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Proposed Tech Specs,Correcting Typos & Making Further Editorial Changes to Proposed Amend 139,dtd 860129
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Issue date:	03/17/1988
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ATTACHMENT II RANCHO SECO TECHNICAL SPECIFICATIONS PAGES AFFECTED BY PROPOSED AMENDHENT NO. 166 1

8803230175 880317

'DR 1

ADOCK 05000312 DCD

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS LIST OF TABLES Iahlst P_ asst 2.3-1 REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS 2-9 166-~

3-3.1 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES 3-22b 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS 3-27 3.5.5-1 ACCIDENT MONITORING INSTRUMENTATION OPERABILITY REQUIREMENTS 3-38b 3.6-1 SAFETY FEATURES CONTAINMENT ISOLATION VALVES 3-40 3.7-1 VOLTAGE PROTECTION SYSTEM RELAY TRIP VALUES 3-43a 3.7-2 VOLTAGE PROTECTION SYSTEM LIMITING CONDITIONS 3-43b 3.12-1 SAFETY RELATED HYDRAULIC SNUBBERS 3-51a-e 3.14-1 FIRE DETECTION INSTRUMENTS FOR SAFETY SYSTEMS 3-55 3.14-2 HATER SUPPRESSION ZONES 3-56b,c 3.14-3 CARBON DIOXIDE SUPPRESSION ZCNES 3-56d,e 3.14-4 FIRE HOSE STATIONS 3-57a,b 3.15-1 RADI0 ACTIVE LIQUID EFFLUENT HONITORING INSTRUMENTATION 3-61 3.16-1 RADI0 ACTIVE GASEQUS EFFLUENT HONITORING INSTRUMENTATION 3-64 3.22-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 3-83 3.22-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS 3-86 IN ENVIRONMENTAL SAMPLES 3.29-1 METEOROLOGICAL MONITORING INSTRUMENTATION 3-95 4.1-1 INSTRUMENT SURVEILLANCE REQUIREMENTS 4-3 4.1-2 HINIMUM EQUIPMENT TEST FREQUENCY 4-8 4.1-3 HINIMUM SAMPLING FREQUENCY 4-9 4.2-1 CAPSULE ASSEMBLY HITHDRAHAL SCHEDULE AT DAVIS-BESSE 1 4-12b 4.6-1 DIESEL GENERATOR TEST SCHEDULE 4-35d 4.6-2 ADDITIONAL RELIABILITY ACTIONS 4-35e 4.14-1 SNUBBERS ACCESSIBLE DURING P0HER OPERATION 4-47d,e 4.17-1 HINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED 4-56 DURING INSERVICE INSPECTION 4.17-2A STEAM GENERATOR TUBE INSPECTION 4-57 4.17-28 STEAM GENERATOR TUBE INSPECTION (SPECIAL LIMITED AREA) 4-57a 4.17-3 OTSG AUXILIARY FEEDRATER HEATER SURVEILLANCE 4-57b,c 4.19-1 RADI0 ACTIVE LIQUID EFFLUENT HONITORING INSTRUMENTATION 4-64 SURVEILLANCE REQUIREMENTS 4.20-1 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 4-66 SURVEILLANCE REQUIREMENTS Proposed Amendment No. 166 ix

RANCHO.SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions ~

1.4 PROTECTION INSTRUMENTATION LOGIC 1.4.1 Instrument Channel An instrument channel is the combination of sensor, wires,' amplifiers and output devices which are connected for the purpose of measuring the value of' a process variable for the purpose of observation, control and/or protection.

An instrument channel may be either analog or digital.

1.4.2 Reactor Protection System The reactor protection system is shown in Figures 7.1-1 and 7.2-2 of the 166-USAR.

It is that combination of protective channels and associated circuitry.which forms the automatic system that protects the reactor by control rod trip.

It includes the four protection channels, their associated instrument channel inputs, manual trip switch, all-rod drive control protective trip breakers and activating relays or coils.

1 Proposed Amendment No. 166 1-2a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions 1.4.3 Protection Channel 166--

A protection channel, as shown in Figure 7.1-1 of the USAR (one of three or one of four independent channels, complete with sensors, sensor power supply units, amplifiers and bistable modules provided for every reactor protection safety parameter), is a combination of instrument channels forming a single digital output to the protection system's coincidence logic.

Each protection channel includes two key-operated bypass switches, a protection channel bypass switch and a shutdown bypass switch.

1.4.4 Reactor Protection System Logic This system utilizes reactor trip module relays (coils and contacts) in all 166-~

four of the protection channels as shown in Figure 7.1-1 of the USAR, to provide reactor trip signals for de-energizing the six. control rod drive trip breakers.

The control rod drive trip breakers are arranged to provide a one-out-of-two-times-two logic.

Each element of the one-out-of-two-times-two logic is controlled by a separate two-out-of-four logic from the four reactor protection channels.

With one channel bypassed and untripped, the two-out-of-four logic functions as a two-out-of-three logic for the three active channels.

1.4.5 53fety Features System Logic This system utilizes relay contact output from individual channels arranged in three analog sub-systems and two two-out-of-three logic sub-systems as 166-~

shown in Figure 7.1-5 of the USAR.

The logic sub-system is wired to provide appropriate signals for the actuation of redundant safety features equipment on a two-of-three basis for any given parameter.

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

1.5 INSTRUMENTATION SURVEILLANCE 1.5.1 Trio Test A trip test is a test of logic elements in a protection channel to verify their associated trip action.

1.5.2 Channel Test A channel test is the injection of an internal or external test signal into the channel to verify its proper response, including alarm and/or trip initiating action, where applicable.

l l

1.5.3 Instrument Channel Check An instrument channel check is a verification of acceptable instrument performance by observation of its behavior and/or state; this verification includes comparison of output and/or state of independent channels measuring the same variable.

Proposed Amendment No. 166 1-3

1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Definitions D.

All automatic containment isolation valves are operable or closed in the safety features position.

E.

The containment leakage satisfies Specificatioti 4.1 and no known d

changes have occurred.

1.8 LICENSEE EVENT REPORTS Defined under Administrative Controls Section 6.9.4.

1.9 TIME PERIODS 16 6-+-

May be extended to a maximum of +251. to accommodate operations scheduling.

The total maximum combined interval time for any three consecutive intervals shall not exceed 3.25 times a single specified surveillance interval.

1.9.1 SHIFT A time period covering at least once per twelve (12) hours.

1.9.2 DAU.Y A time period spaced to occur at least once per twenty-four (24) hours.

1.9.3

.NEEKLY A time period spaced to occur at least once per seven (7) days.

1.9.4 FORTNIGHTLY A time period spaced to cccur once per fourteen (14) days.

1.9.5 MONTHLY A time period spaced to occur at least once per thirty-one (31) days.

1.9.6 0UARTERLY A time period spaced to occur at least once per ninety-two (92) days.

1.9.7 SEMI-ANNUALLY A time period spaced to occur at least once per six (6) months.

1.9.8 ANNUALLY A time period spaced to occur at least once per twelve (12) months.

1.9.9 BIANNUALLY A time period spaced to occur at least once in two (2) years.

1.9.10 REFUELING INTERVAL A time period spaced to occur at least once per eighteen (18) months.

1.10 SAFETY Safety as used in these Technical Specifications shall mean nuclear safety and shall encompass all systems and components that have or may have an effect on the health and safety of the general public.

