic tests can be perfor ned, there is also a low probability that the required emergency action would not be performed when needed.

That is, equipnent reliability is improved by using it for other than emergency functions,

e. (DELETED)

f. Three makeup pumps are installed. One makeup pump is normally op- 1 erating, and one pu=p can be down for maintenance. One pump is 6 required for engineered safeguards.

Table 6-1 Core Flooding System Perfor=ance and Ecu1= ment Data Core Flooding Tanks (*)

Number 2 .

Design Pressure, psig 700 Normal Pressure, psig 600 Design Temperature, F 300 Operation Temperature, F 110 Total Volu=e, ft3 1,h10 Nor=al Water Volume, ft 3 9h0 Material of Construction Carbon Steel-lined Check Valves Number per Flooding Line 2 Size, in. lh Material SS Design Pressure, psig 2,500 Design Temperature, F 650 l Isolation Valves l

Number per Flooding Line 1 Size, in. 14 l

Material SS Design Pressure, psig 2,500 Design Temperature, F 650

(*) Designed to ASME Section III, Class C. l1 .

i'~

• 6-h (Revised 1-8-68)

J' '

0001 047

Table 6-1 (Cont % /

Piping 1

Number of Flooding Lines 2 Size, in. 14 Material ss Design Pressure, psig 2,500 Design Temperature, F 650 6.1 3 1 Failure Analysis The single failure analysis presented in Table 6-2 is based on the assumption that a major loss-of-coolant accident had occurred. It was then assumed that an additional malfunction or failure occurred either in the process of actuating the emergency injection syste=s or as a secondary accident effect. All credible failures were analyted. For example, the analysis includes malfunctions or failures such as electrical circuit or motor failures, stuck check valves, etc.

It was considered incredible that valves would change to the opposite position by accident if they were in the required position when the accident occurred.

In general, failures of the type assumed in this analysis should be unlikely be-esuse a program of periodic testing and service rotation of standby equipment will be incorporated in the Station operating procedures.

The single failure analysis (Table 6-2) and the dynamic postaccident perfor=ance analysis (Section ik) of the engineered safeguards considered capacity reduction as a result of equipment being out for maintenance or as a result of a failure to start or operate properly. This amounts to adding anather factor of conser-O/ vatism to the analyses because good operating practice requires repairing equip-ment as quickly as possible.

Station maintenance setivities will be scheduled so that the required capacity of the engineered safeguards systems will always l1 be available in the event of an accident.

"'he adequacy of equipment sizes is demonstrated by the postaccident perfomance analysis described in Section IL, which also discusses the consequences of achieving less than the max 1=um injection flows. "here is sufficient redundancy in the emergency injection systecs to preclude the possibility of any single credible failure leading to core =elting.

6.132 Emergency Injection Response The emergency high-pressure injection valves are designed to open within 10 sec. 1 One makeup pu=p is nor= ally in operation, and the pipe lines are filled with 6 coolant. The four high-pressure injection lines contain thermal sleeves at their connections into the reactor coolant piping to prevent overstressing of the pipe juncture owing to the 90 F water being injected into these high temperature lines.

• he equipment normally operating is handling 125 F water, and hence will exper-ience no thermal shock when 90 F water is introduced.

Injection response of the core flooding system is dependent upon the rate of re-duction of reactor coolant system pressure . For a maximum hypothetical n:pture, the core flooding system is capable of reflooding the core to the hot spot within a safe period after a rupture has occurred.

Emergency low pressure injection by the decay heat removal system will be deli-  !

q vered within 25 see after the reactor coolant system reaches the actuating pres-  !

