Text

Technical Specifications
	ML19094A226
Person / Time
Site:	Surry
Issue date:	03/17/1972
From:	; Virginia Electric & Power Co (VEPCO)
To:	; US Atomic Energy Commission (AEC)
References
	Download: ML19094A226 (300)
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Vepco

SURRY POWER STATION

-UN ITS 1 AND 2 * -

TECHNICAL SPECIFICATIONS

VIRGINIA ELECTRIC AND POWER COMPANY

. DOCKET NOS. 50-280 AND 50-281

TS-i 3-17-72 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS SECTION TITLE PAGE 1.0 DEFINITIONS TS 1.0-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM TS'2,l-1 SETTINGS 2.1 SAFETY LIMIT, REACTOR CORE TS 2.1-1 2.2 SAFETY LIMIT, REACTOR COOLANT SYSTEM PRESSURE TS 2.2-1 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE TS 2.3-1 INSTRUMENT AT ION 3.0 LIMITING CONDITIONS FOR OPERATION TS 3.1-1 3 . .1 REACTOR COOLANT SYSTEM TS 3.1-1 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM TS 3.2-1 3.3 SAFETY INJECTION SYSTEM TS 3,3-1 3.4 SPRAY SYSTEMS TS 3.4-1 3.5 RESIDUAL HEAT REMOVAL SYSTEM TS 3.5-1 3.6 TURBINE CYCLE TS 3.6-1 3.7 INSTRUMENTATION SYSTEM TS 3.7-1 3.8 CONTAINMENT TS 3.8-1 3.9 STATION SERVICE SYSTEMS TS 3.9-1 3.10 REFUELING TS 3.10-1 3.11 EFFL UEN.T RELEASE TS 3.11-1 3.12 CONTROL ROD ASSEMBLIES AND POWER DISTRIBUTION TS 3.12-1 LIMITS 3.13 COMPONENT COOLING SYSTEM TS 3.13-1 3.14 CIRCULATING AND SERVICE WATER SYSTEMS TS 3.14-1

TS-ii 3-17-72 SECTION TITLE PAGE 3.15 CONTAINMENT VACUUM SYSTEM TS 3.15-1 3.16 EMERGENCY POWER SYSTEM TS 3.16-1 3.17 LOOP STOP VALVE OPERATION TS 3.17-1 3.18 MOVEABLE INCORE INSTRUMENTATION TS 3.18-1 3.19 MAIN CONTROL ROOM VENTILATION SYSTEM TS 3.19-1 4.0 SURVEILLANCE REQUIREMENTS TS 4.0-1 4.1 OPERATIONAL SAFETY REVIEW TS 4.1-1 4.2 REACTOR COOLANT SYSTEM COMPONENT TESTS TS 4.2-1 4.3 REACTOR COOLANT SYSTEM INTEGRITY TESTING TS 4.3-1 FOLLOWING OPENING 4.4 CONTAINMENT TESTS 4.5 SPRAY SYSTEMS TESTS TS 4.5-1 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTING .TS 4.6-1 MAIN STEAM LINE TRIP VALVES e 4.7 4.8 AUXILIARY STEAM GENERATOR FEEDWATER PUMPS TS 4. 7-1 TS 4.8-1 4.9 EFFLUENT SAMPLING AND RADIATION MONITORING SYSTEM TS 4.9-1 4.10 REACTIVITY ANOMALIES TS 4.10-1 4.11 SAFETY INJECTION SYSTEM TESTS TS 4.11-1 4.12 VENTILATION FILTER TESTS TS 4.12-1 4.13 NONRADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM TS 4.13-1 5.0 DESIGN FEATURES TS 5.0-1 5.1 SITE TS 5.0-1 5.2 CONTAINMENT TS 5.2-1 5.3 REACTOR TS 5.3-1 5.4 FUEL STORAGE TS 5.4-1 6.0 ADMINISTRATIVE CONTROLS TS 6.1-1 6.1 ORGANIZATION, SAFETY AND OPERATION REVIEW TS 6.1~1 6.2 ACTION TO BE TAKEN IN THE EVENT OF AN TS 6.2-1 ABNORMAL OCCURRENCE IN STATION OPERATION

TS-iii 3-17-72 e

STATION TITLE PAGE 6.3 ACTION TO BE TAKEN IF A SAFETY LIMIT IS TS 6.3-1 EXCEEDED 6.4 UNIT OPERATING PROCEDURES TS 6.4-1 6.5 STATION OPERATING RECORDS TS 6.5-1 6.6 STATION REPORTING REQUIREMENTS TS 6.6-1

TS 1.0-1 3-17-72 1.0 DEFINITIONS The following frequently used tenns are defined for the unifonn interpretation of the specifications.

A. Rated Power A steady state reactor core heat output of 2441 MWt.

B. Thennal Power The total core heat transferred from .the fuel to the coolant.

C. Reactor Operation

1. Refueling Shutdown Condition When the reactor is subcritical by at least 10% ~k/k and Tavg is

<140°F and fuel is scheduled to be moved to or from the reactor core.

2. Cold Shutdown Condition When the reactor is subcritical by at least 1% ~k/k and T is <200°F.

avg

3. Intennediate Shutdown Condition When the reactor is subcritical by an amount greater than or equal to the margin as specified in Technical Specification Figure 3.12-2 and 200°F <T <547°F.

avg

TS 1.0-2 3-17-72

4. Hot Shutdown Condition When the reactor is subcritical by an amount greater than or equal to the margin specified in Technical Specification Figure 3.12-2 and --r avg is > 54iJF.

5. Reactor Critical When the neutron chain reaction is self-sustaining and keff = 1.0.

6. Power Operation When the reactor is critical and the neutron flux power range instrumentation indicates greater than 2% of rated power.

7. Refueling Operation Any operation involving movement of core components when the vessel head is unbolted or removed.

D. Operable A system or component is operable when it is capable of performing its in-

~ended function within the required range. The system or component shall be con-sidered to have this capability when: (1) it satisfies the limiting conditions for operation defined in Section 3, and (2) it has been tested periodically in accordance with Section 4 and meets its performance requirements.

TS 1.0-3 3-17-72 E. Protective Instrumentation Logic

1. Analog Channel An arrangement of components and modules as required to generate a single protective action digital signal when required by a unit condition. An analog channel loses its identity where single action signals are combined,

2. Logic Channel A logic channel is a group of relay contact matrices which operate in response to the digital output signal from the analog channel to generate a protective action signal.

F. Degree of Redundancy*

The difference between the number of operable channels and the minimum number of channels monitoring a specific parameter which when tripped will cause an automatic system trip.

G. Instrumentation Surveillance

1. Channel Check A qualitative determination of acceptable operability by observation of channel behavior during operation. This determination shall include comparison of the channel with

TS 1. 0-4 3-17-72 other independent channels measuring the same variable.

2. Channel Functional Test Injection of a simulated signal into an analog channel or makeup of the logic combinations in a logic channel to verify that it is operable, including alarm and/or trip initiating action.

3. Channel Calibration Adjustment of channel output such that it responds, with acceptable range and accuracy, to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment action, alarm, or trip, and shall be deemed to include the channel functional test.

H. Containment Integrity Containment integrity is defined to exist when:

1. ~l non~automatic containment isolation valves, except those required for intermittent operation in the performance of normal operational activities, are lqcked closed and under administrative control. Non-automatic containment isolation valves may be opened intermittently for operational activites provided that they are under administrative control and are capable of being closed immediately if required.

2. Blind flang~s are installed where r~quired.

3. The equipment access hatch is properly closed and sealed.

4. At least one door in the personnel air lock is properly closed and sealed.

TS1. 0-5 3-17-72

5. All automatic containment isolation valves are operable or are locked closed under administrative control,

6. The uncontrolled containment leakage satisfied Specification 4.4.

I. Abnormal Occurrence An abnormal occurrence is defined as:

1. Any unit condition that results in exceeding a safety limit or that results in safety system settings less conservative than the limiting safety system settings delineated in these Technical Specifications.

2. Any unit condition that results in violation of a limiting condition for operation as established in these Technical Specifications.

3. Any uncontrolled or unplanned release of radioactivity from the site.

4. Any abnormal degradation of one of the several boundaries which are designed to contain radioactive materials resulting from the fission process.

5. Uncontrolled or unanticipated change in reactivity, except for reactor trip greater than 1%Ll,k/k.

6. Engineered Safeguard System malfunction or other component or system malfunction which rendered or could render the Engineered Safeguard System incapable of performing its intended safety function.

7. Any observed inadequacy in the implementation of administrative or procedural controls during the operation of the facility which would significantly affect the safety of operatio.ns.

8. Occurrences or conditions involving an offsite threat to the safety of operation of the facility, including tornadoes, earthquakes, flooding, repetitious aircraft overflights, attempted sabotage,

TSl.0-6 3-17-72 civil disturbances, etc.

J. Unusual Safety Related Event Any unusual safety related event is defined as:

1. Discovery of any substantial errors in the transient or accident analyses, or in the methods used for such analyses, as described in the Final Safety Analyses Report or in the bases for the Technical Specifications.

2. Any substantial variance, in an unsafe or less conservative direction, from performance specifications contained in the Technical Specifications or from performance specifications, relevant to safety related equipment, contained in the Final Safety Analysis Report.

3. Any condition involving a possible single failure which, for a system designed against assumed single failures, could result in a loss of the capability of the system to perform its safety function .

. K. Qu.adrant Power Tilt The quadrant power tilt is defined as the ratio of the maximum upper excore detector current to the average of the upper excore detector currents or the ratio of the maximum lower excore detector current to the average of the lower excore detector currents whichever is greater. If one excore detector is out of service, the three in-service units are used in computing the average.

e l

TS 1. 0-7 3-17-72 L. Low Power Physics Tests Low power physics tests are tests conducted below 5% of rated power which measure fundamental characteristics of the reactor core and related instrumen-tation.

M. Interim Limits Additional limitations are imposed upon reactor core power distribution beyond previously established design bases consistent with interim bases for core cooling analysis established by the AEC in 1971. Two sets of power distribution parameters are* shown; both sets are to be met.

TS 2.1-1 3-17-72 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT, REACTOR* CORE Applicability Applies to the limiting combinations of thermal power, Reactor Coolant System pressure, coolant temperature and coolant flow when a reactor is critical.

Objective To maintain the *integrity of the fuel cladding.

Specification A.

The combination of reactor thermal power level, coolant pressure, and coolant temperature shall not:

1. Exceed the limits shown in TS Fi~ure 2.1-1 when full flow from three reactor coolant pumps exist.

2. Exceed the limits shown in TS Figure 2.1-2 when full flow from two reactor coolant pumps exist and the reactor coolant loop stop valves in the non-operating loop are open.

3. Exceed the limits shown in TS Figure 2.1-3 when full flow from two reactor coolant pumps exist and the reactor coolant loop stop valves in the non-operating loop are closed.

TS 2.1-2 3-17-72 B. The safety limit is exceeded if the combination of Reactor Coolant System average temperature and thermal power level is at any time above the appropriate pressure line in TS Figures 2.1-1, 2.1-2 or 2.1-3.

C. The reactor thermal power shall not exceed 1220 megawatts thermal until the results of the environmental qualification tests performed on the recirculation spray pump motors have been evaluated and approved in writing by the Atomic Energy Commission.

Basis To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under all operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the reactor coolant saturation temperature. The upper boundary of the nucleate boiling regime is termed Departure from Nucleate Boiling (DNB) and at this point there is a sharp reduction of the heat transfer coefficient, which would result in high clad temperatures and the possibility of clad failure. DNB is not, how-ever, an observable parameter during reactor operation. Therefore, the obser-vable parameters; thermal power, reactor coolant temperature and pressure have been related to DNB through the W-3 correlation. The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB.

e The minimum value of the DNB ratio (DNBR) during steady state operation, normal operational transients and anticipated transients, is limited to 1.30. A DNBR

TS 2.1-3 3-17-72 of 1.30 corresponds to a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions. (l)

The curves of TS Figure 2.1-1 which show the allowable power level decreasing with increasing temperature at selected pressures for constant flow (three loop operation) represent the loci of points of thermal power, coolant system average temperature, and coolant system pressure for which the DNB ratio is not less than 1.30. The area where clad integrity is assured is below these lines. In order to completely specify limits at all power levels, arbitrary constant upper limits of average temperature are shown for each pressure at powers lower than approximately 75% of rated power. The temper-ature limits at low power are considerably more conservative than would be required if they were based upon a minimum DNB ratio of 1.30 but are such that the plant conditions required to violate the limits are precluded by the self actuated safety valves on the steam generators.

The curves of TS Figures 2.1-2 and 2.1-3, which show the allowable power level decreasing with increasing temperature at selected pressures for constant flow (two loop operation), represent the loci. of points of thermal power, coolant system average temperature, and coolant system pressure for which either the DNB ratio is equal to 1.30 or the average enthalpy at the exit of the core is equal to the saturation value. At low pressures or high temperatures the average enthalpy at the exit of the core reaches saturation before the DNB ratio reaches 1.30 and, thus, this arbitrary limit is conservative with respect to maintaining clad integrity. In order to completely specify limits at all

TS 2.1-4 3-17-72 power levels, arbitrary constant upper limits of average temperatures are shown for each pressure at powers lower than approximately 45% of rated power.

The limits at low power as well as the limits based on the average enthalpy at the exit of the core are considerably more conservative than would be required if they were based upon a minimum DNB ratio of 1.30. The plant con-ditions required to violate these limits are precluded by the protection system and the self actuated safety valves on the steam generator. Upper limits o.f 70%

power for loop stop valves open and 75% with loop stop valves closed are shown to completely bound the area where clad integrity is assured. These latter limits are arbitrary but cannot be reached due to the Permissive 8 protection system setpoint which will trip the reactor on high nuclear flux when only two reactor coolant pumps are in service.

Operation with natural circulation or with only one loop in service is not allowed since the plant is not designed for continuous operation with less than two loops in service.

The curves are based on the following nuclear hot channel factors. These hot channel factors, instead of the interim ECCS criteria peaking factors, are utilized, because curves generated with the interim peaking factors would be less conservative.

FN = 2.72 q

N Ft.H 1.58 These hot channel factors are higher than those calculated at full power over the range between that of all control rod assemblies fully withdrawn to maxi-mum allowable control rod assembly insertion. The control rod assembly

TS 2.1-5 3-17-72 insertion limits are covered by Specification 3.12. Adverse power distribution factors could occur at lower power levels because additional control rod assemblies are in the core; however, the control rod assembly insertion limits dictated by TS Figure 3.12-1 ensure that the DNBR is always greater at partial power than at full power.

The hot channel factors are also sufficiently large to account for the degree of malpositioning of part-length control rod assemblies that is allowable before the reactor trip setpoints are reduced and control rod assembly with-drawal block and load runback action may be required. <2 )

The Reactor Control and Protection System is designed to prevent any anti-cipated combination of transient conditions for Reactor Coolant System temp-erature, pressure and thermal power level that would result in a DNB ratio of less than 1.30(3) based on stea~y state nominal operating power levels less than or equal to 100%, steady state nominal operating Reactor Coolant System average temperatures less*than or equal to 574.4°F and a steady state nomin~l operating pressure of 2235 psig. Allowances are made in initial conditions assumed for transient analyses for steady state errors of +2% in power, +4°F in Reactor Coolant System average temperature and _+30 psi in pressure. The combined steady state errors result in the DNB ratio at the start of a transient being 10 percent less than the value at nominal full power operating conditions. The steady state nominal operating parameters and allowances for steady state errors given above are also applicable for two loop operat;i.on except that the steady state nominal operating power level is less than or equal to 60%.

TS 2.1-6 3-17~72 The Commission is presently evaluating the results of the post-loss-of-coolant accident environmental qualification tests performed to determine the acceptability of the inside containment recirculation spray pump motors, Two of the motors are located outside the containment and would not be subjected to the post-loss-of-coolant accident environment. These two motors and their associated pumps provide adequate redundancy up to 50 percent of rated power. Accordingly, operation up to 50 percent of rated power (1220 megawatts thermal) is permitted. However, until the Commission has determined that the recirculation spray pump motors located in the containment are adequate for their intended service, operation above 50 percent of rated power is not permitted.

References

( 1) FSAR Section 3.4

( 2) FSAR Section 3.3

( 3) FSAR Section 14.2

TS FIGURE 2.1-1 3-17-72 SAFETY LIMITS REACTOR CORE THERMAL AND HYDRAULIC, THREE LOOP OPERATION, 100% FLOW 645 .

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TS FIGURE 2.1-3 3-17-72 SAFETY LIMITS REACTOR CORE THERMAL AND HYDRAULIC TWO LOOP OPERATION LOOP STOP VALVES CLOSED 650 + +

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TS 2.2-1 3-17-72 2.2 SAFETY LIMIT, REACTOR COOLANT SYSTEM PRESSURE Applicability Applies to the maximum limit on Reactor Coolant System pressure.

Objective To maintain the integrity of the Reactor Coolant System.

Specification The Reactor Coolant System pressure shall not exceed 2735 psig with fuel assemblies installed in the reactor vessel.

Basis The Reactor Coolant System(l) serves as a barrier which prevents radionuclides contained in the reactor coolant from reaching the environment. In the event of a fuel cladding failure the Reactor Coolant System is the primary barrier against the release of fission products. The maximum transient pressure allowable in the Reactor Coolant System pressure vessel under the ASME Code,Section III is 110% of design pressure. The maximum transient pressure allowable in the Reactor Coolant System piping, valves and fittings under USAS Section B31.l is 120% of design pressure. Thus, the safety limit of 2

2735 psig (110% of design pressure) has been established. < )

0 TS 2.2-2 3-17-72 The nominal settings of the power-operated relief valves at 2335 psig, the reactor high pressure trip at 2385 psig and the safety valves at 2485 psig are established to assure never reaching the Reactor Coolant System pressure safety limit. The initial hydrostatic test has been conducted at 3107 psig to assure the integrity of the Reactor Coolant System.

(1) FSAR Section 4 (2) FSAR Section 4.3

TS 2.3-1 3~17-72 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION Applicability Applies to trip and permissive settings for instruments monitoring reactor power; and reactor coolant pressure, temperature, and flow; and pressurizer level.

Objective To provide for automatic protective action in the event that the principal process variables approach a safety limit.

Specification A. Protective instrumentation settings for reactor trip shall be as follows:

e 1. Startup protection (a) High flux, power range (low set point) -

_ 25% of rated power.

(b) High flux, intermediate range (high set point) - current equivalent to< 25% of full power.

6 (c) High flux, source range (high set point) - neutron flux< 10 counts/sec.

2; Core Protection (a) High flux, power range (high set point) -

< 109% of rated power.

(b) High pressurizer pressure - < 2385 psig.

(c) Low pressurizer pressure - > 1860 psig.

l TS2.3-2 3-17-72 (d) Overtemperature 6T where 6T 0 = Indicated AT at rated thermal power, 0 p T = Average coolant temperature, op T' = 574.4°F p = Pressurizer pressure, psig P' = 2235 psig Kl = A constant = 1.12 (3 loop operation and 2 loop operation with the loop stop valves closed in the inoperable loop) 0.94 (2 loop operation with the loop stop valves in the inoperable loop open;)

K2 = A constant = 0. 0113 K3 = A constant = 0.00056 and f(6I) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be individually selected based on measured instrument response during startup tests such that:

(1) for (qt - qb) within+/- 19%, where qt and qb are the percent power in the top and bottom halves of the core respectively, and qt+ qb is total core thermal power in percent of rated thermal power, f(6I) = O.

(2) for each percent that the magnitude of (qt - qb) exceeds +19% the AT trip set point shall be automatically reduced by 2% of its value at rated po~er.

TS 2.3-3 3-17-72 (e) Overpower ~T

[ K4 - K5 dT K (T - T') - f (~I) ]

dt - 6 where No = Indicated Kr at rated thermal power, Op T = Average coolant temperature, Op T' = Average coolant temperature measured at nominal conditions anfi rated power, OF K4 = A constant = 1.10 Ks =~A for decreasing average temperature constant, for increasing average temperature, 0.2 sec/ OF K6 = A constant (for T > T') = O. 00083; K6:0 (for T< T' )*

f(AI) as defined in (d) above, (f) Low reactor coolant loop flow - ~ 90% of normal indicated loop flow as measure*d at elbow taps in each loop (g) Low reactor coolant pump motor frequency - .:::_ 57.5 Hz (h) Reactor coolant pump under voltage - > 70% of normal voltage

3. Other reactor trip setting (a) High pressuriz.er water level - ..::_ 92% of span (b) Low-low steam generator water level - > 5% of narrow range instrument span (c) Low steam generator water level - > 15% of narrow range instrument span in coincidence with steam/feedwater mismatch flow - < 1. Oxl06 lbs/hr (d) Turbine trip (e) Safety Injection - Trip settings for Safety Injection are detailed in T.S. Section 3.7.

TS 2.3-4 3-17-72 B. Protective instrumentation settings for reactor trip interlocks shall be as follows:

1. The reactor trips on low pressurizer pressure, high pressurizer level, turbine trip, and low reactor coolant flow for two or more loops shall be unblocked when power.::_ 10% of rated power.

2. The single loop loss of flow reactor trip shall be unblocked when the power range nuclear flux.:_ 50% of rated power. During two loop operation with the loop stop valves in the inactive loop open, this blocking setpoint, established by Permissive 8, may be increased to 60% of rated power only after the overtemperature ~T setpoint is adjusted .to the mandatory two loop value. For two loop operation with the loop stop valves of the inactive loop closed, Permissive 8 may be increased to 65% of rated power after the stop valves are closed. The overtemperature ~T setpoint may remain at the value for three loop operation 4uring two loop operation with the inactive loop stop valves closed.

3. The power range high flux, low setpoint trip and the intermediate range high flux, high setpoint trip shall be unblocked when power

< 10% of rated power.

4. The source range high flux, high setpoint trip shall be unblocked when the intermediate range nuclear flux is< 5 x 10-11 amperes.

Basis The power range reactor trip low setpoint provides protection in the power

TS 2.3-5 3-17-72 range for a power excursion beginning from low power. This trip value was used in the safety analysis. (l) The intermediate range high flux, low setpoint and source range high flux, high setpoint trips provide additional protection against uncontrolled startup excursions. As power level. increases, during startup, these trips are blocked to prevent unnecessary plant trips.

The high and low pressurizer pressure reactor trips limit the pressure range in which reactor operation is permitted. The high pressurizer pressure reactor trip is also a backup to the pressurizer code safety valves for overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig). The low pressurizer pressure reactor trip also trips the reactor 3

in the unlikely event of a loss-of-coolant accident. ( )

The overtemperature ~T reactor trip provides core protection against DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided only that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 3 seconds),

and pressure.is within the range between high and low pressure reactor trips.

With normal axial power distribution, the reactor trip limit, with allowance 2

for errors, ( ) is always below the core safety limit as .shown on TS Figure 2.1-1.

If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip limit is auto-(4) (5) matically reduced.

In order to operate with a reactor coolant loop out of service (two-loop operation) and with the stop valves of the inactive loop open, the overtemperature

~T trip setpoint calculation has to be modified by the adjustment of the variable

TS 2.3-6 3-17-72 K1

This adjustment, based on limits for two loop operation, provides sufficient margin to DNB for the aforementioned transients during two loop operation, The required adjustment and subsequent mandatory calibrations are made in the pro-tective system racks by qualified technicians* in the same manner as adjustments before initial startup and normal calibrations for three-loop operation. For two loop operation with the inactive loop stop valves closed, the overtemperature tT trip setpoints used for three loop operation are adequate to protect against DNB for all combinations of pressure, power, coolant temperature, and axial power distribution provided only that the transient is slow with respect to transit delays from the core to the temperature detectors.

The overpower tT reactor trip prevents power density anywhere in the core from exceeding 112% of design power density as discussed in Section 7 and specified in Section 14.2.2 of the FSAR and includes corrections for axial power dis-tribution, change in density and heat capacity of water with temperature 1 and dynamic compensation for piping delays from the core to the. loop temperature detectors. The specified setpoints meet this requirement and inc~ude allowance (2) for instrument errors.

The low flow reactor trip protects the core against DNB in the event of a sudden loss of power to one or more reactor coolant pumps. The setpoint (6) specified is consistent with the value used in the accident analysis.

The underfrequency reactor coolant pump trip protects against a decrease in flow caused by low electrical frequency. The specified setpoint assures a reactor trip signal before the low flow trip point is reached.

As used here, a qualified technician means a technician who meets the re-quirements of ANS-3. He shall have a minimum of two years of working experience in his speciality and at least one year of related technical training.

TS 2.3-7 3-17-72 The high pressurizer water level reactor trip protects the pressurizer safety valves against water relief. Approximately 1154 ft 3 of water corresponds to 7

92% of span. The specified setpoint allows margin for instrument error( )

and transient level overshoot beyond this trip setting so that the trip function prevents the water level from reaching the safety valves.

The low-low steam generator water level reactor trip protects against loss of feedwater flow accidents. The specified setpoint assures that there will be sufficient water inventory in the steam generators at the time of trip to 7

allow for starting delays for the Auxiliary Feedwater System. ( )

The specified reactor trips are blocked at low power where they are not required for protection and would otherwise interfere with normal unit operations. The prescribed setpoint above which these trips are unblocked assures their availability in the power range where needed.

Above 10% power, an automatic reactor trip will occur if two or more reactor coolant pumps are lost. Above 50% power during three-loop operation, an automatic reactor trip will occur if any pump is lost or de-energized.

This latter trip will prevent the minimum value of the DNBR from going below 1.30 during normal operational transients and anticipated transients when only two loops are in operation and the overtemperature ~T trip setpoint is adjusted to the valve specified for three loop operation. During two loop operation with the loop stop valves in the inactive loop open, and the over-temperature ~T trip setpoint is adjusted to the value specified for two loop operation, a reactor trip at 60% power will prevent the minimum value of DNBR from going below 1.30 during normal operational transients and anticipated transients when only two loops are in operation. During two loop operation with the inactive loop stop valves closed, a reactor trip at 65% power will

TS 2.3-8 3-17-72 prevent the minimum DNBR from going below 1.30 during normal operational transients and anticipated transients. For this latter case the overtemperature

~T trip setpoints may remain at the values used for three loop operation.

Although not necessary for core protection other reactor trips provide additional protection. The steam/feedwater flow mismatch in concidence with a low steam generator water level is designed for protection from a sudden loss of the reactor's heat sink. Upon the actuation of the safety injection circuitry, the reactor is tripped to decrease the severity of the accident condition. Upon turbine trip, at greater than 10% power, the reactor is tripped to reduce the severity of the ensuing transient, (1) FSAR Section 14.2.1 (2) FSAR Section 14.2 (3) FSAR Section 14.5 (4) FSAR Section 7.2 (5) FSAR Section 3.2.2 (6) FSAR Section 14.2.9 (7) FSAR Section 7.2

TS 3.1-1 3-17-72 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 REACTOR COOLANT SYSTEM Applicability Applies to the operating status of the Reactor Coolant System.

Objectives To specify those limiting conditions for operation of the Reactor Coolant System which must be met to ensure safe reactor operation.

These conditions relate to: operational components, heatup and cooldown, leakage, reactor coolant activity, oxygen and chloride concentrations, and minimum temperature for criticality.

A. Operational Components Specifications

1. Reactor Coolant Pumps

a. A reactor shall not be brought critical with less than two pumps, in non-isolated loops, in operation.

b. If an unscheduled loss of one or more reactor coolant pumps occurs while operating below 10% rated power (P-7) and results in less than two pumps in service, the affected

TS 3.1-2 3-17-72 plant shall be shutdo~m and the reactor made subcritical by inserting all control banks into the core. The shutdown rods may remain withdrawn.

c. A minimum of one pump in a non-isolated loop, or one residual heat removal pump and its associated flow path, shall be in operation during reactor coolant boron con-centration reduction.

d. Reactor power shall not exceed 50% of rated power with only two pumps in operation and the inactive loop stop valves open unless the overtemperature ~T trip setpoint, K1, for two loop operation, has been set at 0.94, after which power shall not exceed 60%. For two loop operation with the inactive loop stop valves closed, K1 may remain at the value for three loop operation and power shall not exceed 65%.

2. Steam Generator A minimum of two steam generators in non-isolated loops shall be operable when the average reactor coolant temperature is greater than 350oF.

3. Pressurizer Safety Valves

a. One valve shall be operable whenever the head is on the reactor vessel, except during hydrostatic tests.

TS 3.1-3 3-17-72

b. Three valves shall be operable when the reactor coolant average temperature is greater than 350°F, the reactor is critical, or the Reactor Coolant System is not connected to the Residual Heat Removal System.

c. Valve lift settings shall be maintained at 2485 psig +/-1 percent.

4. Reactor Coolant Loops Loop stop valves shall not be closed in more than one loop unless the Reactor Coolant System is connected to the Residual Heat Removal System and the Residual Heat Removal System is operable.

5. Pressurizer The reactor shall be maintained subcritical by at least 1% until the steam bubble is established and necessary sprays and heaters are operable.

Basis Specification 3.1.A-l requires that a sufficient number of reactor coolant pumps be operating to provide coastdown core cooling flow in the event of a loss of reactor coolant flow accident. This provided flow will maintain the DNBR above 1.30. (1) Heat transfer analyses also show that reactor heat equivalent to approximately 10% of rated power can be removed with natural circulation; however, the plant is not designed for critical operation with natural cir-culation or one loop operation and will not be operated with these conditions.

TS 3.1-4 3-17-72 When the boron concentration of the Reactor Coolant System is to be- reduced the process must be uniform to prevent sudden reactivity changes in the reactor. Mixing of the reactor coolant will be sufficient to maintain a uni-form concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place. The residual heat removal pump will circulate the equivalent of the reactor coolant system volume in approximately one half hour.

One steam,generator capable of performing its heat transfer function will provide sufficient heat removal capability to remove core decay heat after a normal reactor shutdown. Because of the low-low steam generator water level reactor trip, normal reactor criticality cannot be achieved without water in e the steam generators in r.eactor coolant loops with open loop stop valves. The requirement for two operable steam generators, combined with the requirements of Specification 3.6, ensure adequate heat removal capabilities for reactor coolant system temperatures of greater than 350°F.

Each of the pressurizer safety valves is designed to relieve 295,000 lbs.

per hr. of saturated steam at the valve setpoint. Below 350°F and 450 psig in the Reactor Coolant System, the Residual Heat Removal System can remove decay heat and thereby control system temperature and pressure. There are no credible accidents which could occur when the Reactor Coolant System is connected to the Residual Heat Removal System which could give a surge rate exceeding the capacity of one pressurizer safety valve. Also, two safety valves have a capacity greater than the maximum surge rate resulting from complete loss of load. <2 )

TS 3.1-5 3-17-72

- The limitation specified in item 4 above .on reactor coolant loop isolation will prevent an accidental isolation of all the loops which would eliminate the capability of dissipating core decay heat when the Reactor Coolant System is not connected to the Residual Heat Removal System.

The requirement for steam bubble formation in the pressurizer when the reactor has passed 1% subcriticality will ensure that the Reactor Coolant System will not be solid when criticality is achieved.

References:

(1) FSAR Section 14.2.9.

(2) FSAR Section 14.2.10

TS 3.1-6 3-17-72 B. HEATUP AND COOLDOWN Specification

1. Unit 1 reactor coolant temperature and pressure and the system heatup and cooldown (with the exception of the pressurizer) shall be limited in accordance with TS Figure 3.1-1.

Heatup:

The 0°F/hr, curve of Figure 3.1-1 may be used for heatup rates of up to 60°F/hr. below an indicated temperature of 304~F and 100°F/hr. above 304°F, Cooldown:

Allowable combinations of pressure and temperature for a specific cooldown rate are below and to the right of the limit lines for that rate as shown in TS Figure 3.1-1. This rate shall not exceed 50°F/hr. for temperatures at or below an indicated temperature of 308°F. For temperatures above an indicated temperature of 3QS°F, the rate shall not exceed 100°F/hr. The limit lines for rates between those shown in TS Figure 3.1-1 may be obtained by interpolation.

2. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the vessel is below 70°F.

TS 3.1-7 3-17-72 e

3.

The pressurizer heatup and cooldown rates shall not exceed 200°F/hr. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320°F.

4. TS Figure 3.1-1 shall be updated periodically in accordance with the following procedures, before the calculated maximum exposure of the vessel exceeds the exposure for which TS Figure 3.1-1 applies.

a. The curve based on 0.25% Cu, in TS Figure 3.1-2 shall be used to predict the increase in transition temperature based on integrated power unless measurements on the most recently examined irradiation specimens show that this is not appropriate. In this case,a new curve having the same slope as the original shall be constructed such that it is above all the applicable data points.

b. At or before the end. of the integrated power period for which TS Figure 3.1-1 applies, the limit lines on the figure shall be updated for a new integrated power period as follows. The total integrated reactor thermal power from startup to the end of the new period shall be converted to an equivalent integrated neutron exposure: The predicted increase in transition temperature at the end of the new period shall then be obtained from TS Figure 3.1-2 as revised by TS 3.1.B.4a above.

TS 3.1-8 3-17-72 c, The limit lines in TS Figure 3,1-1 shall be moved parallel to the temperature axis (horizontally) in the direction of increasing temperature a distance equivalent to the transition temperature increase obtained from TS Figure 3.1-2 as revised less the increment used for the end of the present period.

Basis All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to reactor system temperature and pressure (1) changes. These cyclic loads are introduced by normal unit load transients, reactor trips, and startup and shutdown operation. The number of thermal and loading cycles used for design purposes are shown in Section 4.1 of the FSAR, During unit startup and shutdown, the rates of temperature and pressure are limited. The maximum plant heatup and cooldown rate of 100°F/hr. is consistent with the design number of cycles and satisfies stress limits for (2) cyclic operation.

The allowable pressure vs. temperature is based on a temperature scale relative to the RT The RT is basically the drop weight NDTT of the material, NDT NDT as determined by ASTM E208. However, to assure that this value is conservative, and to guard against the possibility that material with low upper shelf toughness, or with a low rate of increase of toughness with temperature, is not properly evaluated, Charpy tests are also performed. If 35 mils lateral

TS 3.1-9 3-17-72 e expansion or 50-ft-lbs is not obtained at NDTT + 60, the RTNDT is shifted upward until these criteria are met.

This procedure of selecting RTN T assures that the K curve used to D IR calculate allowable pressures will be conservatively applicable to the material.

The procedure for determining the limiting RTNDT for the Reactor System is as follows:

1. Determine the highest RT of the material in the core region of the NDT reactor vessel, using original values and adding to this the predicted shift in RTNDT due to radiation during the service period for which this RTNDT applies, This takes into account the copper content of the material.

2. Examine the data for all other ferritic materials in the reactor system to assure that the RTNDT so selected is the highest in the system. If drop weight data are not available for all materials, the RTNDT of these shall be estimated in a conservative manner using trend data for the materials concerned, 3, For succeeding service periods, the same procedure as given in (1) above will be used unless test data from the surveillance program indicates that this will not be appropriate, In this event, the results of these tests will be used to predict the limiting RT NDT Test results on Tbt-:erial from the Surry Unit 1 reactor vessel is presented in FSAR Table 4.A-1. Using _the above procedure, the highest original RT of the core region plates is +20°F. No drop weight NDTT value is NDT

TS 3.1-10 3-17-72

- available for the core region weld material but on the basis of actual drop weight data on many similar weld materials, plus the actual Charpy values on this material, the drop weight NDTT is estimated to be 0°F, but until more data are obtained, will be assumed to be 40°F, The RTNDT for the first two years of operation will include a conservative estimate of the shift in RTNDT caused by radiation of 100°F. This, added to the original RTNDT of 40°F assumed for the welds, gives a reference RTNDT of 140°F to be used for the first two years of operation, or until the radiation shift is estimated to be over 100°F, In examining the data for the rest of the material in the vessel; as well as the properties for the other ferritic components of the reactor system, it is certain that all other materials will have RTNDT values significantly lower than 100°F.