Proposed Amendment No. 166 1-5

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings of rated power. Adding to this the possible variation in trip set points due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112 percent, which was used in the safety analysis.(4)

A.

Overpower trip based on flow and imbalance The power level trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to accommodate the most severe thermal transient-considered in the design, the loss-of-coolant flow accident from 166- -

high power.

The analysis in USAR Section 14 demonstrates the adequacy of the specified power to flow ratio.

The power level trip set point produced by the power-to-flor: ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases.

The power level trip set point produced by the power to flow ratio provides overpower DNB protection for all modes of pump j

operation.

For every flow rate there is a maximum permissible low flow rate.

Typical power level and low flow rate combinations for the pump situations of Table 2.3-1 are as follows:

1.

Trip would occur when four reactor coolant pumps are operating if power is 106 percent and reactor flow rate is 100 percent, or flow rate is 94.34 percent and power level is 100 percent.

2.

Trip would occur when three reactor coolant pumps are operating if power is 78.8 percent and reactor flow rate is 74.4 percent or flow rate is 70.75 percent and power level is 75 percent.

3.

Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps opeiating) if the power is 51.4 percent and reactor flow rate is 48.5 percent or flow rate is 46.22 percent and the power level is 49 percent.

For safety analysis calculations the maximum calibration and instrumentation errors for the power level were used.

The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded.

These thermal limits are either power peaking kW/ft limits or DNBR limits.

The reactor power imbalance (power in the top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of Figure 2.3-2 are produced.

The power-to-flow ratio reduces the power level trip and associated reactor-power reactor-power-imbalance boundaries by 1.06 percent for a 1 percent flow reduction.

Pro osed Amendment No. 166 r

2-6

RANCKD SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting s

Table 2.3-1 Safety System Settings REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS One Reactor Coolant Pump Four Reactor Coolant Pumps Three Reactor Coolant Pumps Operatine in Each Loop Operating (Nominal Operating (Nominal (NominafOperating Shutdown 4

Operating Power - 100%)

Operating Power - 75%)

Power - 49%)

Bypass

1. Nuclear Power, % of rated, max.

104.9 104.9 104.9 l5.0(3)

2. Nuclear power based on flow (2) 1.06 times flow minus 1.06 times flow minus 1.06 times flow minus.

Bypassed and Imbalance, % of rated, max.

reduction due to reduction due to reduction due to Imbalance (s)

Imbalance (s)

[

3. Nuclear power based on pump (S)

NA NT.

55 Bypassed monitors, % of rated, max.

4. High reactor coolant system 2355 2355 2355 1820(4)'

pressure, psig max.

I

5. Low reactor coolant system Pressure, psig min.

1900 1900 1900 Bypassed

6. Variable low reactor coolant

.12.96 Tout-5334(I) 12.96 Tout-5834(l) 12.96 Tout-5834(l)

Bypassed system pressure, psig min.

t

7. Reactor coolant temp.

F., max.

618 618 618 618 i

8. High Reactor, Building 4

4 4

4 pressure, psig max.

9. Anticipatory reactor trip (6)

Autostop(gjlpressure-Autostop(gjl pressure -

Autostop(gjl pressitee 166*

(Turbine Trip) 50 psig 50 psig 50 psig

10. Anticipatory reactor trip (9)

MFP govegr oil pressure MFPgovegroilpressure MFPgovegroilpressure (Loss of both Main Feed Pumps) 50 psig 50 psig 50 psig

.I (1) T is in degrees Fahrenheit (F).

out 166**

(2) Reactor coolant system flow, %.

'(3) Administratively controlled reduction set only during reactor shutdown.

(4) Automatically set wSen other segments of the RPS (as specified) are bypassed.-

(5) The pump monitors also produce.a trip on: (a) loss of two reactor coolant pumps in one reactor coolant ~1oop,and (b) loss of one or two reactor coolant pumps during two-pump operation.

(6) This trip is disabled below 45% power. (power channel)-

(7) Indicative of a Turbine Trip.

(8). Indicative of loss of both Main Feed pumps.

166**

(9) This trip is disabled below 20% power. (power channel) i Proposed Amendment No. 166 2-9

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The maximum allowable pressure is taken to be the lowest pressure of the three calculated pressures.

The pressure limit is adjusted for the pressure differential between the point of system pressure measurement and the limiting component for all reactor coolant pump combinations. The limit curves were prepared based upon the most limiting adjusted reference temperature of all the beltline region materials.at the end of the fifth effective full power year.

The actual shift in RTNDT of the beltline region material will be established periodically during operations by removing and evaluating, in accordance with Appendix H to 10 CFR 50, reactor vessel material irradiation surveillance specimens installed near the inside wall of this or a similar reactor vessel in the core area.

Because the neutron energy spectra at the specimen location and at the vessel inner wall location are essentially the same, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel.

The limit curves must be recalculated when the ARTNDT determined from the surveillance capsule is different from the calculated ARTNDT for the equivalent capsule radiation exposure.

The unitradiated impact properties of the beltline region materials, requireo by Appendices G and H to 10 CFR 50, were determined for those materials for which sufficient amounts of material were available.

The adjusted reference temperatures are calculated by adding the radiation-induced ARTNDT and the unirradiated RTONT.

The predicted aRTNDT are calculated using the respective neutron fluence and copper and phosphorus contents in accordance with Reg. Guide 1.99.

The assumed RTNDT of the closure head region is 60*F and the outlet nozzle steel forgings is 60*F.

The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME code requirements.

The spray temperature difference restriction based on a stress analysis of the spray line nozzle is imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit.

Temperature requirements for the steam generator correspond with the measured NDTT for the shell.

REFERENCES (1)

USAR paragraph 4.1.E.4 166-~

(2)

ASHE Boiler and Pressure Vessel Code,Section III (3)

USAR paragraph 4.3.8.5 (4)

USAR paragraph 4.3.3 166-(5)

USAR paragraph 4.1.2.8 and 4.3.3 (6)

Analysis of Capsule RSI-8 from Sacramento Municipal Utility District Unit 1 Reactor Vessel Materials Surveillance Program, BAH-1702, February, 1982.

Proposed Amendment No. 166 3-4

RANCH') SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.3 MINIMUM CONDITIONS FOR CRITICALITY Soeci fications 3.1.3.1 The reactor coolant temperature shall be above 525'F except for portions of low power physics testing when the requirements of Specification 3.1.8 shall apply.

3.1.3.2 Reactor coolant temperature shall be above Ductility Transition Terrperatura (DTT) + 10*F.

3.1.3.3 When the reactor coolant temperature is below the minimum temperature specified in 3.1.3.1 above, except for portions of low power physics testing when the requirements of Specification 3.1.8 shall apply, the reactor shall be subcritical by an amount equal to or greater than the calculated reactivity insertion due to depressurization.

3.1.3.4 The reactor shall be maintained subtritical by at least 1 percent ok/k until a steam bubble is formed and an indicated water level between 10 and 316 inches is established in the pressurizer.

3.1.3.5 Except for physics tests and as limited by 3.6.2.1 and 3.5.2.5, safety rod groups shall be fully withdrawn prior to any other reduction in shutdown margin by deboration or regulating rod withdrawal during the approach to criticality.

Following safety rod withdrawal, the regulating rods shall be positioned within their position limits as defined by specification 3.5.2.5 prior to deboration.