Q sure of 200 psig. This anticipated delay time consists of these intervals:

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6-5 (?evised 1-5-cS) 0001 048 l

l l l

a. Total instru=entation lag --

b. Energency power source start

• 1 see-h

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c. Pu=p =ctor startup (frc= the time the pu=p

=otor line circuit breaker c1waes until the pu=p attains full speed) -- = 10 see

d. Injection valve openirg ti=e -- <10 sec

e. Berated water storage tank outlet valves -- <10 see Total (only b and e are additive) = 25 sec 6.1 3 3 Special Features The core flooding nozzles (Figure 3-k7) vill be specially designed to in-i sure that they will safely take the differential te=peratures i= posed by I

the a::cident condition. Special attention also vill be given to the ability of the injection lines to absorb the expansion resulting frc= the recirculating water te=perature.

For =cs: of their routing, the e=ergency injection lines will be outside the reactor and stea= generator shielding, and hence protected frc= mis-siles originating within these areas. The portiens of the injection lines located between the pri=ary reactor shield and the reactor vessel vall are not subject to =1ssile da= age because there are no credible sources of

=issiles in that area. To afford further =issile protection, a high- 6 i pressure injection line connects to each reactor coolant inlet pipe and the two core flooding no::les are located on opposite sides of the reactor vessel.

l All vater used for e=ergency injection fluid vill be =aintained at a mini-

=u= concentration of 2,270 pp= of bcron. The te=perature, pressure, ard level of these tanks will be displayed in the control roc =, and alar =s vill sound when any condition is outside the nor=al li=its. The water vill be periodically sa= pled and enlyzed to insure preper boren concen-

! tration.

l 6.1 3 4 Check Valve Leakage - Cere Flooding Syste=

The action that would be taken u the case cf check valve leakage veuld 1 be a function of the =agnitude of the leakage.

L1=ited check valve leakage vill have no adverw effect on reactor opera-tion. The valves vill be specified to =eet the tightness require =ents of MSS-SP-61. For these valves, this a=ounts to a =ax1=u= pe: ::issible leakage of 114 cc/hr per valve. Two valves in series are provided in each core ficoding line; hence, leakage should be below this value.

Leakage across these check valves can have three effects: (a) it can cause a te=perature increase in the line and core flooding tank, (b) it can cause a level and resultant pressure increase in the tank, and (c) it can cause dilution of the borated water in the core f1 ceding tank.

,re' hb, -t 6-6 (Revised 1-8-68) 0001'049

O Leakage at the rate mentioned abov'e causes insignificant changes in any 1 of these parsmeters. A leakage of 140 cc/hr causes level increase in the tank of less than 1 in./=o. The associated te=perature and pressure increase is correspondingly lov.

If it were assu=ed that the leakage rate is 100 tir.:e3 greater than spec-ified, then there vould still be no significant effect on reactor opera-tion since the level change vould be approximately 2 in./ day. A 2-in.

level change vill result in a pressure increase of approximately 10 pai.

With redundant temperature, pressure, and level indicators and alarms available to monitor the core flooding tank conditions, the most signifi-cant effect on reactor operations is expected to be a = ore frequent sam-pling of tank boric acid concentration.

Ib insure that no temperature increase vill occu in the tank, even at higher leakage rates, the portion of the line between th2 two check valves and the line to the tanks vill be left uninsulated to promote con-vective losses to the building at=osphere.

In su=ma.~/, reactor operation may continue with no adverse effects coin-cident with check valve leakage. Maxi:::um permissible li=1ts on core flooding tank parameters (level, temperature, and boron concentration) vill be established to insure compliance with the core protection cri-teria and final safety analyses.

p All active ec=ponents, as listed in Table 6-3, of the erergency injec-1 Q tion syste=s vill be tested periodically to demonstrate system readiness.

In addition, normally operating components will be inspected for leaks from pump seals, valve packing, flanged joints, and safety valves. ,

l 0001 050 6-6a (Revised 7-21-67)

A O U O

_G

[r Table 6-2 -

Uf Single Failure Analysis-Emergency Injection Component Malfunction Comments and Consequences A. liigh Pressure Injection

1. Pneumatic valve at makeup Valve remains open. When the tank is empty, tank pressure tank outlet. would be less than the high-pressure injection pump suction pressure (with borated water storage tank on the line), thus preventing the release of hydrogen from the tank to the pump suction line.