Since the neutron spectra at the samples and vessel inside radius are identical, the measured (RT)NDT shift for a sample can be supplied with confidence to the adjacent section of reactor vessel for some later stage in plant life. The maximum exposure of the vessel is obtainable from the measured sample data by appropriate application of the calculated azimuthal neutron flux variation.

During cooldown, the thermal stress varies from tensile at the inner wall to compressive at the outer wall. The internal pressure superimposes a tensile stress on this thermal stress pattern, increasing the stress at the inside wall and relieving the stress at the outside wall. Therefore, the limiting stress always appears at the inside wall and the limit line has a direct

TS 3.1-11 3-17-72 dependence on cooldovm rate. This leads to a family of curves for cooldown, as shown in TS Figure 3.1-1. For heatup, the thermal stress is reversed and the location of the limiting stress is a function of heatup rate. The limit lines no longer bear the simple relationship to heatup as they do to cooldown rate. The 0°F/Hr cooldown line on TS Figure 3.1-1 bounds all limit lines for heatup rates up to 60°F/Hr for indicated temperatures at or below 304°F, and 100°F/Hr above 304°F.

TS Figure 3.1-1 defines stress limitations only. For nonnal operation other inherent plant characteristics, e.g., pump parameter and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure ranges.

The heatup and cooldown rate of 100°F/Hr for the steam generator is consistent with the remainder of the Reactor Coolant System, as discussed in the first paragraph of the basis. The stresses are within acceptable limits for the anticipated usage.

Temperature requirements for the steam generator correspond with the measured NDT for the shell. The spray should not be used if the temperature difference between the pressurizer and spray fluid is greater than 320°F. This limit is imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit.

TS 3.1-12 3-17-72

References:

(1) FSAR, Section 4.1.5 (2) ASME Boiler & Pressure Vessel Code,Section III, N-415 (3) ASME Boiler & Pressure Vessel Code,Section III, proposed non-mandatory Appendix G2000

TS 3.1-13 3-17-72

c. Leakage Specifications

1. Detected or suspected leakage from the Reactor Coolant System shall be investigated and evaluated. At least two means shall be available to detect reactor coolant system leakage. One of these means must depend on the detection of radionuclides in the containment.

2. If the leakage rate, from other than controlled leakage sources, such as the Reactor Coolant Pump Controlled Leakage Seals, exceeds 1 gpm and the source of the leakage is not identified within four hours of leak detection, the reactor shall be brought to hot shutdown. If the source of leakage is not* identified within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the reactor shall be brought to a cold shutdown condition.

3. If the sources of leakage are identified and the results of the evaluations are that continued operation is safe, operation of the reactor with a total leakage, other than leakage from controlled sources, not exceeding 10 gpm shall be permitted except as specified in C.4 below.

4. If it is determined that leakage exists through a non-isolable fault which has developed in a Reactor Coolant System component body, pipe ~ell, vessel wall, or pipe weld, the reactor shall be brought to a cold shutdown condition and corrective action taken prior to resumption of unit operation.

5. If the total leakage,other than leakage from controlled sources, e exceeds 10 gpm the reactor shall be placed in the cold shutdown condition.

TS 3.1-14 3-17-72 Basis Leakage from the Reactor Coolant System is collected in the containment or by other systems. These systems are the Main Steam System, Condensate and Feedwater System, the Gaseous and Liquid Waste Disposal Systems, the Component Cooling System, and the Chemical and Volume Control System.

Detection of leaks from the Reactor Coolant System is by one or more of the following:

1. An increased amount of makeup water required to maintain normal level in the pressurizer.

2. A high temperature alarm in the leakoff piping provided to collect reactor head flange leakage.

3. Containment sump water level indication.

4. Containment pressure, temperature, and humidity indication.

If there is significant radioactive contamination of the reactor coolant, the radiation monitoring system provi_des a sensitive indica.tion of primary I~

TS 3.1-15 3-17-72 system leakage. Radiation monitors which indicate primary system leakage include the containment air particulate and gas monitors, the condenser air ejector monitor, the component cooling water monitor, and the steam generator blowdown monitor.

I{._, f e rences FSAR, Section 4.2.7 - Reactor Coolant System Leakage fSAR, Section 14.3.2 - Rupture of a Main Steam Pipe D. Maximum Reactor Coolant Activity

.SJ1._, c'. Lf i cat ions The total specific activity of the reacto;.* coolant due to nuclides with half-lives of more than JO minutes shall not exceed (41 /E)µ Ci/cc whenever the reactor is critical or the average temperature is greater than 500°F, where Eis the average sum of the beta and gamma energies, in Mev, per disintegration, If this limit is not satisfied, the reactor shall be shut down and cooled to S00°F or less within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after detection.

Should this limit be exceeded by 25%, the reactor shall be made subcritical and cooled to 500°F or less within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after detection.

TS 3.1-16 3-17-72 Basis The specified limit provides protection to the public against the potential release of reactor coolant activity to the atmosphere, as demon~trated by the following analysis of a steam generator tube rupture accident.

Rupture of a steam generator tube would allow reactor coolant activity to enter the secondary system. The major portion of this activity is noble gases which would be vented to the containment on high radiation signal at the air ejector. Activity would continue to be released to the containment until the operator could reduce the primary system pressure below the set-point of the secondary relief valves and could isolate the faulty steam generator.

The worst credible set of circumstances is considered to be a double-ended break of a single tube, and this has been analyzed on the b~ses of isolation of the faulty steam generator by the operator within 30 min. after the event, see FSAR Section (14.3.1).

rhe reactor coolant was assumed to be at an equilibrium activity resulting from 1% failed fuel. The airborne release of the activity from the secondary system was assumed to be all of the noble gases that enter the secondary system and a 1.0 x 10-2 fraction of the halogens. Using a xlQ of 8.14 x 10- 4 sec/m 3 , the doses from immersion in a pulse duration cloud at the exclusion

TS 3.1-17 e 3-17-72 boundary would be 0.30 Rem whole body and 0.28 Rem thyroid. Thus, these doses are well below the guidelines suggested in 10CFRlOO.

The basis for the 500°F temperature contained in the Specification is that the saturation pressure corresponding to 500°F, 680.8 psia, is well below the pressure at which the atmospheric relief valves on the secondary side could be actuated.

Measurement of E will be performed at least twice annually. Calculations required to determine E will consist of the following:

1. Quantitative measurement, in units of µCi/cc, of the significant radionuclides with half lives greater than 30 minutes. These nuclides make up at least 95% of the total activity in the reactor coolant.

Table 9.1-5 of the FSAR lists significant radionuclides expected with 1% failed fuel.

2. A determination of the beta and gamma decay energy per disintegration of each nuclide determined in (1) above by applying known decay energies and schemes.

3. A calculation of Eby appropriate weighing of each nuclide's beta and gamma energy with its concentration as determined in (1) above.

E. Minimum Temperature For Criticality Specifications

TS 3.1-18 3-17-72

1. Except during low power physics tests, the reactor shall not be made critical at any temperature above which the moderator temperature coefficient is positive.

2. In no case shall the reactor be made critical with the reactor coolant temperature below DTT + 10°F, where the value of DTT + 10°F is as determined in Part B of this specification.

3. When the reactor coolant temperature is below the minimum temperature as specified in E-1 above, the reactor shall be subcritical by an amount equal to or greater than the potential reactivity insertion due to primary coolant depressurization.

Basis During the early part of the initial fuel cycle, the moderator temperature coefficient is calculated to be slightly positive at coolant temperatures below (1) (2) the power operating range. The moderator coefficient at low temperatures will be most positive at the beginning of life of the initial fuel cycle, when the boron concentration in the coolant is the greatest. Later in the initial cycle and during subsequent reload fuel cycles, the boron concentrations in the coolant will be lower and the moderator coefficients will be either less positive or will be negative. At all times, the moderator coefficient is negative in the

- 2 power opera t ing

range. (l) ( ) The maximum

. temperature at wh.ich t h e mo d era t or coefficient is positive at the beginning of life of the initial fuel cycle with all control rod assemblies withdrawn, is determined during pre-operational physics tests. When control rod assemblies are inserted, the temperature at which the

TS 3.1-19 3-17-72 e moderator coefficient becomes negative is lower so that at the temperature determined during the physics tests and with the operational control rod program, the temperature coefficient is expected to be negative.

The requirement that the reactor is not to be made critical when the moderator coefficient is positive has been imposed to prevent any unexpected power ex-cursion during normal operations as a result of either an increase of moderator temperature or decrease of coolant pressure. This requirement is waived during low power physics tests to permit measurement of reactor moderator coefficient and other physics design parameters of interest. During physics tests, special operating precautions will be taken. In addition, the strong negative Doppler 2

coefficient ( ) ( 3 ) and the small integrated ~k/k would limit the magnitude of e a power excursion resulting from a reduction of moderator density.

The requirement thRt the reactor is not to be made critical with a reactor coolant temperature below DTT + 10°F provides increased assurance that the proper relationship between reactor coolant pressure and temperature will be maintained during system heatup and pressurization whenever the reactor vessel is in the nil ductility transition temperature range. Heatup to this temperature is accomplished by operating the reactor coolant pumps.

If a specified shutdown reactivity margin is maintained (TS Section 3.12), there is no possibility of an accidental criticality as a result of an increase of moderator temperature or a decrease of coolant pressure.

(1) FSAR Figure 3.3-8 (2) FSAR Table 3.3-1 (3) FSAR Figure 3.3-9

TS 3.1-20 3-17-72 F. Maximum Reactor Coolant Oxygen, Chloride and Fluoride Concentration Specification

1. Concentrations of contaminants in the reactor shall not exceed any one of the following limits when the reactor coolant is above 250°F.

Normal Steady-State Transients not to Exceed Contaminant Operation* (PPM) 24 Hours (PPM)

a. Oxygen 0.10 1.00

b. Chloride 0.15 1.50 C, Fluoride 0.15 1.50 e 2. If any one of the normal steady-state operating limits as specified in 3.1.F.1 above are exceeded, or if it is anticipated that they may be exceeded, corrective action shall be taken immediately.

3. If the concentrations of any one of the contaminants can not be controlled within the limits of Specification 3.1.F.l above, the reactor shall be brought to the cold shutdown condition, utilizing normal operating procedures, and the cause of the out-of-specification operation ascertained and corrected. The reactor may then be restarted and operation resumed if the maximum concentration of any of the contaminants did not exceed the permitted transient values. Otherwise, a safety review is required before startup.

TS 3.1-21 3-17-72

4. Concentrations of contaminants in the reactor coolant shall not exceed the following maximum limits when the reactor coolant temperature is below 250°F:

Normal Concentration Transient not to Contaminant (PPM) exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (PPM)

a. Oxygen Saturated Saturated

b. Chloride 0.15 1.5

c. Fluoride 0.15 1.5 If the limits above are exceeded, the reactor shall be immediately brought to the cold shutdown condition and the cause of the out-of-specification condition shall be ascertained and corrected.

5. For the purposes of correcting the contaminant concentrations to meet technical specifications 3.1.F.l and 3.1.F.4 above, increase in coolant temperature consistant with operation of primary coolant pumps for a short period of time to assure mixing of ~he coolant shall be permitted. This increase in temperature to assure mixing shall in no case cause the coolant temperature to exceed 250°F.

6. If more than one contaminant or contaminants transient, which results in contaminant levels exceeding any of the normal steady state operation limits specified in 3.1.F.l or 3.1.F.4, is experienced in any seven consecutive day period, the reactor shall be placed in a cold shutdown condition until the cause of the out-of-specification operation is as-certained and corrected.

TS 3.1-22 3-17-72 Basis:

By maintaining the oxygen, chloride and fluoride concentrations in the reactor coolant below the limits as specified in technical specification 3.1.F.l and 3.1.F.4 the integrity of the reactor coolant system is assured under all operating (1) conditions. If these limits are exceeded, measures can be taken to correct the condition, e.g., replacement of ion exchange resin, or adjustment of the 2

hydrogen concentration in the volume control tank. ( ) Because of the time dependent nature of any adverse effects arising from oxygen, chloride, and fluoride concentration in excess of the limits, it is not necessary to shutdown innnediately if the condition can be corrected. Thus the period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for corrective action to restore concentrations within the limits has been established.

If the corrective action has not been effective at the end of the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, then the reactor will be brought to the cold shutdown condition and the corrective action will continue.

In restoring the contaminant concentrations to within specification limits in the event such limits were exceeded, mixing of the primary coolant with the reactor coolant pumps may be required. This will result in a small heatup of short duration which will not increase the average coolant temperature above 250°F.

More than one contaminant transient, in any seven consecutive day period, that results in exceeding normal steady state operation limits, could be indicative of unforeseen ~hemistry control problems. Such potential problems warrant in-vestigation, correction and measures to insure that the integrity of the reactor coolant system.is maintained.

TS 3.1-23 3-17-72 References (1) FSAR Section 4.2 (2) FSAR Section 9.2 e

UPPER PRESSURIZATION LIMITS FOR HEATUP AND COOLDOHN SURRY UNIT ONE 24-00 2200 2000

- 1800 er.,

Cl..

LLJ 0:: 1600 er.,

er., ~

LLJ 0::

14-00 Cl..

I-z:

1200

<t 0

0 u 1000 0

LLJ I-

<t 800 u

0

z: 600 4-00 200 **+

0 20 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 360 INDICATED COOLANT TEMPERATURE (°F) H

(/)

"rj Figure 3.1-1 Upper pressurization limit for heatup is w I-'*

I OQ

_... i::

depicted by "0°F/Hour" curve. Upper pressurization limits -.J rj for cooldown are dependent as depicted, on the cooldown I ro

-.J u.)

rates. This figure is applicable through an nvt of I\.) f-'

1.1 x 1018 neutrons per cm2 (E> 1 Mev) I f-'

TS Figure 3a l* 2 3-17-72 103 8

l-o I-z 10 2 CC:

IMev)

Figure 3.1-2. Radiation Induced Increase In Transition Temperature

TS 3.2-1 3-17-72 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM Applicability Applies to the operational status of the Chemical and Volume Control System.

Objective To define those conditions of the Chemical and Volume Control System necessary to ensure safe reactor operation.

Specification A. When fuel is in a reactor there shall be at least one flow path to the core for boric acid injection. The minimum capability for boric acid injection shall be equivalent to that supplied from the refueling water storage tank.

B, For one unit operation the reactor shall not be critical unless the following Chemical and Volume Control System conditions are met:

1. Two charging pumps shall be operable.
2. Two boric acid transfer pumps shall be operable.
3. The boric acid tanks (tank associated with the unit plus the common tank) together shall contain a minimum of 4200 gallons of at least 11.5% (but not greater than 13%) by weight boric acid solution at a temperature of at least 145°F.
TS 3.2-2 3-17-72
- 4. System piping and valves shall be operable to the extent of establishing two flow paths to the core; one flow path from the boric acid tanks to the charging pumps and a flow path from the refueling water storage tank to the charging pumps.
5. Two channels of heat tracing shall be operable for the flow paths requiring heat tracing.
6. Recirculation between a unit's Boron Injection Tank and the Boric Acid Tank(s) assigned to the unit shall be maintained.
C. For two unit operation the reactor shall not be critical unless the following Chemical and Volume Control System conditions are met:
1. Two charging pumps shall be operable per unit.
2. Three boric acid transfer pumps shall be operable.
3. When the common tank is in service, it shall be assigned to only one unit at a time. For that unit which has usage of the common tank, the boric acid tanks (unit's tank plus common tank) tog'ether shall contain a minimum of 4200 gallons of at least 11.5% (but not greater than 13%) by weight boric acid solution at a temperature of at least 145°F.
For that unit which does not have usage of the common tank, the unit's own tank shall contain a minimum of 4200 gallons of at least 11.5% (but not greater than 13%) by weight boric acid solution at a temperature of at least 145°F When the common tank is assigned to one unit, valves shall be positioned to establish a flow path to that unit and prevent flow to the other unit,
TS 3.2-3 3-17-72
4. System piping and valves shall be operable to the extent of establishing two flow paths to the core; one flow path from the boric acid tanks to the charging pumps and a flow path from the refueling water storage tank to the charging pumps.
5. Two channels of heat tracing shall be operable for the flow paths requiring heat tracing.
6. Recirculation between a unit's Boron Injection Tank and the Boric Acid Tank(s) assigned to the unit shall be maintained.
D. The requirements of Specifications Band C above may be modified to allow one of the following components to be inoperable at any one time.
If the system is not restored within the time period specified, the reactor shall be placed in the hot shutdown condition. If the requirements of Specification 3.2.B and Care not satisfied within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the reactor shall be placed in the cold shutdown condition.
1. One of the stipulated boric acid transfer pumps may be inoperable for a period not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided immediate attention is directed to making repairs.
2. Two charging pumps may be inoperable subject to the provisions of Specification 3.3-B.
3. One heat tracing circuit may be inoperable for a period not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided immediate attention is directed to making repairs.
TS 3.2-4 3-17-72 Basis The Chemical and Volume Control System provides control of the Reactor Coolant System Boron inventory. This is normally accomplished by using boric acid transfer pumps which discharge to the suction of each unit's charging pumps.
The Chemical and Volume Control System contains four boric acid transfer pumps.
Two of these pumps are normally assigned to each unit but valving and piping arrangements allow pumps to be shared such that 3 out of 4 pumps can service either unit. An alternate (not normally used) method of boration is to use the charging pumps taking suction directly from the refueling water storage tank. There are two sources of borated water available to the suction of the charging pumps through two different paths, one from the refueling water storage tank and one from the discharge of the boric acid transfer pumps.
A. The boric acid transfer pumps can deliver the boric acid tank contents (11.5% solution of boric acid) to the charging pumps.
B, The charging pumps can take suction from the volume control tank, the boric acid transfer pumps and the refueling water storage tank. Reference is made to Technical Specification 3.3.
The quantity of boric acid in storage from either the boric acid tanks or the refueling water storage tank is sufficient to.borate the reactor coolant in order to reach cold shutdown at any time during core life.
Approximately 4200 gallons of the 11.5% solution of boric acid are required to meet cold shutdown conditions. Thus, a minimum of 4200 gallons in the boric acid tank is specified. An upper concentration limit of 13% boric acid in the
TS 3.2-5 3-17-72 tank is specified to maintain solution solubility at the specified low temperature limit of 145°F. For redundancy, two channels of heat tracing are installed on lines normally containing concentrated boric acid solution.
Continuous recirculation between the Boron Injection Tank and the Boric Acid Tank(s) ensures that a unit's Boron Injection Tank is full of concentrated boric acid at all times. The Boric Acid Tank(s), which are located above the Boron Injection Tank(s), are supplied with level alarms, which would annunciate if a leak in the system occurred.
References FSAR Section 9.1 Chemical and Volume Control System
_J
TS 3.3-1 3-17-72 3.3 SAFETY INJECTION SYSTEM Applicability Applies to the operating status of the Safety Injection System.
To define those limiting conditions for operation that are necessary to provide sufficient borated cooling water to remove decay heat from the core in emergency situations.
Specifications A. A reactor shall not be made critical unless the following conditions are met:
1. The refueling water tank contains not less than 350,000 gal. of borated water with a boron concentration of at least 2000 ppm.
2. Each accumulator is pressurized to at least 600 psig and contains 3
a minimum of 934 ft and a maximum of 939 ft 3 of borated.water with a boron concentration of at least 1950 ppm.
3. The boron injection tank and isolated portions of the inlet and outlet piping contains no less than 900 gallons of water with a boron con-centration equivalent to at least 11.5% to 13% weight boric acid solution
TS 3.3-2 3-17-72 at a temperature of at least 145°F.
4. Two channels of heat tracing shall be available for the flow paths.
5. Two charging pumps are operable.
6. Two low head safety injection pumps are operable.
7. All valves, piping, and interlocks associated with the above components which are required to operate under accident conditions are operable.
8. The Charging Pump Cooling Water Subsystem shall be operating as follows:
a) Make-up water from the Component Cooling Water Subsystem shall be available.
b) Two charging pump component cooling water pumps and two charging pump service water pumps shall be operable.
c) Two charging pump intermediate seal coolers shall be operable.
B. The requirements of Specification 3.3-A may be modified to allow one of the following components to*be* inoperable*at any' one time. If the system is not restored to meet the requirements of Specification 3.3-A within the time period specified, the reactor shall initially be placed in the hot shutdown condition. If the requirements of Specification 3.3-A are not satisfied within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months the reactor shall be placed in the cold shutdown condition.
TS 3.3-3 3-17-72
1. One accumulator may be isolated for a period not to exceed 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
2. Two charging pumps per unit may be out of service, provided immediate attention is directed to making repairs and one pump is restored to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If one pump unit is out of service, the standby pump shall be tested before initiating maintenance and once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to assure operability.
3. One low head safety injection pump per unit may be out of service, provided immediate attention is directed to making repairs and the pump is restored to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The other low head safety injection pump shall be tested to demonstrate operability prior to initiating repair of the inoperable pump and shall be tested once every eight (8) hours thereafter, until both pumps are in an operable status or the reactor is shut down.
4. Any one valve in the iafatr Injection ~yst~m mar be inoperable provided repairs are initiated immediately and are completed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Prior to initiating repairs, all automatic valves in the redundant system shall be tested to demonstrate operability.
5. One channel of heat tracing may be inoperable for a period not to exceed 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s*, provided immediate attention is directed to making repairs.
6. One charging pump component cooling water pump or one charging pump service water pump may be out of service provided the pump is
TS 3.3-4 3-17-72
- 7.
restored to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
One charging pump intermediate seal cooler or other passive component may be out of service provided the system may still operate at 100 percent capacity and repairs are completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .
C. The accumulator discharge valves (MOV 865A, B, & C) in non-isolated loops shall be blocked open by de-energizing the valve motor operator when the reactor coolant pressure is greater than 1000 psig.
Basis The normal procedure for starting the reactor is, first, to heat the reactor coolant to near operating temperature by running the reactor coolant pumps.
The reactor is then made critical by withdrawing control rods and/or diluting boron in the coolant. With this mode of startup the Safety Injection System is required to be operable as specified. During low power physics tests there is a negligible amount of energy stored in the system; therefore an accident comparable in severity to the Design Basis Accident is not possible, and the full capacity of the Safety Injection System is not required, The operable status of the various systems and components is to be demonstrated by periodic tests, detailed in TS Section 4.1. A large fraction of these tests are performed while the reactor is operating in the power range. If a component is found to be inoperable, it will be possible in most cases to effect repairs and restore the system to full operability within a relatively short time. A
TS 3.3-5 3-17-72 single component being inoperable does not negate the ability of the system to perform its function, but it reduces the redundancy provided in the reactor design and thereby limits the ability to tolerate additional equipment failures.
To provide maximum assurance that the redundant component(s) wili operate if required to do so, the redundant component(s) are to be tested prior to initiating repair of the inoperable component and, in some cases are to be retested at intervals during the repair period. In some cases, i.e.chargirig pumps, additional components are installed to allow a component to be inoperable without affecting system redundancy. For those cases which are not so designed, if it develops that (a) the inoperable component is not repaired within the specified allowable time period, or (b) a second component in the same or related system is found to be inoperable, the reactor will initially be put in the hot shutdown condition to provide for reduction of the decay heat from the e fuel, and consequent reduction of cooling requirements after a postulated loss-of-coolant accident. After 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months in the hot shutdown condition, if the malfunction(s) are not corrected the reactor will be placed in the cold shutdown condition, following normal shutdown and cooldown procedures.
The Specification requires prompt action to effect repairs of an inoperable component, and therefore in most cases repairs will be completed in less than the specified allowable repair times. Furthermore, the specified repair times do not apply to regularly scheduled maintenance of the Safety Injection System, which is normally to be performed during refueling shutdowns. The limiting times for repair are based on: estimates of the time required to diagnose and correct various postulated malfunctions using safe and proper procedures, the availability of tools, materials and equipment; health physics requirements and the extent to which other systems provide functional redundancy to the system under repair.
TS 3.3-6 3-17-72 Assuming the reactor has been operating at full rated power for at least 100 days, the magnitude of the decay heat production decreases as follows after initiating hot shutdown.
Time After Shutdown Decay Heat, % of Rated Power 1 min. 3.7 30 min. 1.6 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1.3 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 0.75 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 0.48 Thus, the requirement for core cooling in case of a postulated loss-of-coolant accident while in the hot shutdown condition is reduced by orders of magnitude below the requirements for handling a postulated loss-of-coolant accident occuring during power operation. Placing and maintaining the reactor in the hot shutdown condition significantly reduces the potential consequences of a loss-of-coolant accident, allows access to some of the Safety Injection System components in order to effect repairs, and minimizes the exposure to thermal cycling.
Failure to complete repairs within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of going to hot shutdown condition is considered indicative of unforeseen problems, i.e., possibly the need of major maintenance. In such a case the reactor is to be put into the cold shutdown condition.
e
TS 3.3-7 3-17-72 The accumulators (one for each loop) discharge into the cold legs of the reactor coolant piping when Reactor Coolant System pressure decreases below accumulator pressure, thus assuring rapid core cooling for large breaks. The line from each accumulator is provided with a motorized valve to isolate the accumulator during reactor start-up and shutdown to preclude the discharge of the contents of the accumulator when not required. These valves receive a signal to open when safety injection is initiated.
The AEC requires that these valves receive a signal to open when the Reactor Coolant Pressure exceeds a preselected value. Such a feature to satisfy this requirement will be installed prior to startup after the first refueling. In the interim period the valves will be blocked open by de-energizing the valve motor operators when the reactor coolant pressure exceeds 1000 psig.
The operating pressure of the Reactor Coolant System is 2235 psig. and safety injection is initiated when this pressure drops to 650 psig. De-energizing the motor operator when the pressure exceeds 1000 psig allows sufficient time during normal start-up operation to perform the actions required to de-energize the valve. This procedure will assure that there is an operable flow path from each accumulator to the Reactor Coolant System during power operation and that safety injection can be accomplished.
References FSAR Section 9.1 Chemical and Volume Control System FSAR Section 6.2 Safety Injection System
~ FSAR Technical Specifications Section 4.1 FSAR Supplement Addendum S6.25
TS 3.3-8 3-17-72 References FSAR Section 9.1 Chemical and Volume Control System FSAR Section 6.2 Safety Injection System FSAR Technical Specifications Section 4.1 FSAR Supplement Addendum S6.25
TS 3.4-1 3-17-72 3.4 SPRAY SYSTEMS Applicability Applies to the operational status of the Spray Systems.
Objective To define those conditions of the Spray Systems necessary to assure safe unit operation.
Specification A. A unit~s Reactor Coolant System temperature or pressure shall not be made to exceed 350°F or 450 psig, respectively, or the reactor shall not be made critical unless the following Spray .System conditions in that unit are met:
1. Two Containment Spray Subsystems, including containment spray pumps, piping, and valves shall be operable.
2. Four Recirculation Spray Subsystems, including recirculation spray pumps, coolers, piping, and valves shall be operable.
3. The refueling water storage tank shall contain not less than e 350,000 gal of borated water at a maximum temperature as shown in TS Fig. 3.8-1.
TS 3.4-2 3-17-72 If this volume of water cannot be maintained by makeup, or the temperature maintained below that specified in TS Fig. 3.8-1, the reactor shall be shutdown until repairs can be made. The water shall be borated to a boron concentration not less than 2,000 ppm which will assure that the reactor is in the refueling shutdown condition when all control rod assemblies are inserted.
4. The refueling water chemical addition tank shall contain not less than 3,360 gal of solution with a sodium hydroxide con-centration of not less than 18*percent by weight.
S. All valves, piping, and interlocks associated with the above components which are required to operate under accident conditions shall be operable.
B. During power operation the requirements of specification 3.4-A may be modified to allow the following components to be inoperable. If the com-ponents are not restored to meet the requirement of Specification 3.4-A within the time period specified below, the reactor shall be placed in the hot shutdown condition. If the requirements of Specification 3.4-A are not satisfied within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months the reactor shall be placed in the cold shutdown condition using normal operating procedures.
1. One Containment Spray Subsystem may be out of service, provided immediate attention is directed to making repairs and the subsystem can be restored to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The other Con-tainment Spray Subsystem shall be tested as specified in Specification 4.5-A to demonstrate operability prior to initiating repair of the I
inoperable system.
TS 3.4-3 3-17-72
2. One containment spray pump dri.ve, ei.ther motor or turbine, may be out of service, provided immediate attention is directed to making repairs and the drive can be restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If the inoperable drive is uncoupled from its pump, returning the effected spray pump and its remaining drive to an operable status within the time requirement of Specification 3.4.B-l, shall satisfy the requi.rements of Specifi.cation 3.4.A-l.
3. One outside Recirculation Spray Subsystem may be out of service provided immediate attention is directed to making repairs and the subsystem can be restored to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
The other Recirculation Spray subsystems shall be tested as specified in Specification* 4.5-A to demonstrate operability prior to initiating repair of the inoperable system.
4. One inside Recirculation Spray Subsystem may be out of service provi.ded immediate attention is directed to making repairs and the subsystem can be restored to operable'status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .
The other Recirculation Spray subsystems shall be tested as specified in Specification 4.5-A to demonstrate operability prior to initiating repair of the inoperable subsystems.
C. Should the refueiing water storage tank temperature fai.l to be maintained at or below 45°F, the containment pressure and temperature shall be maintained in accordance with TS Fi.g. 3.8-1 to maintain the capability of the Spray System with the higher refueling water temperature. If the containment' temperature and pressure cannot be maintained within the limits of TS Fig. 3.8.:,;l, the reactor shall be placed in the cold shutdown condition.
TS 3.4-4 3-17-72 Basis The Spray Systems in each reactor unit consist of two separate parallel Contain-ment Spray Subsystems, each of 100 percent capacity, and four separate parallel Recirculation Spray Subsystems, each of 50 percent capacity.
Each Containment Spray Subsystem draws water independently from the 350,000 gal capacity refueling water storage tank. The water in the tank is cooled to 45°F or below by circulating the tank water through one of the two refueling water storage tank coolers through the use of one of the two refueling water recirculation pumps. The water temperature is maintained by two mechanical refrigeration units as required. In each Containment Spray Subsystem, the water flows from the tank through a dual drive (motor and turbine) containment spray pump and is sprayed into the containment atmosphere through two separate sets of spray nozzles. The capability of the Spray Systems to depressurize the containment in the event of a Design Basis Accident is a function of the pressure and temperature of the containment atmosphere, the service water temperature, and the temperature in the refueling water storage tank as discussed in Specification 3.8-B.
Each Recirculation Spray Subsystem draws water from the common containment sump. In each subsystem the water flows through a recirculation spray pump and recirculation spray cooler, and is sprayed into the containment atmosphere through a separate set of spray nozzles. Two of the recirculation spray pumps are located inside the containment and two outside the containment in the containment auxiliary structure.
TS 3.4-5 3-17-72 With one Containment Spray Subsystem and two Recirculation Spray Subsystems operating together, the Spray Systems are capable of cooling and. depressurizing the containment to subatmospheric pressure in less than 40 minutes following the Design Basis Accident. The Recirculation Spray Subsystems are capable of maintaining subatmospheric pressure in the containment indefinitely following the Design Basis Accident when used in conjunction with the Containment Vacuum System to remove any long term air inleakage.
In addition to supplying water to the Containment Spray System, the refueling water storage tank is also a source of water for safety injection following an accident. This water is borated to a concentration which assures reactor shutdown by approximately 10 percent ~k/k when all control rod assemblies are inserted and when the reactor is cooled down for refueling.
References FSAR Section 4 Reactor Coolant System FSAR Section 6.3.1 Containment Spray Subsystem FSAR Section 6.3.1 Recirculation Spray Pumps and Coolers FSAR Section 6.3.1 Refueling Water Chemical Addition Tank FSAR Section 6.3.1 Refueling Water Storage Tank FSAR Section 14.5.2 Design Basis Accident FSAR Section 14.5.5 Containment Transient Analysis
TS 3.5-1 3-17-72 3.5 RESIDUAL HEAT REMOVAL SYSTEM Applicability Applies to the operational status of the Residual Heat Removal System.
Objective To define the limiting conditions for operation that are necessary to remove decay heat from the Reactor Coolant System in normal shutdown situations.
Specification A. The reactor shall not be made critical unless:
1. Two residual heat removal pumps are operable.
2. Two residual heat exchangers are operable.
3. All system piping and valves, required to establish a flow path to and from the above components, are operable.
4. All Component Cooling System piping and valves, required to establish a flow path to and from the above components, are operable.
B. The requirements of Specification A may be modified to allow one of the following components (including associated valves and piping) to be in-operable at any one time. If the system is not restored to meet the re-quirements of Specification A within 14 days, the reactor shall be shutdown.
TS 3.5-2 3-17-72
1. One residual heat removal pump may be out of service, provided immediate attention is directed to making repairs.
2. One residual heat removal heat exchanger may be out of service, provided immediate attention is directed to making repairs.
C. Electrical power to the motor operated valve in the line connecting the Residual Heat Removal System to the Reactor Coolant System (MOV 1701 for Unit 1 and MOV 2701 for Unit 2) shall be locked out with the valve in the closed position when the reactor coolant pressure exceeds 465 psig.
Basis The Residual Heat Removal System is required to bring the Reactor Coolant System fron conditions of approximately 350°F and pressures between 400 and e 450 psig to cold shutdown conditions. Heat removal at greater temperatures is by the Steam and Power Conversion System. The Residual Heat Removal System is provided with two pumps and two heat exchangers. If one of the two pumps and/or one of the two heat exchangers is not operative, safe operation of the unit is not affected; however, the time for cooldown to cold shutdown conditions is extended.
'rhe AEC re(!uires that the series motorized valves in the line connecting the RHRS and RCS be provided with pressure interlocks to prevent them from opening when the reactor coolant system is at pressure. Such a feature to satisfy the intent of this criterion shall be installed prior to startup after the first refueling, In the interim period Specification 3.5.C shall be implemented to insure that the RHR system cannot be overpressurized by de-energizing the motor operated valve which is not already pressure interl0cked. The
TS 3.5-3 3-17-72 pressure at which this procedure will be implemented is the same as the set-point of the motor operated valve which is interlocked with reactor coolant pressure.
References FSAR Section 9.3 Residual Heat Removal System
TS 3.6-1 3-17-72 3.6 TURBINE CYCLE Applicability Applies to the operating status of the Main Steam and Auxiliary Steam Systems Objective To define the conditions required in the Main Steam System and Auxiliary Steam System for protection of the steam generator and to assure the capability to remove residual heat from the core during a loss of station power.
- Specification A. A unit's Reactor Coolant System temperature or pressure shall not exceed 350°F or 450 psig, respectively, or the reactor shall not be critical unless the five main steam line code safety valves associated with each steam generator in unisolated reactor coolant loops, are operable.
B. To assure residual heat removal capabilities, the following conditions shall be met prior to the commencement of any unit operation that would establish reactor coolant system conditions of 350°F and 450 psig which would preclude operation of the Residual Heat Removal System.
1. Two of the three auxiliary feedwater pumps shall be operable.
TS 3.6-2 3-17-72
2. A minimum of 96,000 gal of water shall be available in the tornado missile protected condensate storage tank to supply emergency water to the auxiliary feedwater pump suctions.
3. All main steam line code safety valves, associated with steam generators in unisolated reactor coolant loops, shall be operable.
4. System piping and valves required for the operation of the components enumerated in Specification B.l, 2, and 3 shall be operable.
C. The iodine - 131 activity in the secondary side of any steam generator, in an unisolated reactor coolant loop, shall not exceed 9 curies.
D. The requirements of Specification B-2 above may be modified to allow utilization of protected condensate storage tank water with the auxiliary steam generator feed pumps provided the water level is maintained above 60,000 gallons, sufficient replenishment water is available in the 300,000 gallon condensate storage tank, and replenishment of the protected con-densate storage tank is commenced within two hours after the cessation of protected condensate storage tank water cpnsumption.
Basis A reactor which has been shutdown from power requires removal of core residual heat. While reactor coolant temperature or pressure is greater than 350°F or
TS 3.6-3 3-17-72 450 psig, respectively, residual heat removal requirements are normally satis-fied by steam bypass to the condenser. If the condenser is unavailable, steam can be released to the atmosphere through the safety valves, power operated relief valves, or the 4 inch decay heat release line.
The capability to supply feedwater to the generators is normally provided by the operation of the Condensate and Feedwater Systems. In the event of complete loss of electrical power to the station, residual heat removal would continue to be assured by the availability of either the steam driven auxiliary feedwater pump or one of the motor driven auxiliary feedwater pumps and the 100,000 gallon condensate storage tank.
A minimum of 92,200 gallons of water in the 110,000 gallon condensate tank is sufficient for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of residual heat removal following a reactor trip and loss of all off-site electrical power. If the protected condensate storage tank level is reduced to 60,000 gallons, the immediately available repl~nishment water in the 300,000 gallon condensate tank can be gravity-feed to the pro-tected tank if required for residual heat removal. An alternate supply of feed-water to the auxiliary feedwater pump suctions is also available from the Fire Protection System Main in the auxiliary feedwater pump cubicle.
The five main steam code safety valves associated with each steam generator have a total combined capacity of 3,725,575 pounds per hour at their individual set pressure; the total combined capacity of all fifteen main steam code safety valves is 11,176,725 pounds per hour. The ultimate power rating steam flow is 11,167,923 pounds per hour. The combined capacity of the safety valves required by Specification 3.6 always exceeds the total steam flow corresponding to the maximum steady-state power than can be obtained during one, two or three reactor
TS 3.6-4 3-17-72 reactor coolant loop operation.
The availability of the auxiliary feedwater pumps, the protected condensate storage tank, and the main steam line safety valves adequately assures that sufficient residual heat removal capability will be available when required.
The limit on steam generator secondary side iodine - 131 activity is based on limiting inhalation thyroid dose at the site boundary to 1.5 rem after a postulated accident that would resuit in the release of the entire contents of a unit's steam generators to the atmosphere. In this accident, with the halogen inventories in the steam generator being at equilibrium values, 1-131 would contribute 75 percent of the resultant thyroid dose at the site boundary; the remaining 25 percent of the does is from other isotopes of iodine. In the analysisJone-tenth of the contained iodine is assumed to reach the site boundary, making allowance for plate out and retention in water droplets.
The inhalation thyroid dose at the site boundary is given by:
Dose (Rem)= (C) (x/Q) (D 00 /AT) (B.R.)
( . 7 5) (P . F . )
where: C = steam generator I-131 activity (curies) x/Q 8.14 X 10-4 sec/m 3 D00 /AT = 1.48 x 106 rem/Ci for I-131 B.R. = breathing rate, 3.47 x 10-4 m3/sec.
from TID 14844 P.F. = plating factor, 10 Assuming the postulated accident, the resultant thyroid dose is
1. 5 rem.
L
TS 3.6-5 3-17-72
- The steam generator's specific iodine - 131 activity limit is calculated by dividing the total activity limit of 9 curies by the water volume of a steam generator. At full power, with a steam generator water volume of 47.6 M3, the specific iodine - 131 limit would be .18 µCi/cc; at zero power, with a steam generator water volume of 101 M3, the specific iodine - 131 limit would be .089 µCi/cc.
References FSAR Section 4 Reactor Coolant System FSAR Section 9.3 Residual Heat Removal System FSAR Section 10.3.1 Main Steam System FSAR Section 10.3. 2 Auxiliary Steam System FSAR Section 10.3.5 Auxiliary Feedwater Pumps FSAR Section 10.3.8 Vent and Drain Systems FSAR Section 14.3.2.5 Environmental Effects of a Steam Line Break
TS 3.7-1 3-17-72 I
3.7 INSTRUMENTATION SYSTEMS Operational Safety Instrumentation Applicability:
Applies to reactor and safety features instrumentation systems.
Objectives:
To provide for automatic initiation of the Engineered Safety Features in the event that principal process variable limits are exceeded, and to delineate the conditions of the plant instrumentation and safety circuits necessary to ensure reactor safety.
Specification:
A, For on-line testing or in the event of a sub-system instrumentation channel failure, plant operation at rated power shall be permitted to continue in accordance with TS Tables 3.7-1 through 3.7-3.
B. In the event the number of chann.els of a particular sub-system in service falls below the limits given in the column entitled Minimum Operable Channels, or Minimum Degree of Redundancy cannot be achieved, operation shall be limited according to the requirement shown in Column 4 of TS Tables 3.7-1 through 3.7-3.
TS 3.7-2 3-17-72
c. In the event of sub-system instrumentation channel failure permitted by specification 3.7-n, TS Tables 3.7-1 through 3.7-3 need not be observed during the short period of time the operable sub-system channels are tested where the failed channel must be blocked to prevent unnecessary reactor trip.
D. The Engineered Safety Features initiation instrumentation setting limits shall be as stated in TS Table 3.7-4.
E. Automatic functions operated from radiation monitor alarms shall be as stated in T~ Table 3.7-5.
Basis Instrument Operating Conditions During plant operations, the complete instrumentation systems will normally be in service. Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established limits. Safety is not compromised, however, by continuing operation with certain instrumentation channels out of service since provisions were made for this in the plant design. This specification outlines limiting conditions for operation necessary to preserve the effectiveness of the Reactor Control and Protection System when any one or more of the channels is out of service .
.Almost all reactor protection channels are supplied with sufficient redundancy to provide the capability for channel calibration and test at power. Exceptions
TS 3.7-3 3-:,17-72 are backup channels such as reactor coolant pump breakers. The removal of one trip channel on process control equipment is accomplished by placing that channel bis table in a tripped mode; e.g., a two-out-of-three circuit becomes a one-out-of-two circuit. The nuclear instrumentation system channels are not intentionally placed in a tripped mode since the test signal is superimposed on the normal detector signal to test at power. Testing of the NIS power range channel requires: (a) bypassing the Dropped Rod protection from NIS, for the channel being tested; and (b) placing the 6T/T protection CHANNEL avg SET that is being fed from the NIS channel in the trip mode and (c) defeating the power mismatch section of T control channels when the appropriate NIS avg channel is being tested. However, the Rod Position System and remaining NIS channels still provide the dropped-rod protection. Testing does not trip the system unless a trip condition exists in a concurrent channel.
Instrumentation has been provided to sense accident conditions and to (1) initiate operation of the Engineered Safety Features Safety Injection System Actuation Protection against a Loss of Coolant or Steam Break Accident is brought about by automatic actuation of the Safety Injection System which provides emergency cooling and reduction of reactivity.
The Loss of Coolant Accident is characterized by depressurization of the Reactor Coolant System and rapid loss of reactor coolant to the containment.
The Engineered Safeguards Instrumentation has been designed to sense these effects of the Loss of Coolant accident by detecting low pressurizer pressure
TS 3.7-4 3-17-72
- and level and to generate signals actuating the SIS active phase based upon the coincidence of these signals. The SIS active phase is also actuated by a high containment pressure signal brought about by loss of high enthalpy coolant to the containment. This actuation signal acts as a backup to the low pressurizer pressure and level signal actuation of the SIS and also adds diversity to protection against loss of coolant, Signals are also provided to actuate the SIS upon sensing the effects of a steam line break accident. Therefore, SIS actuation following a steam line break is designed to occur upon sensing high differential steam pressure between the steam header and steam generator line or upon sensing high steam line flow in coincidence with low reactor coolant average temperature or low
- steam line pressure.
The increase in the extraction of RCS heat following a steam line break results in reactor coolant temperature and pressure reduction. For this reason protection against a steam line break accident is also provided by coincident low pressurizer pressure and level signals actuating safety injection.
Protection is also provided for a steam line break in the containment by actuation of SIS upon sensing high containment pressure.
SIS actuation injects highly borated fluid into the Reactor Coolant System in order to counter the reactivity insertion brought about by cooldown of the reactor coolant which occurs during a steam line break accident.
TS 3.7-5 3-17-72 Containment Spray The Engineered Safety Features also initiate containment spray upon sensing a high-high containment pressure signal. The containment spray acts to reduce containment pressure in the event of a loss of coolant or steam line break accident inside the containment, The containment spray cools the containment directly and limits the release of fission products by absorbing iodine should it be released to the containment.
Containment spray is designed to be actuated at a higher containment pressure (approximately 50% of design containment pressure) than the SIS (10% of design).
Si.nee spurious actuation of containment spray is to be avoided, it is initiated only on coincidence of high-high containment pressure sensed by 3 out of the 4 containment pressure signals provided for its actuation.
Steam Line Isolation Steam line isolation signals are initiated by the Engineered Safety Features closing all steam line trip valves. In the event of a steam line break, this action prevents continuous, uncontrolled steam release from more than one steam generator by isolating the steam lines on high-high containment pressure or high steam line flow with coincident low steam line pressure or low reactor coolant average tempera-ture. Protection is afforded for breaks inside or outside the containment even when it is assumed that there is a single failure in the steam line isolation system.
Feedwater Line Isolation The feedwater lines are isolated upon actuation of the Safety Injection System
TS 3.7-6 3-17-72 in order to prevent excessive cooldown of the reactor coolant system. This mitigates the effect of an accident such as steam break which in itself causes excessive coolant temperature cooldown.
Feedwater line isolation also reduces the consequences of a steam line break inside the containment, by stopping the entry of feedwater.
Setting Limits
1. The high containment pressure limit is set at about 10% of design containment pressure. Initiation of Safety Injection protects against 2 3 loss of coolant ( ) or steam line break ( ) accidents as discussed in the safety analysis.
2. The high-high containment pressure limit is set at about 50% of design containment pressure. Initiation of Containment Spray and Steam Line Isolation protects against large loss of coolant ( 2 ) or steam line break accidents ( 3 ) as discussed in the safety analysis.
3. The pressurizer low pressure setpoint for safety injection actuation is set substantially below system operating pressure limits. However, it is sufficiently high to protect against a loss-of-coolant accident as shown int h e sa f ety ana 1 ysis.
' ( 2)
The pressurizer low level limit is set sufficiently high to protect against a loss of coolant accident as shown in the accident analysis.
4. The steam line high differential pressure limit is set well below
TS 3.7-7 3-17-72 the differential pressure expected in the event of a large stea~
(3) line break accident as shown in the safety analysis.
5. The high steam line flow limit is set approximately 20% of the full steam flow at no load and at 120% of full steam flow at full load, with the high steam line flow differential pressure setpoint linearly programmed between no load and full load in order to protect against large steam line break accidents. The coincident low T setting limit for SIS and avg steam line isolation initiation is set below its hot shutdown value. The coincident steam line pressure setting limit is set below the full load operating pressure. The safety analysis shows that these settings provide (3) protection in the event of a large steam line break.
e Automatic Functions Operated from Radiation Monitors The Process Radiation Monitoring System continuously monitors selected lines containing or possibly containing, radioactive effluent. Certain channels in this system actuate control valves on a high-activity alarm signal. Additional 4
information on the Process Radiation Monitoring System is available in the FSAR. ( )
Reference (1) FSAR - Section 7,5 (2) FSAR - Section 14.5 (3) FSAR - Section 14.3.2 (4) FSAR - Section 11.3.3
e TABLE 3. 7-1 REACTOR TRIP INSTRUMENT OPERATING CONDITIONS 1 2 3 4 OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI-OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET
1. Manual 1 Maintain hot shutdown
2. Nuclear Flux Power Range 3 2 Low trip setting when 2 Maintain hot of 4 power channels greater shutdown than 10% of full power
3. Nuclear Flux Intermediate Range 1 2 of 4 power channels greater Maintain hot than 10% full power shutdown
4. Nuclear Flux Source Range 1 1 of 2 intennediate rang=io Maintain hot channels greater than 10 shutdown amps
5. Overtemperature llT 2 1 Maintain hot shutdown
6. Overpower /J.T 2 1 Maintain hot shutdown
7. Low Pressurizer Pressure 2 1 3 of 4 nuclear power channels Maintain hot and 2 of 2 turbine load shutdown channels less than 10% of rated power I.,) rl I Cf.l Hi Pressurizer Pressure 2 1 Same as Item 7 above Maintain hot f-'
8.
shutdown I .
-..JI.,)
-....J -....J NI (X)
e TABLE 3.7-1 (Cont'd)
REACTOR TRIP INSTRUMENT OPERATING CONDITIONS 1 2 3 4 OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI-OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET
9. Pressurizer~Hi Water Level 2 1 3 of 4 nuclear power channels Maintain hot and 2 of 2 turbine load shutdown channels less than 10% of rated power
10. Low Flow 2/operable If inoperable loop channels Maintain hot loop are not in service they must shutdown be placed in the tripped mode.
11. Turbine Trip 2 1 Maintain less than 10% rated power
12. Lo Lo Steam Generator 2/non-iso- 1/non- Maintain hot Water Level lated loop isolated loop shutdown
13. Underfrequency 4 KV Bus 2 1 Maintain hot w ~
shutdown I C/)
I-'
14. Undervoltage 4 KV Bus 2 1 Maintain hot
-..J I
-..J
.w
-..J N I shutdown I.D
15. Control rod misalignment Monitor**
a) rod position deviation 1 Log individual rod
TABLE 3.7-1 (Cont'd)
REACTOR TRIP INSTRUMENT OPERATING CONDITIONS 1 2 3 4 OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI-OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET positions once/hour, and after a load change> 10% or after> 30 inches of control rod motion.
b) quadrant power tilt 1 Log individual upper monitor (upper and and lower ion chamber lower excore neutron currents once/hour detectors) and after a load change 10% or after> 30 inches of control rod ,
motion.
16. Safety Injection See Item 1 of TS Table 3.7-2
17. Low steam generator 1/non-iso- Maintain hot shutdown water level with lated loop steam/feedwater 1/non-iso-mismatch flow lated loop
If both rod misalignment monitors (a and b)
WH are inoperable for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or more, the I Cfl I-'
nuclear overpower trip shall be reset to 93 -...JW I
percent of rated power in addition to the -...J -...J NI increased surveillance noted. I-'
0
e TABLE 3.7-2 ENGINEERED SAFEGUARDS ACTION 1 2 3 4 OPERATOR ACTION MIN. IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI -
OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET 1 SAFETY INJECTION
a. Manual 1 0 Cold Shutdown
b. High Containment Pressure 3 1 Cold Shutdown (Hi Setpoint)
c. High Differential Pressure 2/non- 1/non-between any Steam Line and isolated isolated Cold Shutdown the Steam Line Header loop loop
d. Pressurizer Low Pressure 2* 1 Primary Pressure Cold Shutdown and Low Level less than 2000 psig except when reactor is critical
e. High Steam Flow in 2/3 1/steamline *** Reactor Coolant aver- Cold Shutdown Steam Lines with Low T 2 Ta signals 1 age temperature less or Low Steam Line Pres~~~e 2 St~~m Pres- 1 than 547°F during sure Signals heatup and cooldown.
2 CONTAINMENT SPRAY w f-3 I CIJ I-'
a. Manual 2 ** Cold Shutdown -...iw I *
-...J ......
N1
b. High Containment Pressure 3 1 Cold Shutdown I-'
I-'
(Hi Hi Setpoint)
- Each channel has two separate signals
- Must actuate 2 switches simultaneously
- Hith the specified minimum operable channels the 2/3 high steaN flow is already in the trip mode
e TABLE 3.7-3 INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FUNCTIONS 1 2 3 4.
OPERATOR ACTION IF CONDITIONS OF DEGREE COLUMN 1 OR 2 MIN. OF EXCEPT AS CONDI*
OPERABLE REDUN- PERMISSIBLE BYPASS TIONED BY COLUMN 3 FUNCTIONAL UNIT CHANNELS DANCY CONDITIONS CANNOT BE MET
1. CONTAINMENT ISOLATION
a. Safety Injection See Item No. 1 of Table 3.7-2 Cold Shutdown
b. Manual 1 Hot Shutdown
c. High Containment Pressure 3 1 Cold Shutdown (Hi setpoint)
d. High Containment Pressure 3 1 Cold Shutdown (Hi-Hi setpoint)
2. STEAM LINE ISOLATION
a. High Steam Flow in 2/3 lines 1/steamline *** Cold Shutdown and 2/3 Low Tavg or 2 T signals 1 avg 2/3 Low Steam Pressure 2 Steam Pressure. 1 Signals
b. High Containment Pressure 3 1 Cold Shutdown (Hi Hi Level)
c. Manual 1/line Hot Shutdown Wl-'3
3. FEEDWATER LINE ISOLATION I cn r'
"W I *
a. Safety Injection See Item No. 1 of Table 3.7-2 Cold Shutdown "N "I r'
N
With the specified minimum operable channels the 2/3 high steam flow is already in the trip mode
TABLE ::1.* 7-4 ENGINEERED SAFETY FEATURE SYSTEM INITIATION LIMITS INSTRUMENT SETTING NO. FUNCTIONAL UNIT CHANNEL ACTION SETTING LIMIT 1 High Containment Pressure (High Containment a) Safety Injection ~ 5 psig Pressure Signal) b) Containment Vacuum Pump Trip c) High Pressure Contain-ment Isolation d) Safety Injection Contain-ment Isolation e) F.W. Line Isolation 2 High High Containment Pressure (High High a) Containment Spray < 25 psig Containment Pressure Signal) b) Recirculation Spray c) Steam Line Isolation d) High High Pressure Con-tainment Isolation 3 Pressurizer Low Pressure and a) Safety Injection > 1,700 psig Low Level b) Safety Injection Contain-> 5 percent instrument ment Isolation span.
c) Feedwater Line Isolation 4 High Differential Pressure Between any a) Safety Injection < 150 psi Steam Line and the Steam Line Header b) Safety Injection Contain-ment Isolation c) F.W. Line Isolation 5 High Steam Flow in 2/3 Steam Lines a) Safety Injection < 20% (at zero load) of full b) Steam Line Isolation steam flow c) Safety Injection Contain-< 120% (at full load) of full ment Isolation steam flow d) F.W.Line Isolation Coincident with Low T or Low Steam > 541°F T avg avg
µsig steam line w~
Line Pressure > 500 I en 1--1 pressure '1W I *
'1 '1 NI 1--1 I.,;
e TABLE 3.7-5 AUTOMATIC FUNCTIONS OPERATED FROM RADIATION MONITORS ALARM AUTOMATIC FUNCTION MONITORING ALARM SETPOINT -~
