ML18151A509

Forwards Rev 2 to Updated Fsar.Rev 2 Pages Contain Necessary Changes to Bring Updated FSAR Up to Date as of 840101

ML18151A509
Person / Time
Site:	Surry
Issue date:	06/29/1984
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML18141A761	List: ML18151A509
References
349, NUDOCS 8407060272
Download: ML18151A509 (98)

Text

e e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA. 23261 W. L. STEWART VxcE PBESXDENT NucLEAB OPEBATIONS June 29, 1984 Mr. Harold R. Denton, Director Serial No. 349 Office of Nuclear Reactor Regulation NO/JHL:lms U. S. Nuclear Regulatory Commission Docket Nos. 50-280 Washington, D. C. 20555 50-281 License Nos.DPR-32 DPR-37

Dear Mr. Denton:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT NOS. 1 AND 2 REVISION TWO TO THE UPDATED FINAL SAFETY ANALYSIS REPORT Pursuant to 10 CFR 50.71(e), the Virginia Electric and Power Company submits Revision 2 to the Updated Final Safety Analysis Report (UFSAR) for the Surry Power Station Unit Nos. 1 and 2. One signed original and twelve additional copies of the Revision to the UFSAR are enclosed.

The Revision 2 pages contain the necessary changes to bring the UFSAR up to date as of January 1, 1984 which is within six months prior to the date of this letter.

The enclosed Revision 2 pages replace existing pages in the UFSAR and the revised material is identified by the change bar in the margin with the number "2" next to it. The enclosure also contains instructions for inserting the new pages and a revised list of effective pages.

As a duly authorized officer of Vepco, I hereby certify that the information given in the enclosed Revision 2 of the UFSAR accurately presents changes made since the previous submittal. Revision 2 reflects previous information and analyses submitted to the Commission or prepared pursuant to Commission requirements.

Very truly yours, Enclosure cc: Mr. James P. O'Reilly (w/o enclosure)

Regional Administrator Region II A ~A 0

Mr. D. J. Burke (w/o enclosure) _f'_ ~.'"J\\

NRC Resident Inspector Surry Power Station , - ~ri~70602 7294o 629-:----

K ADOCK 050002ao*

PDR

TABULATION OF CHANGES

e e TABULATION OF CHANGES DELETE DATED SUBSTITUTE DATED LEP-7 6/83 LEP-7 6/84 LEP-8 6/83 LEP-8 6/84 LEP-11 6/83 LEP-11 6/84 LEP-13 6/83 LEP-13 6/84 LEP-14 6/83 LEP-14 6/84 LEP-15 6/83 LEP-15 6/84 LEP-16 6/83 LEP-16 6/84 LEP-17 6/83 LEP-17 6/84 LEP-18 6/83 LEP-18 6/84 LEP-19 6/83 LEP-19 6/84 LEP-20 6/83 LEP-20 6/84 LEP-27 6/83 LEP-27 6/84 LEP-28 6/83 LEP-28 6/84 LEP-29 6/83 LEP-29 6/84 KWI Page 43 KWI Page 43 6/84 Plant Drawings Page 2 7/82 Plant Drawings Page 2 6/84 Plant Drawings Page 2a 6/84 Plant Drawings Page 4 7/82 Plant Drawings Page 4 6/84 Plant Drawings Page 6 7/82 Plant Drawings Page 6 6/84 Plant Drawings Page 6a 6/84 Page 4.2-1 7/82 Page 4.2-1 6/84 Page 4.2-5 7/82 Page 4.2-5 6/84 Page 4.2-5a 6/84 Page 4.2-6 7/82 Page 4.2-6 6/84 Page 4.2-6a 6/84 Page 4.2-39 6/83 Page 4.2-39 6/84 Page 5-i 7/82 Page 5-i 6/84 Page 5.2-2 7/82 Page 5.2-2 6/84 Page 5.2-3 7/82 Page 5.2-3 6/84 Page 5.2-3a 6/84 Page 5.2-19 6/83 Page 5.2-19 6/84 Page 5.2-22 6/83 Page 5.2-22 6/84 Page 5.2-23 6/83 Page 5.2-23 6/84 Page 5. 2..,.24 6/83 Page 5.2-24 6/84 Page 5.3-13 7/82 Page 5.3-13 6/84 Page 5.3-14 7/82 Page 5.3-14 6/84 Page 7.5-13 6/83 Page 7.5-13 6/84 Page 7.5-13a 6/84 Page 7.5-21 6/83 Page 7.5-21 6/84 Page 7.5-2la 6/84 Page 7.5-22 6/83 Page 7.5-22 6/84 Page 8.2-2 7/82 Page 8.2-2 6/84 Page 8.4-1 7/82 Page 8.4-1 6/84 Page 8.4-5 7/82 Page 8.4-5 6/84 Page 9-xv 6/83 Page 9-xv 6/84 Page 9.3-12 7/82 Page 9.3-12 6/84 Page 9.6-1 7/82 Page 9.6-1 6/84 Page 9.6-2 7/82 Page 9.6-2 6/84 Page 9.6-3 6/83 Page 9.6-3 6/84

e Page 9.6-5 7/82 Page 9.6-5 6/84 Page 9.6-6 7/82 Page 9.6-6 6/84 Page 9.6-7 7/82 Page 9.6-7 6/84 Page 9.6-8 7/82 Page 9.6-8 6/84 Page 9.6-9 7/82 Page 9.6-9 6/84 Page 9. 6-10 7/82 Page 9.6-10 6/84 Page 9. 6-11 7/82 Page 9. 6-11 6/84 Page 9.6-13 6/83 Page 9. 6-13 6/84 Page 9.6-17 7/82 Page 9.6-17 6/84 Page 9.6-18 7/82 Page 9.6-18 6/84 Page 9.6-20 7/82 Page 9.6-20 6/84 Page 9.6-21 7/82 Page 9.6-21 6/84 Figure 9.6-1 6/83 Figure 9.6-1 6/84 Figure 9.6-2 6/83 Figure 9.6-2 6/84 Figure 9.6-3 7/82 Figure 9.6-3 6/84 Figure 9.6-3a 6/84 Figure 9.6-4 7/82 Figure 9.6-4 6/84 Figure 9.6-4a 6/84 Page 9.8-5 6/83 Page 9.8-5 6/84 Page 9.10-23 6/83 Page 9.10-23 6/84 Page 9. 11-1 7/82 Page 9.11-1 6/84 Page 9.11-2 7/82 Page 9.11-2 6/84 Page 9.ll-2a 6/84 Page 9.11-5 6/84 Page 9.11-6 6/84 Page 9.11-7 6/84 Page 9.11-8 6/84 Page 9.11-9 6/84 Page 9.11-10 6/84 Page 9.11-11 6/84 Page 9.11-12 6/84 Page 9. 11-13 6/84 Page 9.11-14 6/84 Page 9.11-15 6/84 Figure 9. ll-2A 6/84 Figure 9. ll-2B 6/84 Figure 9.ll-2C 6/84 Figure 9. ll-2C 6/84 Figure 9. ll-2D 6/84 Figure 9. ll-2E 6/84 Figure 9. ll-2F 6/84 Figure 10. 3-24 6/83 Figure 10.3-24 6/84 Page 11. 2-6 7/82 Page 11. 2-6 6/84 Page 11. 3-11 7/82 Page 11. 3-11 6/84 Page 11. 3-18 6/83 Page 11. 3-18 6/84 Page 11. 3-22 6/83 Page 11. 3-22 6/84 Page 11. 3-22a 6/83 Page 11. 3-22a 6/84 Page 11.3-41 7/82 Page 11. 3-41 6/84 Page 12-iv 6/83 Page 12-iv 6/84 Page 12. 2-11 7/82 Page 12. 2-11 6/84 Page 12.2-12 6/83 Page 12.2-12 6/84 Page 12.2-12a 6/84 Page 12.2-14 6/83 Page 12.2-14 6/84 Page 12.2-15 6/83 Page 12.2-15 6/84

Page 12.2-16 6/83 Page 12.2-16 6/84 Page 12.2-17 6/83 Page 12.2-17 6/84 Page 12.2-18 6/83 Page 12.2-18 6/84 Page 12.2-19 6/84 Figure 12.2-1 6/83 Figure 12.2-1 6/84 Figure 12.2-2 6/83 Figure 12.2-2 6/84 Page 12.3-1 6/83 Page 12.3-1 6/84 Page 12.3-2 6/83 Page 12.3-2 6/84 Page 12.3-3 6/83 Page 12.3-3 6/84 Page 12.3-4 6/83 Page 12.3-4 6/84 Page 12.3-5 6/83 Page 12.3-5 6/84 Page 12.5-1 7/82 Page 12.5-1 6/84 Page 15-viii 7/82 Page 15-viii 6/84 Page 15.1-1 7/82 Page 15.1-1 6/84 Figure 15.5-17 6/84 Figure 15.5-18 6/84 Figure 15.5-19 6/84

,_.ftr2 St>

-* <f!J"LfO 7 ~b 0?--.7.2 4.2 SYSTEM DESIGN AND OPERATION GENERAL DESCRIPTION coolant system consists of three similar heat transfer loops parallel to the reactor vessel. Each loop contains a steam pump, two loop stop valves, loop piping, and instrumentation.

The pres \izer surge line is connected to one of the loops on the reactor side of a stop valve. Auxiliary system piping connections into the reactor are provided as necessary. A flow diagram of the system is and 4.2-2.

boundary shown on the flow diagram indicates those major the containment. The intersection of indicates a containment penetration.

Reactor coolant and component design data are listed in

• - Table!34.l-1 through 4.1-7.

Ftessure in the system is led by the pressurizer, where water and steam pressure are maintained through the use of sprays and electrical heaters. Steam can be formed by the heate s or condensed by pressurizer spray to minimize pressure variations due to con~ction and expansion of the coolant. Instrumentation to be used in the ~ssure control system is described in Chapter 7. Spring-loaded code safe valves and power-operated relief valves are connected to the pressurizer discharge to the pressurizer relief tank, where the discharged steam condensed and cooled by mixing with water.

4.2.2 COMPONENTS 4.2.2.1 Reactor Vessel The reactor vessel is a cylinder with a hemispherical bottom nd a flanged and gasketed removable upper head. Figure 4.2-3 is a schematic f the reactor vessel. The materials of construction of the reactor vessel are g*ven

SPS UFSAR 4.2-2 in Table 4.2-1. Provision is made for removal of reactor internals if required during reactor life.

Coolant enters the reactor vessel through inlet nozzles in a plane just be.~ow the vessel flange and above the core. The coolant flows downward through the annular space between the vessel wall and. the core barrel into a plenum at the bottom of the vessel, where it reverses direction.

Approximately 95% of the total coolant flow is effective for heat removal from the core. The remainder of the flow includes the flow through the control rod assembly guide thimbles, the leakage across the fuel assembly outlet nozzles, and the flow deflected into the head of the vessel for cooling the upper flange. All the coolant is united and mixed in the upper plenum, and the mixed coolant stream then flows out of the vessel through exit nozzles located on the same plane as the inlet nozzles.

The reactor vessel contains the core support assembly, upper plenum assembly, fuel assemblies, control rod assemblies, surveillance specimens, and incore 'inst~umentation. The reactor vessel internals are designed to direct the coolant flow, support the reactor core, and guide the control rod assemblies in the withdrawn position *.

• The reactor internals are described in detail in Section 3.5, and the general arrangement of the reactor vessel and internals is shown in Fi,gure 3.5-2. Design data are listed in Table*4.1-2.

A one-piece thermal shield, concentric with the reactor core, is located between the core barrel and the reactor vessel. The shield is bolted and welded to the top of the lower core barrel. The shield, which is cooled by the coolant on its downward pass, protects the vessel by attenuating much of the gamma radiation and some of the fast neutrons that escape from the core.

This shield reduces thermal stresses in the vessel that result from heat generated by the absorption of gamma energy. It is illustrated in Figure 3.5-5 and described in Section 3.2.3.

Fifty incore instrumentation nozzles are located on the lower head. The reactor vessel closure head and the reactor vessel flange are joined by 58 e

SPS UFSAR 4. 2-5

• Intermediate and lower shells formed and welded but not welded together, not clad.

Lower head dome welded to transition ring, not clad.

4.2.2.2 Pressurizer The general. arrangement of the pressurizer is shown in Figure .4. 2-4, and the design data are listed in Table 4.1-3.

The pressurizer maintains the required reactor coolant pressure during steady-state operation, limits the pressure changes caused*by coolant thermal expansion and contraction during normal load transients, and prevents the pressure in the reactor coolant system from exceeding the design pressure.

The pressurizer contains replaceable direct immersion heaters, safety and relief valves, a spray nozzle and interconnecting piping, valves, and

• instrumentation. The electric heaters are located in the lower section of the vessel, and maintain the pressure of the reactor coolant system by keeping the .,

water and steam in the pressurizer at system saturation temperature. The heaters are capable of raising the temperature of the pressurizer and contents at approximately 55°F/hr during reactor startup.

The pressurizer is designed to accommodate positive and negative surges caused by load transients. The surge line attached to the bottom of the pressurizer connects the pressurizer to the hot leg of a reactor coolant loop.

During a positive surge, caused by a decrease in load, the spray system, which is fed from the cold leg of a coolant loop, condenses steam in the vessel to prevent the pressurizer pressure from reaching the operating point of the power-operated relief valves. Power-operated spray valves on the pressurizer limit the pressure during load transients. In addition, the spray valves can be operated remote manually from the control room. A small continuous spray flow is provided to ensure that the pressurizer liquid is homogeneous with the coolant and to prevent excess cooling of the spray and surge line piping.

L

SPS UFSAR 4.2-6 During a negative pressure surge, caused by an increase in load, flashing of water to steam and generation of steam. by automatic actuation of the heaters keep the pressure above the minimum allowable limit. Heaters are also energized on high water level during positive surges to heat the subcooled surge water entering the pressurizer from the reactor coolant loop.

A Westinghouse Owners' Group analysis has determined that the minimum re'quirement to maintain natural circulation in a three-loop plant with a 3

pressurizer volume of 1300 ft is 125 kW of heater capacity. Two backup heater groups rated at 250 and 200 kW and their associated controls are energized from redundant emergency buses Hand J, which are capable of being fed from either offsite power or emergency power. The Class IE interfaces for motive and control power are protected by safety grade circuit breakers.

l The pressurizer heaters are not automatically shed from the emergency power sources upon the occurrence of a safety injection actuation signal.

Existing diesel-generator loading indicates that with the pressurizer heaters energized at any time following a LOCA with a loss of offsite power, each diesel generator will be within its continuous kW rating. Procedures have been implemented to instruct the operator in the use of pressurizer heaters in establishing and maintaining natural circulation.

The pressurizer is constructed of carbon steel, with internal surfaces clad with austenitic stainless steel. The heaters are sheathed in austenitic stainless steel.

The pressurizer vessel surge nozzle is protected from thermal shock by a thermal sleeve. A thermal*sleeve also protects the pressurizer spray nozzle connection. A manway is provided in the top portion of the pressurizer.

4 *.2.2.3 Steam Generators A steam-generator repair program was completed at the Surry Power Station in 1979 and 1980 for Units 2 and 1, respectively. The purpose of the program was to repair degradation caused by corrosion-related phenomena and to restore the integrity of the steam generators to a level equivalent to new equipment.

REVISION 1 6/83 SPS UFSAR 4.2-39 Table 4.2-2 e REACTOR COOLANT SYSTEM WATER CHEMISTRY SPECIFICATION Electrical conductivity Determined by the concentration of boric *acid and controlled chemicals present. Expected range is from less than 1 to 40 Mhos/cm at 25°C.

