Text

Amend 157 to License DPR-40,revising TS to Implement Administrative Changes
	ML20058F591
Person / Time
Site:	Fort Calhoun
Issue date:	11/22/1993
From:	Adensam E; Office of Nuclear Reactor Regulation
To:	;
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ML20058F587	List: ML20058F584; ML20058F591; ML20058F595;
References
	NUDOCS 9312080190
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g UNITED STATES E

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NUCLEAR REGULATORY COMMISSION

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WASHINGTON, D.C. 2055& 0001

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OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 157 License No. DPR-40 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by the Omaha Public Power District (the licensee) dated June 17, 1993, as supplemented October 8, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

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Accordingly, Facility Operating License No. DPR-40 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No. DPR-40 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 157, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

The license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Elinor E/Adensam, Assist nt Director for Region IV & V Reactors Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: November 22, 1993 i

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ATTACHMENT TO LICENSE AMENDMENT NO.157 FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Revise Appendix "A" Technical Specifications as indicated below. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

REMOVE PAGES INSERT PAGES 111 111 4

4 1-4 1-4 1-8 1-8 2-15a 2-15a 2-18 2-18 2-20 2-20 2-22 2-22 2-57e 2-57e 2-58 2-58 2-62 2-62 2-64 2-64 2-66 2-66 3-Oa 3-Da 3-1 3-1 3-15 3-15 3-18 3-18 3-21 3-21 3-40 3-40 3-54 3-54 3-56 3-56 3-60 3-60 3-61 3-61 3-62 3-62 3-76 3-76 3-77 3-77 3-79 3-79 3-84 3-84 4-4 4-4 5-1 5-1 5-4 5-4 5-5 5-5 5-6 5-6 5-7 5-7 5-8 5-8 5-Ba 5-9 5-9 5-12 5-12 5-17a 5-17a

o TABLE OF CONTENTS (Continued) i 4.3 Nuclear Steam Supply System (NSSS)................................ 4-3 4.3.1 Reactor Coolant System................................... 4 3 4.3.2 Reactor Core and Control.................................. 4 3 4.3.3 Emergency Core Cooling.................................. 4 3 4.4 Fuel Storage............................................... 4-4 4.4.1 New Fuel Storage....................................... 4-4 4.4.2 Spent Fuel Storage...................................... 4-4 4.5 Seismic Design for Class I Systems................................. 4-5 o

5.0 ADMINISTRATIVE CONTROLS...................................... 5-1 5.1 Responsibility............................................... 5-1 5.2 Organization............................................... 5-1 5.3 Facility Staff Qualifications.....................................5-la 5.4 Training................................................. 5-3 5.5 Review and Audit............................................ 5-3 5.5.1 Plant Review Committee (PRC).............................. 5-3 5.5.2 Safety Audit and Review Committee (SARC)...................... 5-5 1

5.6 Reportable Event Action........................................ 5-9 5.7 Safety Limit Violation......................................... 5-9 5.8 Procedures........................................... -.... 5-9 5.9 Reporting Requirements

.. 5 10 5.9.1 Routine Reports....................................... 5-10 5.9.2 Reportable Events...................................... 5-12 5.9.3 Special Reports....................................... 5-15 5.9.4 Unique Reporting Requirements.............................. 5 15 5.9.5 Core Operating Limits Report............................. 5-17a 5.10 Records Retention........................................... 5-18 5.11 Radiation Protection Program.................................... 5-19 5.12 DELETED 5.13 Secondary Water Chemistry..................................... 5-20 5.14 Systems Integrity............................................ 5-21 5.15 Post. Accident Radiological Sampling and Monitoring..................... 5-21 5.16 Radiological Effluents and Environmental Monitoring Programs............... 5-22 5.16.1 Radioactive Effluent Controls Program......................... 5-22 5.16.2 Radiological Environmental Monitoring Program................... 5-23 5.17 Offsite Dose Calculation Manual (ODCM)............................ 5-25 5.18 Process Con rol Program (PCP)

..................................5-26 6.0 INTERIM SPECIAL TECIINICAL SPECIFICATIONS........................ 6-1 6.1 Limits on Reactor Coolant Pump Operation............................ 6-1 6.2 Use of a Spent Fuel Shipping Cask.................................. 6-1 6.3 Auxiliary Feedwater Automatic Initiation Setpo:nt........................61 6.4 Operation With Less Than 75 % ofIncore Detector Strings Operable........................................... 6-1 iii Amendment No. 32.31,13,5 t,55,57, y, _e n, er, e n,ns,nn.. ".'.'.42 157

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3

'DEFTNrrIONS PROTECTIVE SYSTEMS (Continued)

Engineered Safety Feature Logicm The system which utilizes relay contact outputs from individual instrument channels to provide a dual channel signal to independently initiate the actuation of the engineered safety feature equipment. Two logic subsystems, termed A and B, are provided; each subsystem is composed of four channels wired to provide independent safety feature initiation signals on a 2-out-of-4 basis.

Decree of Redundancy The difference between the number of operable channels and the number of channels which when tripped will cause an automatic system trip.

INSTRUMENTATION SURVEILLANCE Channel Check A qualitative determination of acceptable operability by observation of channel behavior during l

normal plant operation. This determination shall where feasible, include comparison of the channel with other independent channels measuring the same variable.

Channel Functional Test Injection of a simulated signal into the channel to verify that it is operable, including any alarm and/or trip initiating action.

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, alarms, interlocks or trip, and shall be deemed to include the channel functional test.

Source Check Verification of channel response when the channel sensor is exposed to a radioactive source.

4 Amendment No. 86;157 4

1.0 SAFETY LIMITS AND LIMTITNG SAFETY SYSTEM SETITNGS l

r 1.2 Safety Limit. Reactor Coolant System Pressure i

Applicability Applies to the limit on reactor coolant system pressure.

Obiective To maintain the integrity of the reactor coolant system and to prevent the release of significant amounts of fission product activity to the containment.

Srecification The reactor coolant system pressure shall not exceed 2750 psia when fuel assemblies are located within the reactor vessel.

Basis The reactor coolant system sewes as s barrier to prevent radionuclides in the reactor l

coolant from reaching the containment atmosphere.m In the event of a fuel cladding failure, the reactor coolant system is the primary barrier against the release of fission products. Establishing a system pressure limit helps to assure the continued integrity of the reactor coolant system and fuel cladding. 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 B31.1 is 120% of design pressure.

Thus, the safety limit of 2750 psia (110% of the 2500 psia design pressure) has been established.*

The settings and capacity of the main steam safety valves (1000 - 1050 psia) *, the reactor high-pressure trip (s2400 psia) and the reactor coolant system safety valves (2500-2545 psia)

have been established to assure never reaching the reactor coolant system pressure safety limit. The initial hydrostatic test pressure was conducted at 3125 psia (125% of design pressure) to verify the integrity of the reactor coolant system.