2. Pneumatic suction valve Fails to open. Similar valve in other makeup pump 6 for makeup pumps from bo- string will deliver required flow.

i' rated water storage tank.

w 6

-s 3. Makeup pump. Out of maintenance. Two pumps will still be available.

7 Only one pump is required for engi-e neered safeguards.

a .

h. Makeup pump. Fails (stops). Other makeup pump delivets required l6 7 _

flow.

?

os 53 5 Makeup pump isolation Left inadvertently closed. See Item A-h above. Valves will nor-valve. mally be left open since the check valve in each pump discharge will pre-vent backflow. Operating procedures will call for pump isolation valves to be closed only for maintenance.

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_ Table 6-2 (Cont'd)

Component Malfunction Comments and Consequences _

6. Makeup pump discharge Sticks closed. This is considered incredible since check valve. the pump discharge pressure of 2,700 psi at no flow would tend to open even a very tightly stuck check disc.

7 Pressurizer level control Fails to close. No consequences. .l 6 valve.

8. Seal injection control Fails to close. Injection flow through this line would valve. be small compared to the flow through the injection lines due to the high flow resistance of the reactor coolant pump seals.

i' os 9 Pneumatic valve in high- Fails to open. Flow from one pump will go through the 6 g7 pressure injection line. alternate line. Other pump will oper-q ate as normal.

$ 10. See comment on Item A-6.

g, Check valve in injection Sticks closed.

line (inside reactor 4, building).

$s 33 11. Injection line inside reac- Rupture. Flow rate indicators in the four in- l6 tor building. jection lines would indicate the gross difference in flow rates. Check valve in the injection line would prevent additional loss of coolant from the reactor. The line is protected from C-, missiles by reactor coolant system q;) shielding.

C Ln N

O O O

O D O t

Table 6-2 (Cont'd)

Component Mal functi on 6 Comments and Consee"ences

12. Pneumatic valve from de- Inadvertently left open. No significant consequences. A small cay heat coolers. percentage of I.P injection flow will be bypassed to llP suction.

13. Pneumatic valve from de- Fails to open. This valve, if required to be opened, cay heat coolers. need not be opened until about 39 min.

after the accident. This provides time for manual opening.

14. Double manual valves con- Inadvertently left open. Not credible that both valves will in-necting pump lines. advertently be left open.

T T

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r O O O Table 6-2 (Cont'd)

Component Halfunction Conanonts and Consequences

'~

, B. Core Flooding System r,- ,

1. Flooding line check valve. Sticks closed. '1his is considered incredible based on the valve size and opening pressure applied.

C. Decay Heat Removal System

1. Check valve at reactor St.icks closed. 'lhis is considered incredible since vessel. these valves will be used periodically during decay heat removal, and the opening force will be approximately 5,000 pounds, os b 2. Air-operated injection Falla to open. Second injection line will deliver re-valve. quired flow.

3 safety valve. stuck open. loss of injection flow is small since valve is small.

16 Decay heat cooler. Isolation valve left closed. Other heat exchancer will take re-quired injection flow and remove re-quired heat. Valves will be closed only fov caaintenance of heat exchanger.

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y Z Table 6-2 (Cont'd,-

e.

Coranents and Consequences

~T Component Halfunction 5 Decay heat cooler. Massive rupture. Not credible. During rennal decay heat removal operation, heat exchanger will be exposed to higher pressure and approximately the same temperature as the postaccident temperature and pressure.

6. necay heat cooler. Out for maintenance. Remaining heat exchanger will take re-quired injection flow.

7 Decay heat pump isolation Left closed. Remaining pump will deliver required 6 valve. injection flow.

os k 8. Decay heat pump discharge Sticks closed. See comment on Item C-1 above.