MONITOR CHANNEL AT ALARM CONDITIONS REQUIREMENTS µCi/cc
1. Process vent particulate and Stops discharge from containment See Specifications Particulate < 4xl0-S
gas monitors vacuum.systems and waste gas decay 3.11 and 4.9 Gas < 9xlo-2 (RM-GW-101 & RM-GW-102) tanks (Shuts Valve Nos.
RCV-GW-160, FCV-GW-260, FCV-GW-101)
2. Component cooling water Shuts surge tank vent valve See Specification < Twice Background radiation monitors HCV-CC-100 . 4. 9 (RM-CC-105 & RM-CC-106)
3. Liquid waste disposal Shuts effluent discharge See Specifications < 1. Sxlo- 3 radiation monitor valves FCV-LW-104A and FCV-LW-104B 3.11 and 4.9 (RM-LW-108)
4. Condensor air ejector Diverts flow to the containment of See Specification < 1.3 radiation monitors the affected unit 4.9 (RM-SV-111 & RM-SV-211) (Opens TV-SV-102 and shuts TV-SV-103 or opens TV-SV-202 and shuts TV-SV-203)
5. Containment particulate Trips affected unit's purge supply See Specifications Particulate< 9xl0-9 and gas monitors and exhaust fans, closes affected 3.10 and 4.9 Gas < lxlo-5 (RM-RMS-159 &. RM-RMS~l60, unit's purge air butterfly valves RM-RMS-259 & RM-RMS-260) (MOV-VS-lOOA, B, C & Dor MOV-VS-200A, B, C & D)
6. Manipulator crane area Trips affected unit's purge supply See Specifications .::_ 50 mrem/hr monitors (RM-RMS-162 & and exhaust fans, closes affected 3.10 and 4.9 RM-RMS-262) unit's purge air butterfly valves (MOV-VS-lOOA, B, C & Dor MOV-VS-200A, B, C & D
TS 3.8-1 1-17-72
3.8 CONTAINMENT Applicability Applies to the integrity and operating pressure of the reactor containment.
Objective To define the limiting operating status of the reactor containment for unit operation.
Specification A. Containment Integrity and Operating Pressure
1. The containment integrity, as defined in TS Section 1.0, shall not be violated, except as specified in A2, below, unless the reactor is in the cold shutdown condition.
2. The reactor containment shall not be purged while the reactor is operating, except as stated in Specification A.3.
3. During the plant startup, the remote manual valve on the steam jet air ejector suction line may be open, if under administrative control, while containment vacuum is being established. The Reactor Coolant System temperature and pressure must not exceed 350°F and 450 psig, respectively, until the air partial pressure in the containment has been reduced to a value equal to, or below, that specified in TS Figure 3.8-1.
TS 3. 8-2 3-17-72
4. The containment integrity shall not be violated when the reactor vessel head is unbolted unless a shutdown margin greater than 10 percent A k/k is maintained.
5. Positive reactivity changes shall not be made by rod drive motion or boron dilution unless the containment integrity is intact.
B. Internal Pressure
1. If the internal air partial pressure rises to a point 0.25 psi above the present value of the air partial pressure
\
(TS Figure 3.8-1),
the reactor shall be brought to the hot shutdown condition.
- 2. If the leakage condition cannot be corrected without violating the containment integrity or if the internal partial pressure continues to rise, the reactor shall be brought to the cold shutdown condition utilizing normal operating procedures.
3. If the internal pressure falls below 8.25 psia the reactor shall be placed in the cold shutdown condition.
Basis The Reactor Coolant System temperature and pressure being below 350°F and 450 psig, respectively, ensures that no significant amount of flashing steam will be formed and hence that there would be no significant pressure buildup in the containment if there is a loss-of-coolant accident.
TS 3.8-3 3-17-72 The shutdown margins are selected based on the type of activities that are being carried out. The 10 percentAk/k shutdown margin during refueling precludes criticality under any circumstance, even though fuel and control rod assemblies are being moved.
The subatmospheric air partial pressure used for normal operation is maintained between 9.0 and 11.0 psia. The set value of this partial pressure depends on the temperature of the service water which is used to cool the recirculation spray coolers, the ambient air temperature in the containment and the heat sink capability of the refueling water storage tank. The allowable air partial pressure is given in TS Figure 3.8-1 as a function of service water temperature, the refueling water storage tank temperature, and the ambient containment air temperature. If the air partial pressure rises 0.10 psi above this set value, action shall be taken immediately to correct the condition. If the condition cannot be corrected, the reactor will be manually shutdown and cooldown initiated.
Shutdown of the reactor ensures that the containment design pressure of 45 psig will not be exceeded and that depressurization can still be accomplished in the unlikely event of a loss-of-coolant accident. The containment isolation system is not activated by the containment high pressure alarm until the containment high pressure set point is reached.
Figure 3.8-1 has been established to provide the operator with information con-cerning where the air partial pressure must be maintained as a function of RWST
TS 3.8-4 3-17-72 and service water temperature to insure depressurization within 40 minutes following a loss-of-coolant accident. In the event of a loss-of-coolant accident the containment is brought back to subatmospheric conditions by cooling the containment with the containment sprays and recirculation sprays.
Since the containment spray is from the refueling water storage tank (RWST) and the recirculation spray is cooled by service water, the rate at which the containment is cooled becomes a function of the RWST water temperature and the service water temperature. The pressure to which the containment returns for a given RWST water temperature and service water temperature following a LOCA depends on the initial air partial pressure. Thus, for a given combination of RWST water temperature and service water temperature, the containment air partial pressure must be regulated below a certain level to assure that the containment will become sub-atmospheric within 40 minutes following a LOCA.
The lowest set value of air partial pressure in the containment is 9.0 psia.
This corresponds to a total pressure of approximately 9.5 psia at an air temperature of 105°F and 80°F dew point. The shell and dome plate liner of the containment are capable of withstanding an internal pressure as low as 3 psia, and the bottom mat liner is capable of withstanding an internal pressure as low as 8 psia, References FSAR Section 4.3.2 Reactor Coolant Pump FSAR Section 5.2 Containment Isolation FSAR Section 5.2.1 Design Bases FSAR Section 5.2.2 Isolation Design
TS FI G. 3. 8 - I 3-17.:... 72 e II. 2 .----.---~-~--..---~----.--~--..----r----,---...---...,....----,----,
(./)
r_ __
a. 11.0 I
-1
w a: 10.8
J
(./)
(./)
w a:
a.. - ~---->---- - - - - - - -* *-**-* .-R-EF~;LIN~ l~; 4 J-S-T-0 R-A--G-E-t-TA_N_K--l----i
_J I
t-a:
<t 1*- I TEMPFRATURE f-:J:;;:~ -+----+---+----!
t- :
Q_
a::
~~+/-
<i
(!)
z i== :: : t----+---+------+. I] - . ~~-------+--------+
9.*r *tt** -- ,.
a:
w 0.
0 w I
_J ro
<t
~
_J
_J
~
J
~
X
<t
i I-9.0 L----'-----'--~---'---...J.'"-*---'---~--'---__.__ __._ _........._ _.___ __.__ __.
35 45 55 55 75 85 95 105 SERVICE WATER TEMPERATURE - °F NOTES.
MAXIMUM ALLOWABLE OPERATING AIR PARTIAL PRESSURE -I'- 1~1 THE CONTAINMENT AS A FUNCTION OF SERVICE WATER TEMPERATURE AND REFUELING WATER STORAGE TANK TEMPERATURE.
SET POINT VALUE IN CONTAINMENT VACUUM SYSTEM INSTRUMENTATION MINIMUM AVE SERVICE CONTAINMENT WATER TEMPERATURE TEMPERATURE 95°F 95°F 95 75 85 55 75 35 MAXIMUM ALLOWABLE OPERATING AIR PARTIAL PRESSURE SURRY POWER STATION
TS 3.9-1 3-17-72 3.9 STATION SERVICE SYSTEMS Applicability Applies to availability of electrical power for operation of station auxiliaries.
Objective To define those conditions of electrical power availability necessary to pro-vide for safe reactor operation.
Specification A. A unit's reactor shall not be made critical without:
1. All three of the unit's 4,160 v buses energized
2. All six of the unit's 480 v buses energized
3. Both of the 125 v d-c buses energized
4. One b.attery charger per battery operating
5. Both of the 4,160 v emergency buses energized
6. Both of the 480 v emergency buses energized
7. Two emergency diesel generators operable as explained in Section 3.16.
TS 3.9-2 3-17-72 B. The requirements of Specification 3.9-A above may be modifieq for two reactor coolant loop operation to allow one of the unit's 4160 v normal buses and the two 480 v normal buses feed from this 4160 v bus, to be un-available or inoperable.
Basis During startup of a unit, the station's 4,160 v and 480 v normal and emergency buses are energized from the station's 34.5 kv buses. At reactor power levels greater than 5 percent of rated power the 34.5 kv buses are required to energize only the emergency buses because at this power level the station generator can supply sufficient power to the normal 4,160 v and 480 v lines to operate the unit.
Three reactor coolant loop operation with all 4160 v and 480 v buses energized is the normal mode of operation for a unit. Equipment redundacy and bus arrangements, however, allow safe unit startup and operation with one 4160 v normal bus and the two 480 v normal buses feed from this 4160 v bus, unavailable or inoperable.
References FSAR Section 8.4 Station Service Systems FSAR Section 8.5 Emergency Power Systems
TS 3.10-1 3-17-72 3.10 REFUELING Applicability Applies to operating limitations during refueling operations.
Objective To assure that no accident could occur during refueling operations that would affect public health and safety.
Specification A. During refueling operations the following conditions are satisfied:
1. The equipment door and at least one door in the personnel air lock shall be properly closed. For those systems which provide a direct path from containment atmosphere to the outside atmosphere, all automatic containment isolation valves in the unit shall be operable or at least one valve shall be closed in each line penetrating the containment.
2. The Containment Vent and Purge System and the area and airborne radiation monitors which initiate isolation of this system, shall b.e tested and verified to be operable immediately prior to refueling operations.
TS 3.10-2 3-17-72
3. At least one source range neutron detector shall be in service at all times when the reactor vessel head is unbolted. Whenever core geometry or coolant chemistry is being changed, subcritical neutron flux shall be continuously monitored by at least two source range neutron detectors, each with continuous visual indi-cation in the Main Control Room and one with audible indication within the containment.During core fuel loading phases, there shall be a minimum neutron count rate detectable on two operating source range neutron detectors with the exception of initial core loading, at which time a minimum neutron count rate need be established only when there are eight (8) or more fuel assemblies loaded into the reactor vessel.
4. Manipulator crane area radiation levels and airborne activity e levels within the containment and airborne activity levels in the ventilation exhaust duct shall be continuously monitored during refueling. A manipulator crane high radiation alarm or high airborne activity level alarm within the containment will auto-matically stop the purge ventilation fans and automatically close the containment purge isolation valves.
5, Fuel pit bridge area radiation levels and ventilation vent exhaust airborne activity levels shall be continuously monitored during refueling. The fuel building exhaust will be continuously bypassed through the iodine filter bank during refueling procedures, prior to discharge through the ventilation vent.
6. At least one residual heat removal pump and heat exchanger shall be operable to circulate reactor coolant.
TS 3.10-3 3-17-72
7. When the reactor vessel head is unbolted, a minimum boron concen-tration of .2,000 ppm shall be maintained in any filled portion of the Reactor Coolant System and shall be checked by sampling at least once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,
8. Direct communication between the Main Control Room and the refueling cavity manipulator crane shall be available whenever changes in core geometry are taking place.
9. No movement of irradiated fuel in the reactor core shall be accomplished until the reactor has been subcritical for a period of at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months .
- 10. A spent fuel cask or other heavy loads exceeding 110 percent of the weight of*a fuel assembly (not including fuel handling tool) shall not be moved over spent fuel, and only one spent fuel assembly will be handled at one time over the reactor or the spent fuel pit, B, If any of the specified limiting conditions for refueling are not met, refueling of the reactor shall cease, work shall be initiated to correct the conditions so that the specified limits are met, and no operations which increase the reactivity of the core shall be made.
C, After initial fuel loading and after each core refueling operation and prior to reactor operation at greater than 75% of rated power, the moveable.
incore detector system shall be utilized to verify proper power distribution.
Basis Detailed instructions, the above specified precautions and the design of the
TS 3.10-4 3-17-72 fuel handling equipment, which incorporates built-in interlocks and safety features, provide assurance that an accident, which would result in a hazard to public health and safety, will not occur during refueling operations.
When no change is being made in core geometry, one neutron detector is sufficient to monitor the core and permits maintenance of the out-of-function instrumentation. Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. Containment high radiation levels and high airborne activity levels automatically stop and isolate the Containment Purge System. The fuel building ventilation exhaust is divertedthrough charcoal filters whenever refueling is in progress. At least one flow path ts required for cooling and mixing the coolant contained in the reactor vessel so as to maintain a uniform boron concentration and to remove residual heat.
The shutdown margin established by Specification A-7 maintains the core subcritical, even with all of the control rod assemblies withdrawn from the core. During refueling, the reactor refueling water cavity is filled with approximately 220,000 gal of water borated to at least 2,000 ppm boron. The boron concentration of this water is sufficient to maintain the reactor subcritical by approximately 10% ~k/k in the cold shutdown condition with all control rod assemblies inserted and also to maintain the core subcritical by approximately 1% with no control rod assemblies inserted into the reactor.
Periodic checks of refueling water boron concentration assure the proper shutdown margin. Specification A-8 allows the Control Room Operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.
TS 3.10-5 3-17-72 In addition to the above safeguards, interlocks are used during refueling to assure safe handling of the fuel assemblies: An excess weight interlock is provided on the lifting hoist to prevent movement of more than one fuel assembly at a time. The spent fuel transfer mechanism can accommodate only one fuel assembly at a time.
Upon each completion of core loading and installation of the reactor vessel head, specific mechanical and electrical tests will be performed prior to initial criticality.
The fuel handling accident has been analyzed based on the activity that could be released from fuel rod gaps of 204 rods of the highest power assembly with a 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months decay period following power operation at 2550 MWt for 23,000 hours0 days 0 hours 0 weeks 0 months . The requirements detailed in Specification 3.10 provide assurance that refueling plant conditions conform to the operating conditions assumed in the accident analysis.
Detailed procedures and checks insure that fuel assemblies are loaded in the proper locations in the core. As an additional check, the moveable incore detector system will be used to verify proper power distribution. This system is capable of revealing any assembly enrichment error or loading error which could cause power shapes to be peaked in excess of design value.
References FSAR Section 5.2 Containment Isolation
TS 3.10-6 3-17-72 FSAR Section 6.3 Consequence Limiting Safeguards FSAR Section 9.12 Fuel Handling System FSAR Section 11.3 Radiation Protection FSAR Section 13.3 Table 13.3-1 FSAR Section 14.4.1 Fuel Handling Accidents FSAR Supplement : Volume I: Question 3. 2
TS 3.11-1 3-17-72 3.11 EFFLUENT RELEASE Applicability Applies to the controlled release of radioactive liquids and gases from the station.
Objective To establish conditions by which gaseous and liquid wastes containing radio-active materials may be released, and to assure that all such releases are within the concentrations specified in 10 CFR 20. In addition, to assure that the releases of liquid and gaseous radioactive wastes to unrestricted areas are as low as practicable.
Specification A. Liquid Wastes
1. The controlled release rate of radioactive liquid effluents from the station shall be such that the concentration of radionuclides leaving the circulating water discharge.canal shall not exceed the limits specified in 10 CFR 20, Appendix B, for unrestricted areas.
2. The concentration of radionuclides leaving the circulating water dis-charge canal shall not exceed 16 percent of the limits specified in 10 CFR 20, Appendix B, for unrestricted areas, averaged over any calendar quarter.
TS 3.11-2 3-17-72
3. The release of radioactive material in liquid waste, when averaged over any calendar quarter, shall be such that these quantities, if continued at the same release rate for a year, would not exceed 105 curies per year, excluding tritium and dissolved gases, or would not result in an annual average concentration of tritium or an annual average concentration of dissolved gases in the discharge canal in excess of 4 x 10- 5 microcuries per milliliter.
4. Prior to release of liquid wastes, a sample shall be taken and analyzed to demonstrate compliance with A-1 above, using the exist-ing circulating water discharge rate.
5. Liquid waste activity and flow rate shall be continuously monitored and recorded during release by the liquid waste disposal radiation monitor and the liquid waste flow recorder. During liquid waste release the circulating water discharge tunnel radiation monitor shall be operating.
6. Steam generator blowdown activity shall be continuously recorded and monitored by the steam generator blowdown sample monitor system.
7. Prior to and during liquid waste release, circulating water flow shall be established in the discharge tunnel.
8. The equipment installed in the liquid radioactive waste system shall be maintained and operated to process all liquid wastes for which it is designed.
9. The effluent control monitor shall be set to alarm and automatically close the waste discharge valve such that the requirements of specification
TS 3.11-3 3-17-72
A-1 above are met.
B. Gaseous Wastes
1. The controlled release rates of gaseous and airborne particulate wastes originating from station operation shall be limited as follows:
3
< 2,0 X 10 5 m sec where Qi is the controlled release rate (curies per second) of any radioisotope i and (MPC)., in unit of microcuries per cubic centimeter 1
is defined in Colllllln 1, Table II of Appendix B to 10 CFR 20, except that for halogen and particulate isotopes with half-lives greater than 8 days, the values of (MPC). shall be reduced by a factor of 700.
1
2. The release rates of activity, except halogens and particulate with half lives' longer than eight days shall not exceed 16 percent of those specified in specification B-1 above averaged over any calendar quarter.
3. The release rate of Iodine 131, when averaged over any calendar quarter shall be such that if continued at the same release rate for a year would not exceed 0.9 curies per year.
4, Gaseous waste gross and particulate activity and flow rate shall be continuously monitored and recorded during release of radioactive gaseous wastes to the process vent.
5. During release of radioactive gaseous waste to the process vent, the following conditions shall be met:
a. At least one process vent blower shall be operating.
TS 3.11-4 3-17-72
b. The process vent gas monitor and particulate monitor shall be operating.
6. All effluents to be discharged to the atmosphere from the waste gas decay tanks of the gaseous waste disposal system shall be sampled and analyzed to demonstrate compliance with specification B-1 above prior to release via the process vent.
7. Whenever the air ejector discharge monitor is inoperable and the steam generator blowdown monitors indicate an increase in secondary side activity, samples shall be taken from the air ejector discharge and analyzed for gross activity on a daily basis.
8. During normal conditions of plant operation, radioactive gaseous wastes shall be provided a minimum holdup of 60 days except for low radioactivity gaseous wastes resulting from purge and fill operations associated with refueling and reactor startup.
9, The maximum activity to be contained in one gas decay tank shall not exceed 95,400 curies equivalent of Xenon 133.
10. Purging of the containment shall be governed by the following conditions:
a. Containment purge shall be filtered through the high efficiency particulate air filters and charcoal absorbers whenever the concen-tration of iodine and particulate isotopes exceed the occupational MPC inside the containment.
b. Containment purge shall be filtered through the high efficiency particulate air filters and charcoal absorbers whenever irradiated fuel is being handled or any object is being handled over irradiated
TS 3.11-5 3-17-72 fuel in the containment.
Basis The releases of radioactive materials will be kept as low as practicable as required by 10 CFR 50 and will not exceed the concentration limits specified in 10 CFR 20. At the same time, the licensee is permitted the flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided a dependable source of power under unusual operating conditions which may temporarily result in releases in excess of four percent of the concentration limits specified in 10 CFR 20. However, all releases
.must be kept within the concentration limits specified in 10 CFR 20. It is ex-pected that using this operational flexibility under unusual operating conditions, the licensee shall exert every effort to keep levels of radioactive materials e released from the plant as low as practicable and that annual releases will not exceed a small fraction of the annual average concentration limits specified in 10 CFR 20.
The limiting conditions for operation contained in specification A-3 above, which relates to the total number of curies which may be released in liquid effluents in any year, is based on the expected performance of the Surry Power Station assuming both units are operating with 0.25 percent leaking fuel and each unit is experiencing a 20 gallon per day primary to secondary system leak rate.
The formula prescribed in specification B-1 takes atmospheric dilution into account and assures that at the point of maximum ground concentration at the site boundary, the requirements of 10 CFR 20 will not be exceeded. The limit is based on the highest annual average value of X/Q which will occur at the
TS 3.11-6 3-17-72 site boundary (5.0 x 10-6 seconds per meter 3 ). The requirement to reduce the maximum permissible concentrations of 10 CFR 20 by a factor of 700 for halogen and particulate isotopes with half lives greater than eight days con-servatively limits exposures from airborne radioactive materials that may enter terrestrial food chains. Details of the equipment provided and mode of operation of the liquid and gaseous waste disposal systems are presented in sections 11.2.3 and 11.2.5, respectively, of the FSAR, The limit on Xe-133 content of the waste gas decay tanks is based on the maximum content assumed in the analysis of a decay tank failure.
References FSAR Section 2.2.3 Average Atmospheric Dilution FSAR Section 11. 2. 3 Liquid Waste Disposal System FSAR Section 11.2.5 Gaseous Waste Disposal System FSAR Section 11. 3. 3 Process Radiation Monitor Systems FSAR Section 11. 3. 4 Area Radiation Monitor Systems FSAR Section 14.4.2.2 Waste Gas Decay Tank Rupture
TS 3.12-1 3-17-72 3.12 CONTROL ROD ASSEMBLIES AND POWER DISTRIBUTION LIMITS Applicability Applies to the operation of the control rod assemblies and power distribution limits.
Objective To ensure core subcriticality after a reactor trip, a limit on potential reactivity insertions from a hypothetical control rod assembly ejection, and an acceptable core power distribution during power operation.
e Specification A. Control Bank Insertion Limits
1. Whenever the reactor is critical, except for physics tests and control rod assembly exercises, the shutdown control rods shall be fully withdrawn.
2. Whenever the reactor is critical, except for physics tests and control rod assembly exercises, the control rod assembly groups shall be no further inserted than the limits shown. on TS Fig.
3.12-1, 3.12-2, or 3.12-3 for three loop operation and TS Fig.
3.12-4, 3.12-5 or 3.12-6 for two loop operation.
TS 3.12-2 3-17-72
3. The limits shown on TS Figures 3.12-1 through 3.12-6 may be revised on the basis of physics calculations and physics data obtained during unit startup and subsequent operation, in accordance with the following:
a. The sequence of withdrawal of the controlling banks, when going from zero to 100% power, is A, B, C, D.
b. An overlap of control banks, consistent with physics calculations and physics data obtained during unit startup and subsequent operation, will be permitted.
c. The shutdown margin with allowance for a stuck control rod assembly shall exceed the applicable value shown on TS Figure 3.12-7 under all steady-state operating conditions, except for physics tests, from zero to full power, including effects of axial power distribution. The shutdown margin as used here is defined as the amount by which the reactor core would be subcritical at hot shutdown conditions (T >547°F) if all control rod assemblies were tripped, avg-assuming that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon, boron, or part-length rod position.
TS 3.12-3 3-17-72
4. Whenever the reactor is subcritical, except for physics tests, the critical rod position, i.e., the rod position at which criticality would be achieved if the control rod assemblies were withdrawn in normal sequence with no other reactivity changes, shall not be lower than the insertion limit for zero power.
5. The part length rod bank may be moved over the entire travel range, full out to full in, as required by axial power dis-tribution control.
6. Insertion limits do not apply during physics tests or during e periodic exercise of individual rods. However, the shutdown margin indicated in TS Figure 3.12-7 must be maintained except for the low power physics test to measure control rod worth and shutdown margin. For this test the reactor may be critical with all but one full length contro1 rod,expected to have the highest worth, inserted and part length rods fully withdrawn.
B. Power Distr.ibution Limits
1. If the power tilt ratio exceeds 1.10, except for physics test, or if a part-length _or full-length control rod is _more than 15 inches out of alignment with its bank, then within eight hours:
a. The situation shall be corrected, or
TS 3.12-4 3-17-72
b. The hot channel factors shall be determined and maximum allowable power shall be reduced one percent for each percent the hot channel factor exceeds the values of Design Limits Interim Limits N N F = 2.72 1.58 FN = 2.52 F~H m 1.50 q q or
c. Power shall be limited to 75% of rated power for 3 loop operation or 45% of rated power for 2 loop operation.
2. If after a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the power tilt ratio in 1. above is not corrected to less than 1.10, and e a. If design hot channel factors are not exceeded, an evaluation as to the cause of the discrepancy shall be made and reported to the Atomic Energy Commission, within 30 days.
b. If design hot channel factors are exceeded and power is greater than 10%, the nuclear overpower.and overpowerAT trips shall be reduced one percent for each percent the hot channel factor exceeds the design values in the Technical Specifications.
3. If the power tilt ratio exceeds 1.25, except for physics tests, then the reactor shall be put in the hot shutdown condition,
TS 3.12-5 3-17-72 and the Atomic Energy Commission notified per Specification 6.6,Power operation for the purpose of testing is permitted, provided that maximum power levels are less than 50% of rated thermal power.
4. Upon determination of F: and F:H' as required by Specification 4.10.B, allowable power shall be reduced one percent for each per-cent these hot channel factors exceed the limiting values specified in these Technical Specifications. This restriction on allowable power will remain in effect until the anomaly is corrected.
c. Inoperable Control Rods
1. A control rod assembly shall be considered inope.rable if the assembly cannot be moved by the drive mechanism, or the assembly remains misaligned from its bank by more than 15 inches. A full-length control rod shall be considered inoperable if its rod drop time is greater than 1.8 seconds to dashpot entry.
2. No more than one inoperable control rod assembly shall be permitted when the reactor is critical.
3. If more than one control rod assembly in a given bank is out of service because of a single failure external tci the individual rod drive mechanisms, i.e. programming circuitry, the provisions of Specification Cl and 2 shall not apply and the reactor may remain critical for a period not to exceed two hours provided immediate attention is directed toward making the necessary repairs. In the event, the affected assemblies cannot be returned
TS 3.12-6 3-17-72 to service within this specified period the reactor will be brought to hot shutdown conditions.
4. The provisions of Specifications Cl and 2 shall not apply during physics test in which the assemblies are intentionally misaligned.
5. If an inoperable full-length rod is located below the 200 step level and is capable of being tripped, or if the full-length rod is located below the 30 step level whether or not it is capable of being tripped, then the insertion limits in TS Figure 3 .12,...2 apply.
6. If an inoperable full-length rod cannot be located, or if the in-operable full-length rod is located above the 30 step level and cannot be tripped, then the insertion limits in TS Figure 3.12-3 apply.
7. No insertion limit changes are required by an inoperable part-length rod.
8. If a full-length rod becomes inoperable and reactor operation is continued the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within 30 days. The analysis shall include due allowance for nonuniform fuel depletion in the neighborhood of the inoperable rod. If the analysis results in a more limiting hypothetical transient than the cases reported in the safety analysis, the plant power level shall be reduced to an analytically determined part power level which is consistent with the safety analysis.
TS 3.12-7
~ 3-17-72 D. If the reactor is operating above 75% rated power with one excore nuclear channel out of service, the core quadrant power balance shall be determined
1. Once per day, and
2. After a change in power level greater than 10% or more than 30 inches of control rod motion.
The core quadrant power balance shall be determined by one of the following methods:
1. Movable detectors (at least two per quadrant)
2. Core exit thermocouples (at least four per quadrant).
E. Inoperable Rod Position Indicator Channels
1. If a rod position indicator channel is out of service then:
a) For operation between 50% and 100% of rated power, the position of the RCC shall be checked indirectly by core instrum~~tation (excore detector and/or thermocouples and/or movable incore detectors) every shift or subsequent to motion, of the non-indicating rod, exceeding 24 steps, whichever occurs first.
b) During operation below 50% of rated power no special monitoring is required.
TS 3.12-8 3-17-72
2. Not more than one rod position indicator (RPI) channel per group nor two RPI channels per bank shall be permitted to be inoperable at any time.
Basis The reactivity control concept assumed for operation is that reactivity changes accompanying changes in reactor power are compensated by control rod assembly motion. Reactivity changes associated with xenon, samarium, fuel depletion, and large changes in reactor coolant temperature (operating temyerature to cold shutdown) are compensated for by changes in the soluble boron concentration. During power operation, the shutdown groups are fully withdrawn and control of power is by the control groups. A reactor trip occurring during power operation will place the reactor into the hot shutdown condition.
The control rod assembly insertion limits provide for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod assembly remains fully withdrawn, with sufficient margins to meet the assumptions used in the accident analysis. In addition, they provide a limit on the maximum inserted rod worth in the unlikely event of a hypothetical assembly ejection, and provide for acceptable nuclear peaking factors. The limit may be determined on the basis of unit startup and operating data to provide a more realistic limit which will allow for more flexibility in unit operation and still assure compliance with the shutdown requirement. The maximum shutdown margin
TS 3.12-9 3-17-72 requirement occurs at end of core life and is based on the value used in analysis of the hypothetical steam break accident. The rod insertion limits are based on end of core life conditions. Early in core life, less shutdown margin is required, and TS Figure 3.12-7 shows the shutdown margin equivalent to 1.77% reactivity at end-of-life with respect to an uncontrolled cooldown.
All other accident analyses are based on 1% reactivity shutdown margin.
Relative positions of control rod banks are determined by a specified control rod bank overlap. This overlap is based on the considerations of axial power shape control. It is not based on safety criteria, but minimizes possible changes in axial offset accompanying control rod motion.
Positioning of the part-length assemblies is governed by the requirement to maintain the axial power shape within specified limits or to accept an automatic cutback of the overpower ~T and overtemperature ~T set points (see Specification 2.3). Thus, there is no need for imposing a limit on the physical positioning of the part-length assemblies.
The various control rod assemblies (shutdown banks, control banks A, B, C D and part-length rods) are each to be moved as a bank, that is, with all assemblies in the bank within one step (5/8 inch) of the bank position.
Position indication is provided by two methods: a digital count of actuating pulses which shows the demand position of the banks and a linear position indicator, Linear Variable Differential Transformer, which indicates the actual assembly position. The position indication accuracy of the pulse
TS 3.12-10 3-17-72 count is within one step (5/8 inch). The accuracy of the Linear Variable Differential Transformer is approximately +/-_5% of span (+/-_7.5 inches) under steady state conditions.(l) The relative accuracy of the linear position indicator is such that, with the most adverse errors, an alarm is actuated if any two assemblies within a bank deviate by more than 14 inches. In the event that the linear position indicator is not in service, the effects of malpositioned control rod assemblies are observable from nuclear and process information displayed in the Main Control Room and by core thermocouples and in-core movable detectors. Below 50% power, no special monitoring is required for malpositioned control rod assemblies with inoperable rod position indicators because, even with an unnoticed complete assembly misalignment(Part-length or full length control rod assembly 12 feet out of alignment with its bank) operation at 50% steady state power does not result in exceeding core limits.
The specified control rod assembly drop time is consistent with safety analyses that have been performed.(2)
An inoperable control rod assembly imposes additional demands on the operators.
The permissible number of inoperable control rod assemblies is limited to one in order to limit the magnitude of the operating burden, but such a failure would not prevent dropping of the operable control rod assemblies upon reactor trip.
TS 3.12-11 3-17-72 A quadrant to average power tilt will be indicated by the excore detectors.
The excore current tilt is indicated by the arrangement of the current recorders on the control board. Four 2-pen recorders are provided, the pens are grouped so that, in the absence of a tilt, the two ink traces coincide.
Any divergence in the traces indicates a power tilt. Furthermore, a maximum-to-average alarm is provided for the upper and lower sets of excore currents. A power tilt ration of 1.1 is not expected to result in peaking factors greater than design under expected operating conditions.
If, instead of using movable in-core detectors to measure the hot channel factors, the operator chooses simply to reduce power, the specified limit of 75% power maintains the minimum design margin to core safety limits for up to 1.25 tilt ratio. Resetting of the overpower trip set points ensures that the protection system basis is maintained for sustained unit operation.
A tilt' ratio of 1.25 or more is indicative of a serious performance anomaly and a unit shutdown is prudent.
References (1) FSAR Section 7.2.
(2) FSAR Section 14.
TS FIGURE 3.12-1 3-:17-72 FIGURE 3.12-1 CONTROL BANK INSERTION LIMITS FOR NORMAL 3 LOOP OPERATION (b.2) a.a.--~~__:....-..:....~~~~~---~~~~~__;_~~~-,(0.78)
BANK B Cl 0.2 LLl Ii LJ..I VI z:
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~ (0. 76) 0.8 1.0 ~---_.._______..___ _ _......._ _ _ _ _ _ _ __
a.a* 0.2 (0.34)0.4 0.6 0.8 . 1.0 FRACTION OF RATED POWER
TS FIGURE 3 .12-2 3-17-72 FIGURE 3.12-2 CONTROL BANK INSERTION LIMITS FOR 3 LOOP OPERATION WITH ONE BOTTOMED ROD 0.0 I/
/
/
./
0.2 /
/ BANKC /
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0.0 0.2 0.4 0.6 0.8 1.0 FRACTION FULL POWER
TS FIGURE 3.12-3 3-17-72 FIGURE 3.12-3 CONTROL BANK INSERTION LIMITS FOR 3 LOOP OPERATION WITH ONE INOPERABLE ROD 0.0 V V
/ V V BANKC I/
0.2 V /
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~ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ ~ ~ - ~ -- --- l TS FIGURE 3.12-4 3-17-72 FIGURE 3.12-4 CONTROL BANK INSERTION LIMITS FOR 2 LOOP NORMAL OPERATION 0.0 I/
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~NKC 0.2 V
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w t- /v ffi 0.4 /
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1.0 0.0 0.2 0.4 0.6 FRACTION FULL POWER
TS FIGURE 3.12-5 3-17-72 CONTROL BANK INSERTION LIMITS FOR 2 LOOP OPERATION WITH ONE BOTTOMED ROD 0.0 V
/
V
/
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0.2 /
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z V 0 /
t; I/
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ct 0.6 ff: , I/
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[7 0.8 1.0 o.o 0.2 0.4 0.6 FRACTION FULL POWER
TS FIGURE 3.12-6 3-17-72
- CONTROL BANK INSERTION LIMITS FOR 2 LOOP OPERATION WITH ONE INOPERABLE ROD 0.0
/
V
/
0.2 /
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V
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w /v I-e a:: 0.4 w
Cf) z V
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z 0
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ct 0.6 fE 0.8 1.0 0.0 0.2 0.4
0.6 FRACTION FULL POWER
e REQUIRED SHUTDOWN REACTIVITY AS A FUNCTION OF REACTOR COOLANT BORON.CONCENTRATION 2.0 1.8
_.__,~ .. , - - - t - - - - + - - - - + - - - + - - - t * * - ----*---- ---**-------- >------ ---* . . .. - ~~ -* **--*---- f---*---+---~
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TS 3.13-1 3-17-72 3.13 COMPONENT COOLING SYSTEM Applicaliility Applies to the ope.rational status of all subsystems of the Component Cooling Sys:tem. The Component Cooling System consists of the Component Cooling Water Subsystem, Chilled Component Water Subsystem, Chilled Water Subsystem, Neutron Shield Tank Cooling Water Subsystem and Charging Pump Cooling Water Subsystem.
Objective To define. limiting conditions for each subsystem of the Component Cooling System necessary to assure safe operation of each reactor unit of the station during startup, power operation, or cooldown.
Specifications A. When a unit's Reactor Coolant System temperature and pressure exceed 350°F and 450 psig, respectively, or when a unit's reactor is critical operating conditions for the Component Cooling Water Subsystem shall be as follows:
1. For one unit operation, two component cooling water pumps and e
TS 3.13-2 3-17-72 heat exchangers shall be operable.
2. For two unit operation, three component cooling water pumps and heat exchangers shall be operable.
3. The Component Cooling Water Subsystem shall be operable for immediate supply of cooling water to the following components, if required:
a. Two operable residual heat removal heat exchangers.
b. Seal water and stuffing box jacket of two operable residual heat removal pumps.
4. During power operation, Specification A-1, A-2, or A-3 above may be modified to allow one of the required components to be inoperable provided innnediate attention is directed to making repairs. If the system is not restored within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the requirements of Specification A-1, A-2, or A-3, an operating reactor shall be placed in the hot shutdown condition. If the repairs are not completed within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the affected reactor shall be placed in the cold shutdown condition.
B. For each unit whose Reactor Coolant System exceeds a temperature of 350°F and a pressure of 450 psig, or when a unit's reactor is critical,
TS 3.13-3 3-17-72 operating conditions for the Charging Pump Cooling Water Subsystem shall be as follows:
1. Make-up water from the Component Cooling Water Subsystem shall be available.
2. One charging pump component cooling water pump and one charging pump service water pump shall be operating. The spare charging pump component cooling water pump and the spare charging pump service water pump shall be operable.
3. One charging pump intermediate seal cooler shall be operating and the spare charging pump intermediate seal cooler shall be operable.
4. During power operation the requirements of B-1, -2, and -3 above may be modified to allow one o*f the following components to be inoperable at any one time. If the system is not restored to meet the conditions of B-1 above within the time period specified below, the reactor shall be placed in the hot shutdown condition. If the system is not restored to meet the conditions of B-1, -2, and -3 within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the reactor shall be placed in the cold shutdown condition.
a. One charging pump component cooling water pump or one
TS 3 .13-4 3-17-72 charging pump service water pump may be out of service provided the pump is restored to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
b. One charging pump intermediate seal cooler or other passive component may be out of service provided the system may still operate at 100 percent capacity and repairs are completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .
Basis The Component Cooling System is an intermediate cooling system which serves both reactor units. It transfers heat from heat exchangers containing reactor coolant, other radioactive liquids, and other fluids to the Service Water System. The Component Cooling System is designed to (1) provide cooling water for the removal of residual and sensible heat from the Reactor Coolant System during shutdown, cooldown, and startup, (2) cool the containment recirculation air coolers and the reactor coolant pump motor coolers, (3) cool the letdown flow in the Chemical and Volume Control System during power operation, and during residual heat removal for continued purification, (4) cool the reactor coolant pump seal water return flow, (5) provide cooling water for the neutron shield tank and (6) provide cooling to dissipate heat from other reactor unit components.
The Component Cooling Water Subsystem has four component cooling water pumps
.and four component cooling water heat exchangers, Each of the component cooling water heat exchangers is designed to remove during normal operation the entire
TS 3.13-5 3-17-72 heat load from one unit plus one half of the heat load common to both units.
Thus, one component cooling water pump and one component cooling water heat exchanger are required for each unit which is at power operation. Two pumps and two heat exchangers are normally operated during the removal of residual and sensible heat from one unit during cooldown. Failure of a single component may extend the time required for cooldown but does not effect the safe operation of the station.
A Charging Pump Cooling Water Subsystem is provided for each reactor unit.
The subsystem has two charging pump component cooling water pumps, two charging pump service water pumps, two charging pump intermediate seal coolers, a surge tank, and interconnecting valves and piping to provide cooling water to the charging pump lubricating oil coolers and charging pump mechanical seal coolers. Each of the charging pump component cooling water pumps, service water pumps, and intermediate seal coolers has full capacity, providing 100 percent redundancy for this sybsystem.
References FSAR Section 5.3 Containment Systems FSAR Section 9.4 Component Cooling System FSAR Section 15.5.1.2 Containment Design Criteria
TS 3.14-1 3-17-72 3.14 CIRCULATING AND SERVICE WATER SYSTEMS Applicability Applies to the operational status of the Circulating and Service Water Systems.
Objective To define those limiting conditions of the Circulating and Service Water Systems necessary to assure safe station operation.
Specification A. The Reactor Coolant System temperature or pressure of a reactor unit shall not exceed 350°F or 450 psig, respectively, or the reactor shall not be critical unless:
1. The high level intake canal is filled to at least El.+ 18.0 ft at the high level intake structure.
2. Unit subsystems, including piping and valves, shall be operable to the extent of being able to establish the following:
a. Flow to and from one bearing cooling water heat exchanger.
TS 3.14-2 3-17-72
b. Flow to and from the component cooling heat exchangers required by Specification 3.13.
3. At least two circulating water pumps are operating or are operable.
4. At least two emergency service water pumps are operable; these two pumps will service both units simultaneously.
5. Two service water flowpaths to the charging pump service water subsystem are operable.
6. Two service water flowpaths to the recirculation spray subsystems are operable.