Additional assurance that the nuclear steam supply system (NSSS) pressure does not exceed the safety limit is provided by setting the pressurizer power-operated relief valves, consistent with the reactor high pressure trip, and opening the steam system steam dump l

and bypass valves upon receipt of a turbine trip signal.*

1 1-4 Amendment No.157

1.0 SAFETY LfMITS AND LIMTITNG SAFETY SYSTEM SETTINGS 1.3 Limitine Safety System Settines. Reactor Protective System (continued)

(3)

Eigh Pressurizer Pressure - A reactor trip for high pressurizer pressure is provided in conjunction with the reactor and steam system safety valves to prevent reactor coolant system overpressure (Specification 2.1.6). In the event of loss of load without reactor trip, the temperature and pressure of the reactor coolant system would increase due to the reduction in the heat removed from the coolant via the steam generators. The power-operated relief valves are set to operate concurrently with the high pressurizer pressure reactor trip. This setting is below the nominal safety valve setting (2500 psia) to avoid unnecessary l

operation of the safety valves. This setting is consistent with the trip point assumed in the accident analysis.*

(4)

Thermal Margin /lew Pressure Trip - The thermal margin / low pressure trip is provided to prevent operation when the DNBR is less than 1.18, including allowance for measurement error. The thermal and hydraulic limits showm in the Thermal Margin / low Pressure 4 Pump Operation Figure, contained in the COLR, define the limiting values of reactor coolant pressure, reactor inlet temperature, axial shape index, and reactor power level which ensure that the thermal criteria

are not exceeded. 'Ihe low set point of a 1750 psia trips the reactor in the unlikely event of a loss-of-coolant accident.

The thermal margin / low pressure trip set poins shall be set according to the equation given in the COLR for the Thermal Margin / low Pressure Limit.

1-8 Amendment No. 8,20,32, -

47,70,77,92,141,157

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves (continued) a.

With one or more PORV(s) inoperable, within I hour either restore the PORV(s)

I to operable status or close the associatal block valve (s); otherwise, be in at least HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the i

following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i b.

With one or more block valve (s) inoperable, within I hour either restore the l

block valve (s) to operable status or close the block valve (s). Otherwise, be in at least HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Basis The highest reactor coolant system pressure reached in any of the accidents analyzed resulted from a complete loss of turbine generator load without simultaneous reactor trip while operating at 1500 MWt.* This pressure was less than the 2750 psia safety limit and the ASME Section III upset pressure limit of 10% greater than the design pressure.m The reactor is assumed to trip on a "High Pressurizer Pressure" trip signal.

The pressurizer safety valves are required to be calibrated to within i 1% of the specified setpoint value using ASME Section XI test methods. ASME Section XI requires that valves in steam service use steam as the test medium for establishing the setpoint. With the presence of a water-filled loop seal, establishing the valve setpoint with steam may result in in-situ valve actuation at pressures outside the i 1 % tolerance specified. Under transient conditions, it is expected that the valve (s) will actuate at no less than 4% below, nor greater than 6% above, the specified setpoint, which is within the tolerance assumed in the safety analysis.m The power-operated relief valves (PORV's) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.

To determine the maximum steam flow, the only other pressure relieving system assumed operational is the main steam safety valves.

Conservative values for all systems parameters, delay times and core moderator coefficients are assumed. Overpressure protection is provided to portions of the reactor coolant system which are at the highest pressure considering pump head, flow pressure drops and elevation heads.

If no residual heat were removed by any of the means available, the amount of steam l

. which could be generated at safety valve lift pressure would be less than half of the capacity of one safety valve. This specification, therefore, provides adequate defense against overpressurization when the reactor is suberitical.

2-15a Amendment No. 54,145,157

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2.0 LIMITING CONDITIONS FOR OPERATION

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2.2 Chemical and Volume Control System (Continued) dl.

The required BAST volume of Figure 2-11 can be combined between CH-II A and CH-11B when both tanks and LCV-218-3 are operable.

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d2.

When LCV-218-3 is inoperable or the SIRW tank volume is below Technical Specification 2.2(1) minimum, then each BAST must be operable and contain the required volume of Figure 2-11 corresponding to the requirements of the SIRW tank Technical Specification boron concentration.

d3.

When BAST CH-11B is inoperable, then BAST CH-11 A must be operable and contain the required volume of Figure 2-11 and LCV-218-3 must be operable.

d4.

When BAST CH-11 A is inoperable, then BAST CH-11B must be operable and contain the required volume of Figure 2-11 and I.CV-218-3 must be operable.

e.

Ixvel instruments on the inservice BAST shall be operable.

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(3) Modification of Minimum Requirements During power operation, the minimum requirements may be modified to allow any one of the following conditions to exist at any one time. If the system is not restored to meet the minimum requirements within the time period specified, the reactor shall be placed in the hot shutdown condition in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and in the cold shutdown condition within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

a.

One of the operable charging pumps may be removed from service provided two charging pumps are operable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

Both boric acid pumps may be out of service for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided that both BASTS meet the requirement of Figure 2-11.

c.

One level instrument channel on each inservice concentrated boric acid tank may be out of service for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4 d.

One BAST may be removed from service for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided that either of the conditions of 2.2(2)d3 or 2.2(2)d4 above is met.

Bliis The chemical and volume control system provides control of the reactor coolant system boron inventory.* This is normally accomplished by using any one of the three charging pumps in series with one of the two boric acid pumps. An alternate method of boration will be to use the charging pumps directly from the SIRW storage tank. A third method will be to depressurize and use the safety injection pumps. There are two sources of borated water available for

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injection through three different paths.

2-18 Amendment No. 43,103,131,157

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2.0 LIMITING CONDrrIONS FOR OPERATION l

2.3 Emercency Core Cooline System l-Applicability Applies to the operating status of the emergency core cooling system.

Objective f

To assure operability of equipment required to remove decay heat from the core.

Specifications (1) Minimum Recuirements The reactor shall not be made critical unless all of the following conditions are met:

l a.

The SIRW tank contains not less than 283,000 gallons of water with a l

boron concentration of at least the refueling boron concentration at a i

temperature not less than 50"F.

b.

One means of temperature indication (local) of the SIRW tank is operable.

c.

All four safety injection tanks are operable and pressurized to at least 240 psig with a tank level of at least 116.2 inches (67 %) and a maximum level l

of 128.1 inches (74%) with refueling boron concentration.

d.

One level and one pressure instrument is operable on each safety injection i

tank.

e.

One low-pressure safety injection pump is operable on each associated 4,160 V engineered safety feature bus.

f.

One high-pressure safety injection pump is operable on each associated 4,160 V engineered safety feature bus.

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g.

Both shutdown heat exchangers and three of four component cooling heat exchangers are operable.

L h.

Piping and valves shall be operable to provide two flow paths from the SIRW tank to the reactor coolant system.

i.

All valves, piping and interlocks associated with the above components l

and required to function during accident conditions are operable.

HCV-2914,2934,2974, and 2954 shall have power removed from the motor operators by locking open the circuit breakers in the power supply lines to the valve motor operators. FCV-326 shall be locked open.

2-20 Amendment No. 17,32,43,103,117, 119,133,141,157 I

2.0 LIMITING CONDITIONS FOR OPERAT.lON 2.3 Emercency Core Cooline System (Continued)

(3)

Protection Acainst Low Temperature Overpressurization The following limiting conditions shall be applied during scheduled heatups and cooldowns. Disabling of the HPSI pumps need not be required if the reactor vessel head, a pressurizer safety valve, or a PORV is removed.

Whenever the reactor coolant system cold leg temperature is below 32(PF, at least one (1) HPSI pump shall be disabled.

Whenever the reactor coolant system cold leg temperature is below 3127, at least two (2) HPSI pumps shall be disabled.

Whenever the reactor coolant system cold leg temperature is below 2717, all three (3) HPSI pumps shall be disabled.

In the event that no charging pumps are operable, a single HPSI pump may be made operable and utilized for boric acid injection to the core.