O check volve.

n E 9 Decay heat pump. Fails to start. Remaining pump will deliver required 6 injection flow.

>}.

a

10. Stop-check valve at bo- Sticks closed. Alternate line will permit required 6 e rated water storage tank flow.

outlet.

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11. Air-operated valve permit- Fails to open. Tuo lines and valves are provided, but ting suction from reactor need not be actuated until 39 min building sump. after start of accident which provides time for manual operation.

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Table 6-2 (Cont'd)

. Component Halfunction comunents and Consequences

12. Reactor building stap out- Decomes c1c6ged. 'Ihis is considered incredible because let pipe. of the dual sump line arrangement, the size of the lines, and the sump design.

'lhe two recirculation lines take suc-tion from the different portions of ti.c sump. A grating will be provided over the sump, and additional heavy dt.ty strainers will be provided for auction lines.

13. Dual manual valves con- Inadvertently left open. Not credible that valves will inadver- 6 necting suction headers. tently be left open.

p 14. Air-operated valve per- Inadvertently and prematurely See answer to Question 17.6, Amend- 6 p . mitting suction from re- opened after LOCA. ment #h, Supplement 3.

actor building sump.

E 5

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1" O

C LII e

O Table 6-3

, Energency Injection Ecuitnent Performance Testing Makeup Pu=ps One pu=p is operating continuously. The 6 other two pu=ps vill be periodically tested.

High Pressure Injection The re=otely operated stop valve in each Line Valves linc will be opened partially one at a time. The flev devices vill indicate flow through the lines.

Makeup Pu=p Suction Valve The =akeup tank water level vill be raised l6 to equalite the pressure exerted by the storage tank and the borated water storage tank. The valv<.. vill then be opened indi-vidually and cicsed.

Decay Heat Pu=ps In addition to use for shutdown cooling, these pu=ps vill be tested singly by cpen-ing the borated water stcrage tank outlet valves and the bypasses in the berated water stcrage tank fill line. This vill allow water to be pu= ped frc= the berated water stcrage tank through each of the in-jection lines and back to the tank.

ll\ '

Borated Water Storage Ta=k The operational readiness of these valves Outlet Valves vill be established in ec=pleting the pu=p operational test discussed above. During this test, each of the valvec vill be test-ed separately fer flev.

Lev Pressure Injection With pu=ps shut devn and berated water Valves storage tank outlet valves closed, these valves vill be opened and reclosed by oper-ator action.

Valve for Suction Frc= With pu=ps shut devn and borated water Su=p stcrage tank outlet valves closed, these valves vill be opened and reclosed by oper-ater action.

Valves in Cere Floeding Valves can be operated during eact shutdevn Injecticn Lines to deter =ine perfor=ance. Isolation valves vill be closed to contain water in core f1 ceding tanks during shutdevn.