Bases At the beginning of life of the initial fuel cycle, the moderator temperature coefficient is expected to be slightly positive at operating temoeratures with the operating configuration of control rods.

(1) Calculations show that above 525'F the positive moderator coefficient is acceptable.

Since the moderator temperature coefficient at lower taperatures will be 166--

less negative or more positive than at operating temperature, startup and operation of the reactor when reactor coolant temperature is less than 525'F is prohibited except where necessary for low power physics tests.

The potential reactivity insertion due to the moderator pressure coefficient 166-~

that cculd result from depressurizing the coolant from 2185 psia to saturation pressure of 885 psia is approximately 0.1 percent ok/k.

During physics tests, special operating precautions will be taken.

In addition, the strong negative Doppler coefficient (1) and the small integrate ' ok/k would limit the magnitude of a power excursion resulting from a reduction of moderator density.

Proposed Amendment No. 166 3-6

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The requirement that the reactor is not to be made critical below DTT + 10*F provides increased assurances that the proper relationship between primary coolant pressure and temperatures will be maintained relative to the NDTT of the primary coolant system. Heatup to this temperature will be accomplished by operating the reactor coolant pumps.

The DTT at Beginning of Life (BOL) for the most limiting component in the reactor coolant system is less than

+100*F.

If the shutdown margin requirad by Specification 3.5.2 is maintained, there is no possibility of an accidental criticality as a result of a decrease of coolant prenure.

The requirement for pressurizer bubble formation and specified water level when the reactor is less than 1 percent subcritical will assure that the reactor coolant system cannot become solid in the event of a rod withdrawal accident or a start-up accident and that the water level is above the minimum detectable level.

The requirement that the safety rod groups be fully withdrawn before criticality ensures shutdown capability during startup.

This does not prohibit rod iatch confirmation, i.e., withdrawal by group to a maximum of 3 inches withdrawn of all seven groups prior to safety rod withdrawal.

REFERENCES (1)

USAR, section 3 166-Proposed Amendment No. 166 3-7

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.5 CHEMISTRY l

Acolicability 166- -

Applies to the limiting conditions of reactor coolant chemistry for continuous operation of the reactor.

Obiective To protect the reactor coolant system from the effects of impurities in the rcactor coolant.

l l

Soecification l

l 3.1.5.1 The following limits shall not be exceeded for the listed reactor coolant conditions.

l Contaminant Soecification Reactor Coolant Conditions Oxygen as 02 0.10 ppm max above 250*F Chloride as Cl- 0.15 ppm max above cold shutdown conditions Fluoride as F-0.15 ppm max above cold shutdown conditions 3.1.5.2 During operation above 250'F, if any of the specifications in 3.1.5.1 are exceeded, corrective action shall be initiated within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

If the concentration limit is not restored within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after initiation of corrective action, the reactor shall be placed in a cold shutdown condition using normal procedures.

3.1.5.3 During operations between 250*F and cold shutdown conditions, if the chloride or fluoride specifications in 3.1.5.1 are exceeded, corrective action shall be initiated within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to restore the normal operating limits.

If the specifications are not restored within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after initiation of corrective action, the reactor shall be placed in a cold shutdown condition using normal procedures.

3.1.5.4 If the oxygen concentration and either the chloride or fluoride concentration of the primary coolant system exceed 1.0 ppm the reactor shall be immediately brought to the hot shutdown condition using normal shutdown procedures, and action is to be taken immediately to return the system to within normal operation speci fi',;ations.

If specifications given in 3.1.5.1 have not been reached in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , the reactor shall be brought to a cold shutdown condition using normal procedures.

Proposed Amendment No. 166 3-10 3

.I RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.8 LOW POWER PHYSICS TESTING RESTRICTIONS Soecification The following special limitations are placed on low power physics testing.

3.1.8.i Reactor Protective System Requirements A.

Below 1820 psig shutdown bypass trip setting limits shall apply-in accordance with Table 2.3-1.

B.

Above 1900 psig nuclear overpower' trip shall be set at a maximum of 5.0 percent.

3.1.8.2 Startup rate rod withdrawal hold shall be in effect at all times.

3.1.8.3 During low power physics testing, the minimum reactor coolant temperature for criticality shall be 240'F. A minimum shutdown margin of 1 percent ok/k shall be maintained with the highest worth control rod fully withdrawn.

Ilaiel The above specification provides additional safety margins during low power physics testing.

The startup rate rod withdrawal hold is described in 166--

paragraph 7.2.2.1.3 of the USAR and applies to the source and intermediate power ranges.

Proposed Amendment No. 166 3-15b

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 166-390,000 gallons of borated water are supplied for emergency core cooling and Reactor Building spray in the event of a loss-of-coolunt accident.

This amount fulfills requirements for emergency core cooling.

The borated water storage tank minimum volume of 390,000 gallons is based on refueling volume requirements. Heaters maintain the borated water supply at a temperature to prevent freezing.

The boron concentration is set at the amount of baron required to maintain the core 1 percent subcritical at 70*F without any control rods in the core.

This concentration is 1585 ppm baron while the minimum value specified in the tanks is 1,800 ppm boron.

The requirement that one BHST isolation valve shall be open assures a static head to the injection pump not lined up to the makeup tank.

The post accident Reactor Building cooling may be accomplished by two spray units or by a combination of two emergency cooling units and one spray unit. The specified requirements assure that the required post accident components are available.

The spray system utilizes common suction lines with the decay heat removal system.

If a single train of equipment is removed from either system, the other train must be assured to be operable in each system.

When the reactor is critical, maintenance is allowed per Specification 3.3.2 provided requirements in Specification 3.3.3 are met which assure operability of the duplicate components. Operability of the specified components shall be based on the results of testing as required by Technical Specification 4.5.

In the event that the need for emergency core cooling should occur, functioning of one train (one high pressure injection pump, one decay heat removal pump and both core flooding tanks) will protect the core and in the event of a main coolant loop severance, limit the peak clad temperature to less than 2,200'F and the metal-water reaction to less than 1 percent of the clad.

The nuclear service cooling water system consists of two independent, full capacity 100 percent redundant systems, to enst..'e continuous heat removal.l3)

The requirements of Specification 3.3.4 assure that the decay heat removal system will not be overpressurized, resulting in a LOCA that bypasses containment.

Tvo in-series check valves function as a pressure isolation barrier between the higS pressure reactor coolant system and the lower pressure decay heat remo al system extending beyond containment. Valve leakage limits provide asserance that the valves are performing their intended isolation function, Proposed Amendment No. 166 3-22

. =.

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The requirements of Specification 3.3.5 assure that, should all trains of a Safety Features equipment or system specified in -this Section 3.3 become inoperable as defined in Specification 1.3, the reactor will be placed in a cold shutdown condition.

It is necessary for a component or system to have available its r,armal and emergency sources of power. When a system or comoonent is determined to be inoperable solely because its normal or emergency power source is inoperable, it may be considered OPERABLE for the purps a of satisfying the requirements of its applicable Limiting Conditions for Doeration provided its redundant system or component is OPERABLE with an OPERABLE normal and emergency power source.