Basis Tne normal procedure for starting the reactor is to first heat the reactor coolant to near cyting temperature by running the reactor coolant pumps. The reactor is then made cnial by withdrawing CEA's and diluting boron in the reactor coolant. With this mode of start-up, the energy stored in the reactor coolant during the approach to criticality is sub:tantially equal to that during power operation and therefore all engineered safety features and auxiliary cooling systems are required to be fully operable.

The SIRW tank contains a minimum of 283,000 gallons of usable water containing a boron concentration of at least the refueling boron concentration. This is sufficient boron concentration to provide a shutdown margin of 5%, including allowances for uncertainties, with all control rods withdrawn and a new core at a temperature of 6CTF.*

The limits for the safety injection tank pressure and volume assure the required 13nount of water injection during an accident and are based on values used for the accident analyses. The minimum 116 2 inch level corresponds to a volume of 825 ft' and the maximum 128.1 inch level..orresponds to a volume of 895.5 ft'. Prior to the time the reactor is brought critical, the valving of the safety injection system must be checked for correct alignment and appropriate valves locked. Since the system is used for shutdown cooling, the valving will be changed and must be properly aligned prior to start-up of the reactor.

2-22 Amendment No. 17,39,43,47,54,74 "

400,103,133,141,E

2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.4 Power Distribution Limits (Continued)

In order for these objectives to be met, the reactor must be operated consistent with the operating limits specified for margin to DNB.

i The parameter limits given in (5) and the F,7, F,,7 and Core Power Limitations Figure provided in the COLR along with the parameter limits on quadrant tilt and control element assembly position (Power Dependert Insertion Limit Figure provided

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in the COLR) provide a high degree of assurance that the DNB overpower margin i

will be maintained during steady state operation.

The actions specified assure that the reactor is brought to a safe condition.

The reactor coolant pump differential pressure monitoring system may be used to measure flow.

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i 2-57e Amendment No. 32,57,l'1,157 J

2.0 LIMrrING CONDITIONS FOR OPERATION 2.11 Containment Building and Fuel Storace Buildine Crane Applicability Applies to the use of cranes over the reactor coolant system and the spent fuel storage pool.

Obiective To specify restrictions on the use of the overhead cranes in the Containment Building and the Auxiliary Building.

Stiecifications Use of the Containment Building and the Auxiliary Building overhead cranes shall be subject to the following limiting conditions.

(1) The containment polar crane shall not be used to transport loads over the reactor coolant system if the temperature of the coolant or steam in the pressurizer exceeds 225T.

(2) The Auxiliary Building crane shall not be used to move material over irradiated fuel in the fuel storage pool. If the crane interlocks are inoperable or bypassed, the crane operation will be under the direct control of a supenisor.

Basis Loads are not to be allowed over the pressurized reactor coolant system to preclude dropping objects which could rupture the boundary of the reactor coolant system allowing loss of coolant and over-heating of the core.

The Auxiliary Building crane is provided with an electrical interlock system that will normally prevent the trolley from moving over the storage pool.m This minimizes the possibility of dropping an object on the irradiated fuel stored in the pool and resulting in the release of radioactive products. The interlocks may be bypassed under strict administrative control to allow required movement of fuel and material over the pool.

The crane can be used over the equipment hatches and areas located in the north and west ends of the Auxiliary Buildig and over the railroad siding without the interlocks operable since a load, even if dropped, could not fall into the storage pool.

References (1) USAR, Section 14.18 l

2-58 Amendment No.157

2.0 LIMITING CONDITIONS FOR OPERATION 2.14 Encineered Safety Features System Initiation Instpmentation Settings (Continued)

(3) Containment High Radiation (Air Monitoring) (Continued)

Effluent radiation monitor isolation function setpoints will be calculated in accordance with the ODCM.

Process radiation monitor setpoints will be

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calculated in accordance with the tpplicable Chemistry Manual calibration i

procedure.

(4) Low Steam Generator Pressure A signal is provided upon sensing a low pressure in a steam generator to close the main steam isolation valves in order to minimize the temperature reduction in the reactor coolant system with resultant loss of water level and possible addition of react'.vity. The setting of 500 psia includes a.i22 psi uncertainty and was the setting used in the safety analysis.*

Closure of tht; MSIVs (and the bypass valves, along with main feedwater isolation and bypass talves) is accomplished by the steam generator isolation signal which is a logical combination of low steam generator pressure or high containment pressure.

As part of the AFW actuation logic, a separate signal is provided to terminate flow to a steam generator upon sensing a low pressure in that steam generator if the other steam generator pressure is greater than the pressure setting. This is done to minimiz.e the temperature reduction in the reactor coolant system in the event of a main steam-line break. The setting of 466.7 ps*4 includes a +31.7 psi uncertainty; therefore, a setting of 435 psia was used in the safety analysis.

(5) SIRW Tank Low Level Level switches are provided on the SIRW tank to actuate the valves in the safety injection pump suction lines in such a manner so as to switch the water supply from the SIRW tank to the containment samp for a recirculation mode of operation after a period of approximately 24 minutes following a safety injection signal. The switch-over point of 16 inches above tank bottom is set to prevent the pumps from running dry during the time required to stroke the valves and to l

l hold in reserve approximately 28,000 gallons of water of at least the refueling boron concentration. The USAR loss of coolant accident analysis

assumed the l

recirculation started when the minimum usable volume of 283,000 gallons had been pumped from the tank.

2-62 Amendment No. 5,32,43,65,S6,103, 408,133,141,152, 157

C TABL' 1-1 ENGINEERED SAFETY FEATURES SYSTEM INITIATION INSTRUMENT SETTING LIMITS Functio tal Unit Channel Settine Limit

1. liigh Containment Pressure

a. Safety injection A 5 psig

b. Containment Spray *

c. Containment isolation

d. Containment Air Cooler DBA Mode

e. Steam Generator Isolation m

2. Pressurizer lew/Iow Pressure

a. Safety injection 11600 psis m

b. Containment Spray

c. Containment Isolation

d. Containment Air Cooler DBA Mode

3. Containment fligh Radition Containment Ventilation Isolation in accordance with the opplicable Chemistry Manual calibration procedure

4. Low Steam Generator Pressure

a. Steam Line Isolation 1500 psia *

b. Auxiliary Feedwater Actuation 1466.7 psia

5. SIRW tow level Switches Recirculation Actuation 16 inches +0, -2 in. above tank bottom

6. 4.16 KV Emergency Bus Low

a. Ioss of Voltage (2995.2 + 104) volts i

Voltage 1 5.9'**8 seconds f Trip

b. Degraded Voltage A 3825.52 volts

} Tn,p i) Bus I A3 Side (4.8 f,.5) seconds J

2-64 A_ A _;No.4! M,?6, M,357

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2.0 LIMITING CONDITIONS FOR OPERATION 2.15' Instrumentation and Control Systms (Continued) ventilation isolation signals available if the containment ventilation isolation valves j

are closed. If after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time ofinitiating a hot shutdown procedure the inoperable engineered safety features or isolation functions channel has not been i

restored to operable status, the reactor shall be placed in a cold shutdown l

condition within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This specification applies to the high rate trip-wide range log channel when the plant is at or above 10d% power and is operating below 15% of rated power.