B. There shall be an operating service water flow path to and from one operating control area air conditioning condenser and at least one operable service water flow path to and from at least one operable control area air conditioning condenser whenever fuel is loaded in reactor core.
C. The requirements of Specifications A-5 and A-6 may be modified to allow unit operation with only one operable flow path to the charging pump service water subsystem and to the recirculation spray subsystems.
If the affected systems are not restored to the requirements of Specifications A-5 and A-6 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor shall be placed in a hot shutdown condition. If the requirements of Specifications A-5 and A-6 are not met within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the reactor shall be
TS 3.14-3 3-17-72 placed in a cold shutdown condition.
Basis The Circulating and Service Water System are designed for the removal of heat resulting from the operation of various systems and components of either or both of the units. Untreated water, supplied from the James River and stored in the high level intake canal is circulated by gravity through the recircu-lation spray coolers and the bearing cooling water heat exchangers and to the charging pumps lubricating oil cooler service water pumps which supply service water to the charging pump lube oil coolers.
In addition, the Circulating and Service Water Systems supply cooling water to the component cooling water heat exchangers and to the control area air conditioning condensers. The component cooling water heat exchangers are not required during loss-of-coolant accident conditions for unit safety; however, they are normally required for power operation and during a loss-of-station power accident. At least one operating and one operable control area air conditioning condenser is required whenever monitoring of reactor conditions is required.
The long term service water requirement for the loss-of-coolant accident in one unit with simultaneous loss-of-station power and second unit being main-tained in a safe condition is 15,000 gpm.
Three diesel driven emergency service pumps with a design capacity of 15,000 gpm each, are provided to supply water to the high level intake canal during a loss-of-station power incident.
TS 3.14-4 3-17-72 Thus, one operating emergency service water pump is required for both units plus one operable spare, and the third pump can be down for maintenance.
A minimum level of El. +18 ft in the high level intake canal is required to provide design flow of service water through the recirculation spray coolers during a loss-of-coolant accident. If the water level falls below El. +18 ft, the turbines will be automatically tripped and the reactors will be automatically tripped as a result of the turbines being tripped.
References:
FSAR Section 9.9 Service Water System FSAR Section 10.3.4 Circulating Water System FSAR Section 14.5 Loss-of-Coolant Accidents, Including the Design Basis Accident
TS 3.15-1 3-17-72 3.15 CONTAINMENT VACUUM SYSTEM Applicability Applies to the operational status of the Containment Vacuum System.
Objective To define those conditions of the Containment Vacuum System necessary to assure safe station operation.
Specification The unit Reactor Coolant System shall not be made to exceed a temperature or pressure greater than 350° For 450 psig, respectively, or the reactor shall not be critical unless the following unit Containment Vacuum System conditions are met:
A. Following a period when the containment was at atmospheric pressure and containment integrity has been achieved, the Containment Vacuum System steam ejector and its associated piping and valves shall have been operating and have reduced the containment internal pressure to the subatmospheric operating pressure corresponding to an air partial pressure of between 9.0 and 11.0 psia as established in Specification 3.8.
TS 3.15-2 3-17-72 B. One mechanical vacuum pump and one associated flow path shall be operable.
C. The steam ejector and its associated piping and valves shall be secured and isolated once the containment internal pressure is at the subatmospheric pressure used for normal operation.
D. The vacuum pump shall not be operated if the air partial pressure rises 0.25 psi above the preset value of the air partial pressure as determined by Technical Specification 3.8.
Basis The Containment Vacuum System consists of a steam ejector and two mechanical vacuum pumps with the required piping, valves, and instrumentation. It is designed to perform the following functions:
A. Evacuation *of the containment from atmospheric pressure to the subatmospheric pressure used for normal operation.
B. Removal of air from the containment to compensate for containment inleakage during normal operation.
The system, through the use of the steam ejector, is designed to reduce the containment pressure from atmospheric pressure to the subatmospheric pressure used for normal operation in approximately 4 hr, compatible with the unit startup schedule. Each of two mechanical vacuum pumps has the capacity re-quired to maintain the normal subatmospheric operating pressure. However,
TS 3.15-3 3-17-72 due to the low leakage characteristics of the containment, neither pump will operate for long periods of time. The vacuum pumps are capable of being operated from the emergency buses and discharge to the gaseous waste disposal system.
Technical Specification 3.8.B requires that the reactor be brought to the hot shutdown condition if the internal air partial pressure rises 0.25 psi above the preset value of the air partial pressure. Accordingly, the vacuum pump shall not be operated if this condition exists to insure that, in the event of an accident, there will be no release through the vacuum system. If containment vacuum is not established, the requirement that Reactor Coolant System temperature and pressure be no greater than 350°F and 450 psig, respectively, and that the reactor not be critical, will insure that, if a loss-of-coolant accident does occur, no significant pressure buildup in the containment would occur.
References FSAR Section 6.3.2 Containment Vacuum System Technical Specification 3.~ Containment
TS 3.16-1 3-17-72 e 3.16 EMERGENCY POWER SYSTEM Applicability Applies to the availability of electrical power for safe operation of the station during an emergency.
Objective To define those conditions of electrical power availability necessary to shutdown the reactor safely, and provide for the continuing availability of Engineered Safeguards when normal power is not available.
Specification A. A reactor shall not be made critical nor shall a unit.be operated such that the reactor coolant system pressure and temperature exceed 450 psig and 350°F, respectively, without:
1. Two diesel generators (the unit diesel generator and the shared backup diesel generator) operable with each generator's day tank having at least 290 gallons of fuel and with a minimum on-site supply of 35,000 gal of fuel available.
2. Two 4,160 v emergency buses energized.
3. Two 480 v emergency buses energized.
TS 3.16-2 3-17-72
4. Two physically independent circuits from the offsite transmission network to energize the 4,160 and 480 v emergency buses, One of *these sources must be immediately available, i.e. primary source; arid the other must be capable of being made available within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; i.e.
dependable alternate source.
5. Two operable flow paths for providing fuel to each diesel generator.
6. Two batteries, two chargers, and the d.c. distribution systems operable.
B. During power operation or the return to power from hot shutdown conditions, the requirements of specification 3.16-A may be modified by one of the following:
1. One diesel generator and its associated fuel oil pumps and flow paths may be unavailable or inoperable provided the operability of the other diesel generator and its associated fuel oil pumps and flow paths is demonstrated daily. If this diesel generator is not returned to an operable status within 7 days, the reactor shall be brought to a cold shutdown condition.
2. If a primary source is not available, the unit may be operated for seven(7) days provided the dependable alternate source can be operable within 8 ~ours. If specification A-4 is not satisfied within seven(7) days, the unit shall be brought to the cold shutdown condition.
3. One battery may be inoperable for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided the other battery and battery chargers remain operable with one battery charger carrying the d.c. load of the failed battery's supply system. If the
TS 3.16-3 3-17-72 battery is not returned to operable status within the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period the reactor shall be placed in the hot shutdown condition.
If the battery is not restored to operable status within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the reactor shall be placed in the cold shutdown condition.
C. The continuous running electrical load supplied by an emergency diesel generator shall be limited to 2750 kw.
Basis The Emergency Power System is an on-site, independent, automatically starting power source. It supplies power to vital unit auxiliaries if a normal power source is not available. The Emergency Power System consists of three diesel generators for two units. One generator is used exclusively for Unit 1, the second for Unit 2, and the third generator functions as a backup for either Unit 1 or 2. The diesel generators have a continuous 2,000 hour0 days 0 hours 0 weeks 0 months rating of 2750 kw and a two hour rating of 2850 kw. The actual loads using conservative ratings for accident conditions, require approximately 2,320 kw, Each unit has two emergency buses, one bus in each unit is connected to its exclusive diesel generator. The second bus in each unit will be connected to the backup diesel generator as required. Each diesel generator has 100 percent capacity and is connected to independent 4,160 v emergency buses, These emergency buses are*
normally fed from the reserve station service transformers. The normal station service transformers are fed from the unit isolated phase bus at a point between the generator terminals and the low voltage terminal of the main
TS 3.16-4 3-17-72
- step-up transformer, The reserve station service transfonners are fed from the autotransformers in the high voltage switchyard. The circuit which supplies power through either autotransformer is called a "primary source." In the event an autotransformer is inoperable, the remaining one may be cross-tied by a 34.5 bus to all three reserve station service transformers. Thus, a primary source is available to both units even if one of the two autotransformers is out of service.
In addition to the "primary sources," each unit has an additional offsite power source which is called the "dependable alternate source." This source can be made available in eight(8) hours by removing a unit from service, disconnecting its generator from the isolated phase bus, and feeding offsite power through the main step-up transformer and normal station service transformers to the emergency buses.
The generator can be disconnected from the isolated phase bus within eight(8) hours. A unit can be maintained in a safe condition for eight(8) hours with no offsite power without damaging reactor fuel or the reactor coolant pressure boundary.
The diesel generators function as an onsite back-up system to supply the emergency buses. Each emergency bus provides power to the following operating Engineered Safeguards equipment:
A. One containment spray pump B. One charging pump C. One low head safety injection pump
TS 3.16-5 3-17-72 D. One recirculation spray pump inside containment E. One recirculation spray pump outside containment F. One containment vacuum pump G. One motor control center for valves, instruments, control air compressor, fuel oil pumps, etc.
H. Control area air conditioning equipment - four air recirculating units, one water chilling unit, one service water pump and one chilled water circulating pump I. One charging pump service water pump for charging pump intermediate seal coolers and lube oil coolers.
J, One charging pump cooling water pump for charging pump seal coolers.
The day tanks are filled by transferring fuel from any one of two buried tornado missile protected "fuel oil storage tanks, each of 20,000 gal capacity.
Two 100 percent capacity fuel oil transfer pumps per diesel generator are powered* from the emergency buses to assure that an operating diesel generator has a continuous supply of fuel. The puried fuel oil storage tanks contain a seven day supply of fuel, 35,000 gal minimum, for the full load* operation of one diesel generator; in addition, there is an above ground fuel oil storage tank on-site with a capacity of 210,000 gal which is used for trans-ferring fuel to the buried tanks.
TS 3.16-6 3-17-72 If a loss of normal power is not accompanied by a loss-of-coolant accident, the safeguards equipment will not be required. Under this condition the following additional auxiliary equipment may be operated from each emergency bus:
A. One component cooling pump B. One residual heat removal pump C. One motor-driven auxiliary steam generator feedwater pump The emergency buses in each unit are capable of being interconnected under e strict administrative procedures so that the equipment which would normally be operated by one of the diesels could be operated by the other diesel, if required.
References FSAR Section 8.5 Emergency Power System FSAR Section 9.3 Residual Heat Removal System FSAR Section 9.4 Component Cooling System FSAR Section 10.3.2 Auxiliary Steam System FSAR Section 10.3.5 Condensate and Feedwater System e
TS 3.17-1 3-17-72 3.17 LOOP STOP VALVE OPERATION Applicability Applies to the operation of the Loop Stop Valves.
Objective To specify those limiting conditions for operation of the Loop Stop Valves which must be met to ensure safe reactor operation.
Specifications
1. Whenever a reactor coolant loop is isolated, the boron con-centration in the isolated loop shall be maintained at a value greater than or equal to the boron concentration in the active loops. The boron concentration in an isolated loop shall be measured and logged at least 5 days per week.
2. Whenever startup of an isolated reactor coolant loop is initiated, the following conditions shall be met:
a. All the channels, including redundant channels, of the Loop Stop Valve Interlock System of the isolated loop are operable.
In the event this condition is not satisfied, the loop must e remain isolated.
TS 3.17-2 3-17-72 b, The unit shall be in a shutdown condition prior to opening either stop valve and throughout the timing interval required prior to opening the cold leg stop valve.
c. Prior to opening the hot leg valve and again prior to opening the cold leg valve, the boron concentration of the isolated loop must be verified as greater than or equal to the boron concentration in the operating loops.
d. The count rate, as given by the nuclear instrumentation shall be logged every five minutes during the timing interval required prior to opening the cold leg stop valve. Should the count rate increase by more than a factor of two over the initial count rate, the hot leg stop valve should be re-closed and no attempt made to open the stop valves until the reason for the count rate increase has been determined.
e. All three reactor coolant pumps must be running during the time a hot leg stop valve is being opened, when a hot leg stop valve is open and the cold leg stop valve in the same loop is closed, and during the time a cold leg stop valve is being opened; except when the reactor coolant temperature and pressure are equal to or less than 350°F and 450 psig, respectively.
Basis The above specified precautions and the design of the Loop Stop Valve Interlock System(l) provide assurance that no accidental reactivity addition to the core
TS '.1.17-3 3-17-72 could occur during the startup operation of an inactive coolant loop.
The Loop Stop Valve Interlock System will eliminate the possibility of adding reactivity to the core at any significant rate by preventing the cold leg valve opening unless a mixing flow between the isolated 1oop and the remainder of the Reactor Coolant System has existed for 90 minutes, and the temperatures in the cold and hot leg of the isolated loop are respectively within 20°F of the highest cold and hot leg temperatures of the other loops.
The boron concentration in the isolated loop is maintained continuously at least equal to that in the active loops. The verification of the boron con-centration in the inactive loop prior to opening the hot leg and the cold leg valves provides a re-assurance of the adequacy of the boron concentration.
The continuous monitoring of the neutron flux by the nuclear instrumentation will immediately indicate a rapid change in the reactivity status of the core during the mixing flow phase. A slow reactivity change will be noticeable by comparing the count rates which will be logged every five minutes with the initial count rate. The fully withdrawn shutdown rods maintain at any time the capability to shut the unit down if this is required.
Reference (1) FSAR Section 4.2
TS 3.18-1 3-17-72
- 3.18 MOVABLE IN-CORE INSTRUMENTATION Applicability Applies to the operability of the movable detector instrumentation system.
Objective To specify functional requirements on the use of the in-core instrumentation systems, for the recalibration of the excore symmetrical off-set detection system.
Specification A. A minimum of 16 total accessible thimbles and at least 2 per quadrant 1 each of which will accept a movable incore~detector, shall be operable during re-calibration of the excore symmetrical off-set detection system, B. Power shall be limited to 90% of rated power for three loop operation, 54% of rated power for two loop operation with the loop stop valves closed, and 50% of rated power for two loop operation with the loop stop valves open if re-calibration requirements for the excore symmetrical off-set detection system, identified in Table 4.1-1, are not met.
TS 3.18-2 3-17-72 Basis The Movable In-core Instrumentation System (l)has five drives, five detectors, and 50 thimbles in the core. Each detector can be routed to twenty or more thimbles. Consequently, the full system has a great deal more capability than would be needed for the calibration of the excore detectors.
To calibrate the excore detectors system, it is only necessary that the Movable In-core System be used to determine the gross power distribution in the core as indicated by the power balance between the top and bottom halves of the core.
After the excore system is calibrated initially, recalibration is needed only infrequently to compensate for changes in the core, due for example to fuel depletion, and for changes in the detectors.
If the recalibration is not performed, the mandated power reduction assures safe operation of the reactor since it will compensate for an error of 10%
in the excore protection system. Experience at Beznau No. 1 and R. E. Ginna plants has shown that drift due to the core on instrument channels is very slight. Thus limiting the operating levels to 90% of the rated two and three loop powers is very conservative for both operational modes.
Reference (1) FSAR - Section 7.6
TS 3.19-1 3-17-72 3.19 MAIN CONTROL ROOM VENTILATION SYSTEM Applicability Applies to the ability to maintain a positive differential pressure in the main control room.
Objective To specify functional requirements for the main control room ventilation system.
Specification A bottled dry air bank shall be available to pressurize the main control room to a positive differential pressure with respect to adjoining areas of the auxiliary, turbine, and service buildings for one hour. A minimum positive differential pressure of 0.05 inches of water must be maintained when the control *room is isolated under accident conditions. This. capability shall be demonstrated by the testing requirement delineated in Technical Specification
. 4.1.
Basis Following a design basis loss-of-coolant accident the containment will be depressurized to subatmospheric conditions in less than one hour, thus terminating leakage from the conta.inment. The main control room is maintained at a positive differential pressure using bottled.ai:r_dl-lring the period when containment leakage may exist to prevent contamination.
TS 4.0-1 3-17-72 4.0 SURVEILLANCE REQUIREMENTS Surveillance requirements provide for testing, calibrating, or inspecting those systems or components which are required to assure that operation of the units or the station will be as prescribed in the preceding sections.
Specified time intervals may be adjusted plus or minus 25 percent to accommodate normal test schedules.
References FSAR Section 12.9 Inservice Inspection FSAR Section 14.1 Operational Safety Review
TS 4.1-1
- 4.1 OPERATIONAL SAFETY REVIEW 3-17-72 Applicability Applies to items directly related to safety limits and limiting conditions for operation.
Objective To specify the minimum frequency and type of surveillance to be applied to unit equipment and conditions.
Specification A, Calibration, testing, and checking of instrumentation channels shall be performed as detailed in Table 4.1-1.
B. Equipment tests shall be conducted as detailed in Table 4.l-2A.
C. Sampling tests shall be conducted as detailed in Table 4.2-2B.
D. Whenever containment integrity is not required, only the asterisked items in Table 4.1-1 and 4.1-2 are applicable.
E, Flushing of sensitized stainless steel pipe sections shall be conducted as detailed in TS Table 4.1-3.
TS 4.1-2 3 72 F. During or immediately after the flushes required by Specification 4.1.E, water flushed from the pipe section will be analyzed for chloride concen-tration and fluoride concentration. If either the chloride or fluoride concentration exceeds 0.15 ppm, the flush will be continued or resumed until further samples indicate that the chloride concentration and the fluoride concentration in the sensitized pipe, have been reduced to less than 0.15 ppm.
Basis Check Failures such as blown instrument fuses, defective indicators, and faulted amplifiers which result in "upscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system.
Furthermore, such failures are, in many cases, revealed by alarm or annunciator action, and a periodic check supplements this type of built-in surveillance.
Based on experience in operation of both conventional and nuclear unit systems, when the unit is in operation, the minimum checking frequencies set forth are deemed adequate for reactor and steam system instrtllllentation.
Calibration Calibration shall be performed to ensure the presentation and acquisition of accurate information.
The nuclear flux (power level) channels shall be calibrated daily against a heat balance standard to account for errors induced by changing rod patterns and
TS 4.1-3 3-17-72 core physics parameters.
Other channels are subject only to the "drift" errors induced within the instrumentation itself and, consequently, can tolerate longer intervals between calibration. Process systems instrumentation errors resulting from drift within the individual instruments are normally negligible.
During the interval between periodic channel tests and daily check of each channel, a comparison between redundant channels will reveal any abnormal condition resulting from a calibration shift, due to instrument drift of a single channel.
During the periodic channel test, if it is deemed necessary, the channel may be tuned to compensate for the calibration shift. However, it is not expected that this will be required at any fixed or frequent interval.
Thus, minimum calibration frequencies of once-per-day for the nuclear flux (power level) channels, and once each refueling shutdown for the process system channels is considered acceptable.
Testing The minimum testing frequency for those instrument channels connected to the safety system is based on an average unsafe failure rate of 2/5 x 10- 6 failure/hr per channel. This is based on operating experience at conventional and nuclear units. An unsafe failure is defined as one which negates channel operability and which, due to its nature, is revealed only when the channel is tested or when it attempts to respond to a proper signal.
TS 4.1-4 3-17-72 For the specified one month test interval, the average unprotected time is 360 hrs in case of a failure occurring between test intervals. Thus, the probability of failure of one channel between test intervals is 360 x 2.5 x 10- 6 or .9 x 10-3 , Since two channels must fail in order to negate the safety function, the probability of simultaneous failure of two-out-of-three channels 2
is 3(.9 x 10- 3 ) = 2.4 x 10-6, This represents the fraction of time in which each three-channel system would have one operable and two inoperable channels and equals 2.4 x 10
-6 x 8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months per year, or (approximately) 1 minute/year.
It must also be noted that to thoroughly and correctly test a channel, the channel components must be made to respond in the same manner and to the same type of input as they would be expected to respond to during their normal operation. This, of necessity, requires that during the test the channel be made inoperable for a short period of time. This factor must be, and has been, taken into consideration in determining testing frequencies.
Because of their greater degree of redundancy, the 2/4 logic arrays provide an even greater measure of protection and are thereby acceptable for the same testing interval. Those items specified for monthly testing are associated with process-components where other means of verification provide additional assurance that the channel is operable, thereby requiring less frequent testing.
Flushing During construction of the facility, stress relieving of some of the cold bent type 316 stainless steel piping, resulted in its becoming sensitized to potential stress corrosion cracking under certain conditions e.g. low pH in conjunction with high clorides. The systems containing sensitized pipe which remain wet during normal operation and have no flow, e.g. safety injection, may contribute to the buildup of those contaminants which could cause accelerated corrosive
TS 4.1-5 3-17-72 attack of the pipe. In order to insure the continued integrity of the pipe throughout plant life, the affected lines are flushed periodically to remove stagnant water which may contain contaminants.
The flushing requirements delineated in TS Table 4.1-3 insure that a buildup of contaminants will not occur. The specified minimum flush durations, with expected flow rates during flushing, insures that a volume of water greater than the volume contained in the stagnant flow paths listed in Table 4.1-3 will be flushed. The required sampling of flushed lines further ensures that the specified flushing procedures were effective in removing any undesirable contaminants that may have accumulated in the sensitized piping.
The control room ventilation system is required to establish a positive differential pressure in the control room for one hour following a design basis loss-of-coolant accident using a bottled air supply as the source of air. The ability of the system to meet this requirement is tested by pres-surizing the control room using the ventilation system fans and comparing the volume of air required to that stored. The test is conducted each refueling interval (approximately 12 to 18 months), normally coinciding with the refueling outage of either Unit 1 or Unit 2.
- TABLE 4.1-1 MINIMUM FREQUENCIES FOR CHECK, CALIBRATIONS AND
. TESTS OF INSTRUMENT CHANNELS Channel Description Check Calibrate Test Remarks
1. Nuclear Power Range s D (1) BW (2) 1) Against a heat balance standard M (3) 2) Signal to ~T; bistable action AP (3) (permissive, rod stop, trips)
3) Upper and lower chambers for symetric offset by means of the moveable incore detector system.
2. Nuclear Intermediate Range *S (1) N.A. p (2) 1) Once/shift when in service
2) Log level; bistable action (permissive, rod stop, trip)
3. Nuclear Source Range icS (1) N.A. p (2) 1) Once/shift when in service
2) Bistable action (alarm, trip)
4. Reactor Coolant Temperature *S R BW (1) 1) Overtemperature-~T BW (2) 2) Overpower - ~T
s. Reactor Coolant Flow s R M
6. Pressurizer Water Level s R M
7. Pressurizer Pressure(High and low) s R M
8. 4 Kv Voltage & Frequency s R M Reactor protection circuits only
9. Analog Rod Position *S (1,2) R M (3) 1) With step counters (4) 2) Each six inches of rod motion >-3 when data logger is out of w C/l I
I-' .i:,.
service -..J
I I-'
3) Rod bottom bistable action -...i I
4) NA When reactor is in cold shut- N °'
down
TABLE 4.1~1 (Continued)
Channel Description *check Calibrate Test Remarks
10. Rod Position Bank Counters S (1, 2) N .A. N.A. 1) Each six inches of rod motion when data logger is out of service
2) With analog rod position
11. Steam Generator Level S R M
12. Charging Flow N.A. R N .A.
13. Residual Heat Removal Pump Flow N.A. R N.A.
14. Boric Acid Tank Level *D R N.A.
15. Refueling Water Storage Tank Level W R N.A.
16. Boron Injection Tank Level W R N.A.
17. Volume Control Tank Level N.A. R N.A.
18. Reactor Containment Pressure-CLS *D R }1 (1) 1) Isolation Valve signal and spray signal
19. Process and Area Radiation Monitor-ing Systems *D R M
20. Boric Acid Control N.A. R N.A.
21. Containment Sump Level N.A. R N.A.
22. Accumulator Level and Pressure s R N.A.
23. Containment Pressure-Vacuum Pump H
System s R N.A. (/)
24. Steam Line Pressure s R M
i
- TABLE 4.1-1 (Continued)
Channel Description Check Calibrate Test Remarks
25. Turbine First Stage Pressure s R M
26. Emergency Plan Radiation Instruments *M R M
27. Environmental Radiation Monitors 1-M N.A. N.A. TLD Dosimeters
28. Logic Channel Testing N.A. N.A. M
29. Turbine Overspeed Protection N.A. R R Trip Channel (Electrical)
30. Turbine Trip Set Point N.A. R R Stop valve closure or low EH fluid pressure
31. Seismic Instrumentation M SA M
32. Reactor Trip Breaker N.A. N.A. M S - Each Shift M - Monthly D - Daily P - Prior to each startup if not done previous week W - Weekly R - Each Refueling Shutdown NA - Not applicable BW - Every two weeks SA - Semiannually *p;p - After each startup if not done previous week
See Specification 4.lD
T!.i li .1-9 3-17-72 TABLE 4.l-2A MINIMUM FREQUE~CJ FOR EQUIPMENT TESTS FSAR e Description Test Frequency Section Reference
1. Control Rod Assemblies Rod drop times of Each refueling 7 all full length rods shutdown or after at hot and cold con- disassembly or ditions. maintenance requiring the breech of the Reactor Coolant System integrity
2. Control Rod Assemblies Partial movement Every 2 weeks 7 of all rods
3. Refueling Water Chemical Functional .Each refueling 6 Addition Tank shutdown
4. Pressurizer Safety Setpoint Each refueling 4 Valves shutdown
5. Main Steam Safety Valves Setpoint Each refueling shutdown 10
6. Containment Isolation *Functional Each refueling 5 Trip shutdown
7. Refueling System
Fune tional Prior to refueling 9.12 e 8.
Interlocks Service Water System *Functional Each refueling shutdown 9.9
9. Fire Protection Pump Functional Monthly 9.10 and Power Supply
10. Primary System Leakage
Evaluate Daily 4
11. Diesel Fuel Supply *Fuel Inventory 5 days/week 8.5
12. Boric Acid Piping *Operational Monthly 9.1 Heat Tracing Circuits
13. Main Steam Line Trip Functional Monthly 10 Valves
14. Service Water System Functional Each Refueling 9.9 Values in Line Supplying Recirculation Spray Heat Exchangers
15. Control Room *Ability to maintain Each refueling 9.13 Ventilation positive pressure for interval (approx.
System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months using a volume every 12-18 months) of air equivalent to or less than that stored in the bottled air suppy.
See Specification 4.1.D
TS 4.1-10 3-17-72 TABLE 4.1-2B MINIMUM FREQUENCIES FOR SAMPLING TESTS FSAR Section Description Test Frequency Reference
1. Reactor Coolant Liquid Radio-chemical Monthly Samples Analysis (1)
Gross Activity (2) 5 days/week 9.1
Tritimn Activity Weekly 9.1
Chemistry (Cl, F&0 ) 5 days/week 4 2
Boron Concentration Twice/week 9.1 E Determination Semiannually (3)
2. Refueling Water Storage Boron Concentration Weekly 6 Tank Water Sample
3. Boric Acid Tanks *Boron Concentration Twice/week 9.1
4. Boron Injection Tank Boron Concentration Twice/week 6
5. Chemical Additive NaOH Concentration Monthly 6 Tank e 6. Spent Fuel Pit *Boron Concentration Monthly 9.5
7. Secondary Coolant Fifteen minute degassed Weekly 10.3
/3 and'/ activity (4)
8. Stack Gas Iodine and
I-131 and particulate Weekly Particulate Samples radioactive releases(5) 9, Accumulator Boron Concentration Monthly 6.2
See Specification 4.1.D (1) A radiochemical analysis will be made to evaluate the following corrosion products:
Cr 51, Fe 59, Mn 54, Co 58, and Co 60.
(2) A gross beta-gamma degassed activity analysis shall consist of the quantitative measurement of the total radioactivity of the primary coolant in units of µCi/cc.
(3) E determination will be started when the gross gamma degassed activity of radionucli:ies with half-lives greater than 30 minutes analysis indicates > lOµCi/ cc.
(4) If the fifteen minute degassed beta and gamma activity is 10% of that given in Specification 3.6.C, an I-131 analysis will be performed.
(5) If the activity of the samples is 10% or greater of that given in Specification 3.11.B.1. the frequency shall be increased to daily.
TABLE 4.1-3 MINIMUM FREQUENCIES FOR FLUSHING SENSITIZED PIPE FLUSH FLOW PATH MINIMUM FLUSH GENERAL DESCRIPTION DURATION FREQUENCY REMARKS
1. From C.S. Pump CS-P-lA 15 minutes Monthly Run in conjunction with to M.O. Isolation Valves or immediately after pump test required by Specification 4.5.A.l
2. From C.S. Pump CS-P-lB 20 minutes Monthly Run in conjunction with to M.O. Isolation Valves or immediately after pump test required by Specification 4.5.A.1
3. From L.H.S.I.Pump, SI-P-lA, 20 minutes Monthly Run in conjunction with or Discharge Line to MOV 863A immediately after pump test required by Specifica-tion 4.11.B.1
4. S.I. line, from charging pump Flushes to be performed discharge loop fill header to only when R.C. System containment missile barrier, for pressure is greater than flow to 1500 psig.
a. R.C. hot leg loop 1 15 minutes Monthly
b. R.C. hot leg loop 2 10 minutes Monthly
c. R.C. hot leg loop 3 15 minutes Monthly
5. s. I. line, from charging pump discharge Flushes to be performed header to containment missile barrier, only when R.C. System for flow to pressure is greater than
a. R.C. cold leg loop 1 5 minutes Monthly 1500 psig. f-;!
b. R.C. cold leg loop 2 5 minutes Monthly Cf.l w
C, R.C. cold leg loop 3 5 minutes Monthly I 1--'
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TS !1.,L-J 3-17-72 4.2 REACTOR COOLANT SYSTEM COMPONENT INSPECTIONS Applicability Applies to in-service inspection of the reactor vessel and the reactor coolant system pressure boundary.
Objectives To provide assurance of the continued integrity of the reactor vessel and reactor coolant system press.ure boundary.
Specifications A. Prior to initial unit operation a survey using ultrasonic, visual, and surface techniques shall be made to establish pre-operational system integrity and establish baseline data.
B. Following initial unit operation, nondestructive inspections shall be performed as specified in TS Table 4. 2-1. The results obtained from these inspections shall be evaluated after five years of commercial operation and reviewed in light of the technology available at that time.
C. The normal inspection interval is 10 years.
TS 4.2-2 3-17-72 D, Detailed records of each inspection shall be maintained to allow a continuing evaluation and comparison with futurP. inspections.
Bases The in-service inspection program is based on Section XI of the ASME Code for Inservice Inspection of Nuclear Reactor Coolant Systems dated January, 1970, Winter 1970 Addenda, (and the* rules and guidelines given in the AEC document, "In-service Inspection Requirements for Nuclear Power Plants Constructed with Limited Accessibility for In-service Inspection"). Since each unit was designed and partially constructed without the benefit of the ASME In-Service Inspection Code, 100-percent compliance may not be feasible or practical. However, the inspection program was developed by adopting, insofar as practicable, the principles and intent embodied in the Code, It must be recognized that equipment and techniques to perform the inspection are still in development, and to that extent are speculative. In most areas scheduled for test, a detailed pre-service mapping will be conducted using techniques which are believed to be appropriate.for the later operational inspections. The areas specified for inspection are representative of those experiencing relatively high strain and therefore will serve to indicate potential problems before significant flaws could develop. As more experience is gained in operation of pressurized-water reactors, the recollllllended time schedule and location of inspection might be altered, or should new techniques be developed, consideration will be given to incorporate these into this inspection program.
TS 4.2-3
--- 3-17-72 The program defines the examinations to be performed within the first 5 years of unit operation. The inspection sampling and frequency during this period shall be governed by the requirements as specified in the ASME Code for the first third of the 10 year inspection interval. A tentative 10 year program is also included, although the operating experience and results of the first five years will constitute the bases for re-evaluation to take into account changes in technology, equipment, and the Code. This program complies with the intent of the Code but has some variations to take into account unit design as discussed under each inspection category.
The techniques anticipated for in-service inspection include visual inspections, ultrasonics, radiographic, magnetic particle, and liquid penetrant testing of e selected parts during refueling periods and necessary maintenance outages.
The primary pressure boundary covered by this inspection will include the primary reactor coolant system and branch lines 2 inches in diameter and greater. However it should be noted that excluded from volumetric examination will be that piping 4 inches in diameter and smaller and on which no meaningful volumetric examination results can be obtained. The inspection programs for the reactor vessel, the pressurizer, the primary side of the steam ~enerators, piping, pumps, and valves are outlined in TS Table 4.2-1. Each component system is discussed in categories corresponding to Section XI of the ASME Code.
TS 4.2-4 3-17-72
- The following discussions in each category cover the general type of exami-nation planned in TS Table 4.2 and the technical reasons for any exceptions to the Code.
A. Reactor Pressure Vessel Category A - Pressure Containing Welds in Reactor Vessel Belt-Line Region Depending upon development of appropriate equipment, it is intended that these welds be volumetrically examined, when required, using automated ultrasonic techniques. The mechanized device could be used only when the core and lower internals have been removed from the vessel. For this reason, examination may be made at or near the end of the 10-year interval or when the core and lower internals are removed at an earlier date for other reasons.
Category B - Pressure Containing Welds in Remainder of Vessel Dependent upon development of appropriate equipment, the same mechanized ultrasonic techniques used for examination of the welds in the core region would be used to examine the other longitudinal and circumferential welds in the barrel section of the reactor pressure vessel and would be performed at the same time and manner as the welds in Category A. There are no meridional seam welds.
Category C - Pressure Containing Welds, Vessel-to-Flange and Head-to-Flange The head flange weld can be examined using either mechanized or manual ultrasonic techniques. Mechanized techniques are preferred due to the repeatability of such techniques and the ability to record the data.
TS 4. 2-5 3-17-72 e The vessel flange weld can be examined using mechanized ultrasonic techniques.
This examination would also include the ligaments in the flange between the bolt holes.
Category D - Pressure Containing Nozzles in Vessels Depending upon development of appropriate equipment, it is intenped that these welds be volumetrically examined from the ID using automated ultrasonic techniques.
Examination of the coolant outlet nozzle-to-shell welds would be performed from the ID without removal of the core barrel. This examination would also include the integral extension of the nozzle inside the vessel and the inner nozzle radii.
Examination of the coolant inlet nozzles-to-shell welds from the ID requires removal of the core and lower internals. Therefore, these nozzle welds, including the inner radii, will be examined at the same time as the welds in Category A.
Category E Pressure Containing Welds in Vessel Penetrations The only penetrations in this category are the control rod dtive housings in the upper head and the in-core instrumentation penetrations in the lower head.
The control rod drive tubes are welded to the upper head with a partial penetration weld. A thermal sleeve is permanently affixed to the inside of the stub tube prohibiting'volumetric inspection from the ID. Geometry and the partial penetration weld do not allow a meaningful inspection from the O.D. These welds would be visually examined as specified in Category E-2.
TS 4. l-b 3-17-72 Volumetric inspection of the in-core instrumentation penetrations is dependent upon development of acceptable equipment and techniques. Consequently, these welds will be examined as*specified in Category E-2, Category E Pressure Containing Welds in Vessel Penetrations A visual examination for evidence of leakage would be made of the penetrations in the upper head at .the ti.me of the system hydrostatic tests as required by section IS-520 of the Code. The penetrations in the lower head will also be examined for leakage during this test if volumetric inspection techniques are not developed and if radiation levels permit access.
Category F - Pressure Containing Dissimilar Metal Welds Depending upon development of appropriate equipment, it is intended that these welds be volumetrically examined from the ID using automated ultrasonic techniques.
Experience with other units has shown that, in general, these welds can be volumetrically examined with ultrasonic techniques. The feasibility of this examination would be determined on the pre-operational examination, and this will determine the acceptable method to.use in the inservice inspection.
Examination of the coolant outlet nozzle dissimilar metal weld (pipe-to-nozzle weld) would be performed from the ID without removal of the core barrel.
Examination of the coolant inlet nozzle dissimilar metal weld would require removal of the core and lower internals. Therefore these welds would be examined at the same time as the welds in Category A. Exception is taken to performing a visual and surface examination from the O.D. on these welds due to anticipated radiation levels and physical access problems.
TS 4.2-7 3-17-72 Category G - Pressure Retaining Bolting All of the reactor vessel studs are scheduled to be removed at each refueling cycle. They are, thus, available for a volumetric,' surface, and visual examination as may be required. A visual examination and an ultrasonic volumetric examination would be made of each stud on an examination schedule as shown in TS Table 4.2-1 for the 5-year program and the tentative 10-year program.
The ciosure stud nuts woultj. be examined with techniques similar to those for the studs, and the washers would be visually inspected only.
The ligaments between the bolt holes of the reactor pressure vessel would be volumetrically examined at the same time the vessel flange weld is examined.
Category H - Vessel External Supports The reactor pressure vessel is supported on pads, integrally welded to the coolant nozzles. In accordance with Category Hof Table IS-251 of the Code, this examination is covered by Category D.
Category I - Vessel Interior Clad Surfaces Those areas of the vessel cladding normally accessible during refueling periods shall be visually examined during the inspection interval. However, whenever the core barrel and lower internals are removed, the patches in the vessel interior would be visually examined. The examination scheduled in the first 5-year period and tentatively scheduled for the 10-year period is given in TS Table 4.2-1.
TS 4.2-8 3-17-72 Selected areas of the cladding in the upper closure head will be examined duri~g normal refueling periods when the head is removed.
Category N - Interior Surfaces and Internal Components of Reactor Vessels It is proposed that examinations in this category be made by remote television or borescopic examination. A critical examination would be made at the first refueling to detect if any changes have occurred due to initial operation.
The amount of examination to be performed at subsequent refueling outages would depend upon the results of the first examination and those made on comparable pressurized-water systems.
B. Pressurizer Category B - Pressure Containing Welds in Vessels The primary heads on the pressurizer are of cast material and do not contain meridional welds. There are circumferential welds joining the heads to the barrel section and circumferential and longitudinal welds in the barrel section that require examination. The examinations scheduled for the 5-year period and tentatively scheduled for the 10-year period are given in TS Table 4.2-1.
Category D - Pressure Containing Nozzles in Vessels The nozzles are integrally cast into the head and there are, therefore, no welds in this category.
TS 4.2-9 e 3-17-72 Category E - Pressure Containing Welds in Vessel The heater connections and instrument penetrations of the pressurizer meet the exclusion criteria of Section IS-121 of the Code. These penetrations shall be given a visual examination for evidence of leaking at the time of the system hydrostatic test as required by IS-520 of the Code.
Category F - Pressure Containing Dissimilar Metal Welds There are dissimilar metal welds on the nozzles of the pressurizer. Experience with other units has shown that, in general, these welds can be volumetrically examined with ultrasonic techniques. The feasibility of this examination would be determined on the pre-operational examination, and this will determine the acceptable method to use in the in-service inspection. The examinations scheduled for the first 5-year period and tentatively scheduled for the 10-year period are given in TS Table 4.2-1.
Category G - Pressure Retaining Bolting Bolting on the pressurizer below 2 inches in diameter would require visual examination. Excluded from examination are bolting of a single connection whose failure results in ~onditions that satisfy the exclusion criterion of IS-121 of the Code. The examinations scheduled for the 10-year period are given in TS Table 4.2-1.
Category H - Vessel External Supports There are no integrally welded vessel supports on the Surry pressurizer.