Solution pH Determined by the concentration of boric acid and controlled chemicals present. Expected values range between 4.2 (high boric acid concentration) to 10.5 (low boric acid concentration) at 25°C.

Oxygen, ppm, max. 0.1 Chloride, ppm, max. 0.15 Fluoride, ppm, max. 0.15 Hydrogen, cm3 (STP)/kg H o 25.:35 2

Total suspended solid,

• ppm, max.

7 pH control agent (Li 0H) 1.0 0.3 X lQ- 4 to 3,2 X 10- 4 molal strong base alkali ,

equivalent to 0.22 to 2.2 ppm Li 7

Boric acid, as ppm B Design range from Oto -3000

REVISION 1 6/83 SPS UFSAR 4.2-40 Table 4.2-3 RADIATION-INDUCED INCREASE IN TRANSITION TEMPERATURE FOR A302B STEEL Neutron Change in 2

Temp., Exposure, n/cm NDTT, References a Material OF ( >l MeV) OF

1. NRL Report 6160, p. 12 SA302B 450 5 X 1018 140

2. NRL Report 6160, p. 12 SA302B 550 5 X 1018 65 19

3. NRL Report 6160, p. 13 SA302B 490 1.4 X 10 200

4. ASTM-STP 341, P* 226 SA302B 550 6 X 1017 30b

5. ASTM-STP 341, p. 226 SA302B 550 6 X 1017 45

6. ASTM-STP 341, p. 226 SA302B 550 8 X 1018 85b

7. ASTM-STP 341, p. 226 SA302B 550 8 X 1018 100

8. ASTM-STP 341, p. 226 SA302B 550 1.5 X 1019 130b 9.

10.

11.

ASTM-STP 341, p. 226 NRL report 6160, p. 6 Nuclear Science and Engineering SA302B All steels 550 450 1.5 X Various 1019 140 Various 19: 18-38 (1964) SA302B 450 Various Various

12. Quarterly Report of Progress, "Irradiation Effects on Reactor Structural Materials" 19 11-1-64/1-31-64 SA302B 550 3 X 10 120

13. Quarterly Report of Progress, "Irradiation Effects on Reactor Structural Materials" 19 11-1-64/1-31-64 SA302B 550 3 X 10 135 aApplicable. to Figure 4.2-9.

bT ransverse specimens.

SPS UFSAR 5.2-1 5.2 CONTAINMENT ISOLATION 5.2.1 DESIGN BASES The containment isolation system has the following design bases:

1. During incident conditions, at least two barriers exist between the atmosphere outside the containment structure and

a. The atmosphere inside the containment structure.

b. The reactor coolant and connecting systems.

2. The design pressure of all piping and connecting components within the isolation boundary is greater than the design pressure of the containment, 45 psig.

3. The failure of one valve or barrier does not prevent isolation *

• 4.

5.

The operation of the containment isolation system is automatic.*

All isolation valves and equipment are protected from missiles and water jets originating from the reactor coolant system.

6. All remotely actuated and automatically operated isolation valves have their positions indicated in, and can be operated from, the control room.

7. Containment isolation system valves are located so as to require a minimum length of piping between the isolation valves and their penetrations.

8. Special consideration is given to the design of the low-head safety injection and recirculation spray pump inlet lines, in that highly reliable components are used in a single valve arrangement, which is enclosed in a special valve pit.

SPS UFSAR s.2~2 For isolation, the two-barrier valving arrangements consist of the following:

1. Two automatic isolation valves, normally one on each side of the containment wall. In specific cases such as the containment vacuum system the two isolation valves are located outside the containment.

2. An automatic isolation valve and a membrane barrier.

3. An administratively controlled, manually operated valve outside the containment, and a sealed system inside the containment.

4. Two administratively controlled, manually operated valves, one on each side of the containment wall.

5. A sump recirculation pipe and valve arrangement, conservatively designed and fabricated, and enclosed by a special valve pit.

The suction lines for the low-head safety injection pumps and the recirculation-spray pumps are designed to prevent gross system leakage.

major portion of this piping is buried in the reinforced-concrete base mat, The and only a short length of piping exists between the mat and the isolation valve. This valve is equipped with a reliable remote operator. The design of this portion of the installation is compatible with letters from the Advisory Committee on Reactor Safeguards to the U. S. Atomic Energy Comm1ss. i on. 1,2 Provisions for detecting leaks in these suction lines are described in Chapter 6.

A membrane barrier consists of either pipe, tubing, or a component wall.

An incoming line from a centrifugal pump or a surge tank is considered an open line and a check valve is used in incoming lines instead of an automatic isolation valve. However, check valves, singly or in pairs, are not used to provide the only means for isolating a penetrating line.

The criteria applied to the various functional classes of piping to implement the design bases are as follows.

SPS UFSAR 5.2-3

1. Class I piping is open to the outside atmosphere, and it is connected e to the reactor coolant system, or a connecting system, or is open to the containment atmosphere. An-example is the line from the reactor containment sump pumps to the waste drain tanks. For Class I piping, the following is provided for isolation subsequent to a LOCA:

a. Incoming lines with a check valve and an automatic trip valve.

b. Outgoing lines with two automatic trip valves.

2. Class II piping is connected to a closed system outside the contain-ment, and it is connected to the reactor coolant system, or a connecting system, or is open to the containment atmosphere. An example is the excess letdown line. For Class II piping, the -

following is provided for i_solation subsequent to a LOCA:

a. Incoming lines with one automatic trip valve.

b. Outgoing lines with one automatic trip valve *

• 3. Class III piping is connected to open systems outside the contain-ment, and it is separated from the reactor coolant system, or a connecting system, and the containment atmosphere by a closed valve under administrative control or by a membrane barrier. Examples are the component cooling~water lines. For Class III piping, the following is provided for isolation subsequent to a LOCA:

a. Incoming lines with one check valve and a valve under administrative control.

b. Outgoing lines with one automatic trip valve.

4. Class IV piping must remain open after a LOCA. An example is the high-head safety injection/ charging pump header to the reactor coolant system. For Class IV piping, the following is provided for isolation subsequent to a LOCA:

a. Incoming lines with one remote manual valve.

SPS UFSAR 5. 2-4.

b. Outgoing lines with one remote manual valve. This piping is missile-protected.

5. Class V piping is connected to normally closed systems outside the containment, and it is separated from the reactor coolant system; and connecting systems, and the containment atmosphere by a closed valve and/or by a membrane barrier. An example is the service-air line.

For all normally closed Class V lines, a normally closed, manual valve outside of the containment constitutes the isolation barrier.

Where check valves are used as isolation valves, consideration has been given to the ability of these check valves to prevent the leakage of air into the containment when the containment atmospheric pressure is negative.

Check valves in the containment spray and recirculation spray systems are positive-closure check valves. These valves have an external, adjustable counterweight, set to maintain the disk tightly*sealed during certain phases of.accident conditions when the containment atmospheric pressure is slightly negative. These types of check valves are provided in these systems because they are open to the containment atmosphere through the spray nozzles when the systems are isolated after an accident~

Check valves used for isolation purposes in other pipelines, normally those containing water, are ordinary check valves. They do not have the positive closure feature because they are in series with an automatic trip valve or a valve under administrative control. This arrangement would require a double failure; that is, a failure of the automatic trip valve to close and a rupture in the line downstream from the check valve, which would cause the water leg normally holding the check valve closed and sealed to be drained.

This allows outward leakage past the check valve if the check valve fails to seal tightly with a small differential air pressure.

A monitoring arrangement is provided to test the leaktightness of each automatically actuated trip valve and check valve. Valve arrangements for each class of penetration are depicted in Figure 5.2-1.

REVISION 1 6/83 SPS UFSAR 5.2-5 Instrumentation and adjunct control circuits associated with automatic isolation valve closure are fail-safe (initiate closure) upon loss of voltage and/or control air. Most isolation valves are air-to-open, spring-return, and diaphragm-operated, thus providing a fail-safe design; however, certain isola-tion valves have been changed to piston-operated or direct acting, electric 11 solenoid valves. The automatic isolation valves inside the containment will function properly under all containment atmospheric pressures. The air-operated isolation valve closing force is provided by a spring; control air is applied to the diaphragm of the isolation valve to open it. To close the isolation valve, an electrically operated solenoid valve located in the air supply line to the isolation valve operator is actuated to vent the control air applied to the isolation valve diaphragm through the solenoid to the containment atmosphere, causing the spring to close the automatic isolation valve. The spring side of the isolation valve diaphragm is always vented to the containment atmosphere. Under normal containment atmospheric conditions, venting the spring side of the isolation valve diaphragm to the partial vacuum causes a force on the diaphragm tending to open the valve (the

• normal valve position under this containment condition). The loss of control air, however, will force the valve to close under all conditions. I1 Under accident conditions, the containment pressure is positive and the solenoid valve vents the control air to the containment atmosphere. Because both sides of the isolation valve diaphragm are vented, balanced forces on either side of the diaphragm result, allowing the spring to close the automatic isolation valve. Circuits that control redundant automatic valves are redundant in the sense that no single failure will preclude isolation.

Means are provided to periodically test the functioning of the automatic isolation equipment such as the setpoint of sensors, the speed of response, and the operability of fail-safe features. The containment isolation instrumentation is discussed in Section 7.5.

It should be noted that isolation valves actuated by electric motors upon I1 electrical failure fail in the as-is *position.

The air-operated trip valves in the reactor coolant sample system and the I1 residual heat removal sample systems have been replaced with direct acting

REVISION 1 6/83 SPS UFSAR 5.2-6 electric solenoid valves. This change was made to ensure that the valves could be reopened to draw a sample under single failure criteria, after an accident. An additional electric solenoid valve has been placed in the residual heat removal sample line inside containment and functions as an extra j1

'isolation valve. The function of the trip valves has not changed.

The previous steam-generator blowdown trip valves have been replaced with new 2-in., double disk, pressure seal-type gate valves. The valves are of sufficient size to meet the maximum allowable pressure-drop requirement at design flow rate specified for the original trip valves and will minimize the 1 occurrence of cavitation. The pressure seal-type, body-to-bonnet design

'provides an improved seal. Due to increased stroke-length requirements of the new gate* valves, the air-to-open, spring-return diaphragm actuators have been replaced with corresponding piston-operated actuators.

5.2.2 ISOLATION DESIGN The general criteria covering the number and location of isolation valves required to ensure reactor containment integrity during LOCA conditions are provided in Section 5. 2 .1. Table 5. 2-1 summarizes the major piping penetrations through the reactor containment for each fluid system and the specific types of isolation valves that are provided for each penetration.

Valve positions during normal operation, shutdown, and accident conditions are also listed.

Containment isolation is accomplished under the following conditions:

1. Phase-1 isolation is initiated by a safety injection actuation signal. Safety injection is actuated by any one of the following input signals (Section 7.5):

a. High steam-line flow with low steam-line pressure or low-low T

avg

b. High steam-line differential pressure.

SPS UFSAR 5.2-21 Table 5. 2-2 ESSENTIAL SYSTEMS - LEVEL 1 Valve Position After Safety Injection System sistem DescriEtion Actuation a High-head safety injection .to the cold leg Open High-head safety injection to the hot leg Closed Low-head safety injection to the cold leg Open Low-head safety injection to the hot leg Closed Low-head safety injection from the sump Closed Seal-water injection to reactor coolant pump Open b' .

Containment-spray pump discharge Closed * ,../

b Recirculation-spray suction Open b

Recirculation-spray discharge Open

• Service water into the recirculation-spray heat exchanger Closedc aisolation valves designated as "closed" receive a signal to close immediately after safety injection actuation and are opened by the operator or automatic controls at some point in time following a LOCA.

b Valves are opened on high-high c.ontainment pressure. Valves can then be selectively closed after it is established that the system is no longer required.

cValves remain closed on safety injection system (Phase 1) containment isolation. Valves open on a containment depressurization actuation, containment isolation Phase 3.

SPS UFSAR 5.2-22 Table 5.2-3 ESSENTIAL SYSTEMS - LEVEL 2 Mode of Containment System Description Isolation Component cooling from reactor coolant Phase 3 pump thermal barriers Component cooling from containment air . Phase 3 recirculation cooling coils; Component cooling to reactor coolant Phase 3 pump motor Component cooling from reactor coolant Phase 3 pump motor Main steam relief Setpoint pressure Auxiliary feed water (a)

Component cooling water return from Phase 1 residuai heat removal heat exchanger aClosed system inside and check valve outside provide containment isolation.

SPS UFSAR 5.2-23 Table 5 .2-4 NONESSENTIAL SYSTEMS Mode of Containment System Description Isolation Chemical and volume control system, Phase 1 charging mode Charging system letdown Phase 1 Reactor coolant pump seal-water supply Administratively controlled, normally open Reactor coolant pump seal-water return Phase 1 Containment-air radiation monitor Phase 2 sample Pressurizer relief tank gas and liquid Phase 1 space samples Primary coolant hot-leg sample Phase 1

• Primary coolant cold-leg sample Pressurizer vapor space sample Residual heat removal sample Phase 1 Phase 1 Phase 1

• "I Containment instrument-air suction Phase 1 Safety injection accumulator makeup Administratively controlled, normally closed Residual heat removal return to refueling- Administratively controlled, water storage tank normally closed Steam generator wet layup Administratively controlled, normally closed Primary drain transfer discharge Phase 1 Containment sump pump discharge Phase 1

_Steam generator blowdown Phase 1 Service air Administratively controlled, normally closed Primary-grade water Phase 1

SPS UFSAR 5.2-24 Table 5.2-4 (continued)

NONESSENTIAL SYSTEMS Mode of Containment System Description Isolation Reactor coolant loop fill Administratively controlled, normally closed Primary vent header Phase 1 Containment vacuum system suction Phase 1 Nitrogen to pressurizer relief tank Phase 1 Primary vent pot vent Administratively controlled, normally closed Containment leakage monitoring Phase 1 Condenser air ejector vent Phase 1 Containment purge air ducts Administratively controlled, normally closed Containment air ejector suction Administratively controlled, normally closed Pressurizer dead weight, calibrator Administratively controlled, normally closed Refueling cavity purification Administratively controlled, normally closed Accumulator tanks test line Phase 1 Feed water chemical addition (a)

Main steam isolation valves (b)

(shares penetration with main steam relief lines)

Feed water (c)

Auxiliary feed water (c) a Closed system inside provides containment isolation.

b Isolated high steam-line flow with low steam-line pressure, or low-low T

avg C

Closed system inside and check valves inside and outside provide containment isolation.

SPS UFSAR 5.3-13 this subatmospheric pressure. The containment is designed to resist the external pressure w:i,thout new technology. The concept provides for an increase in unit safety through the reduction in possible activity release.

The need for charcoal air recirculation filters and fans operating in a steam environment is eliminated and dependence is placed on a redundancy of simple spray systems.

The advantages of the subatmospheric containment are best realized if pressure is reduced rapidly to a subatmospheric level. A review of various heat removal systems indicates that the injection of cold water into the containment is the most rapid and dependable means of depressurization. This water must be borated when used in conjunction with shim-controlled reactors, since it will ultimately be recirculated to the reactor for core cooling.

Depressurization is accomplished through the combined operation of the con-tainment spray, which provides a cold-water heat sink, and the recirculation spray, which removes heat from the containment.

• Transient heat transfer calculations required to size the spray systems and consider the effort of static heat sinks are well understood.

Figure 5. 3-3 shows typical pressure transient curves for comparable atmospheric and subatmospheric containments following a LOCA. As shown, the pressure in the atmospheric type of containment design drops to a low value, but outleakage continues indefinitely, while outleakage is terminated in the subatmospheric containment type as soon as pressure is reduced below atmospheric.