(3) In the event the number of channels of a particular system in service falls below the limits given in the columns entitled " Minimum Operable Channels" or l

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" Minimum Degree of Redundancy", except as conditioned by the column entitled i

" Permissible Bypass Conditions", the reactor shall be placed in a hot shutdown j

condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; however, operation can continue without containment -

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ventilation isolation signals available if the ventilation isolation valves are closed.

i If minimum conditions for engineered safety features or isolation functions are not j

met within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time of discovering loss of operability, the reactor shall be placed in a cold shutdown condition within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If the-number of operable high rate trip-wide range log channels falls below that given

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in the column entitled " Minimum Operable Chan.7els" in Table 2-2 and the i

d reactor is at or above 10 % power and at or below 15% of rated power, reactor critical operation shall be discontinued and the plant placed in an operational mode allowing repair of the inoperable channels before startup or reactor critical operation may proceed.

If, during power operation, the rod block function of the secondary CEA position

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indication system and rod block circuit are inoperable for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or the plant computer PDIL alarm, CEA group deviation alarm and the CEA sequencing function are inoperable for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the CEAs shall be withdrawn and maintained at fully withdrawn and the control rod drive system mode switch shall be maintained in the off position except when manual motion of CEA Group 4 is required to control axial power distribution.

(4) In the event that any of the following Alternate Shutdown Panel instrumentation

)

or control circuits become inoperable, either restore the inoperable component (s) to operable status within seven days, or be in hot shutdown within the next twelve hours. This specification is applicable in Modes 1 and 2.

1 i

Wide Range Imgarithmic Power (AI-212)

Source Range Power (AI-212)

Reactor Coolant Cold Leg Temperature (AI-185)

Reactor Coolant Hot Leg Temperature (AI-185)

Pressurizer Level (AI-185)

Volume Control Tank Level (AI-185) 2-66 Amendment No. 8,20,5#,55,88,125,157

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3.0 SURVEILLANCE REOUIREMENTS o

3.0.1 Each surveillance requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval.

3.0.2 The surveillance int.ervals are defined as follows:

Notation Illk Frequency S

Shift At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months D

Daily At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months W

Weekly At least once per 7 days BW Biweekly At least once per 14 days M

Monthly At least once per 31 days Q

- Quarterly At least once per 92 days SA Semiannual At least once per 184 days A

Annually At least once per 366 days R

Refueling At least once per 18 months l

P Start up Prior to Reactor Start up, if not completed in the previous week.

Exception to these intervals are stated in the individual Specifications.

3.0.3 The provisions of Specifications 3.0.1 and 3.0.2 are applicable to all codes and standards referenced within the Technical Specifications. The requirements of the Technical Specifications shall have precedence over the requirements of the codes and standards referenced within the Technical Specifications.

3.0.4 Failure to perform a Surveillance Requirement within the allowed surveillance interval, I

defimal by Specifications 3.0.1 and 3.0.2, shall constitute noncompliance with the GPERABILITY requirements for the corresponding Limiting Condition for Operation.

The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Surveillance Requirements do not have to be performed on inoperable equipment.

3-0a Amendment No. 122,129,157

3.0 SURVEILLANCE REOUIREMENTS i

3.1 Lngiamentation and Control Apolicability Applies to the reactor protective system and other critical instrumentation and controls.

r Obiective To specify the minimum frequency and type of surveillance to be applied to critical plant instrumentation and controls.

Specifications Calibration, testing and checking of instrument channels, reactor protective system and engineered safeguards system logic channels and miscellaneous instrument systems and controls shall be performed as specified in Tables 3-1 to 3-3a.

Luis 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 check supplements this type of built-in surveillance.

Based on the District's experience in operation of conventional power plants and on reported nuclear plant experience, a checking frequency of once-per-shift is deemed adequate for reactor and steam system instrumentation. Calibrations are performed to ensure the presentation and acquisition of accurate information.

The power range safety channels are calibrated daily against a calorimetric balance standard to account for errors induced by changing rod patterns and core physics parameters.

Other channels, subject only to the " drift" errors, can be expected to remain within acceptable tolerances if recalibration is performed on a refueling frequency.

l i

3-1 Amendment No. 9,122,125, 157

I

- % t, TABLE 3-3 (Continued)

BilNinfUhl FREOUENCIES FOR CliECKS. CALIBRATIONS AST) TESTING OF bilSCELLANEOUS INSTRUhlENTATION AND CONTROLS Surveillance Channel Descriotion Function Freu mi.cv Surveillance afethod 8.

Dropped CEA Indication n.

Test R

a.

Insert a negative rate of change power signal to all four Power Range Safety Channels to test alarm.

b.

Test R

b.

Insert CEA's below lower electrical limit to test dropped CEA alarm.

9.

Calorimetric Instrumen-a.

Calibrate R

a.

Apply known d!p to feedwater flow tation sensors.

10.

Control Room Ventilation a.

Test R

a.

Check damper operation for DBA mode.

b.

Test R

b.

Check contml room for positive presure.

I 1.

Containment llumidity a.

Test R

a.

Place sensor in a known high humidity Detector stumpkre.

Known pressure of 265 psia applied to 12.

Interlocks-Isolation Valves a.

Test R

a.

on Shutdown Cooling Line both pressure transmitters.

l 13.

Contml Room Thermome-m.

Test R

a.

Compare readmg with calibrated ter thermometer. If not within 3. 27, replace.

3-15 Amendment No. !f,32,!??,157 i

l l

TABLE 3-4 MINIMUM FREOUENCIES FOR SAMPLING TESTS Type of Measurement Sample and Analysis and Analysis Freauenev 1.

Reactor Coolant (a) Power Operation (1)

Gross Radioactivity 1 per 3 days (Operating Mode 1)

(Gamma emitters)

(2)

Isotopic Analysis for (i) 1 per 14 days DOSE EQUIVALENT I-131 (ii)

I per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

whenever

}

the radioactivity exceeds 1.0 pCilgm DOSE EQUIVALENT I-131.

(iii)

I sample between 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following a thermel power change exceeding 15% of the rated thermal power within a 1-bour period.

(3)

$ Determination 1 per 6 months *

(4)

Dissolved oxygen 1 per 3 days and chloride (b)1101 Standby (1)

Gross Radioactivity 1 per 3 days (Operating Mode 2)

(Gamma emitters) liot Shutdows (2)

Isotopic Analysis for (i)

I per 8 hourf"whenever l

(Operating Mode 3)

DOSE EQUIVALENT I-131 the radioactivity exceeds i

1.0 pCi/gm DOSE EQUIVALENT I-131.

(ii)

I sample between 2-8 hours following a thermal power change exceeding 15% of the i

rated thermal power change exceeding 15%

i of the rated thermal power within a 1-hour period.

(3)

Dissolved oxyfen 1 per 3 days and chloride i

3-18 Amendment No. 2S,67,121,1%157 i

3.0 SURVT,7LLANCE REOUIREhfENTS 3.3 Reactor Coolant System and Other Components Subiect to AShfE XI Boiler & Pressure Vessel Code Inspection and Testing Surveillance Aeolicability Applies to in-service surveillance of primary system components and other components subject to inspection and testing according to AShiE XI Boiler & Pressure Vessel Code.

Obiective To ensure the integrity of the reactor coolant system and other components subject to inspection and testing according to ASME XI Boiler & Pressure Vessel Code.

Specifications (1)

Surveillance of the AShiE Code Class 1, 2 and 3 systems, except the steam generator tubes mspection, should be covered by AShiE XI Boiler & Pressure Vessel Code, a.

In-service inspection of ASME Code Class 1, Class 2, and Class 3 components and in-service testing of ASME Code Class 1, Class 2, and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, as required by 10 CFR Part 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a (g)(6)(i).

b.

Surveillance of the reactor coolant pump flywheels shall be performed as indicated in Table 3-6.

c.