REFERENCES (1)

USAR, paragraph 6.2.1 (2)

USAR, paragraph 9.5.2 (3)

USAR, paragraph 9.4.1 Proposed Amendment No. 166 3-22a l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation TABLE 3.3-1 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES Maximum (a)(b)

System Valve h Allowable Leakage Decay Heat Removal RCS-001 15.0 GPM for each valve RCS-002 DHS-015 DHS-016 Footnote:

(a) 1.

Leakage rates less than or equal to 1.0 gpm are considered acceptable.

2.

Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

3.

Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

4.

Leakage rates greater than 5.0 gpm are considered unacceptable.

(b)

Minimum test differential pressure shall not be less than 150 psid.

Proposed Amendment No. 166 166-~

3-22b

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Bases (Continued)

Each reactor protection channel key operated shutdown bypass switch is provided with alarm and lights to indicate when the shutdown bypass switch is being used.

There are four shutdown bypass keys in the control room under the administrative control of the shift supervisor.

The keys will not be used during reactor power operation.

There are four reactor protection channels.

Normal trip logic is two out of four.

Required trip logic for the power range instrumentation channels is two out of three.

The EFIC trip logic is two times one-out-of-two taken twice. Minimum trip logic on other instrumentation channels is one out of two.

The EFIC system is designed to automatically initiate AFH when:

1.

all four RC pumps are tripped, 2.

RPS has tripped the reactor on anticipatory trip indicating loss of main feedwater, 3.

the level of either steam generator is low, 4.

either steam generator pressure is low, or 5.

SFAS ECCS actuation (high RB pressure or low RCS pressare).

The EFIC system will isolate mgin feedwater to any steam generator when the 166-~

pressure goes below 600 psig (i)

The EFIC system is also designed to isolate or feed AFH according to the following logic:

If both SGs are above 600 psig, supply AFH to both SGs If one SG is below 600 psig, supply AFH to the other SG If both SGs are below 600 psig)but the pressure difference between C

the two SGs exceeds 100 psid l, supply AFH only to the SG with 166--

the higher pressure If both SGs arp)below 600 psig and the pressure difference is less than 100 psid(', supply AFH to both SGs 166--

At cold shutdown conditions all EFIC initiate and isolate functions are manually or automatically bypassed.

When pressure in both steam generators is greater than 750 psig, the following bypassed initiation signals will have been automatically reset:

1) Loss of 4 RC pumps, 2) low steam generator pressure, 3) low steam generator level.

166-~

Note (l):

Pressure setpoint tolerances are 600 1 25 psig and 100 1 50 psid.

l l

Proposed Amendment No. 166 3-26

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Table 3.5.1-1 (Continued)

Limiting Conditions for Operation INSTRUMENTS OPERATING CONDITIONS (C)

(A)

(B)

Operator Action if Functional Unit Total Number of Minimum Channels Conditions of Columns A Channels Operable and B Cannot be Met 9.

Reactor Building Purge Isolation 2

1 Operation may continue provided the purge on high radiation inlet and outlet valves of the inoperable channel (s) are closed and their respective breakers de-energized or comply with 3.5.1.2.

At cold shutdown or refueling, each of the purge inlet and outlet valves will be closed.

Emeroency Feedwater Initiation and Control (EFIC) System 1.

AFW Initiation a.

Manual 2 (Note 1) 2 (Note 1)

See Actions 3 and 4.

b.

Low Level, SGA or B (Note 2) 4/SG (Note 1) 3/SG See Actions 1, 2 and 3.

May be bypassed below 750 psig OTSG pressure.

c.

Low Pressure, SGA or B 4/SG (Note 1) 3/SG See Actions 1, 2 and 3.

May be bypassed below 750 psig OTSG pressure.

d.

Loss of MFW Anticipa-tory Reactor Trip 4 (Note 1) 3 See Actions 1, 2 and 3.

Loss of MFW Anticipatory Reactor Trip is effectively bypassed in RPS below 20 percent power.

e.

Loss of 4 RC Pumps 4 (Note 1) 3 See Actions 1, 2 and 3.

May be bypassed below 750 psig OTSG pressure.

f.

Automatic Trip Logic 2 (Note 1) 2 (Note 1)

See Actions 3 and 4.

2.

SG-A Main Feedwater Isolation a.

Manual 2 (Note 1) 2 (Note 1)

See Actions 3 and 4.

b.

Low SGA Pressure (Note 3) 4 (Note 1) 3 See Actions 1, 2 and 3.

May be bypassed below 750 psig OTSG pressure.

c.

Automatic Trip Logic 2 (Note 1) 2 (Note 1)

See Actions 3 and 4.

Note 1 For channel testing, calibration, or maintenance the Total Number of Channels and/or the Minimum Channels Operable may be -

166~~

reduced by one for a maximum of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months providing the remaining channels are OPERABLE.

Note 2 Low level AFW Initiation has a maximum of a 10.0 second delay.

Note 3 Low pressure AFW Initiation has a maximum of a 3.0 second delay.

Proposed Amendment No. 166 3-30a t

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Table 3.5.1-1 (Continued)

INSTRUMENTS OPERATING CONDITIONS (C)

(A)

(B)

Operator Action if Functional Unit Total Number of Minimum Channels Conditions of Columns A Channels Operable and B Cannot be Met i

3.

SG-B Main Feedwater Isolation a.

Manual 2 (Note 1) 2 (Note 1)

See Actions 3 and 4.

b.

Low SGB Pressure (Note 3) 4 (Note 1) 3 See Actions 1, 2 and 3.

May be bypassed below 750 psig OTSG pressure.

c.

Automatic Trip Logic 2 (Note 1) 2 (Note 1)

See Actions 3 and 4.

4.

AFW Valve Commands (Vector) a.

Vector Enable 2 (Note 1) 2 (Note 1)

See Actions 3 and 4.

b.

Vector Module (Note 4) 4 (Note 1) 3 See Actions 1 and 5.

c.

Control Enable 2 (Note 1) 2 (Nott 1)

See Actions 1 and 3.

d.

Control Module 2 (Note 1) 2 (Note 1)

See Actions 1 and 3.

Note 1 For channel testing, calibration, or maintenance the Total Number of Channels and/or the Minimum Channels Operable may be 166-~

reduced by one f or a maximum of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months providing the remaining channels are OPERABLE.

Note 3 Low pressure AFW Initiation has a maximum of a 3.0 second delay.

Note 4 SG Pressure Difference AFW Valve Command (Vector) has a maximum of a 10.0 second delay.

Proposed Ame.dmcat No. 166 3-30b

RANCHO SECO UNIT 1

'i lECHNICAL SPECIFICATIONS Limiting Conditions for Operation The rod position limits are based on the most limiting of the following three criteria:

ECCS power peaking, shutdown margin, and potential ejected rod worth.

Therefore, compliance with the ECCS power peaking criterion is ensured by the rod position limits.

The minimum available rod worth, consistent with the rod position limits, provides for achieving hot shutdown withdrawn remains in the' full out position.(ghest worth control rod that is by reactor trip at any time, assuming the hi O The rod position limits also ensure that inserted rod groups will not contain single rod worths greater than 0.65% ak/k at rated power.

These values have been shown to be safe by the safety analysis of hypothetical rod ejection accident.(2)

A maximum single inserted control rod worth of 1.0% ak/k is allowed by the rod position limits at hot zero power.

A single inserted control rod worth of 1.0% ak/k at beginning-of-life, hot zero power would result in a lower transient peak thermal power and, therefore, less severe environmental consequences than an 0.65% ok/k ejected rod worth at rated power.