i i

l i , e CCk - 6-12 (Revised 1-8-68) j[ 1

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    ,-  'u

6.2 REACTOR BUILDING ATMOSPEERE CCOLING AND '4 ASHING 6.2.1 CESIGN BASES Emergency building at=osphere cooling and washing is provided to limit post-accident building pressures to design values and reduce the postaccident level of fission products in the building atmosphere. Beactor building air recirculation and cooling units, backed up by reactor building sprays, are used for emergency atmosphere cooling. Chemical additives contained in the building sprays are used to reduce postaccident fission pro-duct concentrations in the building atmosphere. 6.2.2 DESCRIPIION The sche =atic flow diagram of the emergency reactor building atscaphere cooling and washing and associated instrumentation is given in Figure 3-5 E=ergency and nor=al cooling are perfer=ed with the sa=e basic units. Each 1 unit contains an e=ergency ecoling coil, a normal cualing coil, and a two-speed fan. For e=ergency cooling, all units vill operate under postaccident conditions with the heat being rej.. cu - .iver water. Each of these units can re=cve 80 x 100 Btu /hr under peak reactor building temperature cendi- l6 ticas. Figure 6-6 shcws the heat exchange characteristics versus building ambient conditions for these units. The design data for the cooling units are shown in fable 6-k. (Dd Table 6-4 Reactor Building Cooling Unit Performance and Equipment Data 1 (capacities are on a per unit basis) Duty Emergency Normal No. Installed 3 3 No. Required 3 2 Type Coil -Finned Tubg Finned Tube Peak Heat Load, Btu /hr 80 x 100 1.25 x 106 Fan Capacity, eft 54,000 108,0C0 Reactor Building At=osphere Inlet Conditions Te=persture, F 281 110 Steam Partial Pressure, psia 50 -- Air Partial Pressure, psia 20 -- Total Pressure, psig 55 Atmospheric Cooling '4ater Flow, gp= 1,780 250 Cooling Water Inlet Te=perature, F 85 85 l6 Cooling Water Outlet Temperature, F 175 95 l Q, Si=ultaneously with the air recirculation cooling, reactor building sprays V ,,,,are supplied with water by two pumps which take suction on the borated 0.: 5

        ,     ' a3.,;

6-13 (Pevised 1-8-68) 0001 058

water storage tank until this coolant source is exhausted. The sodium thiosul-fate che=ical additive required for the reactor building sprays is supplied lh frc= a stcrage tank connected by dual lines containing check valves to the suc-tion of the spray and decay heat re= oval pu=ps. Sufficient sodium thiosulfate ic injected into the berated water to create a 1 wt % concentration in the re- l1

• actor building water inventory.

After the supply from the borated water storage tank is exhausted, the spray pu=ps take suction frc= the reactor building sump recirculation lines. This continued spraying serves to reduce the reactor building at=csphere to the te - perature of the reactor building su=p. Design data for the reactor building spray system components are given in Table 6-5, and the flow characteristics of the reactor building spray pu=ps are given in Figure 6-7 Desigh data for c0=ponents of the reactor building cooling and decay heat re= oval systems used in this phase of engineered safeguards opera-tion are given in Section 9 and supple =ented by Figures 6-3, 6-4, 6-6, and 6-7 of this Section. Table 6-5 Reactor Building Scray Syste= Perfor=ance and Eculp=ent Data (capacities are on a per unit basis ) - Reactor Euilding Spray Pu=ps Number 2 Flow, gp= 1,500 \. Developed Head at Rated Flow, ft 430 Motor Horcepower, hp 250 l1 Material SS Cesig.s Pressure, psi 3C0 resign Te=perature, F 300 Sodium Thiosulfate Tank Number 1 Volu=e, ftx 1,500 Material SS 6 Design Pressure, psi 50 1 Eesign Temperature, F 150 Sodium Thicsulfate Concentration, wt % 30 Spray Header Number 2 l Spray No::les per Header 375 l1 6.2 3 DESIGN EVALUATION The function of cooling the react,r building at=osphere is fulfilled by either of the two =ethods described above, and redundancy of equip =ent within both

   =cthods will provide for protection of building integrity. The reactor build-ing sprays through duplication, basic washing concept, and che=1 cal additive will serve to reduce fission product levels in the building at=osphere.                     l l; 1

1 k.. 6-1k (Revised 1-8-68) 8

             ..)
                         -                                                         0001    059

For the first 30 ko sin following the maximum blowdown loss-of-coolant acci-dent, i.e., during the time that the reactor building spray pumps take their suction from the borated water storage tank, this system provides more than 100 per cent of the heat removal capacity of the reactor building cooling sys-tem. The reactor building spray system design is based on the spray water being raised to the temperature of the reactor building in falling through the steam-air mixture within the building. Detailed evaluation of system performance is presented in Section 14. Each of the follcwing equipment arrangements will provide sufficient heat removal capability to maintain the postaccident reac-tor building pressure below the design value:

a. Reactor building spray sys, tem.

b. All emergency units in the reactor building cooling system. l1

c. Two emergency cooling units and the reactor building spray cystem at one-half capacity.