A surveillance program to monitor radiation-induced changes in the mechanical and impact properties of the reactor vessel materials shall be maintained in accordance with 10 CFR Part 50 Appendix H.m (2)

Surveillance of Reactor Coolant System Pressure Isolation Valves a.

Periodic leakage testing

on each valve listed in Table 2-9 shall be l

accomplished prior to entering the power operation mode every time the plant is placed in the cold shutdown

To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved 1

procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

i 3-21 Amendment No. 46 76,104,142, 157 7

1 3.0 SURVEILLANCE REOUIREMENTS l

3.5-Containment Tests (Continued)

'Ihe total measured leakage rate at a pressure of 60 psig shall be less than

]

0.75 L,.

If local leakage measurements are taken to effect repairs in j

order to meet 0.75 L, acceptance criteria, these measurements shall be taken at a pressure of 60 psig.

If two consecutive Type A tests fail to meet the acceptance criteria, notwithstanding the requirements of the testing frequency, a Type A test

(

shall be performed on a refueling frequency until two consecutive Type l

A tests meet the acceptance criteria, after which time the normal testing -

frequency schedule may be resumed.

e.

Testine Fu s:UCDCY l

A set of three Type A tests shall be performed, at approximately equal intervals during each 10 year service period. The third test of each set shall be conducted when the plant is shutdown for the 10-year in-service j

inspections.

i The performance of Type A tests shall be limited to periods when the plant facility is non+perational and secured in the shutdown condition under administrative control and in accordance with the safety procedures l

defm' ed in the license.

)

i (4)

Containment Penetrations Leak Rate Tests (Tvoe B Tests) a.

Introduction Type B tests are intended to detect local leaks and to measure leakage across each pressure-containing or leakage limiting boundary for the containment penetrations.

b.

Test Methods l

Type B tests shall be performed by local pneumatic pressurization of the l

containment penetrations, either individually or in groups,' at a pressure.

of 60 psig.

Examination shall be performed by halide leak-detection method or by f

other equivalent test methods such as measurement'of the rate of makeup-required to maintain the test volume at 60 psig.

i i

3-40 Amendment No. 95,151, 157

-l

,l

v.

3.0 SURVEILLANCE REOUIREMENTS 3.6 Safety iniection and Containment Cooline Systems Tests i

Applicability Applies to the safety injection system, the containment spray system, the containment cooling system and air filtration system inside the containment.

Obiective To verify that the subject systems will respond promptly and perform their intended functions, if required.

Soecifications (1)

Safety Iniection System System tests shall be performed on a refueling frequency. A test safety feature l

actuation signal will be applied to initiate operation of the system. The safety injection and shutdown cooling system pump motors may be de-energized for this portion of the test.

A second overlapping test will be considered satisfactory if control board indication and visual observations indicate all components have received the safety feature actuation signal in the proper sequence and timing (i.e., the appropriate pump breakers shall have opened and closed, and ali valves shall have completed their travel).

(2)

Containment Spray System a.

System tests shall be performed on a refueling frequency. The test shall l

be performed with the isolation valves in the spray supply lines at the containment blocked closed. Operation of the system is initiated by tripping the normal actuation instrumentation.

b.

At least every five years the spray nozzles shall be verified to be open.

c.

The test will be considered satisfactory if:

(i)

Visual observations indicate that at least 264 nozzles per spray header have operated satisfactorily.

(ii)

No more than one nozzle per spray header is missing.

d.

Undisturbed samples of Trisodium Phosphate Dodecahydrate (TSP) that have been exposed to the same environmental conditions as that in' the mesh baskets shall be tested on a refueling frequency by:

l 3-54 Amendment No. 44,121, 157

~

s 3.0 SURVEILLANCE REOUIREMENTS 3.6 Safety Iniection and Containment Cooline Systems Tests (Continued) g.

Initial laboratory batch tests of charcoal adsorbers shall show190% n.dioactive methyl iodide removal when tested under conditions of195% relative humidity, 1250'F, within 120% of design face velocity and 5 to 15 mg/m inlet methyl 8

%dide concentration. A sample shall be removed for laboratory testing on a refueling frequency or during the next shutdown following 4300 hours0.0498 days 1.194 hours 0.00711 weeks 0.00164 months of charcoal filtering unit operation and following significant painting, fire or chemical release in any ventilation zone communicating with the system. The results of sample tests shall show 185% radioactive methyl iodide removal under the test conditions given above.

Basis The safety injection system and the containment cooling system are principal plant safeguards that are not operated during normal reactor operation.

Complete systems tests cannot be performed when the reactor is operating because a safety injection signal causes containment isolation and a containment spray system test-requires the system to be temporarily disabled. The method of assuring operability of these systems is, therefore, to combine systems tests to be performed during refueling shutdowns in addition to more frequent component tests which can be performed during reactor operation.

The refueling shutdown tests demonstrate proper automatic operation of the safety injection and containment spray systems. A test signal is applied to initiate automatic action and verification made that the components receive the safety injection actuation signals in the proper sequence. The test demonstrates the operation of the valves, pump circuit breakers, and automatic circuitry.mm During reactor operation, the instrumentation which 13 depended on to initiate safety injection and containment spray is generally checked c'aily and the initiating circuits are tested monthly. In addition, the active components (pumps and valves) are to be tested every three months to check the operation of the starting circuits and to verify that the pumps are in satisfactory running order. The test interval of three months is based on the judgement that more frequent testing would not significantly increase the reliability (i.e., the probability that the component would operate when required), yet more frequent tests would result in increased wear over a long period of time. Verification 3-56 Amendment No. Mr24 157 7

3.0 SURVEILLANCE REOUIREMENTS 3.7 Emergency Power System Periodic Tests (Continued) d.

During refueling shutdowns the correct function of all D.C. emergency transfer switches shall be demonstrated by manual transfer of normal D.C.

supply breakers at the 125 volt D.C. distribution panels.

(3)

Emercency Lichting The correct functioning of the emergency lighting system required for plant safe shutdown shall be verified at least once each year.

(4) 13.8 Kv Transmission Line

'Ihe 13.8 Ky transmission line will be energized and loaded to minimum shutdown requirements at each refueling outage following installation.

l Basis The emergency power system provides power requirements for the engineered safety features in the event of a DBA. Each of the two diesel generators is capable of supplying minimum required safety feature equipment from independent buses. This redundancy is a factor in establishing testing intervals. The monthly tests specified will demonstrate operability and load capacity of each diesel generator. These tests are conducted to meet the objectives of NRC Generic Letter 84-15 regarding the issue of reductions in cold fast starts. For this reason, the test verifying a 10 second start will be conducted from ambient conditions once per 184 days for each diesel. Other monthly tests will allow for manufacturer's recommended warm-up to reduce the mechanical stress and wear on the diesel engines. The fuel supply and various controls are continuously monitored and alarmed for off-normal conditions. Automatic starting on l

loss of off-site power and automatic load shedding, diesel connection, and loading will be verified on a refueling frequency. At the same intervals, capability will be verified l

for manual emergency control of these functions from the diesel and switch-gear rooms.

Considering system redundancy, the specified testing intervals for the station batteries should be adequate to detect and correct any malfunction before it can result in system malfunction. Batteries will deteriorate with time, but precipitous failure is extremely unlikely. The surveillance specified is that which has been demonstrated over the years to provide an indication of a cell becoming unserviceable long before it fails.