REVISION 1 6/83 SPS UFSAR 7. 5-13 information for evaluating the conditions necessary to initiate the recirculation mode of operation.

The containment level instrumentation has been changed to meet the requirements of TMI-2, NUREG-0578, Section 2.1.9. The system now utilizes redundant wide range level transmitters and redundant narrow range transmitters installed inside the reactor containment. The transmitters are qualified to IEEE-323-1974 and IEEE-324-1975 standards.

One of the redundant loops for wide-range level and one of the loops for narrow-range level is recorded on the postaccident monitoring recorder in the control room. The second redundant loop for both wide- and narrow-range level can be recorded in the Technical Support Center.

Depending upon the magnitude of the loss-of-coolant incident, information relative to the pressure of the reactor coolant system will be useful to the operator to determine which pumps will be used for recirculation in the event of a small break. The discharge pressure of the charging pumps, as read on instrumentation outside the containment, will serve this purpose.

Consideration has been given to all the instrumentation and information that is necessary for recovery following a loss-of-coolant incident.

Instrumentation external to the reactor containment, such as radioactivity monitoring equipment, will not be affected by this postulated incident, and will be available to the operator.

Seismic Class I, postaccident monitoring panels for Units 1 and 2 have been installed in the control room, in response to NUREG-0578, Sections 2.1.5a, 2.1.6b, and 2.1.9. The panels are designed to IEEE 344-1975 and the original plant separation criteria. The components have been designed to 1 meet, as a minimum, IEEE 323-1971. The panel contains switches and indicators for equipment such as containment isolation valves, RCS vent valves, and postaccident hydrogen indicators.

e

REVISION 1 6/83 SPS UFSAR 7 .5-14 Ii 7.5.2.3 Calibration and Testing The engineered safeguards actuation channels are designed with sufficient redundancy to provide the capability for channel calibration and test during

• power operation. Bypass removal of one actuation channel is accomplished by placing that channel in a tripped mode; i.e., a two-out-of-three matrix logic becomes a one-out-of-two matrix logic. Testing does not trip the system unless a trip condition occurs in a concurrent channel.

7.5.2.3.1 Analog Channel Testing Engineered safeguards analog channel testing (as shown in Figure 7.5-5) is identical to process analog protection channel testing as described in Section 7.2.2.1.4.

7.5.2.3.2 Logic Testing Figures 7.5-2, 7.5-3, and 7.5-4 illustrate the basic logic test scheme .

Test switches are located in the associated relay racks rather than in a single test panel.

the logic matrices:

The following procedures indicate the method of testing *

1. Test of either train A or train Bis made at one time; this is under administrative control.

2. A selection of the function to be tested is made. Figure 7.5-3, for example, illustrates some of these functional matrices.

3. The relay logic test switch is first turned to the test position, which opens the circuit to the master actuating relay (logic test switches as shown in Figure 7. 5-3 or location 1 as shown in Figures 7.5-2 or 7.5-4) and energizes the "on test" labeled lamp (see position 3 of switch c1 or n1 in Figure 7.5-4).

The master actuating relay is removed from this part of the test in order to avoid unintentional starting of the engineered safeguards

REVISION 1 6/83 SPS UFSAR 7.5-21 The pressure sensors that monitor containment conditions subsequent to a LOCA are capable of indicating pressures from O psia to 65 psia. The temperature sensors that monitor containment conditions subsequent to a LOCA are capable of indicating temperatures from 0°F to 300°F. The pressure and temperature sensors that monitor containment conditions subsequent to a LOCA are capable of indicating conditions more severe than those associated with the design basis of the containment. The pressure and temperature conditions for the design basis of the containment are 45 psig and 280°F.

In addition to the above monitors, redundant pressure monitors capable of measuring a pressure range of three times the containment design pressure (0 to 180 psia) have been installed on Unit 1 in response to NUREG-0578, Section 2.1.9, and subsequent NRC clarifications. A similar Unit 2 installa-tion is expected to be operational in 1983. The transmitters were furnished by Rosemont Incorporated and are qualified to IEEE-323-1971 and IEEE-344-1975.

1 Each transmitter has an indicator associated with it. These indicators are mounted on the main control board and provide continuous indication of the containment pressure over the range of Oto 180 psia. One of the redundant loops for the containment pressure measurement is recorded in the control room on the postaccident monitoring recorder. The second redundant loop can be modified to record in the Technical Support Center.

7.5.3.5.1 .Environmental Qualification of Safety-Related Electrical Equipment In response to IE Bulletin 79-0lB, a program was established to review the environmental qualfication of safety-related electrical equipment located inside the containment. Later, IE Bulletin 79-0lB was issued and further defined the scope of the review to include not only equipment inside the containment, but also equipment in areas of the plant where changing environmental conditions (temperature, pressure, humidity, radiation) occur during and as a result of the accident conditions being reviewed.

The IE Bulletin 79-0lB review was submitted in two separate parts, The 1

45-Day Review, reflected equipment qualifications to FSAR commitments. The e review included a list of safety related systems that are required to achieve or support (1) emergency reactor shutdown, (2) containment isolation,

REVISION 1 6/83 SPS UFSAR 7.5-22 (3) reactor core cooling, (4) containment heat removal, (5) core residual heat removal, and (6) pr~vention of significant release of radioactive material to the environment. This list is included on Table 7-. 5-3. Equipment identified as requiring a review were analyzed for conditions of temperature, pressure

. and humidity inside and outside the containment, and for submergence, aging, chemical spray, and radiation.

2 Revision 4 to the 90-day review, was also submitted. It included a list of all electrical equipment required to mitigate an accident and/or safely shut down the plant and that are subjected to a changing environment due to the accidents. The report reflects the updated Status of Qualification of the electrical equipment. Results of the NRC's safety evaluation for the environmental qualification of safety-related equipment at the Surry Power 1 Station are contained in Reference 3.

REVISION 1 6/83 SPS UFSAR 7.5-22a

7.5 REFERENCES

1. Letter from Vepco to NRC,

Subject:

45-Day Response to IE Bulletin 79-0lB, dated June 16, 1980. (Serial No. 527).

2. Letter from Vepco to NRC,

Subject:

Revision 4 of 90-Day Response to IE Bulletin 79-0lB, dated August 24, 1982.

3. Letter from S. A. Varga, NRC, to W. L. Stewart, Vepco,

Subject:

Transmittal of the Safety Evaluation Report for Environmental Qualifi-cation of Safety-Related Equipment at Surry Power station, Unit Nos. 1 1 and 2, dated January 26, 1983

• SPS UFSAR 8.2-1 8.2 DESIGN BASES The electrical systems are designed to supply electrical power tb all essential unit equipment during normal operation and under incident conditions.

The main generator, described in Section 10~3.3, establishes the facility operating limits and requires the plant to be operated between a 0.90 lagging and a 0.95 leading power factor. As system reactive load changes, generator excitation can be adjusted to ensure operation within the required power factor limits. The system grid also has banks of inductors that can be connected in order to adjust the power factor.

Electrical system components vital to unit safety, including the diesel generators, are designed and protected as necessary so that their integrity is not impaired by potential earthquakes, high winds, floods, or disturbances on the external electrical system. Cables, motors, and other electrical equipment required for operation of the engineered safeguards are suitably protected against the effects of either a nuclear system accident or a severe external environmental phenomenon, in order to ensure a high degree of reliability. The enclosures for motors and electrical switchgear are selected to suit the local conditions and are designed in accordance with specifications issued by the National Electrical Manufacturers Association (NEMA).

Essential electrical equipment components are specified to withstand, without loss of function, the maximum conditions expected during normal operating and postaccident environments, and during operation of the safeguard equipment during the accident. It is. expected that the maximum accident conditions within the containment will be 280°F at 45 psig for 30 min. The environmental qualification of safety-related electrical equipment is discussed further in Chapter 7. Should suitable equipment n:ot be available, the detailed plant design incorporates features to modify the environment to be compatible with the equipment.

SPS UFSAR 8.2-2 In the containment, essential electrical components and conductors are protected from the forces generated during an incident by group separation.

By physically separating each group and providing conductor barriers where necessary, the failure of one group does not jeopardize any other group. In the case of multiple instrument channels in one location, such as the channels associated with the single pressurizer, physical separation is carried out as far

• as possible and the circuitry arranged so that multiple instrument failures are always in the safe direction. Electrical cable connections are run from the instrument transmitter to the area outside the crane support wall using the shortest path while providing separation between redundant channels.

The crane wall acts as a further barrier against any forces generated during an incident.

In general, the 4160-V and 480-V switchgear are of metal-clad deadfront construction with closing and tripping control power taken from the station batteries. Each starter or breaker cubicle is isolated from the adjacent cubicle with metal barriers, and each bus section is physically separated from all others. The main feeds to the 4160-V switchgear from the unit station service transformers are shielded single conductors with neoprene jackets installed in ladder type trays, with 1. 25-diameter spacings between conductors. The main feeds to the 4160-V switchgear from the reserve transformer are of the same construction as the unit transformer feeds, except that part of the run length will be in a duct bank. One reserve transformer

. feeder has separate routing to the 4160-V switchgear, physically isolated from all other transfqrmer secondary leads.

All switchgear associated with engineered safeguards equipment is separated from the main switchgear area and is readily accessible in the main control area.. For all leads supplying engineered safety equipment, the cable is 3/c with interlocked armor overall, run in ladder trays or properly mounted and:~*supported when run external to ladder trays, or 3-1/.c triplexed, run in condtiit, with the exceptions of the 480-V equipment supplied from motor control centers and the emergency generator leads. The only 480-V exceptions are for 30-hp motors .and smaller. The emergency generator leads entering the emergency switchgear room from the duct bank have been derated for cable in conduit in accordance with Insulated Power Cable Engineers Association (IPCEA)

SPS UFSAR 8.4-1 8.4 STATION SERVICE SYSTEMS 8.4.1 4160-V SYSTEM Alternating current station service power is distributed from the 4160-V switchgear. This switchgear is energized from the main generator and unit station service transformers during normal operation, or from the reserve station service transformer source during startup, hot standby, or shutdown operation (Section 8.2). The 4160-V system is duplicated for Unit 2.

The 4160-V switchgear is arranged in three independent bus sections.

Each bus section has a capacity of about 30oo* A. Each feeder or motor circuit is protected by overcurrent relays that trip the associated breaker for a sustained overload or fault.

During unit startup, the total ac demand of the unit is supplied from the reserve. station service source. After the unit has attained opera.ting conditions and the turbine generator is synchronized and connected to the

• system, the station service load is transferred to the unit station service transformers.

transformers.

This transfer is performed without a power interruption by momentarily feeding the 4160-V switchgear from both the reserve and unit The reserve station service transformer is then disco~nected and the turbine generator will supply its own auxiliaries.

To include the possibility o.f two-unit simultaneous loading of the reserve station service (RSS) system, a system was designed to reduce the margin to overload on the RSS system. This system provides for automatic load shedding of selected non-safety-related loads from both units. The scheme ensures that the-voltages available on the emergency buses will be within acceptable limits and that the ampere loading on. the transfer buses and breakers will be no greater than 15% over their continuous rating. A manual override switch is provided in the control room to* allow manual restarting of the shed loads under a controlled condition.

Loss of normal supply to any bus section automatically trips the normal

- source breaker and closes the alternate source breaker. Unit operation with two loops is possible with one section of 4160-V switchgear out of service.

SPS UFSAR 8.4-2

_Motors larger than approximately 300 hp are operated at 4160 V and are arranged for across-the-line starting.* One circuit from each bus section feeds two 4160/480-V station service transformers. The 480-V system is described in Section 8.4.2.

Two independent sections of emergency 4160-V bus and switchgear are provided. Each section is sized to carry 100% of the emergency load. These emergency sections are energized from the reserve station service transformer during normal operation, startup, and shutdown. In the event of total loss of offsite power, the emergency 4160-V buses are isolated from the normal supply and energiz~d from the diesel generators, as described in Section 8.5.

A manually operated air circuit breaker location is provided so that a*

4160-V emergency bus section may be connected to the redundant emergency bus section. This feature is used for maintenance or abnormal conditions only,

.and is under administrative control. This breaker is removed from the cubicle and is not installed when the unit is operating.

8.4.2 480-V SYSTEM The 480-V ac station service system distributes and controls power for all 480-V, 240-V, and 120-V ac station service demands. The source of power for the 480-V ac system buses is from the counterpart 4160-V ac system buses.

The 480-V ac system is divided into three double-ended bus sections, and each section is fed from a counterpart 4160-V ac bus through individual 4160/480-V ac station service transformers. This system is shown in Figure 8 .1-1.

The switchgear is metal-clad, with 125-V de operated air circuit breakers, and arranged with six independent bus sections. The 4160/480-V transformers ate air-cooled. The transformers are throat-connected to the sJ:itchgear.

Normal operation is with the bus sections independent of each other.

Motors up to approximately 300 hp are connected to the 480-V switchgear.

Reduced unit output is possible with two 480-V bus sections out of service.

SPS UFSAR 8.4-5 emergency lighting, as shown in Figure 8.4-2. The principal equipment items e in this system are two 60-cell lead-acid batteries, two static battery chargers, and two battery distribution switchboards. A separate battery, battery charger, and distribution switchboard are available for use in the screenwell structure.

The batteries are of the central power station type and are designed for continuous duty. Each battery consists of 60 cells connected in series. Each cell is of the sealed type, assembled in a shock-absorbing, clear plastic container, with covers bonded in place to form a leakproof seal. The batteries are mounted on protected, corrosion-resistant, earthquake-resistant racks for security and to facilitate maintenance. The two battery areas are separated from each other and from the switchgear room.

Normally, the two battery bus sections are operated independently, with the bus tie breaker open. Each charger supplies power for operation of equipment connected to that bus section and maintains a floating charge on its associated battery. The manually operated bus tie breaker provides for parallel operation of the chargers arid batteries or operation with either battery or charger out of service for maintenance.

The four static battery chargers (two per 125-V-dc bus) are ide~tical, each having an output of 200 A at 132-V de with an input of 440-V ac, three-phase. ~ach charg~r is equipped with a de voltmeter, ammeter, ground detector, ac failure relay, and low charging current alarm relay. Loss of ac or low charging current is alarmed in the control room. Battery ground indicators are located in the control room. Battery voltage is indicated to the operator on the main control board and continuously recorded on recorders located in the control room. The battery chargers are energized from

.emergency motor control centers.

The battery distribution switchboards are NEMA Class II metal-clad structures, each with a 125-V de, two-wire underground main bus, and two-pole manually operated air circuit breakers.

SPS" UFSAR 8.4-6 During normal operation, the 125-V de load is fed from the battery chargers with the batteries floating on the systems. Upon loss of station ac power, the entire direct current load is drawn from the batteries. The power load imposed on each battery will be initially high. After the turbine generator has coasted to a stop, the de auxiliary bearing and seal-oil pump motors may be stopped. .This removes a motor load of approximately 75 hp from the batteries. The remaining load consists primarilr of emergency lighting and vital bus inverters. The batteries are sized for 2 hr of operation, after which it is assumed that station power or emergency generation power will be available to energize the battery chargers. The basis for sizing the station batteries for 2 hr without benefit of any station power is a carryover from the criteria used on nonnuclear power stations where emergency generators were not available to provide power to the battery chargers or turbine auxiliaries for safe coastdown. The batteries will be required for approximately 10 sec between loss of station power and the availability of emergency ac power to supply the battery chargers.