TS 4.2-10 3-17-72 Category I - Vessel Interior Clad Surface A vi:1ual inspection of a 36-square-inch area of the clad surface of the pressurizer. would be made at such time as the pressurizer is opened for necessary maintenance. The examinations scheduled for the first 5-year period and tentatively scheduled for the 10-year period are given in TS Table 4.2-1.
C. Steam Generators Category B - Pressure Containing Welds in Vessels The circumferential welds joining the heads to the tube sheet are the only welds that require examination. The examinations scheduled for the 5-year period and tentatively scheduled for the 10-year period are given in TS Table 4.2-1.
Category D - Pressure Containing Nozzles in Vessels The nozzles are integrally cast with the head and there are no nozzle-to-shell welds. The inner nozzle radii of the heads would be volumetrically examined from the ID when the generators are opened for necessary maintenance.
Category F - Pressure Containing Dissimilar Metal Welds The dissimilar metal welds between the nozzles and the main coolant piping join cast carbon steel to cast stainless steel. Volumetric inspection of these welds will be dependent upon meaningful examination techniques and results of the pre-service inspection,
TS 4.2-11 3-17-72 Category G - Pressure Retaining Bolting There is bolting on the steam generators of less than 2 inches in diameter.
Excluded from examination are bolting of a single connection whose failure results in conditions that satisfy the exclusion criterion of IS-121 of the Gode.
Category H - Vessel External Supports The steam generators are supported by pads integrally cast with the lower head.
There are no welds that require examination.
Category I - Vessel Interior Clad Surface e A visual inspection of the clad surface on the primary head of the steam generators from the ID will be made at such time as the generators are opened for necessary maintenance. The examinations scheduled for the first 5-year period and tentatively scheduled for the 10-year period are given in TS"Table 4.2-1.
D. Piping Pressure Boundaries Category F - Pressure Containing Dissimilar Metal Welds The main coolant piping of this plant is stainless steel, as are the pump and valve nozzles. Thus, the only dissimilar metal welds in the main coolant plpa are the attachments between the coolant pipe and the vessels. There are a few other dissimilar metal welds in the piping system, such as piping connections to the pressurizer.
TS 4.2-12 3-17-72 T'.1tse inspections and the pipe-to-nozzle connections on the steam generator are specified in TS Table 4.2-1, alon~ with inspection of the' reactor vessel pipe-to-nozzle welds. The examinations scheduled for the first 5-year period and tentatively scheduled for the 10-year period are given in TS Table 4.2-1.
~ategory J-i & J Pressure Contairtirtg Welds in'Piping The circumferential welds in the main coolant piping, branch, and auxiliary piping systems, will be examined in compliance with the Code assuming access and radiation levels allow. Exception is taken for socket welds and the longitudinal welds in the cast elbows of the main coolant piping until development of acceptable volumetric inspection techniques.
Category G - Pressure Retaining Bolting All bolting in the piping system is below 2 inches in diameter and will be inspected in compliance with the Code.
Category K - Support Members and Structures for Piping Systems The piping systems that contain supports and restraints integrally welded to the pressure containing boundary will be inspected in compliance with the Code.
Members and structures whose structural integrity is relied upon to withstand the design loads and seismic-induced displacements will be inspected as specified in Item 4.6 of TS Table 4.2-1. The examinations scheduled for the 5-year period and tentatively scheduled for the 10-year period are given in TS Table 4.2-1.
TS 4,2-l~
e 3-17-72 E. Pump Pressure Boundary Category F - Pressure Containing Dissimilar Metal Welds There are no nozzle-to-safe end welds between the main coolant piping and the pump suction and discharge nozzles.
Category G - Pressure Retaining Bolting There are bolts 2 inches and larger in diameter on the pumps. If the connection is not broken during the inspection interval, a visual exami-nation would be made and an ultrasonic examination would be made with the bolting under tension. If the bolting connection is broken for any reason,
,, visual and volumetric examination of the bolting would be made and, in addition, a volumetric inspection of the ligaments of the base material would be made insofar as is possible due to the cast structure. The examinations scheduled for the 5-year period and tentatively scheduled for the 10-year period are given in TS Table 4.2-1.
Category K -* Support Members and Structures for Pumps The support members for the pumps consist of a cast foot welded to the cast casing. These supports and the pump supporting structure would be visually examined since there are no acceptable techhiques for ultrasonically inspecting these welds.
Category L - Pressure Containing Welds in Pump Casing
These welds were volumetrically examined using radiographic techniques during the pump manufacture. A simii'ar radiographic technique is theoretically
TS 4.2-14 3-17-72 possible when the pump is disassembled and the impeller has been removed.
However, it is not known if such radiography can be performed in practice due to pump design and the radiation level that will probably exist in this component as well as due to the interference of the swirl guide vanes, A study would be made to see if this is possible and, in addition, experience with other pressurized-water unit in-service inspections would be solicited to determine the feasibility. Thus, this examination is not scheduled, as is shown in TS Table 4.2-1. A visual examination of the available internal surfaces of the pump would be made at such time as the pump is disassembled for maintenance.
Item 5,8 - Primary Pump Flywheel All flywheels shall be visually examined at the first and third refueling.
An in-place ultrasonic volumetric examination of the areas of higher stress concentration at the bore and keyway of each flywheel shall be performed at approximately three (3) year intervals, during the refueling or maintenance shutdown coinciding with the inservice inspection schedule as required by the ASME Boiler and Pressure Vessel Code Section XI, A surface examination of all exposed surfaces and complete ultrasonic volumetric examination shall be conducted at approximately ten (10) year intervals. Removal of the flywheel is not required to perform these examinations.
F. Valve Pressure Boundaries Category F - Pressure Containing Dissimilar Metal Welds There are no dissimilar metal welds between the valves and the piping system in this facility,
TS 4.2-15 3-17-72 Category G - Pressure Retaining Bolting There is bolting 2 inches and larger in diameter on the main coolant stop valves. All other valve bolting is below 2 inches in size. The examinations scheduled for the 5-year period and tentatively scheduled for the 10-year period are given in rs Tabie 4.2-1, Category M Pressure Containing Welds in Valve Bodies There are no valves in this facility with pressure containing valve body welds.
Category M Valve Bodies The internal pressure boundary surfaces of one disassembled valve (3 inches and over in size) in each category and type shall be visually examined during each inspection interval. If the valve shall be disassembled for maintenance any time during the inspection interval, it may be inspected at this time.
Otherwise the examination may be performed at or near the end of each in-spection interval.
Category K Support Members and Structures for Valves There are no integrally-welded supports on the valves subject to this examination.
Category K Supports and Hangers Any valve support whose structural integrity is relied upon to withstand the design loads and seismic-induced displacements would be examined. The examinations
- scheduled for the 5-year period and tentatively scheduled for the 10-year period are given in TS Table 4.2~1.
TS 4.2-16 3-17-72 G. Miscellaneous Inspections Item 7.1 - Materials Irradiation Surveillance Specimens The reactor vessel surveillance program includes eight speciment capsules to evaluate radiation damage based on pre-irradiation and post-irradiation testing of specimens.
Capsule No. 1 shall be removed and examined at the first region refueling.
Capsule No. 2 shall be removed and examined after five years.
Capsule No. 3 shall be removed and examined after ten years.
Capsule No. 4 shall be removed and examined after twenty years.
Capsules No. 5-8 are spares for complementary or duplicate testing.
H. Sensitized Stainless Steel Piping This piping is subject to augmented inspection to assure. piping integrity.
Item 8.1 Sensitized stainless steel piping which is part of section D, category J will be inspected at three times the frequency required by the code. For the re-quired visual inspection, the piping will be pressurized by the procedures de-fined in Table 4.1-3 of Technical Specification 4.1 concerning flushing of sensitized stainless steel piping.
Item 8.2 Sensitized stainless steel piping which is not subject to Section XI of the ASME Code, will undergo visual and surface examination.
The containment and recirculation spray rings, which are located in the overhead of the containment, will be visually inspected. Additionally, sections of the piping will be examined by liquid penetrant inspection when the piping is visually inspected. At least 25 percent of the examinations shall have been
TS 4.2-17 3-17-72 completed by the expiration of one-third of the inspection interval and at least 50 percent shall have be.en completed by the expiration of two-thirds of the inspection interval. The remaining required examinations shall be completed by the end of the inspection interval.
All other piping included in item 8.2 will be visually inspected at least every two years. Sections of this piping will be examined by liquid penetrant inspection when the piping is visually inspected. For the required visual inspection, the piping will be pressurized by the procedures defined in Table 4.1-3 of Technical Specification 4.1 concerning flushing of sensitized stainless steel piping.
TABLE 4.2~*1 SECTION A, REACTOR VESSEL AND CLOSURE HEAD Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During No. Category Areas Methods 5-Year Interval 10-Year Interval Remarks 1.1 A Longitudinal and circum- Volumetric 0% 5% of the length of the The required exa!l!ina-ferential shell welds in circumferential welds; tions may be made at or core region 10% of the length of the near the end of the 10-longitudinal welds year inspection interval.
w'ben the longitudinal and circumferential weld have received an ex-oosure to neutron flue~ce in excess of 1019 nvt (En of 1 Mev or above),
the length of the weld in the high fluence region to be examined shall be in-creased to 50 per cent, L2 B Longitudinal and circum- Volumetric 0% 5% of the length of the The required examinations ferential welds in shell circumferential welds; 107. nay be made at or near the (other than those of of the length of the end of the 10-year in-Category A and C) and longitudinal welds spection interval.
meridional and circum-ferential seam welds in bottom head and closure head (other than those of Category C) 1.3 C Vessel-to-flange and head- Volumetric 33% of the length of 100% of the length of Both of these welds are to-flange circumferential the circumferential the circumferential welds available for examina-welds welds tion during normal re-fueling operations.
1.4 D Primary nozzle-to-vessel Volumetric 100% coverage of a 100% of the remaining The coolant inlet nozzles- H welds and nozzle-to- coolant outlet nozzle- coolant outlet nozzle-to- to-shell welds and radius w UJ vessel inside radiused to-shell weld and inner shell welds and inner sections will be ex- 1 .i:-.
section nozzle radii f-J
radius sections and 100% amined at the s~e time -...JN of the coolant inlet ~s the welds in Category I I
-...J f-J nozzles-to-shell welds and A. NCO radius sections
TABLE 4.2-1 SECTION A. REACTOR VESSEL Ai~D CLOSURE HEAD (Continued)
Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During No. Category Areas* Methods 5-Year Interval 10-Year Interval Remarks 1.5 E-1 Vessel penetrations, in- Volumetric 0% 25% No meaningful vollll!letric cluding control rod drive examinations can be per-penetrations and control formed on the control rod rod housing pressure drive penetrations at this
boundary welds time. Volumetric examination of the instrument penetrations in the bottom head ~ill be performed dependent upon de-velopment of acceptable equip-ment.
1.6 E-2 Vessel penetrations, in- Visual* 0% 25% The examination will be a eluding control rod drive visual examination for evidence penetrations and control of leaking at the time of the rod housing pressure system hydrostatic test at or boundary welds near the end of the inspection interval.
1. 7 F Primary nozzles to safe Visual and 100% of the dissimilar 100% of the re- The dissimilar metal welds of end welds surface and metal weld on one maining coolant out- each nozzle will be volumetri-volumetric coolant outlet nozzle let nozzle dissimilar cally examined at the same time will be volumetrically metal welds and 100% as the nozzle-to-shell weld, examined of the coolant inlet as specified in Category D.
nozzle dissimilar metal welds 1-3 Cf.)
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e TABLE 4.2-1 SECTION A, REACTOR VESSEL Ai"l'D CLOSURE HEAD (Continued)
Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During
...B£..:.. Category Areas Methods 5-Year Interval 10-Year Interval Remarks
1. 5 E-1 Vessel penetrations, in- Volumetric 0% 25% No meaningful volllI!letric cluding control rod drive examinations can be per-penetrations and control formed on the control rod rod housing pressure drive penetrations at this
boundary welds time. Volumetric examination of the instrument penetrations in the bottom head will be performed dependent upon de-velopment of acceptable equip-ment.
1.6 E-2 Vessel penetrations, in- Visual* 0% 25% The examination will be a eluding control rod drive visual examination for evidence penetrations and control of leaking at the time of the rod housing pressure system hydrostatic test at or boundary welds near the end of the inspection interval.
1. 7 F Primary nozzles to safe Visual and 100% of the dissimilar 100% of the re- The dissimilar metal welds of end welds surface and metal weld on one maining coolant out- each nozzle will be volumetri-volumetric coolant outlet nozzle let nozzle dissimilar cally examined at the same time will be volumetrically metal welds and 100% as the nozzle-to-shell weld, examined of the coolant inlet as specified in Category D.
nozzle dissimilar metal welds L,J I .p-f-' 0
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TABLE 4.,2-1 SECTIOR A. REACTOR VESSEL AND CLOSURE HEAD (Continued)
Required Required Extent of Examination Tentative Inspec-Itea hamination Examination Planned During First tion During
....!!!?:. Category Areas Methods 5-Year Interval 10-Year Interval Remarks
1. 8 G-1 Closure etude end nuts Volumetric 33% 100% Vessel studs and nuts and visual would be inspected or surface volumetrically and by surface techniques as required. The ligaments of the vessel flange would be examined at the saae time as the flange welds.
1.9 G-1 Ligaments between threaded stud Volumetric 33% of the vessel-to- 100% of the vessel The ligaments will be holes flange bolt ligaments flange bolt ligaments examined at the same time will be examined will be examined as the flange weld of Item No, 1.3.
1.10 G-1 Closure washers, bushings Visual 33% 100% None 1,11 G-2 Pressure retaining bolting Visual Not applicabie Not applicable There are no pressure retaining bolts less than 2 inches on the Surry Reactor vessels.
1,12 B Integrslly welded vessel Volumetric (See remarks) (See remarks) These welds are covered aupport11 by the examinations of Category D.
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TABLE 4.2-1 SECTION A. REACTOR VESSEL AND CLOSURE HEAD (Continued)
Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During
~ Catego!:l Areas Methods 5-Year Interval 10-Year Interval Remarks 1.13 I-1 Closure head cladding Visual and 2 patches 6 patches During the 10-year period, surface or at least 6 patches (each volumetric 36 square inches) evenly distributed in the vessel head would be examined.
1.14 I-1 Vessel cladding Visual None 6 patches During the 10-year period, at least 6 patches (each 36 square incl:es) evenly distributed in the vessel head would be visually inspected.
1.15 N Interior surfaces and internals Visual A critical examination The inspections made at The examinati0n *will include and integrally welded internal will be made of the the 4th refueling cycle internal support attachments supports interior surfaces made will be repeated at the welded to the vessel whose available by normal 7th and 10th refueling failure may adversely affect refueling operations at cycle core integrity provided the 1st refueling cycle. these are available for This will be repeated at visual examination by the 4th refueling cycle components removed during with the amount of the normal refueling operations.
inspection being dependent upon results of the 1st inspection and that made on other pressurized-water sys~ems.
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e TABLE 4.2-1 SECTION B. PRESSURIZER Required Required Extent of Examination Tentative Inspec-It~ Examination Examination Planned During First tion During
...l!2!. Category Areas Methods 5-Year Interval 10-Year Interval Remarks 2.1 B Longitudinal and circumferential Visual and 5% of the length of the By the end of the 10-year None welds volumetric circumferential wel'd period, 10% of the length joining the lower head of the longitudinal and to the barrel section 5% of the length of the and 10% of the length circumferential weld of the adjoining longi- would be inspected tudinal weld would be examined.
2.2 D Nozzle-to-vessel welds Visual and Not applicable Not applicable There are no nozzle-to-volumetric head welds, as the nozzles are integrally cast with the heads. Instrument and sample nozzles are included in Category E.
2.3 E-1 Heater connections Visual and (See remarks) (See remarks) These connections are surface considered in Item 2.4.
2.4 E-2 Heater connections Visual (See remarks) A cumulative total of 25% The examination will be a of the heater connections visual examination for would be visually examined evidence of leaking at the time of the system hydrostatic test at or near the end of the inspection interval.
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TABLE 4.2-1 SECTION!. PRESSURIZER (Continued)
Required Required Extent of Examination Tentative Inspec-
!tea Examination Examination Planned During First tion During
~ Categot,: Areas Methods 5-Year lnter,,.al 10-Year Interval Remarks Pressure containing dissimilar Volumetric 100% of the weld joining 100% of the weld joining Although this item is aetal welds the surge line to the the surge line to the not required in the Code, outlet nozzle would be outlet nozzle and of the it is felt that dissimilar volumetrically examined weld joining the spray metal welds and stainless by the end of the 5-year line to the nozzle in the steel safe ends on the interval upper head would be pressurizer should be examined examined.
2 * .5 G-1 Pressure retaining bolting Visual and Not applicable Not applicable There is no bolting 2 volumetric inches and larger in diameter on the Surry pressurizer.
2.6. G-2 Pressure retaining bolting Visual 33% by end of 5-year Cumulative 100% by end The bolting below 2 interval of interval inches in diameter would be visually examined, either in place if the bolted connection is not disassembled during the inspection interval, or whenever the bolted connection is disassembled.
The bolting to be examined would include studs and nuts.
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e TABLE 4.2-1 SECTION B. PRESSURIZER (Continued)
Required Required Extent of Examination Item Tentative Inspec-Examination Examination Planned During First tion During
_.!£:.. Category Areas Methods 5-Year Interval 10-Year Interval Remarks 2.7 H Integrally welded vessel supports Visual and Not applicable Not applicable There are no integrally volumetric welded ves'sel supports on the Surry pressurizer.
2.8 I-2 Vessel cladding Visual None 1 patch (36 square inches) A selected patch of the pressurizer cladding would be visually examined when entry into the vessel is made available for necessary maintenance.
SECTION C. STEAM GENERATORS 3.1 B Longitudinal and circumferential Visual and 5% of the length of By the end of the interval, None welds, including tube sheet-to-- volumetric the circumferential 5% of the length of all head or shell welds on the weld joining the of the welds joining the primary side lower primary head to primary heads to the tube the tube sheet of one sheets would be inspected stejllll generator would be inspected 3.2 D Primary nozzle-to-vessel head Visual and (See remarks) (See remarks) There are no nozzle-to-welds and nozzle-to-head inside volumetric radiused section head welds in the primary heads of the steam generators as the nozzles are integrally cast with the head.
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e TABLE 4.2-1 SECTION C. STEAM GENERATORS Required Require'd Extent of Examination Tentative Inspec-Itea Examination Examination Planned During First tion During
~ Category Areas Hethods 5-Year Interval 10-Year Interval Remarks 3.3 F ~rimary nozzle-to-safe end welds Visual and (See remarks) (See remarks) The primary nozzles are surface and a buttered end pre-volumetric paration and the weld is located between the cast nozzle and a cast elbow. Volumetric inspection is dependent upon meaningful examina-tion techniques.
3.4 G-1 Pressure retaining bolting Visual and Not applicable Not applicable There is no pressure-volumetric retaining bolting greater than two inches in diameter on the steam generators.
3.5 G-2 Pressure retaining bolting Visual 33% by end of 5-year Cumulative 100% by end The bolting below 2 inches interval of interval in diameter would be visually examined, either in place if the bolted connection is not disassembled during the inspection interval, or whenever the bolting connection is disassembled.
The bolting to be examined would include studs and nuts.
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e e TABLE 4.2-1 SECTION C, STEAM GENERATORS (Continued)
Required Required Extent of Examination Tentative Inspec-Item Examination Examinatioll Planned During First tion During
.J!2..:. Category Areas Methods 5-Year Interval 10-Year Interval Remarks 3.6 H Integrally welded vessel supports Visual -!lnrl Not applicable Not applicable There ar~ no supports volumetric integrally welded to the steam generator pressure boundary.
3.7 I-2 Vessel cladding Visual None 1 patch (36 square inches) A visual examination would be made only at such time as the primary side is opened for necessary maintenance.
SECTION D, PIPING PRESSURE BOUNDARY 4.1
F Vessel~, pump-, and valve-safe Visual and (See remarks) By the end of the inter- Examination of these ends-to-primary pipe welds and surface and *val, a cumulative 100% welds is covered under safe ends in branch piping welds volumetric of the welds would have Section A, B, and C and been examined explained in the discussion, 4.2 J-1 Circumferential and longitudinal Visual and 5% of the circum- By the end of the inter- None pipe welds and branch pipe " volumetric ferential welds, val, a cumulative 25%
connection welds larger than 4 including one foot of of the circumferential inches in diameter any longitudinal weld welds in the piping on either side of the system would have been butt weld, would be examined, including one examined. foot of any longitudinal weld on either side of the butt welds l,.)
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TABLE 4.2-1 SECTION D. PIPING PRESSURE BOUNDARY Required Required Extent of Examination Tentative Inspec-Itea Examination Examination Planned During First tion During No, Category Areas Methods 5-Year Interval 10-Year Interval Remarks 4.3 G-1 Pressure retaining bolting Visual and Not applicable Not applicable There is no bolting vol11111etric 2 inches and larger in the piping system, 4.4 G-2 Pressure retaining bolting Visual 33% of the bolting By the end of the inter- All bolting is below would be examined val, 100% of the bolting 2 inches in diameter and would be examined would be visually examined, either in place if the bolted connection is not disassembled during the inspection interval, or whenever the bolted connection is dis-assembled. The bolting to be examined would include studs and nuts, 4.5 K-1 Integrally welded supports Visual and 10% of supports would By the end of the inter- Areas subject to volumetric be examined val, a cumulative 25% examination would include of the supports would any integrally welded be examined external support attach-ment which includes the welds to the pressure containing boundary, the base metal beneath the weld zone, and along the support attachment for distance of two base metal thicknesses.
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SECTION D. PIPING PRESSURE BOUNDARY (Continued)
Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During No. Category Areas Methods 5-Yeat Interval 10-Year Interval Remarks 4.6 K-2 Piping support and hanger Visual 33% of the supports By the end of the inter- The support members and would be examined val, a cumulative 100% structures subject to of the supports would examination would include be examined those supports within the system whose structural integrity is relied upon to withstand the design loads and seismic-induced displace-ments.
The support settings of constant and variable spring-type hangers, snubbers, and shock absorbers would be in-spected to verify proper distribution of design loads among the associated support components.
4.7 J-2 Circumferential and long- Visual 50% of the welds By the end of the in-itudinal pipe welds and would be examined spection interval a cumu-branch pipe connections lative 100% of the welds and pipe branch connections would be examined.
4.8 J-1 Socket welds and pipe Visual and 5% of the circum- By the end of the inspection branch connections welds Surface ferential welds and interval, a cumulative 25% w 4 inches in diameter and 5% of the pipe branch of the circumferential I +:-
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TABLE 4,2-1 SECTION E,
PUMP PRESSURE BOUNDARY Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During No. Category Areas --* Methods 5-Year Interval 10-Year Interval Remarks 5.1 L-1 Pump casing velds Visual and None None The only feasible method volumetric known to date to volumetrically inspect these pump weld casings is radiography. It is not known if such radio-graphy can be performed in service due to the design and the radiation level in the co~ponent.
If experience or a study indicates such radiography is possible, the inspection would be performed at the frequency specified in the Code, In any case, a visual inspection would be made at such time as the pump is opened for any purpose, 5.2 L-2 Puq, casings Visual None By the end of the inter- The only pumps involved val, a cumulative 100% in this program are the of the available inner coolant pumps.
surfaces of the required pumps would be examined if the pumps are dis~
assembled for maintenance 1-3 C/.l l,.)
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e e TABLE li,2-1 SECTION E, PUMP PRESSURE BOUNDARY Required Required Extent of Examination Tentative Inspec~
Item Examination Examination Pl~nned During First tion During No. Catego..!Y Areas Methods 5-Year Interval 10-Year Interval Remarks 5.3 F Nozzle-to-safe end welds Visual and Not applicable Not applicable There are no nozzle-volumetric to-safe.end welds on the pumps.
5,4 G-1 Pressure retaining bolting Visual and (See remarks) By the end of the inter- Bolting 2 inches and volumetric val, a cumulative 100% larg*er in diameter would of the bolting would be be examined either in examined place under tension, or when the bolting is removed or when the bolting connection is disassembled for maintenance purposes.
The bolting and areas to be examined would include the studs, nuts. bushings, threads in the base material, and the* flange ligaments between threaded stud holes.
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Required Required Extent of Examination Tentative Inspec-Examination Examination Planned During First tion During Item Remarks*
Areas Methods 5-Year Interval 10-Year Interval No. Category (See remarks) By the end of the interval, Bolting below 2 inches in 5.5 G2 Pressure retaining *Visual a cumulative 100% of the diameter would be visually bolting examined, whenever the bolted bolting would be examined connection is dissembled for maintenance purposes. The bolting to be examined would includ( :;tuds and nuts.
Integrally welded Visual and 10% of the supports By the end of the interval, None*
5.6 K-1 supports volumetric would be visually a cumulative 25% of the examined supports would be visually examined Supports and hangers Visual 33% of the supports Cumulative 100% of the No.ne 5.7 K-2 would be examined supports would be examined Visual and (See remarks)* 100% examination at or near All flywheels shall be visu-allv 5.8 Primary Pump Flywheel examined at the first refueling volumetric the end of the inspection interval and.third refueling. Every three (3) years an in-place ultrasonic volumetric examination of the areas of higher stress concentra-tion at the hore*and keywav of each flywheel shafl*be performed.
At the end of every 10 vears, all flywheels will be subject to a 100% ultrasonic inspection.
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- e TABLE 4.2-1 SECTION F. VALVE PRESSURE BOUNDARY Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During No. Category Areas Methods 5-Year Interval 10-Year Interval Remarks 6.1 M-1 Valve body welds Visual and Not applicable Not applicable All the valves subject volumetric to this inspectio::i.* are cast valves, thus this item is not applicable.
6.2 M-2 Valve bodies Visual (See remarks) By the end of the The valves subject to this in-interval, a cu.~ula- spection may be examined during tive 100% of the maintenance or at the end of each available inner interval.
surfaces of the re-quire*d valves would be.examined.
6.3 F Valve-to-safe end Visual and Not applicable Not applicable There are no valves in this 1,,,*elds volumetric system with dissimilar metal welds.
6.4 G-1 Pressure retaining Visual and (See remarks) By the end of the Bolting 2 inches and larger in bolting volumetric interval, a cumu- diameter would be exa:nined either lative 100% of the in pla_ce under tension or when the bolting 2 inches and bolting is removed or when the
larger would be bolting connection is disassembled examined for maintenance purposes. The bolting in areas to be exa~ined would include the studs, nuts, bushings, threads in the flange ligaments between threaded stud holes.
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Required Required Extent of Examination Tentative Inspec-I tea Examination Exaaination Planned During First tion During
~ Category Areas Methods 5-Year Interval 10-Year Interval Remarks 6.S G-2 Pressure retaining boltins Visual and 33% By the end of the inter- All bolting is below 2 volumetric val, a CU111ulative 100% inches in diameter and of the bolting would be would be visually examined examined, either in place if the bolting connection is not disassembled at the inspection interval, or whenever the bolting connection is disassembled.
The bolting to be examined would include studs and nuts.
6.6 . K-1 Integrally welded supports Visual and Not applicable Not applicable There are no integrally-volumetric welded supports on the valves subject to this examination.
6.7 ~-2 Supports and hangers Visual 33%.of the supports By the end of the inter- The support members and and hangers would be val, a cumulative 100% structures subject to examined of the supports and examination would include hangers would be those supports for piping, examined valves, and pumps within the system boundary, whose structural integrity is relied upon to withstand the design loads and seisl!ic-induced dis-placement*.
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TABLE 4.2d SECTION F, VALVE PRESSURE BOUNDARY (Continued)
Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During No. Category Areas Methods 5-Year Interval 10-Year Interval Remarks 6.7 (Continued) The support settings of constant and variable spring-type hangers, snubbers and shock absorbers would be in-spected to verify proper dis-tribution of design loads among the associated support components.
SECTION G. MISCELLANEOUS INSPECTIONS 7.1 Materials Tensile and Capsule 1 shall be Capsules shall Capsule 4 shall be removed Irradiation Charpy V removed and examined be removed and and examined after 20 years.
Surveillance Notch (Wedge at the first region examined after Capsules 5-8 are extra Open Loading) replacement. Capsule 10 years capsules for complimentary 2 shall be examined or duplicate testing.
after 5 years.
7.2 Low Head SIS Visual (See Remarks) Not applicable This pipe shall be visually Piping Located inspected at each refueling in Valve Pit shutdown.
7.3 Low Pressure Visual and 100% of blades Not applicable Turbine Rotor magnetic Blades particle or dye penetrant
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e TABLE 4.2-1 SECTION H, SENSITIZED STAINLESS STEEL Required Required Extent of Examination Tentative Inspec-Item Examination Examination Planned During First tion During No. Category Areas Methods 5-Year Interval 10-Year Interval Remarks 8.1.1 J-1 Circumferential Visual and Three times the extent* Three times the extent
Extent in this context is interpreted and Longitudinal Volumetric of Section D. Item of Section D, Item No. to mean that inspections will be con-pipe welds and
No. 4.2 of Table 4.2-1. 4.2 of Table 4.2-1. ducted at three times the inspection branch pipe frequency required by Items 4.2,4.7 and connections 4.8 of Table 4.2-1. Thus, those exami-larger than 4 nations planned for the first five year inches in interval for Items 4.2, 4.7 and 4.8 diameter will be performed every 1 2/3 years (generally each normal refueling outage).
8.1. 2 J-2 Circumferential Visual Three times the extent* Three times the extent See Transscript of Hearing (pp. 303-304) and longitudinal of Section D, Item No. of Section D, Item No. and Initial Decision (p. 7, § 10).
pipe welds and 4. 7 of Table 4.2-1. 4, 7 a£ Table 4-. 2-1. Same as above.
branch pipe connections.
8.1.3 J-1 Socket welds and Visual and Three times the extent* Three times the extent Same as above pipe branch Surface of Section D, Item No. of Section.D, Item No, connections welds 4.8 of Table 4.2-1. 4.8 of Table 4-,2-L 4 inches in dia-meter and smaller 8.2.1 Containment and Visual and (See Remarks) (See Remarks) At least 25 percent of the examinations Recirculation Surface shall have been completed by the expira-tion of one-third of the inspection in-terval and at least 50 percent shall have been completed by the expiration of two-thirds of the inspection interval. The remaining required examinations shall be completed by the end of the inspection interval. Surface examination will include 6 patches (each 9 inches square) evenly >-3 distributed around each spray ring. l,.)
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TS 4.3-1 3-17-72 4.3 REACTOR COOLANT SYSTEM INTEGRITY TESTING FOLLOWING OPENING Applicability Applies to test requirements for Reactor Coolant System integrity. In this context, closed is defined as that state of system integrity which permits pressurization and subsequent normal operation after the system has been opened.
Objective To specify tests for Reactor Coolant System integrity after the system is closed following normal opening, modification or repair.
Specification A. Each time the Reactor Coolant System is closed, the system will be leak tested at not less than 2335 psig in conformance with NDT requirements.
B. When Reactor Coolant System modifications or repairs have been made which involved new strength welds on piping and components greater than 2 in. diameter, the new welds will receive both a surface and 100% volumetric non-destructive examination and meet applicable code requirements.
TS 4.3-2 3-17-72
- C. When Reactor Coolant System modifications or repairs have been made which involve new strength welds on piping and components 2 in. diameter or smaller, the new welds will receive a surface examination.
D. Inspection For Leakage
1. Inspection Requirements
a. Prior to reactor startup following each refueling outage, all pressure-retaining components of the reactor coolant pressure boundary shall be visually examined for evidence of reactor coolant leakage while the system is under the test pressure specified in A above.
This examination (which need not require removal of insulation) shall be performed by inspecting (a) the exposed surfaces and joints of insulations, and (b) the floor areas (or equipment) directly underneath these components.
At locations where reactor coolant leakage is normally expected and collected (e.g., valve stems, etc.),the examination shall verify that the leakage collection system is operative and leak-tight.
b. During the conduct of the examinations of (a) above, particular attention shall be given to the insulated areas of components constructed of ferritic steels to detect evidence of boric acid residues resulting from reactor coolant leakage which might have accumulated during the service period preceding the refueling outage.
TS 4.3-3 3-17-72
c. The visual examinations of (a) and (b) above shall be conducted in conformance with the procedures of Article IS-211 of Section XI of the ASME Boiler and Pressure Vessel Code.*
2. Corrective Measures
a. The source of any reactor coolant leakage detected by the examina-tions of D.l.a above shall be located by the removal of insulation where necessary and the following corrective measures applied:
(1) Normally expected leakage from component parts (e.g., valve stems) shall be minimized by appropriate repairs and mainten-ance procedures. Where such leakage may reach the surface of ferritic components of the reactor coolant pressure boundary, the leakage shall be suitably channeled for collection and disposal.
(2) Leakage from through-wall flaws in the pressure-retaining membrane of a component shall be eliminated, either by corrective repairs or by component replacement. Such repairs shall con-form with the requirements of Article IS-400 of Section XI of the ASME Boiler and Pressure Vessel Code.*
b. In the event boric acid residues are detected by the examinations of D.l.b above, insulation from ferritic steel components shall be removed to the extent necessary for examination of the component surfaces wetted by reactor coolant leakage to detect evidence of corrosion.
The following corrective measures shall be applied:
(1) An evaluation of the effect of any corroded area upon the
TS 4.3-4 3-17-72
- structural integrity of the component shall be performed in accordance with the provisions of Article IS-311 of Section XI Code.*
(2) Repairs of corroded areas, if necessary, shall be performed in accordance with the procedures of Article *IS-400 of Section XI Code.*
Basis For normal opening the integrity of the system, in terms of strength, is un-changed. If the system does not leak at 2335 psig (operating pressure plus 100 psi), it will be leaktight during normal operation.
For repairs on piping and components greater than 2 in. diameter, the thorough nondestructive testing gives a very high degree of confidence in the integrity of the system, and will detect any significant defects in and near the new welds.
Significance of repairs on piping and components 2 in. diameter or smaller are relatively minor in comparison. The surface examination assures an adequate standard of integrity. In all cases, the leak test will ensure leaktightness during normal operation.
Experience has shown that corrosion potential might exist under certain conditions when borated fluid has prolonged contact with carbon steel. Undetected leakage of borated reactor coolant may cause such conditions to exist. To detect such leakage or its effects at an early stage the inspection program described in Specification D provides a means of detecting reactor coolant leakage and/or its effects at an early stage.
1970 Edition with Winter 1970 Addenda
TS 4.4-1 3-17-72 4.4 CONTAINMENT TESTS Applicability Applies to containment leakage testing.
To assure that leakage of the primary reactor containment and associated systems is held within allowable leakage rate limits; and to assure that periodic surveillance is performed to assure proper maintenance and leak repair during the service life of the containment.
Specification A. Periodic and post-operational integrated leakage rate tests of the containment shall be performed in accordance with the requirements of proposed 10 CFR 50, Appendix J, "Reactor Containment Leakage Testing For Water Cooled Power Reactors," as published in the Federal Register, Volume 36, No. 167, August 27, 1971.
B. Testing Requirements Type A tests will be performed in accordance with the reduced pressure test program as defined in paragraph III, A.1.(a) of Appendix J.
a. The reference volume method of leakage rate testing will be used as the method for performing the test, The absolute method of leakage rate testing will be used for verification. Test will be conducted I in accordance with the provisions of ANS 7.60.
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TS 4 .. 4-2 3-17-72
b. An initial leakage rate test will be performed at a pressure of 25 psig (Pt) and a second test at 39,2 psig (Pp),
c, The measured leakage rate Lpm shall not exceed the design basis accident leakage rate (La) of 0.1 weight percent per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at pressure Pp'
d. The maximum allowable test leakage rate Lt will be computed in accordance with paragraph III. A.4.
(a)(iii) of Appendix J,
2. Type Band C tests will be performed at a pressure of 39.2 psig (Pp) in accordance with the provisions of Appendix J, section III. B. and C.
c. Acceptance Criteria Type A, Band C tests will be considered to be satisfactory if the acceptance criteria delineated in Appendix J, Sections III.A.5, III, A.7., III.B,3., and III.C.3 are met.
D. Retest Schedule The retest schedules for Type A, B, and C tests will be in accordance with Section III-D of Appendix J, E. Inspection and Reporting of Tests Inspection and reporting of tests will be in accordance with Section V of Appendix J.
TS 4.4-3 3-17-72 Basis The leaktightness testing of all liner welds was performed during construction by welding a structural steel test channel over each weld seam and performing soap bubble and halogen leak tests.
The containment is designed for an accident pressure of 45 psig. The containment is maintained at a subatmospheric air partial pressure which varies between 9 psia and 11 psia depending upon the cooldown capability of the Engineered Safeguards and is not expected to rise above 42.5 psig for any postulated loss-of-coolant accident.
All loss-of-coolant accident evaluations have been based on an integrated containment leakage rate not to exceed 0.1 percent of containment volume per 24 hr, The above specification satisfies the conditions of 10 CFR 50.54(0) which states that primary reactor containments shall meet the containment leakage test re-quirements set forth in Appendix J.
References FSAR Section 5.4 Design Evaluation of Containment Tests and Inspections of Containment FSAR Section 7.5.1 Design Bases of Engineered Safeguards Instrumentation FSAR Section 14.5 Loss-of-Coolant Accident 10 CFR 50 Appendix J (Proposed), "Reactor Containment Leakage Testing for Water Cooled Power Reactors," as published* in the Federal Register, Volume 36, No. 167, August 27,1971.
TS 4.5-1 3-17-72 4.5 SPRAY SYSTEMS TESTS Applicability Applies to the testing of the Spray Systems.
Objective To verify that the Spray Systems will respond promptly and perform their design function, if required.
Specification A. Test and Frequencies
1. The containment spray pumps shall be flow tested at a reduced flow rate at least once per month.
2. All inside containment recirculation spray pumps shall be dry tested at least once per month.
3. The recirculation spray pumps outside the containment shall be flow tested at a reduced flow rate at least once per month.
TS 4.5-2 3-17-72
4. The weight loaded check valves within the containment in the vari~us subsystems shall be tested by pressurizing the pump dis-charge lines with air at least once each refueling period. Verifi-cation of seating the check valves shall be accomplished by applying a vacuum upstream of the valves.
5. All motor operated valves in the containment spray and recircula-tion spray flow path shall be tested by stroking them at least once per month.
6. The containment spray nozzles and containment recirculation spray nozzles shall be checked for proper functioning at least every five years.
7. The spray nozzles in the refueling water storage shall be checked for proper functioning at least monthly.
B. Acceptable Criteria
1. A dry-test of a recirculation spray pump shall be considered satis-factory if the motor and pump shaft rotates, starts on signal, and the ammeter readings for the motor are comparable to the original dry test ammeter readings.
2. A flow-test of a containment spray pump or an outside recirculation spray pump shall be considered satisfactory if the pump starts, and the discharge pressure and flow rate determine a point on the head curve. A check will be made to determine that no particulate
TS 4.5-3 3-17-72 material from the.refueling water storage tank clogs the test spray nozzles located in the refueling water storage tank.
3. The test of each of the weight loaded check valves shall be considered satisfactory if air flows through the check valve, and if sealing is achieved.
4. A test of a motor operated valve shall be considered satis-factory if its limit switch operates a light on the main control board demonstrating that the valve has stroked.
e 5. The test of the containment spray nozzles shall be considered satisfactory if the measured air flow through the nozzles indicates that the nozzles are not plugged.
6. The test of the spray nozzles in the refueling water storage tank shall be considered satisfactory if the monitored flow rate to the nozzles, when compared to the previously established flow rate obtained with the new nozzles, indicates no appreciable reduction in flow rate.
Basis The flow testing of each containment spray pump is performed by opening the normally closed valve in the containment spray pump recirculation line re-turning water to the refueling water storage tank. The containment spray
TS 4.5-4 3-17-72 pump is operated and a quantity of water recirculated to the refueling water storage tank. The discharge to the tank is divided into two fractions, one for the major portion of the recirculation flow and the other to pass a small quantity of water through test nozzles which are identical with those used in the containment spray headers. The purpose of the recirculation through the test nozzles is to assure that there is no particulate material in the refueling water storage tank small enough to pass through pump suction strainers and large enough to clog spray nozzles.