References (1) USAR, Section 7.3.4.2 (2) USAR, Section 8.4.1 (3) USAR, Section 8.3.4 (4) USAR, Section 8.4.2 3-60 Amendment No. 24,111, 157

n 3.0 SURVEILLANCE REOUTREMENTS 3.8 Main Steam Isolation Valves l

Apolicability Applies to periodic testing of the main steam isolation valves.

Obiective To verify the ability of the main steam isolation valves to close upon signal.

Specifications The operation of the main steam isolation valves shall be tested during each refueling outage to demonstrate a closure time of four seconds or less under no-flow conditions.m Basis The main steam isolation valves serve to limit an excessive reactor coolant system cooldown rate and resultant reactivity insertion following a main steam break incident.

Their ability to close upon signal will be verified at each scheduled refueling outage.

References (1) USAR, Section 10.3 i

3-61 Amendment No. 24 157 7

3.0 SURVETLLANCE REOUTREMENTS 3.9 Auxiliary Feedwater System Applicability Applies to periodic testing requirements of the turbine-driven and motor-driven auxiliary feedwater pumps.

l t

Obiective To verify the operability of the auxiliary feedwater (AFW) system and its ability to respond properly when required.

Specifications (1)

The position of valves necessary to ensure auxiliary feedwater flow to the steam generators shall be verified by a monthly inspection. Anytime maintenance is performed on the auxiliary feedwater system which alters valve alignments, an i

operator shall check that the AFW system valves are properly aligned, to ensure AFW flow to the steam generators, and a second operator shall independently verify proper valve alignment.

(2)

The operability of the motor-driven auxiliary feedwater pump and the steam turbine-driven auxiliary feedwater pump shall be confirmed at least monthly. The discharge pressure will be verified to be at least 40 psig above the steam generator pressure at rated steam flow.

(3)

The operability of auxiliary feedwater pumps' steam generator level regulating i

valves HCV-Il07A, HCV-Il07B, HCV-1108A, HCV-Il08B, and auxiliary feedwater cross-tie valve HCV-1384 shall be confirmed at least every three months.

(4)

Following cold shutdown and prior to raising the reactor coolant temperature

{

above 300 F, the motor-driven auxiliary feedwater pump shall be tested to verify the normal flow path for auxiliary feedwate-to the steam generators.

(5)

On a refueling frequency:

a.

Verify that each automatic valve in the flow path actuates to its correct position upon receipt of each auxiliary feedwater actuation test signal.

J b.

Verify that each auxiliary feedwater pump starts as designed automatically l

upon receipt of each auxiliary feedwater actuation test signal.

i l

3-62 Amendment No. 49,54,90,157

3.0 SURVEILLANCE REOUIREMENTS 3.13 RADIOACTIVE MATERIAL SOURCES SURVEILLANCE l

Applicability Applies to leakage testing of byproduct, source, and specia! nuclear radioactive material sources.

L Obiective To assure that leakage from byproduct, source, and special nuclear radioactive material sources does not exceed allowable limits.

Specification Tests for leakage and/or contamination shall be performed by the licensee or by other persons specifically authorized by the NRC or an agreement State, as follows:

1.

Each sealed source, except stanup sources subject to core flux, containing radioactive material, other than Hydrogen 3, with a half-life greater than thirty days and in any form other than gas shall be tested for leakage and/or contamination at intervals of six months.

l 2.

The periodic leak test required does not apply to sealed scurces that are stored and not being used. The sources excepted from this test shall be tested for leakage prior to any use or transfer to another user unless they have been leak tested within six months prior to the date of use or transfer. in the absence of a certificate from a transferor indicating that a test has been made within six months prior to the transfer, sealed sources shall not be put into use until tested.

3.

Startup sources shall be leak tested prior to and following any repair or r

maintenance and before being subjected to core flux.

3-76 Amendment No. 41,122,157

3.0 SURVEILLANCE REOUTREMENTS 3.14 Shock Suppressors (Snubbers)

Applicability This specification applies to all safety-related snubbers.

Snecifications (1)

All hydraulic snubbers shall be visually inspected. As used in this specification,

" type of snubber" shall mean snubbers of the same design and manufacturer, irrespective of capacity. This inspection shall include, but not necessarily be limited to, inspection of the hydraulic fluid reservoir, fluid connections, and linkage connections to the piping and anchor to verify snubber operability. In those locations where snubber movement can be manually induced without disconnecting the snubber, verify that the snubber has freedom of movement and is not frozen up.

Snubbers which appear inoperable as a result of visual inspections shall be classified as unacceptable and may be reclassified acceptable for the purpose of establishing the next visual inspection interval, provided that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespective of type that may be generica'ly susceptible; and (2) the affected snubber is functionally tested in the as-found condition and determined OPERABLE per functional testing acceptance criteria. n11 snubbers found connected to an inoperable common hydraulic fluid reservoir, hall be counted as unacceptable for determining the next inspection interval. A review and evaluation shall be performed and documented tojustify continued operation with an unacceptable snubber. If continued operation cannot be justified, the snubber shall he declared inoperable and the ACTION requirements shall be met. Visual inspections shall be performed in accordance with Table 3-14.

(2)

On a refueling frequency and subject to the conditions below:

l (a)

A representative sample (88) of hydraulic snubbers shall be functionally tested either in-place or in a bench test, i

e l

i 3-77 Amendment No. 27,59,105;445,157

3.0 SURVEILLANCE REOUIREMEN'IS 3.14 Shock Suppressors (Snubbers) (Continued)

If any snubber selected for functional testing either fails to lockup or fails to move, i.e., is frozen in place, the cause will be evaluated. If the cause is a manufacturer or design deficiency, appropriate action shall be taken for snubbers of the same design subject to the same defect to determine if any more defects exist. This testing requirement shall be independent of the requirements stated above for snubbers not meeting the functional test acceptance criteria.

For any snubber (s) found locked up during normal operation or found inoperable following a seismic event, an engineering evaluation shall be performed on the components which are supported by the snubber (s).

The purpose of this engineering evaluation shall be to determine if the components supported by the snubber (s) were adversely affected by theinoperability of the snubber (s)in order to ensure that the supported component remains capable of meeting the designed service. If the engineering evaluation shows the components to be capable of meeting the designed service without the failed snubber, that snubber may be deleted from service per Specification 2.18(4).

(3)

Snubber Service Life Monitorine A record of the service life of each snubber, the date at which the designated service life commences and the installation and maintenance records on which the designated service life is based shall be maintained as required by Specification 5.10.2.m. On a refueling frequency, the installation and maintenance record for l

each snubber ahtll be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review. If the indicated service life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be re-evaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This re-evaluation, replacement or reconditioning shall be indicated in the records.

Enis All safety snubbers shall be operable to ensure that the structural integrity of the reactor coolant system and all other safety-related systems is maintained during and following a seismic or other event initiating dynamic loads. Snubbers excluded from this inspection program are those installed on non-safety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.

The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. The required inspection interval will be based on Table 3-14.

1 3-79 Amendment No. 27,59,10m 157 l

3.0 SURVEILLANCE REOUIREMENTS 3.16 Recirculation Heat Removal System Intecrity Testine Applica'2ility Applies to determination of the integrity of the residual heat removal systems and associated components.