Control rod groups are withdrawn in numerical sequence beginning with Group 1.

Groups 5, 6 and 7 are overlapped 25 percent.

The normal position at power is for Group 7 to be partially inserted.

The Quadrant Power Tilt limits set forth in Specification 3.5.2.4 have been established to prevent the linear heat rate peaking increase associated with a positive quadrant power tilt during normal power operation from exceeding 7.36%.

The limits in Specification 3.5.2.4 are measurement system independent.

The actual o erating limits, with the appropriate allowance for observability and instrumentation errors, for each measurement system are defined in the station operating procedures.

The Quadrant Tilt and axial imbalance monitoring in Specifications 3.5.2.4F and 3.5.2.5, respectively, normally will be perfor:ned in the process computer.

The two-hour frequency for monitoring these qualities will provide adequate surveillance when the computer is out of service.

Allowance is provided for withdrawal limits and reactor power imbalance limits to be exceeded for a period of two hours without specification violation. Acceptable rod positions and imbalance must be achieved within the two-hour time period or appropriate action such as a reduction of power taken.

166-Operating restrictions are included in Technical Specifications 3.5.2.5D.1 and 3.5.2.5D.2 to prevent excessive power peaking by transient xenon. The xenon reactivity must either be beyond the "undershoot" region and asymptotically approaching its equilibrium value at rated power or the l

reactor must be operated in the range of 87% to 92% of the maximum allowable power for a period exceeding two hours in the soluble poison control mode so that the transient peak is burned out at a lower power level.

REFERENCES j

166--

(1) USAR, Section 3.2.2.1.3 i

(2) USAR, Section 14.2.2.4 (3) BAH-1850, October 1984, page 7-5 l

Proposed Amendment No. 166 3-33b

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Condition for Operation 3.5.6 (continued)

Bases The EFIC system is designed to automatically initiate AFH when:

1.

all four RC pumps are tripped, or 2.

RPS has tripped the reactor on anticipatory trip indicating loss of main feedwater, or 3.

the level of either steam generator is low, or 4.

either steam generator pressure is low, or 5.

SFAS ECCS actuation (high RB pressure or low RCS pressure).

The EFIC system will initiate main feedwater isolation to any steam generator as the pressure goes and stays below a minimum setpoint of 575 psig.

The EFIC system is also designed to isolate or feed AFH according to the following vecipr logic.

Setpoints are nominal and subject to instrument 166-~

inaccuracies:('>

If both SGs are above 600 psig, supply AFH to both SGs If one SG is below 600 psig, supply AFH to the other SG If both SGs are below 600 psig but the pressure difference between 166-~

the two SGs exceeds 100 psid, supoly AFH only to the SG with the higher pressure If both SGs are below 600 psig and the pressure difference is less 166-~

than 100 psid, supply AFH to both SGs At cold shutdown conditions all EFIC automatic initiate and isolate functions are manually or autc'.latically bypassed.

Prior to a pressure of greater than 750 psig in both steam generators, the following bypassed initiation signals

{

automatically reset:

1) Loss of 4 RC pumps,

2) low steam generator j

pressure,

3) low steam generator level.

l Bypassing of automatic AFH initiation on Loss of MFH Anticipatory Trip or SFAS actuation is controlled by bypass permissive logic within the RPS and SFAS, respectively.

166-~

Note (1):

Pressure setpoint tolerances are 600 25 psig and 100 50 psid.

1 i

i Proposed Amendment No. 166 3-38h i

l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.7.2 The reactor shall not remain critical unless all of the following requirements are satisfied.

Soecification J.

The switchyard voltage is at least 219 KV.

Acolicability STARTUP through POWER OPERATIONS Action Should the switchyard voltage drop below 219 KV, a.

positive actions, within the District's procedures, will be implemented in an attempt to return the voltage to at least 219 KV.

166--

b.

both diesel generator trains shall be demonstrated OPERABLE within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months by obtaining acceptable results from performing surveillance requirement 4.6.3.A.4.

Should the switchyard voltage not be restored above 219 KV within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , c.

the reactor shall be in HOT SHUTDOHN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOHN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1 Proposed Amendment No. 166 3-42j

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.9 SPENT FUEL POOL Acolicability Applies to the Spent Fuel Pool Cooling System.

Objective To provide for adequate cooling of.the Spent Fuel Pool to ensure that the pool temperature is kept low enough to prevent boiling.

Soecification 3.9.1 One train of the Decay Heat Removal System (DHRS) must be put in service to provide alternate cooling for the Spent Fuel Pool if the bulk coolant temperature reaches 140*F and the Spent Fuel Pool Cooling System is inoperable, and as a supplement to the Spent Fuel Pool Cooling System if a maximum temperature of 180*F is exceeded.

3.9.2 If a train of the DHRS is being used to provide alternate cooling for the Spent Fuel Pool, it shall be declared inoperable for other I

purposes and the provisions of Technical Specification 3.3.2 shall apply unless the reactor is in Cold Shutdown.

3.9.3 Use of the DHRS for Spent Fuel Pool cooling shall be limited to no more than 100 cumulative hours (when not in Cold Shutdown) in any 12-month period.

3.9.4 Reactor shutdown must be initiated within I hour if the Spent Fuel Pool bulk coolant temperature reaches 180*F, and the reactor must be in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

MSES This specification provides a method to ensure that the Spent Fuel Pool bulk temperature does not reach the boiling point.

The use of a train of the 166--

Decay Heat Removal System (DHRS), as required by operating procedures, provides immediate alternate cooling capability to ensure this.

Either train of the DHRS can easily be lined up for Spent Fuel Pool cooling by opening two manual valves (DHS-032 and DHS-055 or 056), one motor operated valve (HV-26047 or 46), and starting the appropriate decay heat pump (P-261A or B).

However, since use of the DHRS train for Spent Fuel Pool cooling effectively removes it from its normal service, an operating duration limit of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months per 12-month period is imposed.

References

[1]

Licensing Report for High Density Spent Fuel Storage Racks for Rancho Seco.

[2]

Time to Boil Calculation, Supplement No. 2 to Thermo-Hydraulic Calculations for Rancho Seco Nuclear Station; Report No. TM-661.

Proposed Amendment No. 166 3-46a

1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation TABLE 3.14-1 FIRE DETECTION INSTRUMENTS FOR SAFETY SYSTEMS Detection Number of detectors in zones-Zone Instrument Location All Required to be OPERABLE Heat Flame Smoke 4

Control and Computer Room Cabinets 0

0 48 5

South East Uncontaminated Area -

0 0

3 Turbine Deck, Auxiliary Building 7

Contaminated Area - Turbine Deck, 0

0 3

Auxiliary Building 11 West Battery Room - Hezzanine Level, 0

0 8

Auxiliary Building 12 West Battery Charger Room -

0 0

4 Hezzanine Level, Auxiliary Building 166-~

13 Hest 400V Switchgear Room -

0 0

8 Hezzanine Level, Auxiliary Building 14 Hest Cable Shaft - Auxiliary Building 0

0 6

15 East Cable Shaft - Auxiliary Building 0

0 6

166--

16 East 480V Switchgear Room -

0 0

6 Hezzanine Level, Auxiliary Building 17 East Battery Charger Room -

0 0

4 Hezzanine Level, Auxiliary Building 19 Communication Room - Hezzanine Level, 0

0 8

Auxiliary Building 20 Electrical Penetration Room -

0 0

21 Hezzanine Level, Auxiliary Building 21 Reactor Building, 20 feet to 40 feet 0

0 3

Level Proposed Amendment No. 166 3-55

~-

' RANCHO SECO UNIT l' TECHNICAL. SPECIFICATIONS Limiting Conditions for Operation 3.18.3 Iodine-131. Tritium and Radionuclides in Particulate Form The. dose or dose commitment to a member of the'public from I-131,

~

from tritium,.and from radionuclides in particulate form with half-lives greater-than eight days in gaseous effluents released at and beyond the site boundary shall be limited to the following:

~

.a.