Due to the physical arrangement of the recirculation spray pumps inside the containment, it is impractical to flow-test them periodically. These pumps are capable of being operated dry for 60 sec and it can be determined that e the pump shafts are turning by rotation sensors which indicate in the Main Control Room. Motor current is indicated on an ammeter in the Control Room, and will be compared with readings recorded during preoperational tests to ascertain that no degradation of pump operation has occurred. The recircula-tion spray pumps outside the containment have the capability of being dry-run and flow-tested. The flow-test of an outside recirculation spray pump is per-formed by closing the suction line valve and the isolation valve between the pump discharge and the containment penetration. This allows the pump casing to be filled with water and the pump to recirculate water through a test line from the pump discharge to the pump casing.
With system flush conducted to remove particulate matter prior to the installa-tion of spray nozzles and with corrosion resistant nozzles and piping, it is not considered credible that a significant number of nozzles would plug during the life of the unit to reduce the effectiveness of the subsystems; therefore,
TS 4.5-5 3-17-72 provisions to air test the nozzles every five years is sufficient to indicate that plugging of the nozzles has not occurred.
The spray nozzles in the refueling water storage tank provide means to ensure that there is no particulate matter in the refueling water storage tank and the Containment Spray Subsystems which could plug or cause deterioration of the spray nozzles. The nozzles in the tank are identical to those used on the containment spray headers.
The monthly flow test of the containment spray pumps and recirculation to the refueling water storage will indicate any plugging of the nozzles by a re-duction of flow through the nozzles.
References FSAR Section 6.3.1 Containment Spray Pumps FSAR Section 6.3.1 Recirculation Spray Pumps
TS 4.6-1 3-17-72 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTING Applicability Applies to periodic testing and surveillance requirements of the Emergency Power System.
Objective To verify that the Emergency Power System will respond promptly and properly when required.
Specification The following tests and surveillance shall be performed as stated:
A. Diesel Generators
1. Tests and Frequencies
a. Manually initiated start of the diesel generator, followed by manual synchronization with other power sources and assumption of load by the diesel generator up to 27 50 dfM-. Thia test. will be.conducted monthly on each diesel generator for a duration of 30 minutes. Normal station operation will not be affected by this test.
TS 4. 6-2 3-17-72
b. Automatic start of each diesel generator, load shedding, and restoration to operation of particular vital equipment, initiated by a simulated loss of off-site power together with a simulated safety injection signal. This test will be conducted at approximately one year intervals normally during reactor shutdown for refueling to assure that the diesel generator will start within 10 sec and assume load in less than 30 sec after the engine starting signal.
c. Availability of the fuel oil transfer system shall be verified by operating the system in conjunction with the monthly test.
d. Each diesel generator shall be given a thorough inspection during each refueling interval utilizin*g the manufacturer's recommendations for this class of stand-by service.
2. Acceptance Criteria The above tests will be considered satisfactory if all applicable equipment operates as designed.
B. Fuel Oil Storage Tanks for Diesel Generators
1. A minimum fuel oil storage of 35,000 gal shall be maintained on-site to assure full power operation of one diesel generator for seven days.
TS 4.6-3 3-17-72 C. Station Batteries
1. Tests and Frequencies
a. The specific gravity,electrolytic temperature,cell voltage of the pilot cell in each 60 cell battery, and the D.C. bus voltage of each battery shall be measured and recorded weekly.
b. Each month the voltage of each battery cell in each 60 cell battery shall be measured to the nearest 0.01 volts and recorded.
c. Every 3 months the specific gravity of each battery cell, the temperature reading of every fifth cell, the height of electrolyte of each cell, and the amount of water added to any cell shall be measured and recorded.
d. Twice a year, during normal operation, the battery charger shall be turned off for approximately 5 min and the battery voltage and current shall be recorded at the beginning and end of the test.
e. Once a year during the normal refueling shutdown each battery shall be subjected to a simulated load test without battery charger. The battery voltage and current as a function of time shall be monitored.
TS 4.6-4 3-17-72 f, Annually connections shall be checked for tightness and anti-corrosion coating shall be applied to interconnections.
2. Acceptance Criteria
a. Each. test shall be considered satisfactory if the new data when compared to the old data indicate no signs of abuse or deterioration.
b. The load test in (d) and (e) above shall be considered satis-factory if the batteries perform within acceptable limits as established by the manufacturers discharge characteristic curves.
Basis The tests specified are designed to demonstrate that the diesel generators will provide power for operation of essential safeguards equipment. They also assure that the emergency diesel generator system controls and the control systems for the safeguards equipment will function automatically in the event of a loss of all normal station service power.
The testing frequency specified will be often enough to identify and correct any mechanical or electrical deficiency before it can result in a system failure. The fuel supply and starting circuits and controls are continuously monitored and any faults are alarm indicated. An abnormal condition in these systems would be 'signaled without having to place the diesel generators them-selves on test.
TS 4.6-5 3-17-72 Station batteries may deteriorate with time, but precipitous failure is extremely unlikely, The surveillance specified is that which has-been demon-strated by experience to provide an indication of a cell becoming unserviceable long before it fails. In addition alarms have been provided to indicate low battery voltage and low current from the inverters which would make. it extremely unlikely that deterioration would go unnoticed.
The equalizing charge, as recommended by the manufacturer, is vital to main-taining the ampere-hour capability of the battery. As a check upon the effectiveness of the equalizing charge, the battery shall be loaded rather heavily and the voltage monitored as a function of time. If a cell has deteriorated or if a connection is loose, the voltage under load will drop excessively indicating the need for replacement or maintenance.
FSAR Section 8.5 provides further amplification of the basis.
References FSAR Section 8.5 Emergency Power System
TS 4.7-1 3-17-72 4.7 MAIN STEAM LINE TRIP VALVES Applicability Applies to periodic testing of the main steam line trip valves.
Objective To verify the ability of the main steam line trip valves to close upon signal.
Specification A. Tests and Frequencies
1. Each main steam trip valve shall be tested for full closure under cold conditions approximately once a year during each refueling shutdown.
2. Each main steam trip valve shall be in-service tested for partial closure at least once a month.
B. Acceptance Criteria
1. A full closure test of main steam trip valves shall be considered satisfactory if the valve closes fully in 5 sec or less.
TS 4.7-2 3-17-72
2. A partial closure in-service test of a main steam trip valve shall be considered satisfactory if the valve can be stroked at least 5 degrees from its full open position.
Basis The main steam trip valves serve to limit an excessive Reactor Coolant System cooldown rate and resultant reactivity insertion following a main steam line break accident. Their ability to close fully shall be verified at each scheduled refueling shutdown. A closure time of 5 sec was selected since this is the closure time assrnned in the safety evaluation. The in-service testing of partial valve stroke will take place to verify the freedom of the valve disc to function as required. A limit switch in the test circuit prevents e the valve disc from entering the flow stream and slamming the valve shut during in-service testing.
TS 4.8-1 3-17-72 4.8 AUXILIARY FEEDWATER SYSTEM Applicability Applies to periodic testing requirements of the Auxiliary Feedwater System.
Objective To verify the operability of the auxiliary steam generator feedwater pumps and their ability to respond properly when required.
Specification e A. Tests and Frequency
1. Each motor driven auxiliary steam generator feedwater pump shall be flow tested for at least 15 minutes on a monthly bas1s to demonstrate its operability.
2. The turbine driven auxiliary steam generator feedwater pump shall be flow tested for at least 15 minutes on a monthly basis to demonstrate its operability.
3. The auxiliary steam generator feedwater pump discharge valves shall be exercised on a monthly basis.
TS 4.8-2 3-17~72 B. Acceptance Criteria These tests shall be considered satisfactory if control board indication and subsequent visual observation of the equipment demonstrate that all components have operated and sequenced properly.
Basis On a monthly basis the auxiliary steam generator feedwater pumps will be tested to demonstrate their operability by recirculation to the 110,000 Gallon Condensate Storage Tank.
The capacity of any one of the three feedwater pumps in conjunction with the water inventory of the steam generators is capable of maintaining the plant in a safe condition and sufficient to cool the unit down.
Proper functioning of the steam turbine admission valve and the ability of the feedwater pumps to start will demonstrate the integrity of the system.
Verification of correct operation can be made both from instrumentation within the Main Control Room and direct visual observation of the pumps.
References FSAR Section 10.3.1 Main Steam System FSAR Section 10.3.2 Auxiliary Steam System
TS 4.9-1 3-17-72 4.9 EFFLUENT SAMPLING AND RADIATION MONITORING SYSTEM Applicability Applies to the periodic monitoring and recording of radioactive effluents.
Objective To ascertain that radioactive releases are maintained as low as practicable and within the limits set forth in 10CFR20.
Specification
- A. Procedures shall be developed and used, and equipment which has been installed to maintain control over radioactive materials in gaseous and liquid effluents produced during normal reactor operations, including expected operational occurrences, shall be maintained and used to keep levels of radioactive materials in effluents released to unrestricted areas as low as practicable.
B. All effluents to be discharged to the atmosphere from the waste gas decay tanks of the Gaseous Waste Disposal System shall be sampled prior to release via the process vent. Effluent from the Liquid Waste Disposal System shall be continuously monitored by the circulating water discharge tunnel monitor and, periodically, be sampled upstream of the point where e it is discharged into the circulating water discharge tunnel.
TS 4.9-2 3-17-72 C. The grqss activities of all gaseous and airborne particulate effluents released from the Gaseous Waste Disposal System and the Ventilation Vent System and the gross activity of all liquid effluent released from the Liquid Waste Disposal System and steam generator blowdown shall be measured and recorded continuously while they are being discharged.
D. All radiation monitor channels shall be checked, calibrated and tested as indicated in Table 4.1-1.
E. The environmental program given in Table 4.9-1 shall be conducted.
Basis The test and calibration requirements are specified to detect possible equip-ment failures and to show that maximum permissible release rates are not exceeded. All the radiation monitors except the recirculation spray cooler service water outlet monitors operate continuously and the operator observes that these instruments are performing daily. In addition, the check source for each operating channel is tripped daily from the Main Control Room to verify instrument response. All the monitors for a particular unit will be calibrated on a periodic basis, and normally during the refueling shutdown of that unit. Experience with instrument drift and failure modes indicates that the above specified test and calibration frequencies are adequate.
The environmental survey incorporates measurements to provide background data and measure possible plant effects. Samples collected at points where concentrations of effluents in the environment are expected to be the greatest
TS 4.9-3 3-17-72 will be compared with samples collected concurrently at points expected to be essentially uneffected by station effluents. The latter samples will provide background measurements as a basis for distinguishing significant radioactivity introduced into the environment by the operation of the station from that due to nuclear detonations and other sources.
This schedule will ensure that changes in the environmental radioactivity can be detected. The materials which first show changes in radioactivity are sampled most frequently. Those which are less affected by transient changes but show long-tenn accumulations are sampled less frequently.
Data on the composition, quantity, frequency, etc. of releases, dilution factors obtained, and measured concentrations in food and other organisms (if any are observed) should make it desirable to review and re-evaluate this program periodically.
e TABLE 4.9-1 ENVIRONMENTAL MONITORING PROGRAM Sample Type Number of Sampling Points Frequency Type of Analysis
1. WATERS A. James River 2 Semi-Annual Gamma Isotopic & Tritium on composit of Bimonthly Samples, Upstream & Downstream of Station B. Wells 4 Semi-Annual Gross Alpha, Gross Beta, Tritium C. Surface Water 4 Semi-Annual Gross Alpha, Gross Beta, Tritium D. Precipitation 2 Semi-Annual Gross Beta & Tritium on composit of Monthly Samples.
11. AIR A. Particulate 6 Monthly Gross Alpha, Gross Beta B. Radiogas 6 Quarterly mrem Exposure 111. BIOTA A. Crops* Annual Gamma Isotopic B. Fowl Annual Gamma Isotopic
-I C. Oyster w ti) 3 Quarterly Gamma Isotopic I f-'
-..J I
-I:-
D. Clam 3 Quarterly Gamma Isotopic -..J I.O I
N -I:-
E. Crab Annual Gamma Isotopic F. Fi sh*t, Semi-Annual Gamma Isotopic
Three different crops are sampled - corn, peanuts and soybeans.
~~Two types of fish are sampled - carnivore arid bottom feeder.
e TABLE 4.9-1 .CONT'D.
ENVIRONMENTAL MONITORING PROGRAM Sample Type Number of Sampling Points Frequency Type of Analysis IV. SILT 5 Semi-Annual Gamma Isotopic V. SOIL 6 Annual Gamma Iso topic VI. MILK 4 Quarterly 1-131 , Cs-137, Sr-90, Calcium
-I w en I
I-' ~
'-I
I \.0
'-I I N \J"1
TS 4.10-1 3-17-72 4.10 REACTIVITY ANOMALIES Applicability Applies to potential reactivity anomalies.
Objective To require evaluation of applicable reactivity anomalies within the reactor.
Specification A. Following a normalization of the computed boron concentration as a function of burnup, the actual boron concentration of the coolant shall be compared monthly with the predicted value. If the difference between the observed and predicted steady-state concentrations reaches the equivalent of one.
percent in reactivity, an evaluation as to the cause of the discrepancy shall be made and reported to the Atomic Energy Commission per Section 6.6 of these Specifications.
B. During periods of power operation at greater than 10% of power, design N N peaking factors, F and FllH' shall be determined monthly using data from q
limited core maps. If these factors exceed values of Design Limits *Interim Limits FN = 2.72 N ;;: 1.58 FN = 2.52 FN = 1.50 q FllH q llH an evaluation as to the cause of the anomaly shall be made.
TS 4.10-2 3-17-72
- Basis BORON CONCENTRATION To eliminate possible errors in the calculations of the initial reactivity of the core and the reactivity depletion rate, the predicted relation between fuel burnup and the boron concentration necessary to maintain adequate control characteristics, must be adjusted (normalized) to accurately reflect actual core conditions. When full power is reached initially, and with the control rod assembly groups in the desired positions, the boron concentration is measured and the predicted curve is adjusted to this point. As power operation proceeds, the measured boron concentration is compared with the predicted concentration, and the slope of the curve relating burnup and reactivity is compared with that predicted. This process of normalization should be completed after about 10% of the total core burnup. Thereafter, actual boron concentration can be compared with prediction, and the reactivity status of the core can be continuously evaluated. Any reactivity anomaly greater than 1% would be un-expected, and its occurrence would be thoroughly investigated and evaluated.
The value of 1% is considered a safe limit since a shutdown margin of at least 1% with the most reactive control rod assembly in the fully withdrawn position is always maintained.
PEAKING FACTORS A thermal criterion in the reactor core design specifies that "no fuel melting during any anticipated normal operating condition" should occur. To meet the above criterion during a thermal overpower of 112% with additional margin for design uncertainties, a steady state maximum linear power is
TS 4.10-3 3-17-72
- selected. This then is an upper linear power limit determined by the maximum central temperature of the hot pellet.
The peaking factor is a ratio taken between the maximum allowed linear power density in the reactor to the average value over the whole reactor. It is of course the average value that determines the operating power level. The peaking factor is a constraint which must be met to assure that the peak linear power density "does not exceed the maximum allowed value.
During normal reactor operation, measured peaking factors should be significantly lower than design limits. As core burnup progresses, measured designed peaking factors are expected to decrease. A monthly determination of N N e F and FA q oH is adequate to *ensure that core reactivity changes with burnup have not significantly altered peaking factors in an adverse.direction.
TS 4.11-1 3-17-72 4.11 SAFETY INJECTION SYSTEM TESTS Applicability Applies to operational testing of the Safety Injection System.
Objective*
To verify that the Safety Injection System will respond promptly and perform its design functions, if required.
e Specification A. Safety Injection System
1. System tests shall be performed during reactor shutdowns for refueling. The test shall be performed in accordance with
_the following procedure:
With the Reactor Coolant System pressure less than or equal to 450 psig and temperature less than or equal to 350°F, a test safety injection signal will be applied to initiate operation of the system. The charging and low head safety injection pumps are immobilized for this test.
TS 4.11-2 3-17-72
2. The test will be considered satisfactory if control board indication and/or visual observations indicate that all the appropriate components have received the safety injection signal in the proper sequence. That is, the appropriate pump breakers shall have opened and closed, and all valves, required to establish a safety injection flow path to. the Reactor Coolant System and to isolate other systems from this flow path, shall have completed their stroke.
B. Component Tests Pumps e 1. The low head safety injection pumps and charging pumps shall be operated at intervals not greater than one month,
2. Acceptable levels of performance for the low head safety injection pumps shall be that the pumps start, reach their required developed head on recirculation flow and the control board indications and/or visual observations indicate that the pumps are operating properly.
3. In addition to the Safety Injection System, the charging pumps form an integral part of the Chemical and Volume Control System (CVCS),
and are operated on a routine basis as part of this system. If these pumps have performed their design function as part of the routine operation of the eves, their level of performance will be deemed acceptable as related to the Specification.
TS 4.11-3 3-17-72 Valves
1. The refueling water storage tank outlet valves shall be tested in performing the pump tests.
2. The accumulator check valves shall be checked for operability during each refueling shutdown.
3. All valves required to operate on a safety injection signal shall be tested for operability each refueling shutdown.
Basis
- Complete system tests cannot be performed when the reactor is operating be-cause a safety injection signal causes containment isolation. The method of assuring operability of these systems is therefore to combine systems tests to be performed during refueling shutdowns, with more frequent component tests, which can be performed during reactor operation.
The systems tests demonstrate proper automatic operation of the Safety Injection System. With the pumps blocked from starting, a test signal is applied to initiate automatic action and verification is made that the components receive the safety injection signal in the proper sequence. The test demonstrates the operation of the valves, pump circuit breakers, and automatic circuitry.
During reactor operation, the instrumentation which is depended on to initiate safety injection is checked periodically and the initiating circuits are tested in accordance with Specification 4.1. In addition, the active components (pump
TS 4.11-4 3-17-72 and valves) are to be tested monthly to check the operation of the starting circuits and to verify that the pumps are in satisfactory running order. The test interval of one month is based on the judgment that more frequent testing would not significantly increase the reliability (i.e. the probability that the component would operate when required), yet more frequent testing would result in increased wear over a long period of time, Other systems that are also important to the emergency cooling function are the accumulators and the Containment Depressurization System. The accumulators are a passive safeguard. In accordance with Specification 4.1 the water volume and pressure in the accumulators are checked periodically.
Reference FSAR Section 6.2 Safety Injection System
TS 4.12-1 3-17-72 4.12 VENTILATION FILTER TESTS Applicability Applies to the testing of particulate and charcoal filters in safety related air filtration systems.
Objective To verify that leakage efficiency and iodine removal efficiency are within acceptable limits.
Specification A. Tests and Frequencies
1. The charcoal filters in the'Auxiliary Building filter banks, control room emergency filter banks, and relay room emergency filter banks shall be tested for leakage efficiency at least once every 12-18 month period, normally during refueling shutdown using an in-place Freon-112 (or equivalent) test method.
2. The particulate filters in the Auxiliary Building filter, control room emergency filter banks, and relay room emergency filter banks shall be tested for leakage efficiency at least once every 12-18 month period, normally during refueling shutdown using an in-place DOP test method.
3. A carbon sample will be removed from one of the banks once every third year and subjected to chemical analysis to determine the iodine removal capability.
4. Instrumentation, equipment, and procedures shall generally conform to
TS 4.12-2 3-17-72 the reconnnendations in ORNL-NSIC-65, "Design, Construction and Testing of High-Efficiency Air Filtration Systems for Nuclear Application",
C. A. Burchsted and A. B. Fuller, Oak Ridge National Laboratory, USAEC, January, 1970.
B. Acceptance Criteria
1. The in-place leakage tests on charcoal units using Freon-112 (or equivalent) shall be conducted on each filter bank. Removal of 99.0%
of the Freon-112 (or equivalent) shall constitute acceptable performance.
2. The in-place leakage tests on the particulate filters using DOP shall be conducted on each train. Removal of 99.5% of the DOP shall constitute acceptable performance.
3. Chemical analysis of the carbon sample shall be performed to demonstrate the iodine adsorption capability of the charcoal. Verification of an elemental iodine removal efficiency of 99.0% shall constitute acceptable performance.
Basis The purpose of the Auxiliary Building Filter Banks is to provide standby capa-bility for removal of particulate and iodine contaminants from any of the venti-lation systems in the auxiliary building, fuel building, decontamination building, safeguards area adjacent to the containments, and the reactor con-tainments (during shutdown) which discharge through the ventilation vent and could require filtering prior to release. The exhausts of the above systems can be diverted, if required, through the Auxiliary Building filter banks remotely from the control room. The Safeguards Area exhaust is automatically diverted through the filter banks in the event of a LOCA (diverted on high-high contain-
TS 4.12-3 3-17-72 ment pressure). The fuel building exhaust is aligned to continuously exhaust through the filters during spent fuel handling in the spent fuel pool.
The purpose of the control and relay room emergency filter banks is to provide emergency ventilation for the control and relay rooms during accident conditions.
The off-site dose calculations for LOCA and fuel handling accidents, assume only 90% iodine removal efficiency for the air passing through the charcoal filters. Therefore, the demonstration of 99% efficiency will assure the required capability of the filters is met or exceeded.
System components are not subject to rapid deterioration, having lifetimes of many years, even under continuous flow conditions. The tests outlined above provide assurance of filter reliability and will insure early detection of conditions which could cause filter degradation.
Instrumentation, equipment, and procedures for testing shall generally conform to the recommendations in ORNL-NSIC-65, which is considered to be the best available guide for design, construction, and testing of particulate and iodine removal filter systems.
Reference FSAR Section 9.13, Auxiliary Ventilation Systems
TS 4 .13-1 3-17-72
-* 4.13 NONRADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Applicability The nonradiological environmental monitoring program applies to the monitoring of the temperature-salinity distribution and the biological variables in the 10-mile segment of the James River Estuary centered at Hog Island.
Objective The objective of the program is to determine (1) the relationship between the thermal discharge and the physical-chemical characteristics of the water mass within the 10-mile tidal segment of the James River; (2) the planktonic, nektonic, and benthic characteristics of this segment; and (3) the effects of the operation of the Surry Power Station on the physical, chemical, and and biological variables of the James River Estuary.
0 Specification A. A monitoring program shall be conducted to determine the relationship between the thermal discharge and the physical-chemical characteristics of the water mass within the 10-mile tidal segment centered at Hog Island.
1. The monitoring program shall encompass the segment of the James River Estuary which extends from below the intake of the Surry Power Station upstream to the southern shore of Jamestown Island as shown on TS Figure 4.13-1.
2. Temperature shall be continuously monitored and recorded at instrument towers located throughout the estuarine segment approximately as shown in TS Figure 4.13-1. Near-surface
- and/or bottom temperatures shall be monitored as indicated on TS Figure 4.13-1.
TS 4.13-2 3-17-72
3. The horizontal and vertical salinity structure of the tidal segment shall be determined at monthly intervals as follows:
Cruises shall be conducted at slack before flood tide. A four (4) station transect shall be established between the intake structure and Skiffes Creek just before low slack water. Temperature and salinity data shall be collected at two (2) meter intervals from surface to bottom. The cruise shall continue up the channel with same slack stations established at approximately two (2) mile intervals. The second transect shall be made near the upper limits of the segment, the exact locati.on of which shall be based upon the salinity regime of the system. The approximate:locations of the sampling stations and the. cruise route are. shown on TS Figure 4 .13-1.
4. Mid-depth temperature and salinity shall be continuously, monitored and recorded at the intake of the Surry Power Station.
5. Mid-depth temperature shall be continuously monitored and recorded in the discharge canal.
6. The freshwater discharge at the Richmond gaging stations shall be recorded as a daily average and the corresponding input for Hog Island calculated.
7. If chlorine is used in the condenser-cooling system, chlorin~
demand in the intake canal shall be monitored. The chlori-nation schedule shall be based on plant operating data rather than on a fixed time basis. During chlorination, chlorine shall be monitored at the end of the discharge .groin by use
TS 4.13-3 3-17-72 of analytical methods that are sensitive to chlorine at concentrations much less than the concentration required for control and that will differentiate among the various chlorine containing compounds which constitute the residual chlorine. The concentration of residual chlorine at the point of discharge to the James River shall not be greater than 0.1 mg/liter.
B. A biological monitoring program that ls closely related to the physical and chemical monitoring programs shall be conducted to determine the planktonic, nektonic, and benthic characteristics of the tidal segment centered at Hog Island and to determine biological changes that occur as a result of the operation of the Surry Power Station.
1. Plankton- Water samples for plankton analyses shall be collected at each of six (6) stations as indicated in TS Figure 4.13-2.
Samples shall also be collected in the intake and discharge canals.
a. Phytoplankton samples shall be analyzed quantitatively in terms of sample volume to determine both the dominant genera of the community and the chlorophyll "a" content.
The samples shall be taken at monthly intervals.
b. Zooplankton samples shall be analyzed quantatively in terms of sample volume to determine generic composition, life history stage and, where possible, species. The sampling interval shall be approximately monthly, taking into consideration life-history information about important species in the area which have planktonic stages in their life histories.
TS 4.13-4 3-17-72
2. Attached Benthic Connnunity - Fouling plates that are 125- by 75-nnn asbestos boards shall be suspended 1 m above the bottom at the instrument tower locations shown in TS Figure 4.13-2.
Two vertical and two horizontal plates shall be suspended at each indicated location. One of each pair shall be removed and replaced at quarterly intervals; the other pair shall be left in place for one year before being removed and replaced.
The benthic connnuni.ties attached to the plates shall be analyzed for species composition and diversity.
3. Epibenthos - Replicate benthic grab samples shall be collected at the stations shown in TS Figure 4.13-2. Collection shall be made on a quarterly basis, except during June, July, and August when they shall be made monthly. Population character~
istics such as species composition, diversity, evenness, redundancy, and richness shall be determined. The data shall be analyzed to detect changes in specific components of the epibenthic connnunity including the brackish water clam Rangia cuneata and blue crab Callinectes sapidus.
4. Nekton - The nektonic species of organisms present in th.e tidal segment shown in TS Figure 4.13-2 shall be qualitatively and quantitatively sampled by seining or trawling at the indicated locations in the estuarine segment. Samples shall be collected at monthly intervals except that during periods of active or passive migrations, sampling at three selected stations shall be often enough to establish. relative popu-lation levels. These sampling intervals shall be based on
TS 4.13-5 3-17-72 life history and distribution information that indicates when species of special interest are likely to be in or passing through this segment of the estuary. These species of special interest shall include Anchovy, Atlantic Menhaden, Blueback Herring, Channel Catfish, Atlantic Croaker, Spot, Striped Bass, and White Perch. The samples shall be analyzed for species composition, size, and life history stages.
C. The programs described in Specifications A and B shall commence on the day Unit No. 1 is licensed to operate. Where installation of towers and/or purchase of equipment is necessary and/or involves authorization by other agencies, the affected portion of the program shall be imple-mented at the earliest practicable time, but not later than six (6) months after the license is issued.
D. The dat~ obtained from the programs defined in Specifications A and B shall be analyzed as they are collected and shall be compared with model and analytical predictions and with preoperational data. A report of the results of this evaluation shall be forwarded to the Directorate of Licensing (DL) at the end of each six month period or fraction thereof terminating on June 30 and December 31. Such reports are due within 60 days after the end of each reporting period and shall be submitted with the Routine Operating Report described in Technical Specification 6.6.
A final report summarizing the results of the program shall be submitted sixty (60) days following the third anniversary of the date Unit No. 2 is licensed to operate. If on the basis of such semiannual and final reports it is established that no major adverse environmental impact e
TS 4.13-6 3-17-72 has resulted or is likely to result from continued operation of Unit.
Nos. 1 and 2 then the program shall be terminated. Otherwise it shall continue until a semiannual report does establish that no impact has resulted or is likely to result. If on the basis of any semiannual report or the final report*it is established that the results of the monito.ring program are inconclusive, either whole or in part, the licensee shall propose reasonable changes to the program designed to yield conclusive results and implement such changes when they are approved by DL.
E. Fish killed on the traveling screens at the station or by operating effects of the Surry Power Station shall be identified by species, size, and quantity, and the data shall be recorded in tabular form. These data shall be transmitted to DL semiannually. Significant mortalities of fish that may be relat;.ed to operation of the station shall be reported to Region II, Directorate of Regulatory Operations within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Data concerning significant fish mortalities and the probable cause shall be included in a more detailed report to DL within 10 days.
Basis Excess temperature distributions and alteration of density flows in the tidal segment have been predicted from data developed from model studies for two-unit operation. Surface isotherms for wind conditions of 5 MPH have been plotted for differen~ stages in the tidal cycle. The*data collected under
TS 4.13-7 3-17-72 e Specification A will permit an evaluation o.f the predictions and provide the basis for describing the parameters which may have environmental significance. The surface and bottom records combined with profile data will also permit cross-sectional as well as longitudinal physical-chemical evaluations to be made of the tidal segment.
The tidal segment encompassing the Surry site is in the vicinity of the mean transition zone between fresh and saltwater. During periods when the freshwater inflow~ as measured at the head of the coastal physio-graphic province exceeds approximately 12,000 cfs for an extended period, the water in the reach is fresh. At lower flows, the water becomes brackish and during extreme drought conditions the salinity on the
discharge side of the point may reach 11 ppt. Since it is not feasible to take direct measurements of fresh water inflow, calculational methods will be used t_o predict the flow from data which is available at gaging stations.
The condenser is cleaned by a mechanical system and it is expected that it will not be necessary to use chlorine to maintain condenser clean-liness. In the remote event it becomes necessary to utilize chlorine, its use will be regulated by need as demonstrated by a change in operating parameters. Residual chlorine will be monitored at the point of discharge to the James River and shall not exceed 0.1 mg/liter. This concentration should have no effect on river organisms.
The post-operational non-radiological monitoring program is designed to evaluate biological populations in which the species and number of
TS 4-.13-8 3-17-72 individuals present at a given time is influenced by:
1. Seasonal and natural distribution patterns
2. Salinity as influenced by freshwater inflow
3. Ecological "salinity gradient zone" characteristics
4. Geological "turbidity maximum" zone influence The high level of natural statistical associated variation characteristics of samples collected from the segment influence the confidence limits that can be assigned to population parameters.
The nonradiological environmental monitoring will maximize effort in these areas most likely to measure the effects of station operation on the tidal segment.
The data collected under this program will be discussed with the appropriate State and Federal Agencies havingregulatory authority in the area.
TS FIGURE 4.13-1 3-17-72 TEMPERATURE AND SALINITY MONITORING STATIONS HOG ISLAND e
~
SU~
STATION 0 JAMES RIVER
,000 0 IOOO . ZOOO 3000 Ya,,,
LEGEND:
6 Monthly Salinity - Temperature Profile Station 0 Continuous Salinity - Temperature Monitoring Station
[] Near Surface Temperature Monitoring Station
Near Surface and Bottom Temperature Monit,oring Station Boat Cruise
TS FIGURE 4.13-2 3-17-72 BIOLOGICAL SAMPLE STATIONS
6 6
0
-- Do 0
6
............_..........................._o_ _ _ __.___ _ ____,z Naul1ctt'Mtl11 JAMES RIVER 0 IOOO . 1000 SOOD 'Tar,1 LEGEND:
0 Trawl (Nekton)
Seine (Nekton) e* D Plankton
6 Fouling Plates Benthos
TS 5-1.1 3-17-72 5.0 DESIGN FEATURES 5.1 SITE Applicability Applies to the location and boundaries of the site for the Surry Power Station.
Objective To define those aspects of the site which will affect the overall safety of the installation.
Specification The Surry Power Station is located in Surry County, Virginia, on property owned by Virginia Electric and Power Company on a point of land called Gravel Neck which juts into the James River. It is approximately 46 miles SE of Richmond, Virginia, 17 miles NW of Newport News, Virginia, and 25 miles NW of Norfolk, Virginia. The minimum distance from a reactor centerline to the site exclusion boundary as defined in 10CFRlOO is 1,650 ft. This is the distance for Unit 1, which is controlling.
References FSAR Section 2.0 Site FSAR Section 2.1 General Description
TS 5.2-1 3-17-72 5.2 CONTAINMENT Applicability Applies to those design features of the reactor containment structures and con-tainment systems relating to operational and public safety.
Objective To define ths significant design features of the reactor containment structures and containment systems.
Specifications A. Structure
1. A containment structure completely encloses each reactor and Reactor Coolant System and assures that an acceptable upper limit for leakage of radioactive materials to the environment is not exceeded even if gross failure of a Reactor Coolant System occurs. Each. structure provides biological shielding for both normal operation and accident situations. Each containment structure is designed for an internal subatmospheric pressure of 8 psia.
TS 5.2-2 3-17-72
2. Each containment structure is designed for a reactor operating at the ultimate rated thermal power of 2546 MWt.
3. Each containment structure is designed to withstand an internal design pressure of 45 psig acting simultaneously with: (1) loads resulting from an Operational Basis Earthquake having a hori-zontal ground acceleration of 0.07 g at zero period with an assumed structural damping factor of 5 percent, or (2) loads resulting from a Design Basis Earthquake having a horizontal ground acceleration of 0.15 g at zero period with an assumed structural damping factor of 10 percent.
B Containment Penetrations
1. All penetrations through the containment structure for pipe, electrical conductors, ducts, and access hatches are of the double barrier type.
2. The automatically actuated isolation valves are designed to close as outlined below. The actuation system is designed such that no single component failure will prevent contain-ment isolation if required. Refer to Table 3.7-4 in the Technical Specifications for set point values of the signals.
a. A safety injection signal closes all trip valves which are located in normally open lines connecting the reactor coolant loops and penetrating the containment.
TS 5.2-3 3-17-72 e
b. A high containment pressure isolation signal closes the automatic trip valves in all normally open lines penetrating the containment which are not required to be open to control containment pressure to perform an orderly reactor shut down without actuation of the consequence limiting safe-guards in case of a small Reactor Coolant System leak.
c. A further rise in containment pressure, indicating a major loss-of-coolant accident, produces a containment high-high pressure_ isolation signal which closes all normally open lines which penetrate the containment which have not been closed by 2-b above.
d. Isolation can be accomplished manually from the control in the Main Control Room if any of the automatic signals fail to actuate the'above valves.
c. Containment Systems
1. Following a loss-of--coolant accident, the Containment Spray Subsystems distribute at least 2,600 gpm borated water spray containing sodium hydroxide for iodine removal within the containment atmosphere. The Recirculation Spray Subsystems recirculate at least 3,500 gpm of water from the containment sump.
TS 5.2-4 3-17-72
2. No part of the Containment Ventilation System is designed for continued operation during a total loss-of-coolant accident.
It may, however, continue to operate with small Reactor Coolant System leaks until the Containment Spray System is initiated.
References FSAR Section 5.2 Containment Isolation FSAR Section 5.3 Containment Systems FSAR Section 5.4 Design Evaluation FSAR Section 7.2.2 Safeguards Initiation and Containment Isolation FSAR Section 15.2.4 Seismic Design
'\
FSAR Section 15.5.1 Containment Structure Technical Specification Section 3.3 Safety Injection System
TS 5.3-1 3-17-72 5.3 REACTOR Applicability Applies to the reactor core, Reactor Coolant System, and Safety Injection System.
Objective To define those design features which are essential in providing for safe system operations.
Specifications A. Reactor Core
1. The reactor core contains approximately 176,200 lbs. of uranium dioxide in the form of slightly enriched uranium dioxide pellets.
The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods. All fuel rods are pressurized with helium during fabrication.
The reactor core is made up of 157 fuel assemblies. Each fuel assembly contains 204 fuel rods.
2. The average enrichment of the initial core is 2.51 weight per cent of U-235. Three fuel enrichments are used in the initial core. The highest enrichment is 3.12 weight per cent of U-235.
_j
TS 5.3-2 3-17-72
3. Reload fuel will be similar in design to the initial core.
The enrichment of reload fuel will not exceed 3.60 weight per cent of U-235.
4. Burnable poison rods are incorporated in the initial core.
There are 816 poison rods in the form of 12 rod clusters, which are located in vacant control rod assembly guide thimbles.
The burnable poison rods consist of pyrex glass clad with stain-less steel.
5. There are 48 full-length control rod assemblies and 5 part-
\
length control rod assemblies in the reactor core. The full-length control rod assemplies contain a 144 inch-length of silver-indium-cadmium alloy clad with stainless steel. The part-length control rod assemblies contain a 36 inch length of silver-indium-cadmium alloy with the remainder of the stainless steel sheath filled with Al 2o3 *
6. The initial core and subsequent cores will meet the following performance criteria at all times during the operating lifetime.
a. Nuclear hot channel factors:
Deslgn Limits Interim Limits FN = 2.72 2.52 q
N F L\H = 1. 58
TS 5.3-3 3-17-72
b. Moderator temperature coefficient negative in the power operating range.
c. Capable of being made subcritical in accordance with Specification 3.12 A.3.C
7. Up to 10 grams of enriched fissionable material may be used either in the core or available on the plant site, in the form of fabricated neutron flux detectors for the purposes of monitoring core neutron flux.
B. Reactor Coolant System
1. The design of the Reactor Coolant System complies with the code requirements specified in Section 4 of the FSAR.
2. All piping, components and supporting structures of the Reactor Coolant System are designed to Class 1 seismic requirements, and have been designed to withstand:
a. Primary operating stresses combined with the Operational seismic stresses resulting from a horizonal ground acceleration of 0.07g and a simultaneous vertical ground acceleration of 2/3 the horizonal, with the stresses maintained within code allowable working stresses.
b. Primary operating stresses when combined with the Design Basis Earthquake seismic stresses resulting from a horizonal ground acceleration of 0.15g and a simultaneous vertical ground
TS 5.3-4 3-17-72 acceleration of 2/3 of the horizonal, with the stresses such that the function of the component or system shall not be impaired as to prevent a safe and orderly shutdown of the unit.
3. The total liquid volume of the Reactor Coolant System, at rated operating conditions, is approximately 9300 cubic feet.
TS 5, 4-1 3-17-72 5.4 FUEL STORAGE Applicability Applies to the design of the new and spent fuel storage areas.
Objective To define those aspects of fuel storage relating to prevention of criticality in fuel storage areas; to prevention of dilution of the borated water in the reactor; and to prevention of inadvertent draining of water from the spent e fuel storage area.
Specification A, The reinforced concrete structure and steel superstructure of the Fuel Building and spent fuel storage racks are designed to withstand Design Basis Earthquake loadings as Class I structures. The spent fuel pit has a stainless steel liner to ensure against loss of water.
B, The new and spent fuel storage racks are designed so that it is impossible to insert assemblies in other than the prescribed loca-tions. Both new and spent fuel is stored vertically in an arrp~r with a distance of 21 inches between assemblies to assure k . < 0.90, eff-'-