Obiective To verify that the leakage from the residual heat removal system components is within acceptable limits.

t Specifications (1) a. The portion of the shutdown cooling system that is outside the containment, and the piping between the containment spray pump suction and discharge isolation valves, shall be examined for leakage at a pressure no less than 250 psig. This shall be performed on a refueling frequency.

l

b. Piping from valves HCV-383-3 and HCV-383-4 to the suction isolation valves of the low pressure safety injection pumps and containment spray pumps and to the high pressure safety injection pumps shall be examined for leakage at a pressure no less that 82 psig. This shall be performed at the testing frequency specified in (1)a. above.

c. The portion of the high pressure safety injection (HPSI) system that is located outside the containment and downstream of the HPSI pumps shall be examined for leakage when subjected to the discharge pressure of a HPSI pump operating in the minimum recirculation mode. This test shall be performed at the frequency specified in (1)a. above. The leakage contribution from this section shall be the observed leakage from this piping at the test pressure nultiplied by the square root of the ratio 1500/P, where P is the test discharge pressure (in psig) of the operating HPSI pump.

d. Visual inspection of the system's components shall be performed at the frequency specified in (1)a. above to uncover any significant external leakage to atmosphere (including leakage from valve stems, flanges, and pump seals). The leakage shall l

be measured by collection and weighing or by any other equivalent method.

(2) a. The sum of leakages from section (1)a, (1)b, and (1)c above shall not exceed 1243 cc/ hour.

i

b. Repairs shall be made as required to maintain leakage within the acceptable limits.

3-84 Amendment No. 87,122,128,135, 157

t i

4.0 DESIGN FEATURES 4.4 Fuel Storace 4.4.1 New Fuel Stomee The new unirradiated fuel bundles will normally be stored in the dry new fuel storage l

rack with an effective multiplication factor of less than 0.9. The new fuel storage l

rack is located 18'-9" above the main floor of Room 25A which provides for adequate drainage and precludes flooding of the new fuel storage rack.

New fuel may also be stored in shipping containers or in the spent fuel pool racks which have a maximum effective multiplication factor of 0.95 with Fort Calhoun Type C fuel and unborated water.

The new fuel storage racks are designed as a Class I structure.

4.4.2 Spent Fuel Storage Irradiated fuel bundles will be stored prior to off-site shipment in the stainless steel

{

lined spent fuel pool. The spent fuel pool is normally filled with borated water with a concentration of at least the refueling boron concentration.

The spent fuel racks are designed u a Class I structure.

l Normally the spent fuel pool cooling system will maintain the bulk water temperature of the pool below 12WF. Under other conditions of fuel discharge, the fuel pool

{

water temperature is maintained below 14TF.

The spent fuel racks are designed and will be maintained such that the calculated effective multiplication factor is no greater than 0.95 (including all known uncertainties) assuming the pool is flooded with unborated water. The racks are i

divided into 2 regions. Region 1 and 2 cells are surrounded by Boral. Acceptance-criteria for fuel storage in Regions 1 and 2 are delineated in Section 2.8 of these l

Technical Specifications.

?

1 i

i 4-4 Amendment No. 43,43,75,103, i

133,141,155, 157

5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1.1 The Manager - Fort Calhoun Station shall be responsible for overall facility operation i

and shall delegate in writing the succession to this responsibility during his absence.

5.2 Organization 5.2.1 Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.

a.

Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organizational charts, functional descriptions of departmental responsibilities and relationships, andjob i

descriptions for key personnel positions, or in equivalent forms of documentation.

These requirements shall be documented in the USAR.

b.

The Manager - Fort Calhoun Station shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

c.

The Senior Vice President - shall have corporate responsibility for overall plant responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

t d.

The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

5.2.2 Plant Staff He plant staff organization shall be as described in Chapter 12 of the USAR and shall function as follows:

a.

The minimum number and type of licensed and unlicensed operating personnel required onsite for each shift shall be as shown in Table 5.2-1.

l 5-1 Amendment No. 53,76,78,101,115; 449 142;157 7

5.0 ADMINISTRATWE CONTROLS Responsibilities 5.5.1.6 The Plant Review Committee shall be responsible for:

l a.

Review of (1) Administrative Controls Standing Orders and changes thereto, (2) procedures required by Specification 5.8 and requiring a 10 CFR 50.59 safety evaluation, and (3) proposed changes to procedures required by Specification 5.8 and requiring a 10 CFR 50.59 safety

[

evaluation; b.

Review of all proposed tests and experiments that affect nuclear safety.

c.

Review of all proposed changes to the Technical Specifications.

d.

Review of all proposed changes to the Core Operating Limits Repon.

l e.

Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety, f.

Investigation of all violations of the Technical Specifications and shall j

prepare and forward a repon covering evaluation and recommendations to prevent recurrence to the Division Manager - Nuclear Operations and to the Chairperson of the Safety Audit and Review Committee.

j g.

Review of facility operations to detect potential safety hazards.

l i

h.

Performance of special reviews and investigations and reports thereon as i

requested by the Chairperson of the Safety Audit and Review Committee.

l l

i.

Review of the Site Security Plan and implementing procedures and shall submit recommended changes to the Chairperson of the Safety Audit and l

[

Review Committee.

1 j.

Review of the Site Emergency Plan and implementing procedures and shall submit recommended changes to the Chairperson of the Safety Audit l

and Review Committee.

k.

Review of all Reportable Events.

Authority l

5.5.1.7 The Plant Review Committee shall:

t a.

Recommend in writing to the Manager - Fort Calhoun Station approval or disapproval of items considered under 5.5.1.6(a) through (e) above.

5-4 Amendment No. 9,19,84,99,115,141, H b 157 j

5.0 ADMINISTRATIVE CONTROIS 5.5.1.7 b.

Render determinations in writing with regard to whether or not each item considered under 5.5.1.6(b) through (f) above constitutes an unreviewed safety question.

c.

Provide immediate written notification to the Division Manager - Nuclear Operations and the Safety Audit and Review Committee of disagreement between the Plant Review Committee and the Manager - Fort Calhoun Station; however, the Manager - Fort Calhoun Station shall have responsibility for resolution of such disagreements pursuant to 5.1.1 above.

Records 5.5.1.8 The Plant Review Committee shall maintain written minutes of each meeting and copies shall be provided to the Division Manager - Nuclear Operations and Chairperson of the Safety Audit and Review Committee.

l

~

5.5.2 Safety Audit and Review Committee (SARC)

Function 5.5.2.1 The Safety Audit and Review Committe shall function to provide the independent review and audit of designated activities in the arcas of:

a.

nuclear power plant operation b.

nuclear engineering c.

chemistry and radiochemistry d.

metallurgy e.

instrumentation and control f.

radiological safety g.

mechanical and electrical engineering h.

quality assurance Composition 5.5.2.2 The Safety Audit and Review Comm'ttee shall be mmposed of:

Chairperson: Division Manager - Atc>ar Scr ' es l

Member:

Senior Vice President Member:

Vice President

[

Member:

Division Manager - Nuclear Operations Member:

DMsion Manager - Production Engineering Member:

Manager - Fort Calhoun Station Member:

Other qualified OPPD personnel or consultants as required and as determined by the SARC Chairperson 5-5 Amendment No. SS,93,99,101, 445,119,132,141,149, 157

F i

e i

~

5.0 ADMINISTRATIVE CONTROLS f

Altemates 5.5.2.3 Alternate members shall be appointed in writing by the Chairperson of the Safety l

I Audit and Review Committee to serve on a temporary basis; however, no more than two alternates may panicipate in the Safety Audit and Review Committee activities at any one time.

Consultants l

5.5.2.4 Consultants shall be utilized as determined by the Safety Audit and Review Committee Chairperson to provide expen advice to the Safety Audit and Review l

l Committee.

Meeting Frecuency 5.5.2.5 The Safety Audit and Review Committee shall meet at least once every six months.