The reactor building spray system shares the suction lines from the borated l6 water storage tank and the tank itself with the high and low pressure injection safeguards . 6.2 3 1 Failure. Analysis A single failure analysis has been made on all' active components of the systems kA used to show that the failure of any single active component will not prevent fulfilling of the design functions. This analysis is shown in Table 6-6. As-cumptionc inherent in this analysis are the same as those presented in 6.1 3 in regard to valve functioning, failure types, etc. Results of full and par-tial performance of these cafeguards are presented in Section 14 under analy-sis of postaccident conditions.

'\ J f;a 0'001 060
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(.1 , 6-15 (Revised 1-d-68)

a  ; 5-

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                                                                'Ibble 6-6 Single Failure Analysis-Reactor Building Atmosphere Cooling and Washi.'t; Component                           Malfunction                   Comnents and Consequences

1. Reactor buildity; spray nozzles. Clogged. Large number of nozzles (375 on each I of two headers) renders clogging of significant number of nozzles as in-credible.

2. Reactor building spray header. Rupture. 'Dris is considered incredible due to low operating pressure differential.

3 Check valve in spray header Sticks closed. 'Diis is considered incredible due to line. large opening force available at pump shutoff head. 14 Air-operated valve in spray Fails to open. Second header delivers 50 per cent P header line. flow.

   "os 5   Spray pump isolation valve.        Ief t closed.                    Flow and cooling capacity reduced to g                                                                                50 per cent of design. In combination j                                                                                with emergency coolers,150 per cent f                                                                                of total design requirement is still
   &                                                                                provided.

[ 6. Reactor building spray plunp. I tila to start. Flow and cooling capacity reduced to 7 O 50 per cent of design. In combination f with emergency coolers, 150 per cent of total design requirement is still provided. 7 Nonnal and emergency cooling Stops. Emergency cooling by the other operat-unit fan. ing units with supplemental cooling g v by the sprays. ON O O O

                                                                       ^
                     -O                                                O                                                  O q.

Tj Table 6-6 (Cont'd) Component Malfunction Comments and Consequences kb 8. Normal and emergency cooling Rupture of emergency cooling The tubes are designed for 200 psi and

           ~
            ./L        unit.                         coil.                       300 F which exceeds maximum operating conditions. Tubes are protected against credible missiles. lience, rupture is not considered credible.

9 Normal and emergency cooling Rupture of casing and/or Consideration will be given during de-unit. ducts. tailed design to the dynamic forces resulting from the pressure buildup during a postaccident situation. The units will also be inspectable ar.d protected against credible missiles. Cooling with these units will be sup-plemented by the sprays. i' 10. Normal and emergency cooling Rupture of system piping. Rupture is not considered credible Il units. since all piping is Schedule 40, per-mitting an allowable working pressure of at least 500 pai at 650 F for all sizes. Piping is inspectable and pro-tected from missiles. Maximum actual internal pressure will be less than 200 psi at temperatures below 300 F.

11. Air-operated valve at inlet Sticks closed. Flow will be periodically established penetration. through the line to check the opera-tional capability f system. Such tests will show if valve is malfunc-tioning.

12. Air-operated valve at outlet Fails to open. Comments for Item 11 apply.

penetration. C) - Q Q

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                                           *p t                             -

Table 6-6 (Cont'd) 7 ..