During any calendar quarter to 7.5 mrem to any organ.

b.

During any calendar year to 15 mrem to any organ.

Aeolicability At all times Action Hith the calculated dose or dose commitment from the release of I-131, tritium, and radionuclides in particulate form with half-lives greater than.eight days in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days a Special Report.

This Report will identify the cause(s) for exceeding the limit and define the corrective actions to be taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

t Raitti t

This specification is provided to implement the requirements of Sections II.C, i

l III. A and IV.A of Appendix I,10 CFR Part 50.

The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I.

The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that i

the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is'unlikely to be substantially underestimated.

For individuals who may at times.be within the site boundary, the occupancy of the individual will be sufficiently low to i

compensate for any increase in the atmospheric diffusion factor above that for the restricted area boundary.

The OOCH calculational methods for calculating i

the doses due to the actual release rates of the subject materials are j

required to be consistent with the methodology provided in Regulatory Guide i

1.109, "Calculating of Annual Doses to Man from Routine Releases of Reactor 166- -

Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, 1

Appendix I, Revision 1, October 1977 and Regulatory Guide 1.111. "Methods for l

Estimating Atmospheric Transport and Disperion of Gaseous Effluents in Routine Releases from Light-Hater-Cooled Reactors", Revision 1, July 1977.

These i

equations also provide for determining the actual doses based upon the j

historical average atmospheric conditions.

i Proposed Amendment No. 166 3-75

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.20 GAS STORAGE TANKS The quantity of radioactivity contained in each waste gas decay tank shall be limited to 135,000 curies of noble gases (considered as Xe-133).

Acolicability At all times Action When the reactor coolant system activity reaches the limit of Technical Specification 3.1.4, sample the online waste gas decay tank daily to ensure that the limit of 135,000 curies equivalent Xe-133 is not exceeded.

Bates Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tanks contents, the resulting total body exposure to an individual at the nearett exclusion area boundary will not exceed 500 mrem.

This is consistent with Standard Review Plan 15.7.1, "Haste Gas System Failure".

Potential atmospheric releases from a waste gas decay tank are evaluated 166-assuming design coolant activities (see page 140-25 Vol. VIII USAR).

Based on primary coolant activity as shown in USAR Table 14D-7, the decay tank is assumed to hold the activity associated with the off-gas from one reactor coolant system degassing with no credit taken for decay.

Calculation of the limiting decay tank activity based on the coolant activity limit of Technical Specification 3.1.4 yields a maximum decay tank inventory 166-~

of 98,414 curies (Ref. USAR Table 140-23).

In order for the decay tank inventory to reach the limiting condition for operation, coolant activity would have to exceed the Technical Specification 3.1.4 limit on coolant activity and this would require a reactor shutdown, thus preventing a further increase in gaseous activity.

Therefore, it is conservative to require that the online waste gas decay 166-~

tank _be sampled daily upon reaching the coolant limiting activity value (43/E) to ensure the 135,000 curies equivalent Xe-133 is not exceeded. Once the coolant is below the limiting activity, there is no requirement to sample waste gas decay tanks except for discharging.

Proposed Amendment No. 166 3-79

RANCHO SECO UNIT'l TECHNICAL SPECIFICATIONS Surveillance 1ards TABLE 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Descriotion Check Test Calibrate Remarks 166**-

20.

High pressure injection, NA R

NA Reactor Building isolation,

.and Reactor Building emer-gency cooling Channel A manual trip.

166-~

21.

High pressure injection NA R

NA Reactor Building isola-tion, and Reactor Build-ing emergency cooling Channel 8 manual trip.

22.

Low pressure injection NA R

NA Channel A manual trip a

23.

Low pressure injection NA R

NA Channel B manual trip 24.

Reactor Building spray NA R

NA pump Channel A manual trip 25.

Reactor Building spray NA R

NA pump Channel B manual trip 26.

Reactor Building spray NA R

NA valves Channel A manual trip Proposed Amendment No. 166 4-6

RANCK3 SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarks 27.

Reactor Building spray NA R

NA valves Channel B manual trip Process Instrumentation 28.

Core flooding tanks a.

Pressure channels D

NA R

b.

Level channels D

NA R

166*

29.

Pressurizer level channels D

NA R

Equivalent to Item 49, but excludes SPDS.

30.

Pressurizer temperature S

NA R

channels 31.

Make-up tank level D

NA R

channe'es 32.

High pressure injection NA NA R

flow channels 33.

Low pressure injection NA NA R

flow channels 34.

Borated water storage W

NA R

sank level indicator Proposed Amendment No. 166 4-7

RANCKD SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarks 35.

Spray additive tank a.

Level channel W

NA R

36. Concentrated boric acid storage tank a.

Level channel W

NA R

b.

Temperature channel M

NA R

37.

Steam generator water W

NA R

level 38.

Control rod absolute

$(1)

NA R(2)

(1) Check with relative position position indicator (2) Calibrate rod misalignment channel 166**

39.

Control rod relative S(1)

NA R(1)

(1) Check with absolute posi-position tion indicator 40.

Reactor Building NA NA R

temperature 41.

Reactor Building emergency NA NA R

sump level alarm Proposed Amendment No. 166 4-7a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveilltnce Strndtrds TABLE 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarks 42.

Reactor Building drain accumulation tank level NA NA R

43.

Incore neutron detectors M(1)

NA NA (1) Check functioning, including functioning of computer read-out and/or recorder readout.

44.

a.

Process and area radi-ation monitoring system W

M Q

b.

Containment Area Monitors W

NA R

166**

45.

Deleted 46.

Environmental air monitors M(1)

NA R

(1) Check functioning 47.

Strong motion accelerometer Q(1)

NA R

(1) Battery check 48.

Deleted 49.

Pressurizer Water Level (l)

M NA R

50.

Auxiliary feedwater Flow Rate M

R 166-*

51.

Deleted 52.

EMOV Power Position Indicator (Primary Detector)

M NA R

53.

EMOV Position Indicator (Backup Detector)

M NA R

T/C or Acoustic 54.

EMOV Block Valv9 Position Indicator M

NA R

55.

Safety Valve Position In-dicator (Primary Detector)

M NA R

T/C Proposed Amendment No. 166 4-7b

RANCK3 SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarks

68. AN Initiation a.

Manual NA M

NA b.

Low Level SGA or B M (1)

R (1)

(1) Include time delay module.

c.

Low Pressure SrA or 3 S

H (1)

R (1)

(1) Include time delay module.

166~

(2)

(2) Calibrate pressure transmitters annually.

~

d.

Loss of M W Anticipa-tory Reactor Trip S

M NA e.

Loss of 4 RC Pumps S

M NA f.

SFAS Actuation S

R NA g.