8.4.5 LIGHTING SYSTEM Normal lighting for turbine areas, reactor containments, auxiliary building, fuel building, and service buildings is provided from local lighting cabinets located in the area of service. These cabinets are fed from a double-ended lighting. switchboard that is energized from two independent 250-kVA, single-phase, 4160-240/120-V dry type, self-ventilated transformers.

Normally the two buses of the double-ended switchboard are separate. They are capable of being tied together if one transformer fails.

Normal lighting for the office building and remote areas is supplied through local 480-120/240-V, single-phase, dry type transformers. Emergency lighting for remote areas is provided by local self-contained, battery-powered emergency lighting units.

Emergency lighting for turbine areas, auxiliary building, fuel building, and service buildings is in the form of incandescent units and is normally de-energized. These lights are automatically switched to the de system upon sensing loss of voltage on the lighting switchboard. Emergency lighting for

REVISION 1 6/83 SPS UFSAR 9-xv I1 CHAPTER 9: AUXILIARY AND EMERGENCY SYSTEMS LIST OF FIGURES (continued)

Figure Title 9.8-2 Compressed Air System - Unit 2 9.9-1 Service Water System 9.10-1 Fire Protection System - Arrangement 9.11-1 Well-water System 9 .11-2 Flash Evaporator System 9.12-1 Fuel Transfer System 9.13-1 Ventilation Arrangement - Primary Plant Systems, Sheet 1 9.13-2 Ventilation Arrangement - Primary Plant Systems, Sheet 2 9.13-3 Ventilation Arrangement - Service Building, Sheet 1 9 .13-4 Ventilation Arrangement - Service Building, Sheet 2 9.14-1 Decontamination Facility

• SPS UFSAR 9.3-11 Table 9.3-2 (continued)

RESIDUAL HEAT REMOVAL SYSTEM DESIGN DATA Residual heat exchangers (continued)

Tube (reactor coolant)

Design temperature, °F 400 Design pressure, psig 600 6

Design flow rate, lb/hr 2 X 10 Design inlet temperature, °F 140 Design outlet temperature, °F 124 Material Austenitic stainless steel

SPS UFSAR 9.3-12 Table 9.3-3 RESIDUAL HEAT REMOVAL SYSTEM CHEMISTRY GUIDELINES Chemistry Parametera Requirement pH at 25°C 4.0 - 4.5 Conductivity at 25°C, µmhos <l to 40 Turbidity, ppm 1.0 max B, ppm  ::::::2500 Cl-, ppm 0.15 max F, ppm 0.15 max o , ppm 0.10 max 2

aSampling is performed when the system is in operation.

SPS UFSAR 9.6-1 9.6 SAMPLING SYSTEM The sampling system, as shown in Figures 9.6-1 through 9.6-4, is designed to provide primary and secondary fluid and gaseous samples for laboratory analysis. Conductivity, pH, and dissolved oxygen are monitored at different stages of the secondary cycle. The sampling system also has the capability of obtaining and analyzing postaccident liquid and gaseous samples.

9.6.1 DESIGN BASES 9.6.1.1 Routine Samples Process fluids and gases are representatively sampled for testing to obtain data from which performance of the station, equipment, and systems may be determined.

Routine samples of process fluids and gases associated with both the primary and secondary systems are either taken periodically or are continuously monitored. Two general types of samples are obtained by the sampling system: high-temperature samples (greater than 150°F) such as the reactor coolant system samples, and low-temperature samples (less than or equal to 150°F) such as the high-level waste drain tank samples. The various routine samples taken are listed in Table 9.6-1.

9.6.1.2 High Radiation Sampling System The high radiation sampling system is designed to obtain and analyze representative samples of reactor coolant, the containment atmosphere, and the containment sump in a timely fashion after an accident. Prompt sampling and analysis of reactor coolant and containment atmosphere samples can provide information needed to assess and control the course of an accident. The system provides the ability to obtain grab samples from each reactor coolant hot leg, each reactor coolant cold leg, the residual heat removal system, the

SPS UFSAR 9.6-2 chemical and volume control system mixed-bed demineralizer effluent, the containment sump, and the containment atmosphere, all within 1 hr after an accident. The system has the capability to cool and depressurize samples at high temperature and high pressure to allow grab sampling and in-line chemical an.ilysis.

The system also provides the means to remotely dilute reactor coolant and containment sump samples by a factor of 1000 to reduce the personnel exposure levels that would otherwise be associated with postaccident sampling. This initial dilution also reduces the exposure that would be associated with subsequent manual dilutions, if required.

The diluted and undiluted liquid grab samples and the containment air samples are put into specially designed transfer carts with integral shielding. Placement of the samples inside the shields can be accomplished with minimal operator exposure because the cart is integrally designed to nest within the sample panel. The transfer carts facilitate ease of sample movement to designated areas for isotopic or chemical analysis with low operator exposure.

A feature of the sampling system is the ability to strip reactor coolant of dissolved gases for grab sampling and analysis.

An in-line chemical analysis panel is included to facilitate remote measurement of important chemical parameters with a minimum of manual action or exposure to the operator. This chemical analysis panel has the capability to measure primary coolant pH, boron, oxygen concentration, and hydrogen concentration, as well as containment hydrogen concentration. The capability for in-line chloride measurement is also provided. Each parameter is either indicated or recorded on a remote-control panel located in a separate area of the station.

The high radiation sampling system panels are located within existing space in the auxiliary building. The reactor coolant is drawn from sample system lines outside of containment, upstream of the sample system coolers.

REVTSION 1 6/83 SPS UFSAR 9.6-3 Controls are provided to prevent postaccident samples from being inadvertently e introduced to the normal sample room, Sample liquid resulting from recirculation, purging, and drainage can be routed to the high radiation sampling system waste tank, from which the fluid can be pumped or displaced with nitrogen back to the containment sump.

Connections are provided to recirculate, purge, and drain nonaccident liquid samples via normal sample system flow paths for purposes of operator training and periodic equipment testing.

The containment atmosphere sample panel has the capability to take suction from within the hydrogen monitor system. Motive force for the containment atmosphere sample panel is provided by an integral nitrogen eductor. The discharge of the containment atmosphere panel is routed back to the containment via the existing hydrogen monitor system piping.

The high radiation sampling system and components are designated non-safety-related, and are considered Quality Group D and nonseismic, as defined in Regulatory Guide 1. 26, with one exception: the electrical isolation breakers that allow manual tie-in to the station emergency bus in the event of failure of normal system power are safety-related.

9.

6.2 DESCRIPTION

9.6.2.1 Routine Samples All the sample lines coming from within the containment contain high-temperature samples, with the exception of the pressurizer relief tank sample. Where two or more samples join into a common header (i.e., the primary coolant cold-leg samples), each individual sampling line has a 1 solenoid-operated valve in the line that can be remotely operated from a control board in the auxiliary building sampling room. The primary coolant hot-leg and cold-leg samples flow through delay coils before penetrating the containment. These delay coils permit sufficient decay of nitrogen-16 so that these samples can be handled in the sampling room.

REVISION 1 6/83 SPS UFSAR 9.6-4 All sample lines penetrating the containment have two automatically operated valves in the line, one just inside and one just outside the 1

e containment. These trip valves close on receipt of a safety injection signal.

The high-temperature samples pass through sample coolers located in the auxiliary building sampling room. These coolers cool the high-temperature samples to a temperature low enough for safe handling. Sample flows leaving the cooler are manually throttled and can be directed to a purge line or to the sampling sink. The pressurizer vapor space samples, in addition, pass through capillary tubes that limit the flow of steam.

The sampling lines from sampling points outside the containment but inside the auxiliary building also discharge to the auxiliary building sampli~g sink, with the exception of the liquid waste evaporator sample. This sample line is located near the evaporator, as it requires heat tracing.

Sample lines from sampling points in the turbine building discharge to one of the turbine building sample sinks (one for each unit). The high-temperature samples also pass through sample coolers and are manually throttled. In general, samples can either be directed to a purge line or to the sampling sink. The main steam samples also pass through capillary tubes.

The purge flows of the various samples are discharged to the volume control tank, the vent and drain system, or elsewhere, as appropriate. The radioactive samples in the auxiliary building sampling room discharge into hooded sampling sinks.

In addition to the above facilities for periodic sampling, there are facilities for continuous radiation, pH, and conductivity monitoring of the steam-generator blowdown samples; oxygen, pH, and conductivity monitoring of the condensate pump discharge; and pH and conductivity monitoring of the feedwater. Radiation monitors in the steam-generator' blowdown sample line detect primary-to-secondary leaks in the steam generators. Monitoring of the condensate pump discharge is required for detecting tube leaks in the condensers.

SPS UFSAR 9.6-5 9.6.2.2 High Radiation Sampling System Representative postaccident liquid and gas samples from either reactor unit can be routed to one common high radiation sample system. Samples can be received from the sources listed in Table 9.6-2. The tie-in locations for all

.reactor coolant samples are outside the containment, upstream of the sample system coolers. Since the reactor coolant sample lines are combined into common headers inside containment, one common hot-leg sample and one common cold-leg sample for each unit is routed to the high radiation sampling system liquid sample panel.

The motive force for all reactor coolant samples is primary system pressure. An environmentally qualified containment sump pump is provided to obtain containment sump samples. The motive force for containment atmosphere samples is a nitrogen eductor contained within the containment air sample panel.

• The high radiation sampling system for liquid samples is designed so that samples will be recirculated to purge incoming lines and ensure that the grab samples are representative.

this operation.

The line volume will be purged five times during If the primary system is at operating pressure, this recirculation liquid can be purged to the containment sump without intermediate collection and pumping by the high radiation sampling system waste tank and pump subsystem. For system test and operator training, liquid samples can be recirculated via the normal sample pathways to the appropriate volume control tank or high-level drain tank purge headers.

The high radiation sampling system is comprised of five subsystems.

These are

1. Liquid sample panel and coolers.

2. Containment atmosphere sample panel.

3. Chemical analysis panel.

4. Waste tank and pump.

5. Process control panel.

SPS UFSAR 9.6-6 9.6.2.2.1 Liquid Sample Panel and Coolers The liquid sample panel and coolers perform multiple functions:

,c* l. Sample cooling to about 135°F during the recirculation mode and about 120°F during the grab sample mode.

2. Sample depressurization.

3. Liquid degassing to obtain a representative dissolved gas sample.

• 4. Liquid degassing to the extent necessary to allow in-lin/ chemical analysis downstream.

5. Provides undiluted liquid grab sample inside a shielded transfer cask.

6.

7.

Provides diluted (1000 to 1) liquid grab sample inside a shielded transfer cask.

Provides diluted dissolved gas grab sample inside a shielded syringe.

8. Provides integral shielding to minimize operator exposure while working in front of the panel.

9. Provides a ventilated cabinet, held below atmospheric pressure, to contain potential subsystem leakage. Cabinet ventilation is connected to the auxiliary building HVAC system.

The liquid sample subsystem is divided into three modules, based upon the pressure of the incoming liquid. A reactor coolant module handles hot-leg, cold-leg, and residual heat removal system samples. A demineralizer module handles the chemical volume and control system mixed-bed demineralizer effluent samples. A radwaste module handles the containment sump samples.

SPS UFSAR 9.6-7 The liquid sample subsystem contains provisions for flushing with station primary-grade water. The flush water is routed to the high radiation sampling system waste tank.

9.6.2.2.2 Containment Air Sample Panel The containment air sample panel performs the following functions:

1. Provides the motive force to obtain a representative grab sample of containment atmosphere. A nitrogen eductor is provided that is capable of operation when the containment pressure is either slightly negative or at the maximum postaccident pressure.

2. Provides four shielded sample bombs to obtain containment atmosphere samples on a preprogrammed timer sequence.

3. Provides a motive force by a nitrogen eductor to deliver containment air sample flow to the chemical analysis panel for atmospheric analysis to determine the hydrogen concentration.

4. Provides a means to purge and backflush containment air sample lines back to the affected containment.

5. Provides an integrally shielded panel front to minimize postaccident operator dose rates.

6. Provides a ventilated cabinet held below atmospheric pressure to contain potential subsystem leakage. Cabinet ventilation is connected to the auxiliary building HVAC system. v?

9.6.2.2.3 Chemical Analysis Panel The chemical analysis panel performs the following functions:

1. Accepts a preconditioned, cooled, depressurized and degassed, liquid sample from the liquid sample subsystem for postaccident chemical

SPS UFSAR 9.6-8 analysis for boron, pH, dissolved hydrogen and dissolved oxygen, and hydrogen concentration in postaccident containment atmosphere e samples. Surry also has the capability for in-line chloride analysis of postaccident liquid samples.

2. Provides remote readout of chemical analysis panel parameters on the remote process control panel of the high radiation sampling system.

3. Provides an integrally shielded panel front to minimize postaccident operator dose rates.

4. Provides a ventilated cabinet held below atmospheric pressure to contain potential subsystem leakage. Cabinet ventilation is connected to the auxiliary building HVAC system.

Table 9.6-3 lists the types of instrumentation to be used for determination of postaccident chemical parameters. Instrumentation has been selected based upon the following criteria:

1. The ability to measure accurately the full anticipated range of parameters.

2. The ability to withstand high radiation fields.

3. The ability to reproduce results after calibration.

4. The ability to measure chemical parameters with small sample volumes.

The chemical analysis panel is designed with built-in instrument calibration equipment. Instrument calibration will be performed by station personnel on a periodic basis to maintain a ready condition and to minimize instrument drift.

SPS UFSAR 9.6-9 9.6.2.2.4 Waste Tank, Pumps, and Evacuating Compressor The waste _tank and pumps collect and return system purge and flush liquids to containment. The waste tank and pumps will be bypassed during those periods of line purging when primary system pressure has sufficient motive force to return the purge volume directly to the containment without intermediate collection and pump-out. The liquid sample purge return lines to the containment are routed to the containment sump. The waste tank is sized to hold the volume of liquid residue generated by the acquisition of two postaccident samples.

Two 100%-capacity waste tank pumps are provided to purge the tank contents back to the containment. A nitrogen purge connection is provided to force the contents of the tank back to the containment in the event of pump failure, and also to maintain a nitrogen blanket in the waste tank to preclude accumulation of hydrogen.

The waste tank will be held under a slight vacuum at all times by an evacuating compressor, and will be nitrogen-blanketed. An evacuating compressor is provided to maintain the tank vacuum. A bleed and feed system will control the evacuating compressor and nitrogen purge flow. The evacuating compressor discharges to the containment via the same flow path as the containment return line from the containment air sample panel.

Tables 9.6-4, 9.6-5, and 9.6-6 provide design data for the waste tank, the waste tank pumps, and the evacuating compressor, respectively.

9.6.2.2.5 Process Control Panel The process control panel performs the following functions:

1. Provides remote location in the service building in a low dose rate area for operation of all high radiation sampling system remotely operated valves, with the exception of the routine sample system containment isolation valves, which are operated from the control room.

SPS UFSAR 9.6-10

2. Provides space for chemical analysis panel instrument indicators and recorders.

The process control panel contains a complete system graphic display for the other four subsystems. A communication system is provided between the sample panel area in the auxiliary building, the process control panel in the service building, and the control room.

9.6.2.2.6 Instrumentation Application The chemical analysis panel measured parameters are indicated and recorded on the remote process control panel. Parameters measured are boron concentration, pH, dissolved oxygen, chloride, dissolved hydrogen, and containment air hydrogen concentration. Local flow and pressure indication are on the face of the liquid sample, containment atmosphere, and chemical analysis panels to enable the operator to manually align and adjust system flows.

The process control panel permits remote operation of all high radiation sampling system automatic valves, including routine containment sample system valves, which are normally operated from a panel in the routine sample room.