Ooorum i

5.5.2.6 A quorum of the Safety Audit and Review Committee shall consist of the Chairperson or his designated alternate and a majority of the Safety Audit and l

Review Committee members including altemates. No more than a minority of the quorum shall have line responsibility for the operation of the nuclear plant.

Feview 5.5.2.7 The Safety Audit and Review Committee shall review:

The safety evaluations for 1) procedures, equipment or systems and 2) a.

tests or experiments completed under the provision of section 50.59,10 CFR, to verify that such actions did not constitute an unreviewed safety question.

b.

Proposed changes to procedures, equipment or systems which involve an unreviewed safety c,uestion as defined in section 50.59,10 CFR.

l f

t 5-6 Amendment No. 9,19,60, 157 t

s 5.0 ADMINISTRATTVE CONTROLS Proposed tests or experiments which involve an unreviewed safety 5.5.2.7 c.

question as defined in 10 CFR 50.59.

d.

Proposed changes in Technical Specifications or licenses.

Violations of applicable statutes, codes, regulations, orders, Technical c.

Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.

f.

Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.

g.

All Reportable Events.

h.

Any indication of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.

i.

Reports and meeting minutes of the Plant Review Committee.

The Chairperson of the Safety Audit and Review Committee (SARC) may l

designate subgroups, special working committees, or audit teams as he deems necessary in order to carry out the responsibilities of the SARC.

These subgroups, committees, or audit teams will perform the S ARC responsibilities and report on their activities for review at the next regularly scheduled SARC meeting following any group's action.

Audit 5.5.2.8 Audits of facility activities shall be performed under the cognizance of the Safety Audit and Review Committee. These audits shall encompass:

The conformance of facility operation to provisions contained within the l

a.

Technical Specifications and applicable license conditions at least once per year.

b.

The performance, training and qualifications of the entire facility staff at least once per year.

The results of actions taken to correct deficiencies occurring in facility l

c.

equipment, structures, systems or method of operation that affect nuclear safety at least once per six months.

d.

The performance of activities required by the Quality Assurance Program to meet the criteria of Appendix B,10 CFR Part 50, at least once per two years.

5-7 Amendment No. 60;997157

5.0 ADMINISTR ATIVE CONTROLS 5.5.2.8 e.

The Fort Calhoun Station Emergency Plan and implementing procedures at least once every twelve months.

f.

The Site Security Plan and implementing procedures at least once every twelve months.

g.

The Safeguards Contingency Plan and implementmg procedures at least once every twelve months.

h.

The Radiological Effluent Program including the Radiological Environmental Monitoring Program and the results thereof, the Offsite Dose Calculation Manual and implementing procedures, and the Process -

Control Program for the solidifications of radioactive waste at least once -

per 2 years.

i.

Any other area of facility operation considered appropriate by the Safety Audit and Review Committee or the Senior Vice President.

j.

An independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either qualified off-site licensee personnel or an outside fire protection firm.

k.

An inspection and audit of the fire protection and loss prevention program by an outside qualified fire consultant shall be performed at intervals no greater than 3 years.

Authority 5.5.2.9 The Safety Audit and Review Committee shall report to and advise the Senior Vice President on those areas of responsibility specified in Sections 5.5.2.7 and 5.5.2.8.

Records 5.5.2.10 Records of Safety Audit and Review Committee activities shall be prepared, approved and distributed as indicated below:

a.

Minutes of each Safety Audit and Review Committee meeting shall be prepared, approved and forwarded to the Senior Vice President within 30 days following each meeting.

l b.

Reports of reviews encompassed by Section 5.5.2.7e, f, g, h, and i above shall be prepared, approved and forwarded to the Senior Vice President within 30 days following completion of the review.

j c.

Audit reports encompassed by Section 5.5.2.8 above shall be forwarded to the Senior Vice President and to the responsible management positions l

designated by the Safety Audit and Review Committee within 30 days after completion of the audit.

5-8 Amendment No. 93d01,115,119,132,157

5.0 ADMINISTRATIVE CONTROIS 5.6 Reoortable Event Action 1

5.6.1 The following actions shall be taken in the event of a REPORTABLE EVENT:

I The Commission shall be notified pursuant to the requirements of 10 CFR a.

50.72, if applicable.

I b.

Each Reportable Event shall be reviewed by the Plant Review Committee and submitted to the Chairperson of the Safety Audit and Review l

Committee and the Division Manager - Nuclear Operations.

Submit reports of Reportable Events pursuant to the requirements of c.

Specification 5.9.2.

5.7 Safety Limit Violation 5.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a.

The unit shall be place.

at sst HOT SHUTDOWN within I hour.

b.

The Safety Limit Violations shall be reported to the Division Manager -

Nuclear Operations and the Chairperson of the Safety Audit and Review l

Committee (SARC) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c.

A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the Plant Review Committee. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

d.

The Safety Limit Violation Report shall be submitted to the Chairperson l

of the Safety Audit and Review Committee and the Division Manager -

Nuclear Operations within 14 days of the violation.

5.8 Erocedures 5.8.1 Written procedures and administrative policies shall be established, implemented and maintained that meet or exceed the minimum requirements of sections 5.1 and 5.3 of ANSI N18.7-1972 and Appendix A of USNRC Regulatory Guide 1.33 except as provided in 5.8.2 and 5.8.3 below.

5.8.2 Each procedure of Specification 5.8.1, and changes thereto, and any other procedure or procedure change that the Manager - Fort Calhoun Station determines to affect nuclear safety, shall be reviewed and approved as described below, prior to implementation.

5-9 Amendment No. 9,19,38,S4,"),

H6d49;157

5.9.1 Continued work and job functions,2' e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance),

waste processing, and refueling outages. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge 7

measurements. Small exposures totallbg less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

c.

Monthly Operating Report. Routine reports of operating statistics and shutdown i

experience shall be submitted on a monthly basis to the U.S. Nuclear Regulatory i

Commission, Document Control Desk, Mail Station PI-137, Washington, D. C.

20555, with a copy to the appropriate Regional Office, no later than the fifteenth l

of each month following the calendar month covered by the report. This monthly report shall also include a statement regarding any challenges or failures to the pressurizer power operated relief valves or safety valves occurring during the su', ject month.

5.9.2 Reportable Event A Licensee Event Report (LER) shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Mail Station PI-137, Washington, D. C. 20555 with a copy to Region IV of the NRC, within 30 days after discovery of any event meeting the requirements of 10 CFR Part 50.73.

5 This tabulation supplements the requirements of f 20.407 of 10 CFR Part 20.

5-12 Amendment No. 9,24,35,85,99,119,157 (Next page is 5-15)

....m i

5.9.5 Core Oocrating Limits Reoort a

a.

Core Operating limits shall be established and documented in the Core Operating Limits Report (COLR) before each reload cycle or any remaining part of a reload l

cycle.

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approvM by the NRC as follows:

1.

OPPD-NA-8301-P-A, " Reload Core Analysis Methodology Overview" (Latest NRC approved revision as stated in the COLR).

2.

OPPD-NA-8302-P-A, "Neutronics Design Methods and Verification" (Latest NRC approved revision as stated in the COLR).

3.

OPPD-NA-8303-P-A, " Transient and Accident Methods and Verification" (I2 test NRC approved revision as stated in the COLR).

c.

The core operating limits shall be determined so that all applicable limits of the safety analysis are met. The Core Operating Limits Report, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Region IV Administrator and Senior Resident Inspector.

5-17a Amendment No. 441,144,157

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ML20058F591

Text

Attachment:

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