Automatic Trip Logic M

NA 166*

h.

Bypasses S

M R (1)

(1) Calibrate pressure transmitters annually.

~

69. SGA Main Feedwater Line I sol a ti on a.

Marual NA M

NA b.

Automatic Trip Logic S

M NA

70. MB Main Feedwater Line Isolation a.

Manual NA M

NA b.

Automatic Trip Logic S

M NA Proposed Amendment No. 166 4-7d

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Sureeillonce Standards TABLE 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarks

71. AFW Valve Commands (Vector) a.

Vector Enable S

M NA 166*

b.

SGA Pressure Low S

M R(1)

(1) Calibrate pressure transmitters annually.

c.

SGB Pressure Low S

M R(1)

(1) Calibrate pressure transmitters annually.

d.

SG Pressure Difference 166-SGA Pressure >

S M (1)

R(1)(2) (1) Include time delay module.

SGB Pressure (2) Calibrate pressure transmitters SGB Pressure >

annually.

SGA Pressure S

M (1)

R(1)(2) (1) Include time delay module.

(2) Calibrate pressure transmitters annually.

72. AFW (pnir21 Valve Control a.

Manual / Auto NA M

NA in Manual 73.

SG Level Control a.

Setpoint Selection NA M

NA b.

Control Enable NA M

NA C.

Module Response NA M (1)

R (1) Confirm External Controller Settings

74. ADV Control Valve Control a.

Manual / Auto NA M

NA in Manual

75. SG Pressure Control 4.

Module Response NA M (1)

R (1) Confirm External Controller Settings

76. pglup_In_itrument Air hypoly System a.

Pressurt D

NA NA Table Notationi S = Each shift M = Monthly P = Prior to each startup if not done previous week D = Daily Q = Quarterly R = Once during the refueling interval W = Veekly SY = Semiannual Proposed Amendment No. 166 4-7e

RANCHO SECO UNIT l TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarks

82. Spray pond Water Temperature D

NA R

83. Spray Pond Water Level D

NA R

(matgency Shutdown Instrumentation

84. Wide Range OTSG Level M

NA R

85. Wide Range OTSG Pressure M

NA R

166*

86. Pressurizer Level M

NA R

87. Wide Range Reactor Coolant Hot Leg Temperature M

NA R

88. Wide Range Reactor Coolant Cold Leg Temperature M

NA R

89. Wide Range Reactor Coolant Pressure M

NA R

90. Source Range Neutron Flux Indicator

NA NA R

91. Makeup Tank Level M

NA R

S = Each Shift M = Monthly P = Prior to each startup if not done previous week D = Daily Q = Quarterly R - Once during the refueling interval W = Weekly SY = Semiannual

Deenergized and disconnected except when control room habitability is lost.

Proposed Amendment No. 166 4-79

RANCHO SECO UNIT.1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-3 HINIMUM SAMPLING FRE00ENCY Item Check Freauency 1.

Reactor coolant

a. Radio-chemical gnglysisll)

H 166-~

E determinationl4>

Semiannually

b. Gross activity (1) (3) 3/ week

c. Tritium radioactivity H

d. Chemistry (C1 and 0 )

3/ week 2

e. Boron concentration 2/ week

f. Fluoride H

2.

Borated water Boron concentration H and after each storage tank makeup water sample 3.

Core flooding tank Boron concentration (3)

H and after each water sample makeup 4.

Spent fuel storage Boron concentration H and after each water sample makeup 5.

Secondary coolant

a. Gross activity (3)

Heekly

b. Iodine analysis (2)(3)

Weekly 166-~

6.

Concentrated boric Boron concentration (3) 2/ week and after acid tank each makeup 166-~

7.

Spray additive Na0H concentration (3)

Q and after tank each makeup 166-~

8.

Blowdown from Gross activity (3) y cooling towers (1)Hhen radioactivity level is greater than 20 percent of the limits of Technical Specification 3.1.4, the sampling frequency shall be increased to a minimum of once each day.

(2)Hhen gross activity increases by a factor of two above normal, an iodine analysis will be made and performed thereafter when the gross activity increases by ten percent.

(3)Not performed during cold shutdown.

1664 (4)E determination will be started when gross beta-gamma activity analysis indicates greater than 10pCi/ml and will be redetermined each 10pC1/mi increase in gross beta-gamma activity analysis. A radio chemical analysis for this purpose shall consist of a quantitative measurement of 951. of radionuclides in reactor coolant with half lives of >30 minutes.

l Proposed Amendment No. 166 4-9

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Bases Irradiation surveillance provides the capability of determining the radiation-induced changes in the mechanical and impact properties in the region of the reactor vessel surrounding the core.

Test specimens of base metal, deposited weld metal and the heat-affected zone-are installed in capsule assemblies placed inside the vessel.

In accordance with the schedules of Table 4.2-1 specimens will be removed; and a series of drop weight tests, Charpy impact tests and tension tests will be conducted.

Threshold neutron flux detectors and maximum temperature detectors will be installed with the specimens. Changes in nil-ductility transition temperature will be determined, and appropriate alteration to plant operating parameters will be made.

To assure the availability of adequate surveillance data for the Rancho Seco No. I reactor vessel, a program has been developed to monitor the irradiation of the surveillance specimen capsules at the Davis Besse No. 1 reactor, and compare this to the irradiation of the Rancho Seco No, I reactor vessel.

Fluence estimates which are conservative in the appropriate direction are used for this comparison.

The frequency of monitoring varies depending on the known neutron fluence lead factor between the capsules and the reactor vessel.

This provides ample time for anticipating problems and initiating corrective action should operation of the host reactor be 166--

seriously delayed.

For the purpose of Technical Specification 4.2.1.2, the definition of Regulatory Guide 1.16, Revision 4 (August 1975) applies for the term "commercial operation".

Cumulative reactor utilization factor is defined as:

[(Cumulative thermal megawatt hours since attainment of commercial operation at 100% power) x 100] + [(licensed thermal power) x (cumulative hours since attainment of commercial operation at 100% power)).

A preoperational examination was mado which included all the items in ASHE Code Class I systems that would normally be completed throughout the Inspection interval.

This survey established initial system integrity and provided a baseline for future testing.

Specification 4.2.2.1 ensures that inservice inspection of ASHE Code Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves will be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a.

Relief from any of the above requirements has been provided in writing by the Commission and is not a part of these Technical Specifications.

Proposed Amendment No. 166 4-12

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards provided local leakage rate measurements are made before and after repair to demonstrate that the leakage rate

1 reduction achieved by repairs reduces the ove.all measured integrated leak rate to an acceptable value.

4.4.1.1.7 Report of Test Results Each integrated leak rate test will be the subject of a 166-summary technical report in accordance with 10 CFR 50, App. J Section V.B and Specification 6.9.5M.

The report will include any deterioration noted in the containment surfaces inspection described in Specification 4.4.1.4, a description of test methods used and a summary of local leak detection tests. Sufficient data and analysis shall be included to show that a stabilized leak rate was attained and to identify all significant required correction factors such as those associated with humidity and barometric pressure, and all significant errors such as those associated with instrumenta-166-tion sensitivities and data scatter.

The report shall be submitted to the Director of Nuclear Reactor Regulation, NRC approximately three months after the conduct of each test.