                                                     'T                 Component                     Malfunction            Comments and Consequences 155 Pouer operated valve at sodi- Fails to open.           Alternate check valve will permit flov  6 um thiosulfate storage tank                            requir ed for sprays, outlet.

a H CD b. 51 Y m 9 0% O) v C O Os L.r4 O O O

(% 6.2.3.2 Reactor Building Cooling Response 1 Air recirculation established during ner=al operation through two of the three building ventilation units continues under accident conditions. In addition the third unit will be p3 aced in operation for the accident condition. Alter-nate cooling coils in these ventilation units, supplied with e=ergency cooling water, are placad into service after reactor building pressure increases to h psi. Cooling continues utilizing these coils until the building pressure reaches near-atmospheric, and the decay heat removal system is placed into eme - gency service, recirculat ng and cooling fluid fran the reactor building sump. The reactor building spray sfstem vill likewise be activated by a single pa-rameter signal. Two of three signals signifying high reactor building pressure vill start both of the reactor building spray pu=ps, open the reacter building spray inlet valves , and open the suction valves from the borated water storage tank. The system co=penents may also be actuated by operator action frca the control rocm for performance testing. The total time to delivery of reactor building sprays is approximately 1 min 3 after building pressure reaches 10 psi. 6.2.3.3 Seecial Features The casing design for the ventilation units vill be of a conventional nature unless additional analysis shows the possibility of pressure wave collapse. In that event, quick, inward-opening hinged doors, or other protective features, will be incorporated into the design to maintain pestaccident operability. The

  \ ( ~')   ventilation units are located outside the concrete shield for the reactor ves-sel, steam generators, and reactor coolant pumps at an elevation above the water level in the bottom of the reactor building at postaccident conditions.

In this location, the systems in the reactor building are protected from cred-ible missiles and frem ficoding during postaccident operations. Also, this location provides shielding so that the design radiation dose level is 25 mrem /hr and allevs for maintenance and repair, and inspections to be performed during power operation. The spray headers of the reactor building spray system are located cutside and above the reactor and steam generator concrete shield. During operation, a shield also provides missile protection for the area t=nediately above the re-actor vessel. The spray headers are therefore protected from missiles origi-nating within the shield. The spray pu=ps are located outside the reactor building and are thus available for operative checks during Station operation. 6.2.4 TESTS AND INSFECTIONS Active ecmponents of the ventilation units vill normally be in service. Valv-ing en the emergency coils can be periodically cycled, thus placing the coils into service periodically during operation. l l

          $      ! i

6-19 (Revised 11-6-67) 0001 064 t

l l 1 l i The active components of the reactor building spray system vill be tested on a regular schedule as follows: Reactor Building These pu=ps will be tested singly by opening the valves in Spray Pumps the test line and the borated water storage tank outlet valves. Each pump in turn vill be started by Station opera-tor action and checked for flow establishment to each of the spray headers. Flow will also be tested through each of the borated water storage tank outlet valves by operatit.g these valves. - Borated Water These valves vill be tested in performing the pump test Storage Tank listed above. Outlet Valves Reactor Building With the pumps shut down and the borated water storage tank Spray Injection outlet valves closed, these valves vill each be opened and Valves closed by operator action. Reactor Building Under the conditions specified for the previous test, and Spray Noz:les with the reactor building spray valves closed, low pressure 3 air vill be blevn through the test connections. 1 6.3 ENGINEERED SAFEGUARDS LEAKAGE AND RADIATION CONSIDERATIONS 6.

3.1 INTRODUCTION

The use of normally operating equipment for engineered safeguards functions and I ') the location of some of this equipment outside the reactor building require that consideration be given to direct radiation levels hfter fission products have accumulated in these systems with leakage from these systems. Although the engineered safeguards equipment is designed for control room operation fol-loving an accident, long-term postaccident operation could necessitate manual operation of certain valves. The shielding for components of the Engineered Safeguards is designed to provide protection for personnel to perform all operations necessary for mitigation of the accident within the limits of 10 CFR 100 in the event of an MHA. 6.3.2 SLM1ARY OF POSTACCIDENT RECIRCULATION AND LEAKAGE CONSIDERATICNS Following a less-of-coolant accident and exhaustion of the borated water storage tank, reactor building su=p recirculation to the reactor vessel and the reactor l building sprays is initiated. While the reactor auxiliary ;, stems involved in the recirculation ec= plex are closed to the auxiliary building atmosphere, leakage is possible through compo-nent fisnges, seals, instrumentation, and valves. p ;.} ',; ti 6-20 (Revised 11-6-67) 0001 065 O ' -

The leakage sources considered are:

a. Valves.