Isotopic analysis of reactor coolant and containment atmosphere samples will be available within 1 hr of sample acquisition. The postulated activity concentration of postaccident samples is far in excess of the capabilities of normal counting equipment and geometries. Thus, sample dilution will be required prior to analysis. The liquid sample subsystem provides a 1000 to 1 dilution of reactor coolant samples. However, depending upon the accident condition, additional final dilution can be accomplished in a shielded fume hood. The diluted sample can then be analyzed by existing laboratory counting equipment.

The liquid sample subsystem provides a shielded syringe sample of diluted reactor coolant gases that can also be further diluted, if necessary, in the adjacent shielded fume hood. These samples can then be analyzed in existing laboratory counting equipment.

SPS UFSAR 9. 6-11 The containment atmosphere samples are collected in 1-ml shielded sample bombs in the containment atmosphere sample panel. These samples will be isotopically analyzed by a Ge detector, which measures through a 0.25-in.

aperture in the sample vessel lead shield. The 1-ml sample shield apertures are designed to allow measurement in several orientations. Halides and noble gases can be analyzed together. Successive analyses of containment air samples collected on a known time sequence enable the operator to determine the extent of the accident and the effectiveness of the containment spray*

system.

Design conditions of the various sampling panels are given by Tab le 9. 6-8.

9.6.3 DESIGN EVALUATION 9.6.3.1 Routine Samples

• If a critical sampling line becomes inoperable due to some malfunction, there is at least one alternate path that can be used to obtain a similar periodic sample, or for continuous monitoring. For example, if the condensate pump discharge sample line becomes inoperative, condensate can be monitored continuously for conductivity with the local condensate hot-well sampling lines. If one of the steam-generator blowdown radiation monitors malfunctions, a second similar radiation monitor in each unit can be used. If one of the steam-generator blowdown sampling lines becomes inoperative, the condenser air ejector radiation monitor provides indication of a steam-generator primary-to-secondary-side leak.

9.6.3.2 High Radiation Sampling System The high radiation sampling system equipment is designated Quality Group D, nonseismic, as defined in Regulatory Guide 1.26. Seismic failure will not damage station safety-related equipment or the building structures.

REVISION 1 6/83 SPS UFSAR 9.6-12 Electrical power supply is from the normal buses; however, a manual tie-in to the station emergency bus is provided in the event o"f loss of normal e power.

The air-operated trip valves in the residual heat removal sample lines and the reactor coolant system hot-leg and cold-leg sample lines have been replaced with direct-acting solenoid valves. This ensures that the valves can 1 be reopened to draw the sample, under the single-failure criterion after an accident.

The air-operated valves associated with the remaining sample lines (e.g., pressurizer relief tank, gas space sample, etc.) are furnished with dedicated instrument air accumulators so that the ability to open the valves remotely will be available in the event that the station instrument air system is temporarily inoperable. System interlocks are provided throughout to perform the following basic functions:

1. To ensure that samples obtained after an accident can only be returned to the affected containment.

applied to system purge and flush fluids.

A similar philosophy is

2. To ensure that postaccident sample fluid cannot inadvertently enter the routine sample system.

Permanent system connections to the station nitrogen system are provided,

. along with a nitrogen bottle backup system.

Redundant waste tank pumps are provided to pump postaccident samples back to the affected containment. Nitrogen can be used to empty the waste tank in the event of dual pump failure or loss of electric power.

System flush water is obtained from the station's primary-grade water system. Primary-grade water connections to the system are quick-disconnect type. After each use of flush water, the system will be disconnected to minimize the possibility of primary-grade water contamination by postaccident

REVISION 1 6/83 SPS UFSAR 9. 6-13 f1 samples. Each sample acquisition will be followed by a flush to keep background radiation levels to a minimum, in accordance with the ALARA concept.

A shielding analysis has been performed to ensure that operator exposure while obtaining and analyzing a postaccident sample will be less than 3 rem whole-body and 18.75 rem to the extremities. Operator exposure will be accumulated while entering and exiting the sample panel area, operating sample panel manual valves, positioning the grab sample into the shielded transfer carts, and performing additional manual sample dilutions, if required, for isotopic analysis. The major sources of operator exposure are from

1. General auxiliary building background from components not associated with the high radiation sampling system. Operator exposure is limited by the stay time associated with sample panel manual operations, and by selecting entrance and exit routes to the sample room via the lowest dose rate paths .

• 2. Direct radiation from sample lines shielded sample and analysis panels.

that are routed behind the Operator exposure is limited by the integral shielding located in the front of each of the system sample analysis panels. This shielding consists of up to 6 in. of lead shot poured into panel front sections.

3. Backscatter from the walls and roof behind and above the shielded sample and analysis panels. Operator exposure is limited by positioning the panel in an orientation such that the distance from the back of the panel to the nearest wall is maximized to the greatest extent practicable.

For the worst-case assumption of obtaining and analyzing a 1-hr reactor coolant sample, the maximum operator exposure will be less than 2. 5 rem whole-body and 15 rem to the extremities.

In the postaccident state, the high radiation sampling system is capable of obtaining and analyzing samples within time periods listed in Table 9.6-7.

REVISION 1 6/83 SPS UFSAR 9.6-14 11 9.6.4 9.6.4.1 TESTS AND INSPECTIONS Routine Samples Host components are used regularly during power operation, cooldown, and/or shutdown, thus providing assurance of the availability and performance of the system. The continuous monitors are periodically tested, calibrated, and checked to ensure proper

• instrument response and operation of alarm functions.

9.6.4.2 High Radiation Sampling System The high radiation sampling system is designed to be used un.der postaccident conditions, and will not be used regularly during power operation, cooldown, and/or shutdown. Therefore, the system will be tested and maintained on a regular schedule to ensure that all system components are in the ready condition. Station personnel will undergo regular training sessions to ensure familiarity with the function and operation of the system .

The chemical analysis panel instrumentation will be recalibrated and tested on a regular basis to ensure the accuracy and readiness of the instruments.

• SPS UFSAR 9.6-17 Table 9 .6-2 HIGH RADIATION SAMPLING SYSTEM SAMPLE POINTS Number of Sample Points Sample Source For Each Reactor Reactor Coolant

. a Hot leg 4 1ocat1ons Cold leg 3 locationsa RHR loop 2 locationsa eves mixed-bed demineralizer outlet 1 location Containment sump 1 location Containment atmosphere 1 location

• aOne common header from outside the containment is routed to the high radiation sampling system. One sample inlet valve per header is environmentally qualified to be operable after an accident.

SPS UFSAR 9.6-18 Table 9 .6-3 CHEMICAL ANALYSIS PANEL INSTRUMENTATION Instrument or Range of Parameter Method Measurement I. Reactor Coolant and Containment Sump

1. Boron Selective ion 200-2,000 ppm electrode

2. pH Probe, Cole- 1-13 Parmer, or equiv-alent

3. Dissolved oxygen Probe, Yellow 1-20 ppm Springs Instrument

4. Dissolved hydrogen Gas chromatograph, 10-2,000 cc/kg base*line, or equiv-alent

5. Chloride Ion chromatograph, 0-20 ppm Dionex, or equiv-alent II. Containment Atmosphere

1. Hydrogen Gas chromatograph, 0-10%

Baseline, or equiv-alent Table 9. 6-4 HIGH RADIATION SAMPLING SYSTEM WASTE TANK Quantity per station 1 Capacity 17 gal Material of construction Stainless steel Code ASME VIII Design pressure 150 psig Design temperature 150°F

1-SPS UFSAR 9.6-19 Table 9.6-5 HIGH RADIATION SAMPLING SYSTEM WASTE TANK PUMPS Quantity per station 2 Capacity 5 gpm Material of construction Stainless steel Shaft seal Double, mechanical Table 9.6-6 HIGH RADIATION SAMPLING SYSTEM EVACUATING BELLOWS COMPRESSOR Quantity per station 1 Capacity 2 scfm Discharge pressure (max) 40 psig Material of construction Stainless steel Motive device Reciprocating bellows

SPS UFSAR 9.6-20 Table 9.6-7 HIGH RADIATION SAMPLE SYSTEM SAMPLING TIME Sample Source Analyze for Time After Accident I. Reactor coolant Radionuclidesa Extract and analyze (10 Ci/cc to 20 Ci/cc) sample within 1 hr, except spectral,

1. Hot leg Boron which is 2 hr (200-2000 ppm)

2. Cold leg Total gas (oxygen and

3. RHR loop hydrogen)

4. RSC letdown pH (demineralizer (1-14) outlet)

II. Containment atmosphere Radionuclides,a Within 2 hr after hydrogen accident and once per shift thereafter III. Containment sump Same as reactor coolant

~ormal to TID-14844 concentrations.

Same as reactor coolant

SPS UFSAR 11. 3-41 Table 11.3-8 AREA RADIATION MONITORING LOCATIONS, NUMBER, AND RANGES Channel Location (Number) Range (mR/hr) 7 Containment high-range gamma (2) 0.1-10 7

Manipulator crane (2) 0 .1-10 7

Reactor containment area (2) 0.1-10 7

Incore instrument transfer area (2) 0.1-10 7

New fuel storage area (1) 0.1-10 7

Fuel pit bridge (1) 0.1-10 7

Auxiliary building control area (1) 0.1-10 7

Solid waste drum storage and handling area (1) 0.1-10 7

Sample room (1) 0.1-10 7

Main control room (1) 0.1-10 7

Laboratory (1) 0.1-10 7

Spare (2) 0.1-10 7

Decontamination area (1) 0.1-10

REVISION 1 6/83 SPS UFSAR 12-iii CHAPTER 12: CONDUCT OF OPERATIONS TABLE OF CONTENTS (continued)

Section Title 12.8 REVIEW AND AUDIT OF OPERATIONS . * . . . . . * * . . . . . . . 12.8-1

12. 9 EMERGENCY PLANNING. 12. 9-1 12.9 References 12.9-2

REVISION 1 6/83 SPS UFSAR 12-iv CHAPTER 12: CONDUCT OF OPERATIONS LIST OF FIGURES Figure Title 12.2-1 Station Organization, Surry Nuclear Power Station 12.2-2 Corporate Nuclear Operations Organization

SPS UFSAR 12. 2-11 12.2.2.3.2.4 Chemistry Supervisor. The Chemistry Supervisor has the responsibility of ensuring that all routine chemical analyses and evaluations are properly performed during the phases of station operation. He is directly responsible to the Superintendent - Technical Services, and coordinates his efforts with the Superintendent - Operations, the Engineering Supervisors, and the Supervisor of Health Physics.

He also ensures that the chemical treatment of all liquid systems is properly maintained in the station to minimize corrosion products and carryover. All activities in these areas are coordinated with other station groups to ensure full awareness of problems.

He shall have a high school diploma or equivalent and a minimum of 4 years' experience in chemistry. His group shall include individuals with a minimum of 5 years' experience in chemistry, of which a minimum of 1 year shall be*in radiochemistry. A minimum of 2 years of this 5 years' experience should be related technical training. A maximum of 4 years of this 5 years' experience may be fulfilled by related technical or academic training .

Suitable organizational depth should exist to provide for an absence of the principal.

12.2.2.3.2.5 Instrument Supervisor. The Instrument Supervisor reports directly to the Superintendent - Technical Services and coordinates the efforts of his group with the operating maintenance groups.

He is responsible for maintaining adequate and accurate instrumentation status of all station systems, and recalibrating instrumentation and controls.

He is responsible for training his group. He also writes, and when necessary amends, the instrumentation procedures. He is further responsible for maintaining calibrated test instrumentation for tests and experimental work.

He shall have a high school diploma or equivalent and a minimum of 4 ,years' experience in instrumentation. His group shall include individuals e with a minimum of 5 years' experience in instrumentat'ion and control, of which

REVISION 1 6/83 SPS UFSAR 12.2-12 a minimum of 6 months shall be in nuclear instrumentation and control.

minimum of 2 years of this 5 years' experience should be related technical A

e training. A maximum of 4 years of this 5 years' experience may be fulfilled by related technical or academic training. Suitable organizational depth should exist to provide for an absence of the principal.

12.2.2.3.3 Superintendent - Maintenance The Superintendent - Maintenance has the overall responsibility of administering station maintenance activities and stores operations. He reports directly to the Assistant Station Manager. 11 He is responsible for -the maintenance necessary to ensure safe and reliable operation of all station systems not specifically assigned to other groups, and when necessary he directs approved station system modifications.

He is responsible for the implementation of safe working practices associated with the Mechanical, Electrical, and Maintenance Services, and coordinates 1 their group activities with both Operations and Technical Services.

He is responsible for warehouse maintenance, and procurement of spare parts and consumables for the station.

He has the . responsibility of writing, and amending where necessary, maintenance procedures where such documentation is necessary. He is responsible for implementing training programs within his group.

He is responsible for keeping proper maintenance records and files.

He shall have a minimum of 7 years of responsible power plant experience or applicable industrial experience, a minimum of 1 year of which shall be nuclear power plant experience. A maximum of 2 years of the remaining 6 years of power plant or industrial experience may be fulfilled by satisfactory completion of academic or related technical training on a .one-for-one time basis. Further, he should have familiarity with nondestructive testing, craft knowledge, and an understanding of electrical, pressure vessel, and piping codes.

SPS UFSAR 12.2-13 He is a member of the Station Nuclear Safety and Operating Committee.

e 12.2.2.3.3.1 Electrical Supervisor and Supervisor Mechanical Maintenance. The Electrical Supervisor and Supervisor - Mechanical Maintenance are responsible to the Superintendent - Maintenance for supervising all electrical/mechanical maintenance activities. They are responsible for the electrical/mechanical maintenance necessary to ensure safe and prolonged operation of all station electrical/mechanical systems not specifically assigned to other groups and, when necessary, they supervise completion of approved station electrical/mechanical system modifications.

They are responsible for the implementation of safe working, practices within their departments.

Where necessary, they will write and amend maintenance procedures and will keep proper maintenance records and files.

They shall have high school diplomas or equivalent, and a minimum of 4 years of experience in the crafts or disciplines they supervise.

12.2.2.3.3.2 Supervisor - Maintenance Services. He is responsible to the Superintendent - Maintenance for the warehouse, storeroom, procurement of consumables, and maintenance of stock levels for spare parts. He is also responsible for supervision of the laborer force and janitorial services for the station.

He is responsible for supervising all warehousing and storeroom functions. He is responsible for ensuring that all material is properly receipted, stored, and accounted for. He ensures that proper issues are made and the control of all material is maintained. He is responsible for inspections of materials, as required.

He is responsible for the implementation of safe working practices within his department, He is responsible for keeping proper records and files.

REVISION 1 6/83 SPS UFSAR 12.2-14 1

12.2.2.4 Projects e 12.2.2.4.1 Superintendent - Projects The Superintendent - Projects is responsible to the Station Manager for coordination of station activities with Multiple Power Projects and any other agency performing construction or design change activities in compliance with applicable procedures, plans, *and schedules. He is also responsible for the follow-up of activities, including flushing, checkout of systems, and functional testing.

12.2.2.4.2 Supervisor - Planning He is responsible to the Superintendent - Maintenance for the preparation and administration of schedules for*all maintenance and design change activi-ties during unit outages. He provides, through the use of computer data, historical documentation of all maintenance performed on each component installed in the plant.

12.2.2.5 Supervisor - Health Physics The Supervisor - Health Physics is responsible for radiological control of all work conducted on the site. He reports directly to the Station Manager and coordinates his efforts with the Superintendent Operations, Superintendent - Maintenance, Chemistry Supervisor, and the Engineering Supervisors.

He is responsible for keeping records of radiological exposure to all persons working or visiting within the station's restricted areas. This includes the organization of written reports required for Company or regulatory purposes.

He is responsible for conducting regular surveys of the station, and records background radiation levels.