even if unborated water were used to fill the spent fuel storage pit or the new fuel storage area.
TS 5.4-2 3-17-72 C. Whenever there is spent fuel in the spent fuel storage pit, the pit shall be filled with borated water at a boron concentration not less than 2,000 ppm to match that used in the reactor cavity and refueling canal during refueling operations.
D. The only drain which can be connected to the spent fuel storage area is that in the reactor cavity. The strict step-by-step procedures used during refueling ensure that the gate valve on the fuel transfer tube which connects the spent fuel storage area with the reactor cavity is closed before draining of the cavity commences. In addition, the procedures require placing the bolted blank flange on the fuel transfer tube as soon as the reactor cavity is drained.
References FSAR Section 9.5 Fuel Pit Cooling System FSAR Section 9.12 Fuel Handling System
TS 6.1-1 3-17-72 6.0 ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION, SAFETY AND OPERATION REVIEW Specification A, The Station Manager shall be responsible for the safe operation of the facility. The Station Manager shall report to the Superintendent -
Production Operations. The relationship between this supervisor and other levels of company management is shown in TS Figure 6.1-1.
B. The station organization shall conform to the chart as shown in TS Figure e 6.1-2.
1. Qualifi.cations with regard to education and experience and the technical specialties of key supervisory personnel will meet the minimum acceptable levels described in ANSI Nl8.l "Selection and Training of Nuclear Power Plant Personnel," dated March 8, 1971.
The key supervisory personnel are as follows:
a) Manager b) Superintendent-Station Operations c) Operating Supervisor d) Supervisor-Electrical Maintenance e) Supervisor-Mechanical Maintenance f) Supervisor-Engineering Services
- g) h)
Chemistry and Health Physics Supervisor Shift Supervisor
TS 6.1-2 3-17-72
2. Retraining and replacement training of station personnel shall be in accordance with ANSI Nl8.1 "Selection and Training of Nuclear Power Plant Personnel," dated March 8, 1971.
3. The following requirements supplement the applicable regulations of 10 CFR 50.54:
Condition Minimum Complement
1. One unit operating 1 SOL, 2 LO, 2 AO
2. One unit fueled 1 SOL, 1 LO, 1 AO and shutdown**
3. One unit operating and 1 SOL*, 3 LO, 2 AO one unit shutdown
4. Both units fueled and 1 SOL, 2 LO, 1 AO shutdown**
5. Both units operating 2 SOL, 3 LO, 2 AO Note:
SOL Senior Licensed Operator as defined by 10 CFR 55.4(e)
LO Licensed Operator as defined by 10 CFR 55.4(d)
AO Auxiliary Operator
When the shutdown unit is undergoing refueling or startup, 1 additional SOL will be added to this shift complement to ensure supervision of these activities.
ALO for each fueled unit shall be in the control room and a SOL shall be on site. For each SOL in the control room, the requirement to have a LO in the control room shall be waived.
C. Organization units to provide a continuing review of the operational and safety aspects of the nuclear facility shall be constituted and have the authority and responsibilities outlined below:
1. Station Nuclear Safety and Operating Committee
a. Membership
1. Chairman - Manager Vice Chairman - Superintendent-Station Operation
TS 6.1-3 3-17-72
- 3.
4.
Operating Supervisor Supervisor-Electrical Maintenance
5. Supervisor-Mechanical Maintenance
6. Supervisor-Engineering Services
7. Chemistry and Health Physics Supervisor
b. Qualifications: The qualifications of the regular members of the Station Nuclear Safety and Operating Committee with regard to the combined experience and technical specialties of the individual members shall be maintained at a level at least equal to those described in Section 6.1, B.1. of these Specifications.
c. Meeting frequency: As called by the Chairman but not less than monthly.
d. Quorum: Chairman or Vice Chairman, Chemistry and Health Physics Supervisor or his designee, and three others to provide a quorum of five members.
e. Responsibilities
1. Periodically review all proposed normal, abnormal, and emergency operating procedures and all proposed maintenance procedures. Review proposed changes to those procedures, and any other proposed procedures or changes thereto as e determined by the Station Manager to affect nuclear safety.
TS 6.1-4 3-17-72
2. Review all proposed test and experiment procedures and results thereof when applicable.
3. Review proposed changes to Technical Specifications.
4. Review all proposed changes or modifications to systems or equipment that would require a change in established procedures, or which would constitute a design change.
5. Periodically review all operations to detect any potential safety hazards.
6. Investigate all reported instances of departure from Technical Specification.limits, such investigations to include review, evaluation and recommendations to prevent recurrence, to the Station Manager, Superintendent-Production Operations and to the Chairman of the System Nuclear Safety and Operating Committee.
7. The Station Nuclear Safety and Operating Committee shall make tentative determinations as to whether or not proposals considered by the Committee involve unreviewed safety questions. This determination shall be subject to review and approval by the System Nuclear Safety and Operating Committee.
8. 'Review all abnormal occurrence reports.
TS 6.1-5 3-17-72 9, Perform special reviews and investigations and render reports thereon as requested by the Chairman of the System Nuclear Safety and Operating Committee.
10. Initiate periodic drills to test the effectiveness of the emergency procedures.
f. Authority
1. The Station Nuclear Safety and Operating Committee shall advise the Manager on all matters affecting the safe operation of the facility.
2. The Station Nuclear Safety and Operating Committee shall recommend to the Station Manager approval or disapproval of proposals under items e(l) through (4) above.
a) In the event of disagreement between the recommendations of the Station Nuclear Safety and Operating Committee and the actions contemplated by the Station Manager, the course determined by the Station Manager will be followed with immediate notification to the Superintendent-Production Operations and the Chairman of the System Safety and Operating Committee.
g. Records Minutes shall be kept of all meetings of the Station Nuclear
TS 6.1-6 3-17-72 Safety and Operating Committee and copies shall be sent to the Superintendent-Production Operations and to all members of the Station and System Nuclear Safety and Operating Committees.
h. Procedures Written administrative procedures for committee operation shall be prepared and maintained describing the method of submission, and the content of presentations to the committee, provisions for the use of subcommittees; review and approval by members of written committee evaluations and recommendations; the distri-butions of minutes; and, such other matters as may be appropriate.
2. System Safety and Operating Committee
a. Membership
1. Chairman and Vice Chairman appointed by name by the Vice President-Power, which may be an individual listed in Item 2.
2. Five members of the Production Department system office staff (refer to TS Fig. 6.1-1) who are experienced in utility operation and procedures:
Director - Production Operations and Maintenance Superintendent - Production Operations Director - Power Station Design
TS 6.1-7 3-17-72 Director - Nuclear Services Supervisor - Nuclear Design
3. Manager of each nuclear generating station operating on the Virginia Electric and Power Company system, or his designee. In matters or consideration of proposals pertinent to a particular station, the Manager of this station shall serve as a non-voting member of the committee. In matters pertaining to other stations, Station Managers will serve as voting members.
4. At least one qualified non-company affiliated technical consultant. Duly appointed consultant members shall have equal vote with permanent members of the Committee.
b. Qualifications The minimum qualifications of the Company members of the System Safety and Operating Committee will be:
an engineering graduate or equivalent with combined nuclear and conventional experience in power station design and/or operation of eight years, with at least two years involving the direction of nuclear operations or design activity.
c. Consultants e The committee shall have the authority to call technically
TS 6.1-8 3-17-72 qualified personnel from within the Virginia Electric and Power Company organization or from any other consultant source.
d. Quorum: Either the Chairman or Vice Chairman and two thirds of the other members shall constitute a quorum.
e. Meeting frequency: As required by the Chairman but not less than quarterly.
f. Responsibilities
1. Review proposed changes to the operating license including Technical Specifications.
2. Review minutes of meeting of the Station Safety and Operating Committee(s) to determine if matters considered by that committee involve unreviewed or unresolved safety questions.
3. Review matters including proposed changes or modifications to systems or equipment having safety significance referred to it by the Station Nuclear Safety and Operating Committee or by the Station Manager.
4. Conduct periodic review of station operations.
5. Review all reported instances of departure from Technical Specification limits and report findings and recommendations to prevent recurrence to the Manager-Power Production.
TS 6.1-9 3-17-72
6. Perform special reviews and investigations and render reports thereon as requested by company management, or as the committee deems necessary.
7. Review proposed tests and experiments and results thereof when applicable.
8. Review abnormal performance of plant equipment and anomalies.
9. Review unusual occurrences and incidents which are reportable under the provisions of 10 CFR 20 and 10 CFR 50.
10. Review of occurrences if Safety Limits are exceeded.
g. Authority
1. Recommend approval of proposed changes to the operating license, including Technical Specifications, for submission to the A.E.C.
2. Recommend approval of proposed changes or modifications to systems or equipment, provided such changes or modifications do not involve unreviewed safety questions.
h. Records Minutes shall be recorded of all meetings of this Committee.
TS 6.1-10 3-17-72 Copies of the minutes shall be forwarded to the Vice President-Power Production, all members of the committee and any others that the Chairman may designate.
i. Procedures Written administrative procedures for committee operation shall be maintained describing the method of submission and the content of presentations to the committee; provisions for use of subcommittee evaluations and recommendations; distribution of minutes; and, such other matters as may be appropriate.
TS FIG. 6.1-1 3-17-72 e
POWER PRODUCTION DEPARTMENT ORGANIZATION CHART VIRGINIA ELECTRIC AND POWER COMPANY VICE PRESIDENT POWER I
I MANAGER MANAGER SYSTEM NUCLEAR L-.......
SAFETY & OPERATING FUEL RESOURCES POWER PRODUCTION e COMMITTEE DIRECTOR DIRECTOR DIRECTOR PRODUCTION POWER STATION OPERATIONS AND NUCLEAR MAINTE.NANCE SERVICES DESIGN I
SUPERINTENDENT SUPERINTENDENT SUPERINTENDENT SUPERVISOR .,
PRODUCTION PRODUCTION PRODUCTION NUCLEAR MAINTENANCE OPERATIONS ENGINEERING DESIGN.
MANAGER SURRY
VffiGINIA ELECTRIC AND POWER COMPANY*
SURRY POWER STATION ORGANIZATION CHART STATION NUCLEAR SAFETY & OPERATING STATION MANAGER COMMITTEE
SUPERINTENDENT STAT ION OPERATIONS ESL I I I OPERATING SUPERVISOR SUPERVISOR SUPERVISOR CHEMISTRY AND
.sUPERV ISOR ELECTRICAL MECHANICAL ENGINEERING HEALTH PHYSICS*
MAINTENANCE MA I NTENANCE SERVICES SUP ERV I SOR ESL E C l I I
";'c;':
SH I FT INSTRUMENT SUPERVISOR ENGINEER SUP ERV I SOR SL E I
CONTROL ROOM OPERATOR OL LEGEND I
";"* SL - SENIOR LICENSE ASS ISTANT OL - OPERATOR 1 S LICENSE 1-3
(/)
CONTROL ROOM E- GRADUATE ENGINEER tzj OL OPERATOR C- COLLEGE GRADUATE H
- SENIOR LICENSE WILL BE OBTAINED w.0 I
I WITHIN 18 MONTHS AFTER INITIAL I-' (J\
--.J *
-;':"'k I CRITICALITY OF UNIT NO. l --.J f--'
I AUXILIARY ** - SHIFT COMPLEMENTS FOR DIFFERENT NN OPERATOR STATION CONDITIONS ARE DETAILED IN SPECIFICATION 6. 1-B.3
TS 6.2-1 3-17-72 e 6.2 ACTION TO BE TAKEN IN THE EVENT OF AN ABNORMAL OCCURRENCE IN STATION OPERATION Specification A. Any abnormal occurrence shall be reported immediately to and promptly reviewed by the Chairman of Station Nuclear Safety and Operating Committee or his designee, Superintendent - Production Operations and the Chairman of the System Nuclear Safety and Operating Committee.
B. The Shift Supervisor on duty and subsequently the Operating Supervisor shall prepare a report for each abnormal occurrence. This report shall include an evaluation of the cause of the occurrence and also recommenda-e tions for appropriate action to prevent or reduce the probability of a re-currence. Immediate corrective action shall be taken to correct the anomaly.
C. Copies of all such reports shall be submitted to the Superintendent*-
Station Operations, the Station Manager, who also serves as the Chairman of the Station Nuclear Safety and Operating Committee, Superintendent-Production Operations, and to the Chairman of the System Nuclear Safety and Operating Committee for review and approval of any recommendation.
D, The Vice President - Power shall report the circumstances of any abnormal occurrence to the AEC as specified in Section 6.6 of these Specifications.
TS 6.3-1 3-17-72 6.3 ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED Specification A. Should a safety limit (see Section 2.0 of the Technical Specifications) be exceeded, the reactor shall be shutdown and reactor operation shall only be resumed in accordance with the authorization within 10 CFR 50.36 (c)(l)(i).
B. An immediate report of the incident shall be made to the Station Manager, Superintendent - Production Operations and the Chairman of the System Nuclear Safety and Operating CoIIU11ittee.
C. The Station Manager shall promptly report the circumstances to the AEC as specified in Section 6.6 of these Specifications.
D. A complete analysis of the incident together with recommendations to prevent recurrence shall be prepared by the Shift Supervisor and the Operating Supervisor. A preliminary written report shall be submitted to the Superintendent Station Operations, Station Manager who is also the Chairman of the Station Nuclear Safety and Operating Committee, Superintendent - Production Operations, and the Chairman of the System Nuclear Safety and Operating CoIIU11ittee within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the incident.
Appropriate analyses or reports will be submitted to the AEC by the Vice President - Power as specified in Section 6.6 of these Specifications.
TS 6.4-1 3-17-72 6.4 UNIT OPERATING PROCEDURES Specification A. Detailed written procedures with appropriate check-off lists and instructions shall be provided for the following conditions:
1. Normal startup, operation, and shutdown of a unit, and of all systems and components involving nuclear safety of the station.
2. Calibration and testing of instruments, components, and systems involving nuclear safety of the station.
3. Actions to be taken for specific and foreseen malfunctions of systems or components including alarms, primary system leaks and abnormal reactivity changes.
4. Release of radioactive effluents
5. Emergency conditions involving potential or actual release of radioactivity
6. Emergency conditions involving violation of industrial security.
7, Preventive or corrective maintenance operations which would have an effect on the safety of the reactor.
8. Refueling operations.
B. Radiation control procedures shall be provided and made available to all station personnel. These procedures will show permissible radiation exposure. This radiation protection program shall be organized to meet the requirements of 10 CFR 20 and/or the following provisions:
TS 6.4-2 3-17-72 e 1. The intent of 10 CFR 20.203(c)(2)(iii) shall be implemented by satisfying the following conditions:
a. The entrance to each radiation area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr shall be barricaded and conspicuously posed.
b, The entrance to each radiation area in which the intensity of radiation is equal to or greater than 1000 mrem/hr shall be provided with locked barricades to prevent unauthorized entry into these areas. Keys to these locked barricades shall be maintained under the administrative control of the Shift Supervisor.
c. All such accessible high radiation areas shall be surveyed by Health Physics personnel on a routine schedule, as determined by the Chemistry and Health Physics Supervisor, to assure a safe and practical program.
d, Any individual entering a high radiation area shall have completed the indoctrination course designed to explain the hazards and safety requirements involved, or shall be escorted at all times by a person who has completed the course.
e. Any individual or group of individuals permitted to enter a high radiation area per l.d. above, shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area.
TS 6.4-3 3-17-72 f, Entrance to areas with radiation levels in excess of 1 R/hr shall require the use of the "buddy system," whereby a minimum of two individuals maintain continuous visual and/or verbal communication with each other; or other mechanical and/or electrical means to provide constant conununication with the individual in the area shall be provided.
g, A Radiation Work Permit system shall be used to authorize and control any work performed in high radiation areas.
h, All buildings or structures, in or around which a high radiation area exists, shall be surrounded by a chain-link fence.The entrance gate shall be locked under administrative control, or continuously guarded to preclude unauthorized entry.
i. Stringent administrative procedures shall be implemented to assure adherence to the restriction placed on the entrance to a high radiation area and the radiation protection program associated thereto.
I 2. Pursuant to 10 CFR 20.103(c)(l) and (3), allowance can be made for the use of respiratory protective equipment in conjunction with activities authorized by the operating license in determining whether individuals are exposed to concentrations in excess of the limits specified in Appendix B, Table I, Column 1, of 10 CFR 20, subject to the following limitations:
a, The limits provided in 10 CFR 20.103(a) and (b) are not exceeded,
b. If the radioactive material is of such form that intake through the skin or other additional route is likely, individual exposures to radioactive material shall be controlled so that the radioactive
TS 6.4-4 3-17-72 content of any critical organ from all routes of intake averaged over seven (7) consecutive days does not exceed that which would result from inhaling such radioactive material for forty (40) hours at the pertinent concentration values provided in Appendix B, Table I, Column 1, of 10 CFR 20.
11
c. For radioactive materials designated 11 Sub 11 in the Isotope 11 column of Appendix B, Table I, Column 1 of 10 CFR 20, the concentration value specified is based upon exposure to the material as an external radiation source. Individual exposures to these materials shall be accounted for as part of the limitation on individual dose in §20.101
3. In all operations in which adequate limitation of the inhalation of radioactive material by the use of process or other engineering controls is impracticable, an individual shall be permitted to use respiratory protective equipment to limit the inhalation of airborne radioactive material provided:
a) The limits specified in B.2 above are not exceeded.
b) Respiratory protective equipment is selected and used so that the peak concentration of airborne radioactive material inhaled by an individual wearing the equipment does not exceed the pertinent concentration values specified in Appendix B, Table I, Column 1, of 10 CFR 20. The concentration of radioactive material that is inhaled when respirators are worn can be determined by dividing the ambient airborne concentration by the protection factor specified in this specification. If the intake of radioactivity is later deter-mined by other measurements to have been different than that initially estimated, the later quantity shall be used in
TS 6.4-5 3-17-72 evaluating the exposures.
c) Each respirator user is advised that he may leave the area at any time for relief from respirator use in case of equipment malfunction~ physical or psychological discomfort, or any other condition that might cause reduction in the protection afforded the wearer.
d) A respiratory protective program adequate to assure the requirements of this specification are met is maintained.
Such a program shall include:
(1) Air sampling and other surveys sufficient to identify the hazard, to evaluate individual exposures, and to permit proper.selection of respiratory protective equipment.
(2) Written procedures to assure proper selection, super-vision, and training of personnel using such protective equipment.
(3) Written procedures to assure the adequate fitting of respirators; and the testing of respiratory protective equipment for operability.
(4) Written procedures for maintenance to assure full effectiveness of respiratory protective equipment, including issuance, cleaning and decontamination, inspection, repair and storage.
(5) Written procedures for proper use of respiratory pro-tective equipment including provisions for planned
limitations on working times as necessitated by operational conditions.
TS 6.4-6 3-17-72 (6) Bioassays and/or whole body counts of individuals, and other surveys, as appropriate, to evaluate individual exposures and to assess protection actually afforded.
e) Equipment approved by the U. s. Bureau of Mines is used.
Equipment not approved under u. s. Bureau of Mines Approval Schedules may be used only if the equipment has been evaluated and it can be shown on the basis of reliable test information or manufacturer's data, that the material and performance characteristics of the equipment are at least equal to those afforded by U. S. Bureau of Mines approved equipment of the same type.
f) Protection factors in excess of those specified in B.4 of this specification shall not be used.
4. The type of respiratory protective equipment shall be selected and used based on the protection factor required to satisfy the limits specified in B.2 above. No protection factor is to be used to reduce the concen-tration of tritium oxide or noble gases. The protection factor to be used on other particulates, vapors, and gases are delineated in"Table TS 6.4-1.
5. Specification B.2, 3. and 4 shall not preclude the use of respiratory protective equipment in emergencies.
6. Specification B.2, 3, and 4 shall be superceded by changes to 10 CFR 20.103.
All procedures described in A and B above, and changes thereto, shall be reviewed by the Station Safety and Operating Committee and approved by the Station Manager prior to implementation.
TS 6.4-7 3-17-72 e D, All procedures described in A and B above shall be followed.
E, Temporary changes to procedures which do not change the intent of the original procedure may be made, provided such changes are approved by the Operating Supervisor or Superintendent Station Operation. Such changes will be docl.llllented and subsequently reviewed by the Station Nuclear Safety and Operating Committee and approved by the Station Manager.
F. Practice of site evacuation exercises shall be conducted annually, following emergency procedures and including a check of communications with off-site report groups. An annual review of the emergency plan will be performed.
G. The industrial security program which has been established for the station shall be implemented, and approprtate investigation and/or corrective action
. shall be taken if the provisions of the program are violated. However, use of additional.guards* to provide surve+/-llance in areas normally monitored by electronic devices is authorized in lieu of these devices for the first six (6) months of operation. Deployment of additional guards shall be in accordance with the provisions delineated in the letter from the Virginia Electric and Power Company to the Atomic Energy Commission dated May 12, 1972. An annual review of the program shall be performed.
e TABLE TS 6, 4-1 PROTECTION FACTORS FOR RESPIRATORS PROTECTION FACTORS~/
PARTICULATES AND VAPORS AND GASES EXCEPT DESCRIPTION MODES1/ TRITIUM oxrnEl/
I. AIR-PURIFYING RESPIRATORS Facepiece, half-mask 4/ 6/ NP 5 Facepiece, full 6/ - - NP 100 II. ATMOSPHERE-SUPPLYING RESPIRATOR
1. Air Line respirator Facepiece, half-mask CF 100 Facepiece, full CF 1,000 Facepiece, full 6/ D 500 Facepiece, Full - PD 1,000 Hood CF 5/
Suit CF Jj
2. Self-contained breathing apparatus (SCBA)
Facepiece, full 2_/ D 500 Facepiece, full PD 1,000 Facepiece, full R 1,000 1-3 w Cl)
III. COMBINATION RESPIRATOR I Any combination of air- Protection factor for . . . °'
I .i:,,.
purifying and atmosphere- type and mode of operation -...i I N 00 supplying respirator as listed above.
TS 6.4-9 3-17-72 FOOTNOTES TO TABLE TS 6.4-1 1./ See the following symbols:
CF: continuous flow D : demand NP: negative pressure (i.e., negative phase during inhalation)
PD: pressure demand (i.e., always positive pressure)
R: recirculating (closed circuit) 2:._/ (a) For purposes of this specification the protection factor is a measure of the degree of protection afforded by a respirator, defined as the ratio of the concentration of airborne radio-active material outside the respiratory protective equipment to that inside the equipment (usually inside the facepiece) under conditions of use. It is applied to the ambient airborne concentration to estimate the concentration inhaled by the wearer according to the following formula:
Ambient Airborne Concentration Concentration Inhaled= Protection Factor (b) The protection factors apply:
(i) only for trained individuals wearing properly fitted respirators used and maintained under supervision in a well-planned respiratory protective program.
(ii) for air-purifying respirators only when high efficiency
[above 99.5% removal efficiency by U. S. Bureau of Mines type dioctyl phthalate (DOP) test] particulate filters and/or sorbents appropriate to the hazard are used in atmospheres not deficient in oxygen.
(iii) for atmosphere-supplying respirators only when supplied with adequate respirable air.
]_/ Excluding radioactive contaminants that present an absorption or submersion hazard. For tritium oxide approximately half of the intake occurs by absorption through the skin so that an overall protection factor of not more than approximately 2 is appropriate when atmosphere-supplying respirators are used to protect against tritium oxide. Air-purifying respirators are not recommended for use against tritium oxide. See also footnote J../, below, concerning supplied-air suits and hoods.
TS 6.4-10 3-17-72
!!_I Under chin type only. Not recommended for use where it might be possible for the ambient airborne concentration to reach instantaneous values greater than 50 times the pertinent values in Appendix B, Table I, Column 1 of 10 CFR, Part 20.
ii Appropriate protection factors must be determined taking account of the design of the suit or hood and its permeability to the contaminant under conditions of use. No protection factor greater than 1,000 shall be used except as authorized by the Commission,
!ii Only for shaven faces.
NOTE 1: Protection factors for respirators, as may be approved by the U. S. Bureau of Mines according to approval schedules for respirators to protect against airborne radionuclides, may be used to the extent that they do not exceed the protection factors listed in this Table. The pro-tection factors in this Table may not be appropriate to circumstances where chemical or other respiratory hazards exist in addition to radioactive hazards. The selection and use of respirators for such circumstances should take into account approvals of the U. S. Bureau of Mines in accordance with its applicable schedules.
NOTE 2: Radioactive contaminants for which the concentration values in Appendix B, Table 1 of this part are based on internal dose due to inhalation may, in addition, present external exposure hazards at higher concentrations. Under such circumstances, limitations on occupancy may have to be governed by external dose limits.
TS 6.5-1 3-17-72 6.5 STATION OPERATING RECORDS Specification A. Records and logs relative to the following items shall be retained for 5 years, unless a longer period is required by applicable regulations.
1, Records of normal plant operation, including power levels and periods of operation at each power level.
2. Records of principle maintenance activities, including inspection, repair, substitution or replacement of principle items of equipment pertaining to nuclear safety.
3. Record of abnormal occurrences.
4. Record of periodic checks, inspections and calibrations performed to verify that surveillance requirements are being met.
5, Records of any special reactor test or experiments pursuant to 10 CFR 50.59.
6, Records of changes made in the Operating Procedures pursuant to 10 CFR 50.59.
-* 7. Records of shipment of radioactive material.
TS 6.5-2 3-17-72 B. Records relative to the following items shall be retained for the life of the plant.
1. Records of changes made to the plant and plant drawings as de.scribed in the FSAR pursuant to 10 CFR 50.59.
2. Records of new and spent fuel inventory and assembly histories.
3. Records of plant radiation and contamination surveys.
4. Records of off-site environmental monitoring surveys.
5. Records of radiation exposure of all plant personnel, and others as required by 10 CFR 20.
6. Records of radioactivity in liquid and gaseous wastes released to the environment.
7. Primary records of inservice inspections including, but not limited to radiographs (of welds which are radiographed), photographs of the scope traces for welds which are tested by ultrasonic techniques, and photographs of the surface of those welds inspected by visual or surface examinations.
TS 6.6-1 3-17-72 6.6 STATION REPORTING REQUIREMENTS Specification A. Routine Operating Reports - A routine operating report shall be sub-mitted in writing to the Deputy Director for Reactor Projects~ Directorate of Licensing, U. S. Atomic Energy Commission, Washington, D. C. 20545 at the end of each six month period ot fraction thereof terminating on June 30 and December 31. Such reports are due within 60 days after the end of each reporting period. The following information, summarized on a semi-annual basis, except as noted, shall be provided:
1. Operations Summary A summary of operating experience occurring during the reporting period that relates to the safe operation of the facility, including a summary of:
a. changes in facility design,
b. performance characteristics (e.g., equipment and fuel performance) ,
c. changes in procedures which were necessitated by (a) and (b) or which otherwise were required to improve the safety of operations,
d. results of surveillance tests and inspections required by the licensee's technical specifications,
e. the results of any periodic containment leak rate tests performed during the reporting period,
f. a brief summary of those changes, tests and experients requiring authorization from the Commission pursuant to
10 CFR 50.59(a), and
i TS 6.6-2 3-17-72
g. any changes in the plant operating organization which involve positions for which minimum qualifications are specified in the technical specifications.
2. Power Generation A summary of power generated during the reporting period and the cumulative total outputs since initial criticality, including:
a. gross thermal power generated (in MWH)
b. gross electrical power generated (in MWH)
c. net electrical power generated (in MWH)
d. number of hours the reactor was critical
e. number of hours the generator was on-line
f. histogram of thermal power vs. time
3. Shutdowns Descriptive material covering all outages occurring during the reporting period. For each outage, information shall be provided on:
a. the cause of the outage,
b. the method of shutting down the reactor; e.g., _scram, automatic rundown, or manually controlled deliberate shutdown,
c. duration of the outage,
d. plant status during the outage; e.g., cold shutdown or hot standby,
e. corrective action taken to prevent repetition, if apprqpriate.
TS 6.6-3 3-17-72
4. Maintenance A discussion of safety-related maintenance (excluding pre-ventive maintenance) performed during the reporting period on systems and components that are designed to prevent or mitigate the consequences of postulated accidents or to prevent the release of significant amounts of radioactive material. Included in this category are systems and components which are part of the reactor coolant pressure boundary defined in 10 CFR § 50.2(v), part of the engineered safety features, or associated service and control systems that are required for the normal operation of engineered safety features, part of any reactor protection or shutdown system, or part of any radioactive waste treatment handling and disposal system or other system which may contain significant amounts of radioactive material. For any malfunction for which corrective maintenance was required, information shall be provided on:
a. The system or component involved, b, the cause of the malfunction,
c. The results and effect on safe operation,
d. corrective action taken to prevent repetition,
e. precautions taken to provide for reactor safety during repair
f. time required for performing maintenance
TS 6.6-4 3-17-72
5. Changes, Tests and Experiments A summary of all changes in the facility design and procedures that relate to the safe operation of the facility shall be included in the Operations Summary section of the semi-annual report. Changes, tests and experiments performed during the reporting period that require authorization from the Commission pursuant to 10 CFR 50.59(a) and those changes, tests and experiments that do not require Commission authorization pursuant to§ 50.59(a) should be addressed. The report shall include a brief description and a summary of the safety evaluation for those changes, tests, and experiments, carried out without prior Commission approval, pursuant to the requirements of
§ 50.59(b) of the Commission's regulations, that "The licensee shall furnish to the Commission, annually or at such shorter intervals as may be specified in the license, a report containing a brief description of such changes, tests and experiments, including a summary of the safety evaluation of each."
6. Radioactive Effluent Releases Data shall be reported in the form given in Appendix A of U.S.A.E.C. Safety Guide No. 21 for water cooled nuclear power plants, entitled "Measuring and Reporting of Effluents from Nuclear Power Plants," dated December 29, 1971, or in equivalent form. Effluent data shall be summarized on a monthly basis except that when the majority of the activity is released as batches and there are less than 3 batches per
TS 6,6-5 3-17-72 e month, each batch shall be reported. Estimates of the error associated with each six month total shall be reported. Specifically, the following data shall be reported.
a. Gaseous Releases (1) total radioactivity (in curies) releases of noble and activation gases.
(2) maximum noble gas release rate during any one-hour period.
(3) total radioactivity (in curies) releases, by nuclide, based on representative isotopic analyses performed.
(4) percent of technical specification limit.
b. Iodine Releases (1) total (I-131, I-133, I-135) radioactivity (in curies) released.
(2) total radioactivity (in curies) released, by nuclide, based on representative isotopic analyses performed.
(3) percent of -technical specification limit.
c. Particulate Releases (1) gross radioactivity (6,t) released (in curies) excluding background radioactivity.
(2) gross alpha radioactivity released (in curies) excluding background radioactivity.
(3) total radioactivity released (in curies) of nuclides with half-lives greater than eight days.
e (4) percent of technical specification limit.
r TS 6.6-6 3-17-72
d. Liquid Releases (1) gross radioactivity ~,i) released (in curies) and average concentration released to the un-restricted area.
(2) total tritium and alpha radioactivity (in curies) released and average concentration released to the unrestricted area.
(3) total dissolved gas radioactivity (in curies) and average concentration released to the unrestricted area.
(4) total volume (in liters) of liquid waste released.
(5) total volume (in liters) of dilution water used prior to release from the restricted area.
(6) the maximum concentration of gross radioactivity (6,Y) released to the unrestricted area (averaged over the period of release).
(7) total radioactivity (in curies) released, by nuclide, based on representative isotopic analyses performed.
(8) percent of technical specification limit for total activity released.
7. Solid Radioactive Waste(Sumrnarized Monthly)
a. Total amount of solid waste packaged (in cubic feet).
b. Estimated total radioactivity (in curies) involved.
c. Dates of shipment and disposition (if shipped off-site)
8. Fuel Shipments Information relative to each shipment of new and spent fuel shall be provided, including the following:
TS 6.6-7 3-17-72
a. Date of shipments b, Number of elements shipped
c. Identification number and enrichment of elements shipped.
d. Activity level at surface of each shipping cask con-taining spent fuel.
9. Environmental Monitoring a) Descriptive material covering the off-site environmental surveys performed during the reporting period including information on:
(1) The number and types of samples taken; e.g., air, soil, fish, etc.
(2) The number and types of measurements made; e.g.
dosimetry.
(3) Location of the sample points and monitoring stations.
(4) The frequency of the surveys (5) A summary of survey results, including:
(a) number of locations at which activity levels are found to be significantly greater then local backgrounds.
(b) highest, lowest, and the annual average con-centrations or levels of radiation for the sampling point with the highest average and description of that point with respect to the site.
TS 6.6-8 3-17-72 b) If levels of station contributed radioactive materials in environmental media indicate the likelihood of public intakes in excess of 3% of those that could result from continuous exposure to the concentration*values listed in Appendix B, Table II of 10 CFR 20, estimates of the likely resultant exposure to individuals and to population groups, and assumptions upon which estimates are based shall be provided.
c) If a particular sample or measurements indicate statistically significant levels of radioactivity above established or concurrent backgrounds, the following information shall be provided:
(1) The type of analysis performed; e.g.,alpha, beta, gama and/or isotopic.
(2) The minimum sensitivity of the monitoring system.
(3) The measured radiation level or sample concentration.
(4) The specific times when samples were taken and measurements were made.
(5) An estimate of the likely resultant exposure to the public if it exceeds 10 mrem.
B. Non-Routine Reports
1. Abnormal Occurrence Reports - A notification shall be made within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephone or telegraph to the Director, Region II Compliance Office, followed by a written report within 10 days to the Deputy Director for Reactor Projects, Directorate of Licensing (carbon copy to the Director, Region II, Directorate of Regulatory Operations) in the event of an abnormal occurrence. Abnormal
TS 6.6-9 3-17-72 occurrences are defined in Section 1 of these Technical Specifica~
e tions. The written report on abnormal occurrences, and to the extent possible, the preliminary telephone or telegraph notification shall:
a. describe, analyze and evaluate safety implications,
b. outline the measures taken to assure that the cause of the condition is determined, and
c. indicate the corrective action(including any changes made to the procedures and to the quality assurance program) taken to prevent repetition of the occurrence and of similar occurrences involving similar components or systems.
d. relate any failures or degraded performance of systems and components for the incident to similar equipment failures that may have previously occurred at the facility. The evaluation of .the safety implications of the incident should consider the cumulative experience obtained from the record of previous failures and mal-functions of the affected srstems and components or of similar equipment.
2. Reporting of Unusual Safety Related Events A written report shall be forwarded within 30 days to the Deputy Director for Reactor Projects, Directorate of Licensing and to the Director, Region II Directorate of Regulatory Operations in the event of an unusua~ safety related event. Unusual safety related events are defined in Section 1 of these Technical Specifications.
3. Radioactive Effluents If the experienced rate of release of radioactive materials in liquid and gaseous wastes, when averaged over any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period is such that the concentrations of radionuclides at the site boundary
TS 6.6-10 3-17-72
- exceed 4% of the limits in Specification 3.11-A.l or 3.11-B-1, a written report shall be submitted to the Deputy Director for Reactor Projects, Directorate of Licensing and to the Director,Region II Directorate of Regulatory Operations within 30 days following the release. This report shall identify the cause(s) of the activity release and describe the proposed program of action to reduce such C. Jnservice Inspection Reports Notification shall be made to the Director, Region II Directorate of Regulatory Operations at least 60 days before scheduled refueling periods.
This notification shall include the details of the inservice inspection planned for such period, including a detailed schedule identifying specific welds to be inspected and identifying the specific techniques to be em-ployed.
Within 90 days after completion of inservice inspection activities, a written report shall be forwarded to the Deputy Director for Reactor Projects, Directorate of Licensing, and to the Director, Region II, Directorate of Regulatory Operations covering the scope of the examina-tions' conducted and the results thereof. In addition, a preliminary report of the findings of the inservice inspection activities shall be filed with the Deputy Director for Reactor Projects, Directorate of Licensing, and the Director, Region II Directorate of Regulatory Operation before the facility is returned to power.
D, Special Reports
a. Startup Report - A summary report of unit startup and power escalation testing and the evaluation of the results from these test programs shall be submitted when a unit is initially placed
TS 6.6-11 3-17-72 in service and when the unit has been modified to an extent that the nuclear, thermal, or hydraulic performance of the unit may be significantly altered. The test results shall be compared with design predictions and specifications. Startup reports shall be submitted within 60 days following commencement of commercial power operation (i.e., following synchronization of the turbo-generator to produce commercial power).
b. First Year Operation Report - A report shall be submitted within 60 days after completion of the first year of operation (the first year begins with the synchronization of the turbogenerator to pro-duce commercial power). This report may be incorporated into the semiannual operating report and shall cover the following:
(1) an evaluation of plant performance to date in com-parison with design predictions and specifications (2) a reassessment of the safety analysis submitted with the license application in light of measured operating characteristics when such measurements indicate that there may be substantial variance from prior analyses.
(3) an assessment of the performance of structures, systems*
and components important to safety (4) a progress and status report on any items identified as requiring additional information during the operating license review or during the startup of the facility, including items on which additional information was re-quired as conditions of the license and items identi-fied in the licensee '*s sta,rtup report
TS 6,6-12 3-17-72
c. Containment Leak Rate Test - Each containment integrated leak rate test shall be the subject of a summary technical report. Upon completion of the initial containment leak rate test specified by proposed Appendix J to 10 CFR 50, a special report shall, if that Appendix is adopted as an effective rule, be submitted to the Deputy Director for Reactor Projects, Directorate of Licensing, USAEC, Washington, D. C. 20545, and other containment leak rate tests specified by Appendix J that fail to meet the acceptance criteria of the appendix, shall be. the subject of special sununary technical reports pursuant to Section V.B of Appendix J.
"B. Report of Test Results The initial Type A test shall be subject of a summary technical report submitted to the Commission approximately 3 months after the conduct of the test. This report shall include a schematic arrangement of the leakage rate measurement system, the instrumentation used, the supple-mental test method, and the test program selected as applicable to the initial test, and all subsequent periodic tests. The report shall contain an analysis and interpretation of the leakage rate test d,ata to the extent necessary to demonstrate the acceptability of the contain-eent's leaka~e rate in meeting the acceptance criteria."
"For periodic tests,*leakage rate results of Type A, B, and C tests that meet the acceptance criteria of Sections III.A.7, III.B.3, and III.C.3 respectively shall be reported in the licensee's periodic operating
TS 6. 6-13 3-17-72 report. Leakage test results of Type A, B, and C tests that fail to meet the acceptance criteria of Sections III.A.7, III.B.3, and III.C.3 respectively shall be reported in a separate summary report*that includes an analysis and interpretation of the test data, the least-squares fit analysis of the test data, the instru-ment error analysis, and the structural conditions of the containment or components, if any, which contributed to the failure in meeting the acceptance criteria.
Results and analyses of the supplemental verification test employed to demonstrate the validity of the leakage rate test measurements shall also be included."
d. In-service Inspection Evaluation - A special summary technical report shall be submitted to the Deputy for Reactor Proj~cts, Directorate of Licensing, USAEC, Washington, D.C. 20545 after 5 years of operation. This report shall contain an evaluation of the results of the in-service inspection program and will qe reviewed in light of the technology available at that time.
e. Initial Containment Structural Test - A special summary technical report shall be submitted to the Deputy Director for Reactor Projects Directorate of Licensing, USAEC, Washington, D.C., 20545 within 3 months after completion of the test. This report will include a summary of the measurements of deflections, strains, crack width, crack patterns observed, as well as comparisons with predicted values of acceptanc~ criteria.

ML19094A226

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