(1) Dise leakage when valve is on recirculation complex boundary. (2) Stem leakage. (3) Bonnet flange leakage.

b. Flanges.

c. Pump stuffing boxes.

While leakage rates have been assumed for these sources, maintenance and periodic testing of these systems will preclude all but a small percentage of the assumed amounts. With the exception of the boundary valve dises , all of the potential leakage paths may be examined during periodic tests or normal operation. The boundary valve dise leakage is retained in the other closed systems and there-fore vill not be released to the auxiliary building. While valve stem leakage has been assumed for all valves, the manual valves in the recirculation complex are backseating. 6.3.3 LEAKAGE ASSUMPTIONS

w. Source Quantities

a. Valves - Process (1) Dise leakage 10 cc/hr/in. of nominal disc diameter.

(2) Sten leakage 1 drop / min (3) Bonnet flange 10 drops / min

b. Valves - Instrumentati g Bonnet flange and stem 1 drop / min

c. Flanges 10 drops / min

d. Pu=e Stuffing Boxes 50 drops / min For the analysis , it was assumed that the water leaving the reactor building 3 was less than 200 F ..

vh.en recirculation occurs. l I _/ l

0001 066-y '. < - t ,.; 6-21 (Revised 11-6-67) l

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(DELETED) 3 6.3.4 DESIGN SASIS LEAKAGE The design basis leakage quantities derived from these assumptions for post-accident sump recirculation are tabulated in Table 6-7 6.3 5 LEAKAGE AliALYSIS CONCLUSIONS It may be concluded from this analysis (in conjunction with the discussion and analysis in 14.2.2.4.5) that leakage from Engineered Safeguards equipment cut-side the reactor building does not pose a public safety problem. B Table 6-7 Leakage Quantities to Auxiliary Building Atmosehere No. of Per Source, Total 3 Leakage Source Sources drons/ min ec/hr

a. Pumn Seals Decay heat pumps 2 50 Spray pumps 2 50 300 l6 300

                                                                                           )

b. Flanges (* lik 10 3,320

c. Process Valves 35 1 105

d. Instrumentation Valves 25 1 75

e. Valve Seats at Soundaries 11 (b) 750 Total k,850 l6 (a) Assumes process and boundary val ves, and process compo-nents are flanged.

Assumes 10 cc/hr/in. of nominal disc diameter. O h[ $i;tii 6-22 (Revised 1-8-68) Or101 067

i I l 6.3.6 SYSTD4 INTEGRITY In addition to the proper selection of materials, fabrication quality 5 control, and design, additional =easures have been taken to further increase the integrity of the engineered safeguards systems, i.e. High Pressure Injection, Decay Heat Removal System, Core Flooding System, and the Reactor Building Spray System. Provisions are included for the physical separation of each engineered safeguard systen and also the physical separation of redundant compenents within any system. Figure 1-3 shows the physical separation provided for these rystems and their redundant components. The entire train of the systems is included within the confines of the separation barriers and ensures that the failure of any system or redundant component will have no detrimental effect on any other system or component. The drain lines from the reactor building sump frem the point where they leave the building up to and including their first isolation valve are enclosed within a barrier. Leak detection equipment is also pro-vided within the individual cubieles which vill permit isolation of that particular portion of the system which may have failed. l The combination of double barriers, physical separability, and leak de-tection methods precludes the loss of function of any safeguard system. O , 5001 068 . 6-22a (Revised 12-22-67)

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