REVISION 1 6/83 SPS UFSAR 12.2-1s I1 He has the responsibility of determining the radiation levels of all areas where work is to be conducted when it is anticipated that such radiation may exist, and he has the responsibility of establishing, recording, checking, and suitably posting all areas where sources of radiation exist.

He has the responsibility of keeping records and checking all radioactive material releases and shipments from the station.

He has the responsibility of writing and amending the Health Physics Manual and ensuring that all station personnel receive instructions in the implementation of these procedures.

In general, the Supervisor - Health Physics directs the activities of his group to minimize the exposure of station personnel to excessive accumulation doses of radiation and to prevent the spread of radioactive contamination.

He should have a bachelor's degree or the equivalent in a science or engineering subject, including some formal training in radiation protection .

(A master's degree may be considered equivalent to 1 year of professional experience, and a Ph.D. degree may be considered equivalent to 2 years of professional experience where course work related to radiation 1s involved.)

At least 3 years of this professional experience should be in applied radiation protection work in a nuclear facility dealing with radiological problems similar to those encountered in nuclear power stations, preferably in an actual nuclear power station.

He is a member of the Station Nuclear Safety and Operating Committee.

12.2.2.6 Supervisor - Nuclear Training He is responsible to the Station Manager for conducting all training necessary for qualification of employees as reactor and senior reactor operators in accordance with requirements established by the NRC. He is also responsible for conducting general employee training for all employees.

Records of training are maintained under his supervision. He is responsible for supervising the operation of the nuclear control room simulator.

REVISION 1 6/83 SPS UFSAR 12.2-16 12.2.2.7 Supervisor - Administrative Services The Supervisor - Administrative Services is responsible to the Station Manager for ensuring proper security, fire protection, employee safety, records management, administrative support, emergency planning, personnel administration, cost accounting, and budgeting for the station.

1 12.2.2.7.1 Fire Marshall He is responsible to the Supervisor - Administrative Services for the administration of the station's fire detection and suppression programs.

12.2.2.7.2 Supervisor - Safety He is responsible to the Supervisor - Administrative Services for station safety and coordination of training schedules. He performs safety inspections, ensures compliance with OSHA and other safety requirements, administers the First Aid Team, prepares reports pertaining to safety, and coordinates all safety matters with station supervision.

12.2.2.7.3 Coordinator - Emergency Planning He is responsible to the Supervisor - Administrative Services for the administration and coordination of the Station Emergency Plan. He coordinates emergency preparedness activities between station departments, corporate support organizations, local political subdivisions, and emergency response groups. He assists in developing and conducting emergency preparedness training programs and emergency drills. The responsibility for corporate I 1 emergency planning is discussed in Section 12.9.

12.2.2.7.4 Staff Assistant - Personnel He is responsible to the Supervisor - Administrative Services for the coordination of station personnel policies. He provides assistance to the station staff on employee relations, benefits, payroll, and other personnel 1 matters. He coordinates recruitment, selection, and processing of new employees. He implements and administers EEO, Affirmative Action, and appraisal programs, and maintains station personnel and payroll records.

REVISION 1 6/83 SPS UFSAR 12.2-17 12.2.2.7.5 Records Management Supervisor He is responsible to the Supervisor - Administrative Services for supervision of station records and station administrative functions. He ensures compliance with all regulatory requirements in these areas.

12.2.2.7.6 Business Systems Supervisor He is responsible to the Supervisor - Administrative Services for the administration of accounting activities, and monitors cost control and budget activities.

12.2.2.8 Station Quality Assurance The Station Quality Assurance staff performs quality assurance functions for the Station Manager as specified in the Nuclear Power Station Quality Assurance Manual (NPSQAM), but is responsible to the corporate Quality Assurance Organization

• 12.2.2.9 Manager - Quality Assurance The Manager - Quality Assurance has supervisory control over the quality J.

assurance personnel assigned to the power station.

He is responsible to the Executive Manager - Quality Assurance, and is supported by his staff.

He communicates with the Station Manager and his staff on quality assurance matters.

He ensures compliance with the NPSQAM.

He has the Supervisor - Quality Control, Operations and Maintenance, and Supervisor - Quality Control, Nuclear Multiple Power Plants (MPP) Projects reporting directly to him.

REVISION 1 6/83 SPS UFSAR 12.2-18 12.2.2.9.1 Supervisor - Quality Control Operations and Maintenance e

He is responsible to the _Manager - Quality Assurance.

He serves in an advisory capacity to the Station Nuclear Safety and Operating Committee.

He established a comprehensive system of planned and periodic internal audits to ensure compliance with the NPSQAM.

He inspects operating and maintenance activities at the power station, including testing, methods of operation, and modifications to systems, components, or structures where applicable.

He reviews and concurs with, if acceptable, procurement documents, and modifications, maintenance, and repair procedures.

He is supported by assigned quality control inspectors.

12.2.2.9.2 Supervisor - Quality Control MPP Projects He is responsible to the Manager - Quality Assurance and is supported by 11 assigned quality control inspectors.

He performs specific quality assurance functions for the Station Manager as specified in the NPSQAM for applicable MPP Projects (e. g,, receipt inspection, procedure review, etc,).

12.2.2.10 Security Department The security organization is separate from the site organization. The Director - Nuclear Security at Corporate headquarters reports directly to senior corporate management with liaison to the Station Manager. The Station 1 Security Supervisor reports to the Director - Nuclear Security (see Figure 12.2-1). The security organization is responsible for administration and coordination of the station security program to ensure compliance with the Security Plan, NRC regulatory requirements, and Company policy.

STATION ORGANIZATION SURRY NUCLEAR POWER STATION EXECUTIVE STATION NUCLEAR.

SAFETY OPERATING COMMITTEE

~----,-----

I STATION MANAGER ------, MANAGER QUALITY I ASSURANCE DIRECTOR I NUCLEAR SECURITY 8

! L__ -- --~

~-------------------------~ MANAGER STATION QUALITY SECURITY ASSURANCE SUPERVISOR ..

ASSIST ANT STATION MANAGER I I SUPERINTENDENT SUPERINTENDENT SUPERINTENDENT SUPERINTENDENT OPERATIONS MAINTENANCE TECHNICAL SERVICES PROJECTS I

I ISUPERVISOR NUCLEAR TRAINING I SUPERVISOR HEALTH PHYSICS SUPERVISOR PLANNING OPERATIONS COORDINATOR SHIFT SUPERVISOR - SUPERVISOR MECHANICAL MAINTENANCE SUPERVISOR CHEMISTRY -- INSTRUMENT SUPERVISOR ENGINEERING SUPERVISOR SUPERVISOR ADMINISTRAJIVE SERVICES ASSISTANT SHIFT SUPERVISOR

- SUPERVISOR ELECTRICAL MAINTENANCE ENGINEERING SUPERVISOR PERF. a, TEllTl I

-- ENGINEERING SUPERVISOR DESIGN CHANGE FIRE MARSHALL BUSINESS SYSTEMS SUPERVISOR

~

CONTROL ROOM OPERATOR

- SUPERVISOR MAINTENANCE SERVICES REACTOR ENGINEER - ENGINEERING SUPERVISOR SAFETY ENG.

STAFF RECORDS MANAGEMENT SUPERVISOR SUPERVISOR SAFETY H

C/.l H CONTROL ,_.

z0 ROOM OPERATOR NDE GROUP SHIFT TECHNICAL

~COORDINATOR EMERGENCY PLANNING N

,_. TRAINEE ADVISOR N I

CX) w LABOR DEPARTMENT STORES SUPERVISOR STAFF ASSISTANT PERSONNEL

CORPORATE NUCLEAR OPERATIONS ORGANIZATION EXECUTIVE VICE PRESIDENT c.o.o.

SENIOR VICE PRESIDENT POWER I

VICE PRESIDENT NUCLEAR OPERATIONS*

EXECUTIVE MANAGER QUALITY ASSURANCE I CORPORATE

---

:---STATION-_,...... - ___ ___....__

DIRECTOR MANAGER MANAGER STATION

• ADMINISTRATIVE. NUCLEAR OPERATIONS QUALITY ASSURANCE SERVICES* MANAGER ..... *sUP!'ORT C/l e'(j C/l DIRECTOR DIRECTOR DIRECTOR DIRECTOR DIRECTOR DIRECTOR q OPERATIONS & CHEMISTRY t,:j

,TRAINING EMERGENCY SAFETY TECHNICAL C/l MAINTENANCE EVALUATION ANALYSIS & HEALTH SERVICES SUPl'OIH PLANNING

& CONTROL & CONTROL PHYSICS  ?;;

I I I I I I t,:j I II I I-'*

I I I OQ SIJPE RV ISOR SUPERVISOR ~

ENGINEERING NUCLEAR FUEL

~

ti NUCLEAR SUPERVISOR AND HEALTH ro

<: TRAINING (SAFETY ENG.I OPERATIONS PHYSICS H ~

C/l N H

0 N

~ I N

°'

CX) w

• Responsible For Corporate Fire Protection Program.

REVISION 1 6/83 SPS UFSAR 12.3-1 12.3 TRAINING Personnel to staff the Surry Power Station have been selected to ensure that each individual possesses the education, training, and experience necessary to satisfactorily perform his assigned functions. To augment the formal education, training, and experience received by station personnel prior to being stationed at Surry, numerous training programs have been instituted to specifically familiarize employees with the Surry facility. The training programs are administered by the Corporate Nuclear Operations Department, and actual training is performed by corporate, site, and contract personnel from vendor companies.

The principal objectives of the training program are to ensure initial qualification and requalification of station personnel through effective training, to accommodate future growth, to comply with applicable regulations, 1 and utilize the training information contained in relevant guidance documents, including

• 1.

2.

Administrative Policy Manual for Surry Power Station.

10 CFR 50 and 55, and applicable Regulatory Guides.

3. ANSI 18.1-1971.

4. ANSI/ANS 3.1-1981 - Selection and Training of Nuclear Station Personnel.

5. Station Safety Analysis Reports.

6. OSHA and other regulatory requirements.

7. Institute of Nuclear Power Operations (INFO) guidelines.

8. NRC inspections and INFO evaluations.

e

REVISION 1 6/83 SPS UFSAR 12.3-2

9. Technical Specifications.

1 e

10. Quality Assurance Manual.

The station personnel may be qualified through participation in occupational training, which consists of General Station Traini~g, Technical Training, and Employee Development Training, implemented by departmental programs by a combination of formal job training, on-the-job training, and special training. The types of training include

1. Occupational Training, which includes all training efforts intended to develop job knowledge, skills, and employee development required for competent performance of assigned duties.

2. Basic Training, which is designed to provide an understanding of fundamentals, basic principles, and procedures involved in the work to which the employee is assigned.

3. Advanced Training, which addresses topics journeymen or supervisors.

typically taught to

4. Special Training, which is site or equipment specific.

5. Periodic Retraining (Requalification), which is designed to maintain the levels of occupational knowledge, skills, and employee develop-ment required to perform job duties. The Requalification Program 1 reinforces previous training and knowledge, and has essentially the same outline as the initial training program.

6. Backfit Training, which is designed to remedy deficiencies in an employee's background.

The training is conducted using-one or more of the following methods:

1. Formal Job Training, typically classroom training techniques directed at specific job skills and knowledge.

e

_J

REVISION 1 6 / 83 SPS UFSAR 12.3-3

2. On-the-Job Training, conducted under the direction of appropriately experienced personnel.

3. Self-Study Training, where job skills and knowledge may be obtained on an individual basis.

4. Classroom Training, which is formal training utilizing a variety of instructional techniques and media, and requires the tr*ainees to demonstrate their comprehension of the material through discussions, tests, and/or skills performance.

5. Laboratory Training, which provide_s actual hands-on experience in simulated job situations. The laboratory experiences are designed to provide structured and supervised methods of practicing the concepts, principles, and information taught in the classroom. Laboratory training is similar to on-the-job training.

6. Task Training, which is designed to assist the trainee in becoming proficient in learning the basic to advanced job tasks.

7. In-House Training, which is training conducted by an employee of Vepco.

8. Vendor Training, which is training conducted by someone external to Vepco.

The station manager is responsible for timely and effective qualification of assigned personnel. He is assisted by the Training Services Section in the Nuclear Operations Department and the Supervisor of Nuclear Training at the Surry Power Station.

The Training Services Section standardizes programs to meet station requirements, audits station programs, develops methods and materials in support of station programs, evaluates and arranges for vendor training e programs for off site or onsite presentations, and evaluates the overall effectiveness of the program.

REVISION 1 6/83 SPS UFSAR 12.3-4 The station, through the Supervisor of Nuclear Training, e

1. Identifies training requirements, schedule, and types of training needed.

2. Schedules training consistent with station and regulatory requirements.

3. Conducts specific training segments on site.

4. Maintains records of employee qualification, training, and experience.

5. Ensures the training and qualification of station personnel.

6. Makes applications for and maintains licenses and proper certifications required for station personnel.

The program descriptions, course outlines, and training facilities necessary for implementing the station and department training program are coordinated by the Nuclear Training Supervisor, who is directly responsible for the operation of the control room simulator. Each station department has an assigned training coordinator to assist in meeting department training needs and scheduling of training. The Control Room Simulator Technicians are under the direct supervision of the Nuclear Training Supervisor.

The training programs have been updated in response to specific guidance 1

contained in NUREG-0737 regarding upgraded reactor operator and senior 1 reactor operator training programs and training for mitigating core damage.

REVISION 1 6/83 SPS UFSAR 12.3-5

12.3 REFERENCES

1. U. S. Nuclear Regulatory Commission, NUREG-0737, "Clarification of TMI Action Plan Requirements," October 31, 1980 .

SPS UFSAR 12.5-1 12~5 HEALTH PHYSICS Established administrative controls ensure that all procedures and requirements related to radiation protection are followed by all station

}personnel. The Health Physics Section at the Surry plant is responsible for the various radiological safety aspects in the operation of the station.

Health Physics will conduct periodic surveys, calibrations, sample analysis, and other procedures as required to ensure that acceptable guidelines and principles are not violated. As required by Technical Specifications, Section 6.4, detailed written procedures are maintained to govern Health Physics operations. The procedures are also reviewed and updated at reguiar intervals as require!! by Technical Specifications. The Health Physics procedures are contained in two documents and supplemented by administrative p~ocedures. The two documents are the Health Physics Manual and the Respiratory Protection Manual.

12.5.1 HEALTH PHYSICS MANUAL

• The Health Physics Manual consists of three.sections, each containing a number of parts. Section 1 is distributed to all personnel at the station and contains the general rules governing the radiation protection program, as well as an outline of the basic requirements set forth in 10 CFR 20, "Standards for Protection Against Radiation," and also provides for the provisions of

. 10 CFR 19, "Notice~ *rnstruct:j.ons, and Reports to Workers; Inspection." A knowledge of the contents of Section 1 as a minimum is the responsibility of

.. each employee at the station.

Section 2 contains detailed procedural information concerning the radiation protection program, and is for general use of supervisory personnel,

.,who . are to ensure that their subordinates are familiar with the appropriate

• ~;parts concerning an individual's job duties.

Section 3 contains procedural details and information of interest mainly to Health Physics and Chemistry personnel. Subjects in this section include

1. Personal Dosimetry - dosimetry, bioassay, equipment operation.

SPS UFSAR 12.5-2

2. Radioactive Waste - gaseous, liquid fuel - sampling, handling, release, storage, receipt and shipment, processing.

3. Health Physics Survey - station, air sampling, equipment calibration and operation, ALARA program, monitors.

4. Health Physics Count Room - instrument operation and calibration, standard source preparation.

5. Health Physics Environmental Sampling - collection, analysis.

12.5.2 RESPIRATORY PROTECTION MANUAL The Respiratory Protection Manual provides guidance and technical information to guide the administration and implementation of the respiratory protection program. It is designed to meet the requirements outlined in 10 CFR 20 .103, "Exposure of Individuals to Concentrations of Radioactive Materials in Restricted Areas"; Regulatory Guide 8 .15, "Acceptable Programs for Respiratory Protection," October 1962; "Rules and Regulations Governing the Use of Personal Protective Equipment," as adopted by the Safety and Health Codes Commission of the Commonwealth of Virginia, Revision 1, 1974; the Occupational Safety and Health Act (OSHA) 29 CFR 1910 .134, "Respiratory Protection"; the National Institute. for Occupational Safety and Health (NIOSH); the Mine Enforcement and Safety Administration (MESA) 30 CFR 11, "Respiratory Protective Devices"; and NUREG 0041, "Manual of Respiratory Protection Against Airborne Radioactive Materials," October 1976. The manual and its program is designed to protect individuals against the hazards of airborne radioactivity, with additional guidance to overcome the effects of working in a hazardous atmosphere, or for ensuring that the atmosphere is capable of supporting life without respiratory equipment. The manual sections include management policy, respiratory hazards, protection factors, respirator types, respirator testing, specific equipment types, emergencies, quality assurance, and training.

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SPS UFSAR 15-vii CHAPTER 15: STRUCTURES AND CONSTRUCTION LIST OF FIGURES Figure Title 15.1-1 Site Plan 15.1-2 Plot Plan

15. 1-3 Machine Location, Reactor Containment, Elevation 47 ft 4 in.

15.1-4 Machine Location, Reactor Containment, Elevation 18 ft 4 in.

15.1-5 Machine Location, Reactor Containment, Elevation 3 ft 6 in.

15 .1-6 Machine Location, Reactor Containment, Elevation 27 ft 7 in.

15.1-7 Machine Location, Reactor Containment Vertical Section, Sheet 1 15.1-8 Machine Location, Reactor Containment Vertical Section, Sheet 2 15.1-9 Machine Location, Reactor Containment Vertical Section,

• 15.1-10 15 .1-11 15.1-12 Sheet 3 Auxiliary Building Arrangement, Sheet 1 Auxiliary Building Arrangement, Sheet 2 Auxiliary Building Arrangement, Sheet 3 15.1-13 Auxiliary Building Arrangement, Sheet 4 15.1-14 Fuel Building Arrangement, Sheet 1 15.1-15 Fuel Building Arrangement, Sheet 2 15.1-16 Control & Relay Room, Service Area 15.4-1 Quality Assurance Project Organization 15.4-2 Quality Assurance Functional Organization for Westinghouse Nuclear Energy Systems 15.4-3 Nuclear Energy Systems Functional Groups Quality Assurance Flow Chart, 3 Sheets .

15.5-1 Reactor Containment Waterproofing 15.5-2 Containment Loading Plot, Sheet 1 15.5-3 Containment Loading Plot, Sheet 2 15.5-4 Containment Loading Plot, Sheet 3 e 15.5-5 Reinforcing Details, Equipment Access Hatch Opening

SPS UFSAR 15-viii CHAPTER 15: STRUCTURES AND CONSTRUCTION LIST OF FIGURES (continued)

Figure Title 15 .5-6 Reinforcing Details, Sections Through Ring Beam, Equipment Access Hatch 15.5-7 Reinforcing Details, Personnel Hatch Opening 15.5-8 Reinforcing Details, Sections Through Ring Beam, Personnel Hatch 15.5-9 Wall and Mat Joint 15.5-10 Section - Typical Bridging Bar, Sheet 1

15. 5-11 Section - Typical Bridging Bar, Sheet 2 15.5-12 Typical Electrical Penetration Sleeve With Flanges 15.5-13 Typical Piping Penetrations, 2 Sheets 15.5-14 Personnel Hatch Assembly 15 .5-15 Typical Liner Details 15.5-16 Containment Loading Plot 15.6-1 Reactor Neutron Shield Tank Assembly 15.6-2 Steam Generator Support Assembly 15.6-3 Reactor Coolant Pump Supports, General Arrangement 15.6-4 Pressurizer Support

SPS UFSAR 15.1-1 15.1 STRUCTURES AND MACHINERY ARRANGEMENT e

The site arrangement, plot plan, and the general arrangement of equipment within the principal Class I structures are shown on the Figures listed in the following tabulation:

Item Figure Site Plan 15.1-1 Plot Plan 15.1-2 Containment Structure and Containment Auxiliary Structures 15.1-3 through 15.1-10 Auxiliary Building 15.1-11, -12, and -13 Fuel Building 15.1-14 and -15 Control Area 15 .1-16

SPS UFSAR 9.6-21 Table 9 .6-8 HIGH RADIATION SAMPLING SYSTEM SAf1PLING PANEL DESIGN CONDITIONS I. Process

1. Pressure (psig, max) Reactor coolant sampling 2485 Sump sampling 75 Containment atmosphere 60

2. Temperature (psig, max) Reactor coolant sampling 700 Sump sampling 220 Containment atmosphere 310 II. In-containment ambient

1. Pressure (psia) 9-60

2. Temperature (OF) 310

3. Relative humidity (%) 0-100 III. Outside containment ambient

1. Pressure Atmospheric

2. Temperature (OF)40-120

3. Relative humidity (%) 0-100 7

4. Radiation (rads) 1 X 10 IV

• Design life (years) 40

REVISION 1 6/83 SPS UFSAR 9.8-5 in each unit's instrument air receiver being backed up by the three service. air receivers.

3. Instrument air backup between the two units. This is provided by means of cross-connecting lines between the two units at the main headers.

4. Instrument air backup to containment instrument air subsyste~. In the event of the loss of both containment instrument air compressors and receivers, containment instrument air can be supplied from the instrument air system by opening the manually operated valves in the cross-connect line provided.

5. Check va.lves at the containment instrument air receiver outlets. In the event of the loss of one containment instrument air receiver, the valve closes, preventing containment instrument air from escaping from the other receiver .

• 6. The instrument air compressors and the containment instrument air compressors and their driers are connected to the emergency power system.

The containment instrument air system is non-safety-related because the components requiring containment instrument air are not necessary for safe shutdown. The majority of loads on this system are spring-diaphragm-type air-operated valves, which use spring force to maintain the valves in a fail-safe condition. The remaining loads are the personnel airlock inner-door locking device, containment recirculation damper controls, and the reactor head inflatable seal in the head storage area. The compressors, accessories, and piping upstream from the first containment isolation valve and associated pipe support outside containment are not seismically qualified.

9.8.4 TESTS AND INSPECTIONS Testing of this system is unnecessary because of its normal day-to-day operation. Inspection is performed in accordance with normal station maintenance procedures.

REVISION 1 6/83 SPs* UFSAR 9.10-23 11 The separate cable spreading and routing area provided for each unit is bounded on all sides by concrete or concrete blocks, which provide a fire barrier surrounding the three adjoining spaces within a unit. Each area is provided with a total flooding automatic carbon dioxide suppression system and a separate fire detection system that alarms in the control room. The automatic carbon dioxide system is backed up by manual suppression capability using portable extinguishers located in the area, and water hoses from yard hydrants and the hose stati.ons in the turbine building. The areas can be accessed at one end from the outside yard via the motor control center rooms and a spiral staircase down to the outside containment penetration vaults, and at the other end from the turbine building via one of the emergency switchgear rooms. Smoke and heat can be exhausted up the spiral staircases and through the doors of the motor* control center rooms to the outside. Adequate floor space is available to permit access by fire fighters to all locations within the areas.

9.10.4.4 Battery Rooms

• There are four 125-V de station battery rooms. A separate battery is provided for each division of each unit's safety-related equipment.

battery is housed in a separate battery room approximately 9 ft by 14 ft, Each located within or adjacent to the associated division's emergency switchgear room. At least one battery for each unit is required for safe shutdown from the control room or remote shutdown panels.

The combustible material in the rooms consists of the plastic battery cases and the battery power cable insulation.

An unmitigated fire in a battery room could disable one of the station batteries, which would result in a loss of automatic or remote control of a single division of safety-related equipment, including equipment used for safe shutdown. Such a fire would not prevent safe shutdown, however, since redundant equipment controlled from the other unit battery would still be available.

REVISION 1 6/83 SPS UFSAR 9.10-24 Each battery room is bounded on all sides by concrete or concrete block, which provides an adequate fire barrier. Ventilation du_cts that penetrate the e barrier are provided with fire dampers and, with the exception of battery room No. 2B, the doors to the rooms are 3-hr fire-rated. Manual actions by plant personnel are relied upon to detect and suppress a fire. The battery rooms themselves and the areas used for access to the battery rooms are relatively uncongested, with adequate space for manual fire fighting. Portable carbon dioxide extinguishers are located nearby in the emergency switchgear rooms, and a 150-lb wheeled, dry-chemical extinguisher is located in the Unit 2 divi-sion J emergency switchgear room. Also, 1.5-in. hose stations are located in the turbine building within reach of all the battery rooms.

The fire potential in a battery enclosure is virtually negligible because battery hydrogen generation is negligible. The lead cadmium batteries would require more than 5 days' overcharge in a Sealed room without ventilation to attain the burnable tp.reshold concentration of 4% hydrogen. Hydrogen generation, under normal charge or discharge, is less than 10% of this overcharge rate.

9.10.4.5 Cable Tray Rooms

• The two cable tray rooms are large, open spaces directly above the control room. Each room is used as a cable-spreading area primarily for non-safety-related cables of its respective unit. There are, however, a small number of safety-related cables, and a concrete cubicle containing the reactor protection system trip breakers and switchgear, located in each of the rooms.

The safety-related cables are those associated with the interlocks between the reactor coolant pumps and the rod control system, and the interlocks between the main feedwater pumps and the motor-driven auxiliary feedwater pumps. The feed pump interlocks are part of a system that may be used for safe shutdown, since they function to provide the automatic start capability by the motor-driven auxiliary feedwater pumps. These pumps, however, may be started manually at the control board.

An indicating panel monitoring reactor coolant hot-leg temperature and wide-range pressure (pressurizer pressure), as well as the levels of the 1 pressurizer and steam generators, has been installed in the cable tray area

SPS UFSAR 9 .11-1 9.11 WATER SUPPLY AND TREATMENT SYSTEMS 9.11.1 WELL-WATER SUPPLY SYSTEM The well-water supply system provides makeup water to the fire protection and domestic water storage tanks, the hydropneumatic tank in the potable water system, and the fire protection system. The well-water supply system is shown on Figure 9.11-1.

There are three cased water wells located south of the site, with depths of 418, 420, and 420 ft about 400 to 1000 ft apart (wells B, C, and Eon Fig-ure 15.1-1). Each well has a 200-gpm submersible pump discharging to a well-water storage tank. Each well pump has a separate underground discharge line that is interconnected at the storage tank. Centrifugal-type well-water transfer pumps deliver water from the storage tank to consuming systems as required.

The well-water supply system is designed to be automatically or manually controlled.

9.11.2 DOMESTIC WATER SUPPLY SYSTEM A 4000-gal hydropneumatic tank, located in the fire-pump house, is provided for the domestic water supply system. Pressure in the hydropneumatic tank is maintained at 40 to 60 psig by a pressure system, consisting of a pressure-level regulator, air compressor, and related controls and accessories. Hypochlorinator equipment provides a means of chlorinating the domestic water supply. Piping from the hydropneumatic tank supplies cold water to safety showers, drinking water coolers, hot-water storage tanks, and domestic cold water throughout the station.

Domestic water supply component design data are given in Table 9.11-1.

9.11.3 FLASH EVAPORATOR SYSTEM The flash evaporator system is shown on Figure 9 .11-2. The system consists of equipment for the production of high-purity water by de-aerating,

SPS UFSAR 9 .11-2 evaporating, cooling, prefiltering, demineralizing, and postfiltering river water or service water for makeup to the various station systems.

Supplementary chemical feed equipment is provided for feedwater conditioning chemicals. High-purity water is pumped to the primary-water storage tanks (Section 9.1) for reactor plant makeup, and to the condensate storage tank for secondary plant makeup. The flash evaporator and the polishing demineralizer are designed to operate automatically after manual initiation. Evaporation is started manually on low-level alarm from the condensate storage tank and proceeds until an alarm is actuated indicating high level in the condensate storage tank. On high-effluent conductivity, another alarm is actuated and the evaporator effluent is blown down to waste.

The flash evaporator is designed to provide a normal net distillate output of 125 gpm when using sixth-point extraction steam, and a maximum net output of 220 gpm when using steam supplied from the auxiliary steam system.

The prefilter is of the cartridge type. The cartridge-type postfilter unit is designed to remove any suspended material larger than 5 min the demineralizer product stream.

The flash evaporator, with an integral vacuum de-aerator, is constructed for a test pressure of 22.5 psig to evaporate fiitered river water and service water and deliver distillate with less than 7 ppb of dissolved oxygen and less than 25 ppb of dissolved solids (such as Na), excluding iron and copper.

The demineralizer vessel is a rubber-lined steel pressure tank with type 316 stainless steel internal distributing underdrain, constructed in accordance with the ASHE Code for unfired pressure vessels with 100-psig working pressure. The vessel was tested hydraulically at 150 psig and loaded with a mixed cation/anion exchange resin. The unit is designed to produce from a minimum capacity of 50 gpm up to a maximum of 460 gpm of reactor-grade demineralized water.

Flash evaporator system component design data are given in Table 9.11-2.

The fire protection system is discussed in Section 9.10.

SPS UFSAR 11.2-5 steel tanks designed according to Section III. C of the ASME Boiler and Pressure Vessel Code.

11.2.3.1.2 Low-Level Waste Drain Tanks Two low-level waste drain tanks are provided. Each tank has a capacity of 2874 gal. Level indicators are provided. These are stainless steel tanks designed according to Section III.C of the ASME Code.

11.2.3.1.3 Waste Disposal Evaporator and Auxiliaries (Installed But No Longer Used)

One externally heated, forced-circulation evaporator with a feed and distillate capacity of 6 gpm is provided. The evaporator shell is fabricated from a high-nickel alloy in accordance with Section III. C of the ASME Code.

Internals are fabricated from an austenitic stainless steel not susceptible to stress cracking *

• The external heat source is a shell and tube steam reboiler fabricated on the tube side from a high-nickel alloy and on the shell side from carbon steel. Distillate is condensed in a water-cooled shell and tube condenser fabricated from austenitic stainless steel. The reboiler, shell, and tube condenser are all fabricated in accordance with Section III.C of the ASME Code, and TEMA Standards.

The condensed distillate is held in the distillate accumulator. This tank is fabricated from austenitic stainless steel in accordance with Section III.C of the ASME Code.

A distillate cooler is provided to further cool the distillate. The tube side ot the distillate cooler is fabricated from austenitic stainless steel and the shell side from carbon steel, in accordance with Section III.C of the ASME Code.

SPS UFSAR 11. 2-6 11.2.3.1.4 Waste Disposal Evaporator Test Tanks Two waste disposal evaporator test tanks, each of 3000-gal capacity, with level indicators, are provided. These tanks are stainless steel and designed according to Section III.C of the ASME Code.

11.2.3.1.5 Waste Disposal Demineralizer Fifteen waste disposal demineralizers are provided for waste processing.

The demineralizers are fabricated from austenitic stainless steel. Demineral-izer subsystem connections and interconnections are made with flexible hoses.

The demineralizer vessels can be connected in series, parallel, or series-parallel configurations. Ion-exchange and/ or filter media loading is determined by waste stream chemical and isotopic characteristics.

11.2.3.1.6 Waste Disposal Filters Liquid waste effluent filters and the distillate demineralizer filters will be cartridge-type pressure filters. The vessels are fabricated from austenitic stainless steel, in accordance with Section III.C of the ASME Code.

The filter elements are of the synthetic fiber disposal type. Filter cartridges are desig~ed for removal as a single basket assembly. Contaminated drain tank filters are provided to remove lint and other laundry waste matter that could be radioactive. This filter is operated on a precoat-filter-backwash cycle.

11.2.3.1.7 Pumps Centrifugal frame-mounted pumps with single or double mechanical seals are provided. The waste disposal evaporator bottoms pump is a canned pump.

One pump is provided for each tank with cross ties where appropriate, such as on high-level waste drain tank pumps. External cooling and seal water is supplied to radioactive pump seals as required.

11.2.3.1.8 Ion-Exchanger System Components The ion-exchanger system and components are described in Section 11.2.3.2.

SPS UFSAR 11.3-11 11.3.2.9.2 Recovery Phase The design basis for certain systems and their associated shielding did not consider postaccident recovery operations, that is, postaccident cleanup of highly radioactive fluids. These are the waste disposal system, the boron recovery system, the containment purge system, and the letdown and charging portions of the eves. As a result, these systems will not be used for postaccident cleanup operations.

The activity levels (based on Regulatory Guide 1.4 and TID-14844) of the influent to the liquid waste disposal system or to the boron recovery system 3 3 are approximately 2 x 10 µCi/cm after 6 months of radioactive decay. The area radiation dose rates from concentrated waste and from waste storage tanks would severely limit access to parts of the auxiliary building and would hinder the operation of both units. Since the radioactive waste disposal systems are common to both Units 1 and 2, the use of these systems for the cleanup of waste in the accident-affected unit would preclude the normal use

• of the radioactive systems for the non-accident-affected unit

• There is extensive piping for the above-listed systems throughout the auxiliary building. The resulting dose rate if all these systems operated simultaneously would severely limit freedom of access for required operations.

Shielding for the piping and components would be very difficult, and in some cases impossible, to install, because of the physical arrangement of the piping and components.

11.3.2.9.3 Postaccident Sampling Capability The Surry Power Station has the capability to sample the reactor coolant, containment sump, and the containment atmosphere. The reactor coolant sample can be taken within 1 hr of an accident, and analyzed within another hour.

The containment atmosphere sample can be taken with the containment monitoring system. Provisions are included for personnel exposure control. A detailed discussion of the postaccident sampling system is contained in Chapter 9.

SPS UFSAR 11.3-12 11.3.3 PROCESS RADIATION MONITORING SYSTEM The process radiation monitoring system continuously monitors selected lines containing, or possibly containing, radioactive effluents. Lines through which waste liquids and gases are discharged to the environment are also monitored. The function of this monitoring system is to warn personnel of increasing radiation levels that could result in a radiation health hazard and to give early warning of a system malfunction. An audible alarm has been incorporated to indicate the loss of power to the system and thus detector malfunction. The process radiation monitoring system serving both units is comprised of 30 channels (Table 11.3-6) and 4 spare channels.

Each channel has a readout in the control room, and selected channels, as indicated in Table 11.3-6, have a readout at the detector location. In addition, each channel has an audible and visual alarm for radiation levels in excess of preset values, as well as a visual alarm for detector malfunction.

The output from all channels is recorded on strip-chart recorders that produce a continuous record of radiation levels and radioactive discharges from the station. Each channel has its own power supply and check source~ remotely operated from the control room, thus making it completely independent of any other channel. The adjustment of alarm setpoints, voltage, power, and other variables is made from the control room. The entire system is designed to be fail-safe, with emphasis on system reliability and availability. Certain channels, as indicated in the following text, actuate control valves on a high-activity alarm signal.

The expected concentrations of radionuclides in the process streams monitored by the ventilation vent monitors, component cooling water monitors, component cooling heat exchanger service water monitors, condenser air ejector monitors, steam-generator blowdown monitors, and recirculation spray cooler heat exchanger service water monitors are natural background radioactivity.

The sensitivity of these detectors ensures that abnormal plant conditions will be detected before they cause a hazard to the operators or to the general public.

REVISION 1 6/83 SPS UFSAR 11.3-17 11 assembly failure. This is a two-stage monitoring system consisting of a row-range channel and a high-range channel. There is one such system for each unit. After being withdrawn from the letdown line, the sample is passed through a delay line to allow N-16 to decay, then enters a sampler consisting of two gamma scintillation detectors surrounded by 7 to 8 in. of lead, and is finally discharged to the volume control tank. Both detectors sit on a 1/2-in. removable stainless steel tube, providing flow through the sampler.

Shielded lead plugs are used to convert the two detectors into either high- or low-range letdown monitors. Normally, the low-range detector will be sitting on the 1/2-in. tubing and the high-range detector will be sitting on the shielded lead plug. In the event of a fuel element failure, the activity released could be sufficient to raise the coolant activity level above 3

1.0- µCi/cm gross fission products. This causes the high-range monitor to 3

begin to indicate activity level at 10-l µCi/cm , providing a one-decade overlap. At this point, the high-range channel provides the activity data, and the low-range monitor can be converted into a high-range mqnitor by inserting a shielded lead plug .

• 11.3.3.11 Circulating Water Discharge Tunnel Monitors Each of these identical channels (one per unit) monitors the effluent (service water, condenser circulating water, and liquid waste) in the circu-lating water discharge tunnel beyond the last point of possible radioactive material addition. A gamma scintillation detector slides into a capped pipe, which is then inserted directly into the discharge tunnel and acts as a well.

At the top of the pipe is a waterproof support assembly that encloses a check source. The entire device is waterproof.

11.3.3.12 Ventilation Vent Sample Particulate and Gas Monitors A high-range/low-range gas and particulate monitor is used to obtain grab samples of the ventilation flow stream. In addition, the eighth channel of the ventilation multipoint radiation monitor is continuously selected to monitor the flow stream. This multipoint monitor functions as described in Section 11.3.3.3, except that the other seven channels are no longer used.

REVISION 1 6/83 SPS UFSAR 11.3-18 11 11.3.3.13 High-Range Postaccident Radiation Monitors Methods for monitoring high-level releases of noble gases, -iodine, and particulates have been developed and implemented. All potential releases are monitored.by instrumenting the ventilation vent stack, the process vent stack, and the main steam header discharge. The waste gas decay tank and hydrogen purge exhaust are discharged through the process vent stack. The auxiliary building, decontamination building, fuel building, and safeguards area exhausts are discharged through the auxiliary building ventilation vent stack. The containment purge system, which is common to both units, dis-charges through the ventilation vent before and during refueling outages. The main condenser air ejectors normally discharge through the ventilation vent, but flow is diverted to containment if the Technical Specification radiation level limit is exceeded.

Noble gas releases are monitored by a high-range monitor system installed on each discharge line. The system uses three detectors to cover the range 4

from 10- 5 to 10 R/hr.

The high-range noble gas radiation monitors on the ventilation and 7 5 3 process vents have a range of 10- to 10 µCi/cm (Xe-133) under normal back-ground conditions (less than 1 mR/hr). The reduction of effluent detector sensitivity under maximum background conditions will not exceed the normal effluent instrument range. A multidetector system with an automatic back-ground correction feature and sufficient range overlap is provided to ensure complete coverage for all expected background conditions. Equivalently shielded effluent and background subtract detectors in conjunction with a digital-based system are needed to obtain the required sensitivities.

Accident particulate and iodine releases are determined by retrieving fixed filters for laboratory analysis. The filters are shielded to provide personnel protection during removal and reinstallation. Several filters in parallel provide for continuous sampling during filter removal.

High-range monitors, with a usable range of 0.01 mrem/hr to 10,000 R/hr, are installed on all main steam lines. The 10, 000-R/hr maximum reading 9 3 corresponds to a noble gas concentration of 5.4 x 10 µCi/cm

• The main steam

REVISION 1 6/83 SPS UFSAR 11.3-2111 11.3.4.2 Containment Particulate Monitors This channel continuously withdraws a sample from the containment atmosphere into a closed, shielded system exterior to the containment. The sample is passed through a moving filter paper with a collection efficiency of 99% for particle sizes greater than 1.0 The amount of deposited activity is continuously scanned by a lead-shielded beta scintillation detector with a 9 3 sensitivity of 1 x 10- µCi/cm for I-131 in a background of 0.75 mR/hr. The sample system, which is common to both the particulate and gas monitors, includes a pump with a 1. 5-hp motor, a flow meter, automatic pressure protecting valves, a flow regulating valve, and isolation valves. The pump and motor are located inside the containment. A sample point is available for taking a sample of the containment atmosphere after an incident for spectrum analysis in the laboratory. During refueling, if the 1-131 concentration

-9

• 3 exceeds 9 x 10 µCi/cm , a high-activity alarm automatically trips the containment purge air supply and exhaust fans and closes.the purge system butterfly valves, thus isolating the purge system. The counting rate of the 11 3 limiting isotope, I-131, is 8.7 x 10 cpm/µCi/cm .

11.3.4.3 Containment Gas Monitors This channel takes the continuous containment atmosphere sample, after it has passed through the particulate filter paper, and draws it through an inline, easily removable, charcoal cartridge arrangement to the containment gas monitor assembly, which is a fixed-volume, lead-shielded sampler enclosing a beta scintillation detector. The sensitivity of this detector is 5 x

-6 3 10 µCi/cm for Kr-85 in a background of 0.75 mR/hr. The sample activity is measured, and then the sample is returned to the containment.

-5 3 During refueling, if the Kr-85 concentration exceeds 1 x 10 µCi/cm, a high-activity alarm automatically trips the containment purge air supply and exhaust fans and closes the purge system butterfly valves, thus isolating the purge system.

A purge valve arrangement blocks the normal sample flow to permit purging the detector with a clean sample for calibration. Purged gases are discharged

REVISION 1 6/83 SPS UFSAR 11.3-2211 to the containment. Protection and isolation are provided as described in Section 11.3.4.1.

The counting rates of the limiting isotopes, I-131, Xe-133, and Kr-85, 7 7 8 3 are 8.5 x 10 , 3.96 x 10 , and 1.04 x 10 cpm/µCi/cm , respectively.

11.3.4.4 Other Area Radiation Monitoring Equipment This equipment consists of fixed-position, ion-chamber-type gamma detectors and associated electronic equipment. These channels warn personnel of any increase in radiation level at locations where personnel may be expected to remain for extended periods of time. All monitors have an 7

eight-decade display (O.l to 10 mR/hr), and for better resolution can also be set on any three consecutive decades within this range. The channels and their ranges are listed in Table 11.3-8.

In addition, if the dose rate at the manipulator crane area monitor exceeds 50 mR/hr during refueling, the alarm automatically trips the containment's purge air supply and exhaust fans and* closes the purge system butterfly valves, thus isolating the containment from the environment.

• 11.3.4.5 High-Range Postaccident Containment Monitors A high-range reactor containment area monitor is located outside the containment structure. The detector is permanently mounted and aimed at the 7

personnel hatch. The monitor has a range of 0.1 to 10 mR/hr to measure the expected high gamma dose rate in the containment following a LOCA.

An additional set of eight high-range containment radiation monitors is installed at separate locations on the crane wall above the operating deck level inside containment. The monitors are single ion chamber detectors that 7

measure photons over the range of 1 to 10 R/hr. The system is sensitive to photon energies from 60 keV to 3 MeV, with +/-20% accuracy for 0.1-to 3-MeV photons.

The readout for the in-containment monitors is located in the control room and consists of a rate meter and strip chart recorder that starts at a

REVISION 1 6/83 SPS UFSAR ll.3-22a 11 preset value. Each redundant monitor is powered by a separate vital instrument bus.

The in-containment monitors meet the Seismic Class 1 requirements of 5

Regulatory Guide 1.100, and will withstand the LOCA conditions specified by 6

Regulatory Guide 1.89.

11.3.5 ENVIRONMENTAL SURVEY PROGRAM A report on the preoperational radiological surveillance program covering the period from May 1968 through June 1970 was submitted to the Atomic Energy Commission (AEC).

Comments made by the U.S. Fish and Wildlife Service concerning the preoperational phase of the station were taken into account by

1. Holding a conference on June 21, 1968, to discuss the pre- and postoperational surveys for the Surry Power Station. Repre-sentatives from the following agencies were present:

a. Federal Water Pollution Control Authority.

b. Bureau of Commission of Fisheries, Radiobiological Laboratory.

c. Virginia State Water Control Board.

d. Commonwealth of Virginia, Bureau of Industrial Hygiene.

.. ~*.

SPS UFSAR 5.2-19 Table 5.2-1 (continued)

MAJOR PIPING PENETRATIONS THROUGH THE REACTOR CONTAINMENT STRUCTURE Containmen~ Additional Nominal a Isolation Isolation Closed S:tetems Line Line Statue Isolation Requirementea a b Line Signal 8

Valves Fluid Temperature Inside Outside Normal Shutdown Incident Inside Outside Service Size (in.~ ~

Manual (AC) Liquid Cold No No 20 V Closed Int Closed Blind flange Fuel transfer tube Auto trip SI Gas Cold No No 6 I Open Open Closed Check Air ejector condenser vent Manual (AC) Gas Cold Yes Yes 8 V Closed Int Closed Remote Containment vacuum ejector suction manual Auto trip SI Manual Gas Cold No No 2 I Open Open Closed Auto trip Primary vent header Manual (AC) Manual ,Liquid Cold No No 0.75 V Closed- Closed Closed Safety injection accumulator test Int Manual (AC) Manual Gas Cold Yes Yes l V Closed Open-Int Closed Manual Primary vent pot Auto trip SI Gas Cold No No l I Open Open-Int Closed Auto trip Nitrogen relief Auto trip SI Gas Cold No No 2 I Open Open Closed Auto trip Containment instrument air compressor suction Auto trip HH Gas Cold No No 2 I Open Open Closed Check Containment instrument air compressor discharge 8Key: Int c intermittent service; SI* safety injection; HH

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APERTURE

• STEAM GENERATOR BLOWDOWN COOLERS CARD (TYP.) &0104G PDIS CL~OI ICLltl

!IOI04C. &DIO?>C NOTES I (1) ALL PRESSURE VENT DRAIN & TEST CONNECTIONS SHALL BE "-vos-soc UNLESS OTHERWISE NOTED.

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FH*1"7A (J-2) VCS*60A V03*60C CLl51

£ 5E£ AOMIN 9?1 ____.)

REFERENCE DWGS.

FLOW DIAGRAM CONDENSATE FM 17A&B FLOW DIAGRAM MAIN STEAM FM 14A FLOW DIAGRAM CIRCULATING & SE'lVICE WATER FM 21A FLOW DIAGRAM SAMPLING SYSTEM FM 320 FLOW DIAGRAM VENT & DRAIN SYSTEM FM 100A&8 FLOW DIAGRAM STM GEN BLOWDOWN FM 1228 FLOW DIAGRAM STM GEN BLOWDOWN 11548 FM 122A FLDW DIAGRAM STM GEN RECIRC & TRANSFER FM 137A VALVE OPERATING NUMBERS STM GEN SLOWDOWN FM 124A

(*118TD*Z5*151 COtJT.QJ 11'!>4&*Fl1*1'11.A (F*&)

~ Av.U.able On STEAM GENERATQR LOWDOWN TANI< STEAM GENERATOR BLOWDOWN SUBSYSTEM UNIT 1

.. Aperture Card 8* 3 0 6 3 OO 1-5 1 -;L 7 REVISION 1 6/8')

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