4.4.1.2 Local Leakage Rate Tests 4.4.1.2.1 Scope of Testing The local leak rate shall be measured for each of the following components:

(1)

Personnel hatch (2)

Emergency hatch (3)

Equipment hatch seals (4)

Fuel transfer tube seals (5)

Fuel transfer tube shroud bellows (6)

Reactor Building normal sump drain line (7)

Reactor coolant pump seal water outlet line (8)

Reactor coolant pump seal inlet line (9)

Reactor Building equalizing line (10)

Decay Heat removal inlet lines (11)

Reactor Building spray inlet lines (12)

High pressure injection lines (13)

Electrical penetrations (14)

Reactor Building purge inlet line (15)

Reactor Building purge outlet line (16)

Reactor Building atmosphere sample lines (17)

Letdown to purification demineralizer line (18)

Pressurizer relief tank gas sample line (19)

Reactor coolant system vent header Proposed Amendment No. 166 4-17

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.4.1.3 liolation Valve Functional Tests Remotely operated Reactor Building isolation valves shall be stroked to the position required to fulfill their safety function in accordance with requirements of Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55(a)(g), except where specific written relief 4

has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

166-~

4.4.1.4 Containment Surfaces Insoection A visual examination of the accessible interior and exterior surfaces of the containment structure and its components shall be 166-~

performed prior to any integrated leak rate test, to uncover any evidence of deterioration which may affect either the containment's structural integrity or leak-tightness.

The discovery of any significant deterioration shall be accompanied by corrective actions in accord with acceptable procedures, nondestructive tests, and inspections, and local testing where practical, prior 166~

to the conduct of any integrated leak rate test.

Such repairs shall be reported as part of the test results per Specification 6.9.5H.

4.4.1.5 Reactor Buildina Modifications Any major modification or replacement of components affecting the Reactor Building integrity shall be followed by either an integrated leak rate test or a local leak test, as appropriate, and shall meet the acceptance criteria of 4.4.1.1.5 and 4.4.1.2.3 respectively.

Bases The Reactor Building is designed for an aternal pressure of 59 psig and a steam-air mixture temperature of 286 F.

Prior to initial operation, the containment will be strength tested at 115 percent of design pressure. The containment will also be leak tested prior to initial operation at Pp and

)

P (52 psig and 26 psig, respectively). These tests will verify that the ion satisfies the 1akageratefromReactorBuildingpressurizp2) relationships given in the specification. (1 The performance of a periodic integrated leakage rate test during plant life i

provides a current assessment of potential leakage from the containment in case of an accident that would pressurize the interior of the containment.

166-~

The minimum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months was specified for the integrated leakage rate test to help stabilize conditions and thus improve accuracy and to better evaluate data scatter. The frequency of the periodic integrated leakage rate test is keyed to the refueling schedule for the reactor, because these tests can best be performed during refueling shutdowns.

j i

Proposed Amendment No. 166 i

4-19 j

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards The specified frequency of periodic integrated leakage rate tests is based on three major considerations.

First is the low probability of leaks in the liner, because of conformance of the complete containment to a 0.10 percent leakage rate at 52 psig during pre-operational testing and the absence of any significant stresses in the liner during reactor operation.

Second is the more frequent testing, at 52 psig of those portions of the containment envelope that are most likely to develop leaks during reactor operation (penetrations and isolation valves) and the low value 0.06 percent of leakage that'is specificed as acceptable from penetrations and isolation valves.

Third is the tendon stress surveillance program which provides assurance that an important part of the structural integrity of the containment is maintained.

More frequent testing of various' penetrations is specified as these locations are more susceptible to leakage than the Reactor Building liner due to the mechanical closure involved.

Particular attention is given to testing those penetrations with resilient sealing materials, penetrations that vent directly to the Reactor Building atmosphere, and penetrations that connect to the reactor coolant system pressure boundary.

The basis for specification of a total leakage rate of (0.075 percent) from penetrations and isolation valves is that approximately three quarters of the allowable integrated leakage rate should be from those sources in order to provide assurance that the integrated leakage rate woul, remain within the specified limits during the intervals between integratea leakage rate tests.

Valve operability tests are specified to assure proper closure or opening of the Reactor Building isolation valves to provide for isolation of functioning of safety features systems. Valves will be stroked to the position required to fulfill their safety function unless it is established that such testing is not practical during operations.

The airlock seals are tested at 10 psig because that is the manufacturer's recommended pressure for reverse flow through the seals.

The extrapolation formula is derived assuming laminar, incompressible flow and provides conservative leak rates.

This specification complies with the Appendix J to 10 CFR 50 as published in the Federal Register on February 23, 1973, with the exemptions to Appendix J granted July 13, 1977.

REFERENCES (1)

USAR, Paragraph 5.2.1.1.1 (2)

USAR, Section 14 t

Proposed Amendment No. 166 4-20

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Soecification.(Continued)

B.

The grease coverage will be noted along with the temperature to accumulate a record of grease variation versus temperature.

4.4.2.5 Reoorts 166-~

A report covering the results of each inspection will be filed with the plant quality assurance records.

If any significant or critical deterioraticn is noted by this inspection, it will be reported to the NRC as a reportable occurrence in accordance with 166-~

Tecnnical Specification 6.9.58.

Should this be necessary, the initial report may be made within 10 days of the completion of the tests and the detailed report may follow within 90 days of the completion of the tests.

i 166-4 5

I 1

Proposed Amendment No. 166 4-23 i

I RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.5.3 DECAY HEAT REMOVAL SYSTEM AND REACTOR BUILDING SPRAY SYSTEM LEAKAGE golicability Applies to Decay Heat Removal System and the Reactor Building Spray System leakage.

Obiective To prevent significant offsite exposures by maintaining a preventive leakage rate for the Decay Heat Removal System and the Reactor Building Spray System.

Specification 4.5.3.1 Acceotance Limit The maximum allowable leakage from the components (which include valve stems, flanges and pump seals) in the Decay Heat Removal System and the Reactor Building Spray System shall not exceed a 166-sum total of 0.63 gallons per hour (gph) for both systems.

4.5.3.2.A J111 - Decav Heat Removal System During each refueling interval, the following tests of the Decay Heat Removal System shall be conducted to determine leakage:

1.

The portion of the Decay Heat Removal System, except as specified in (2), that is outside the containment shall be tested either by use in normal operation or by hydrostatically testing at 450 psig.

2.

Piping from the containment emergency sump to the decay heat removal pump suction isolation valve shall be pressure tested at no ;ess than 52 psig as a containment local leak rate test under paragraph 4.4.1.2.

3.

Visual inspection shall be made for excessive leakage from components of the system. Any excessive leakage shall be measured by collection and weighing or by another equivalent method.

4.5.3.2.8 Int - Reactor Buildina Soray System During each refueling interval, the, following tests of the Reactor Building Spray System shall be conducted in order to determine leakage:

Proposed Amendment No. 166 4-32 7

i

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.5.3.2.8 1.

The section of the syste;a that is-downstream of the pump suction isolation valve shall be. tested by use in normal operation or by hydrostatically testing at 180 psig.

2.

The section of the system from the containment emergency 166-~

sump isolation valve to the pump suction isolation valve shall be tested at no less than 52 psig as a containment local leak rate test under paragraph 4.4.1.2.

3.

Visual inspection shall be made for excessive leakage from components of the system. Any excessive leakage shall be measured by collecting and weighing or by anothiir equivalent method.

Bases The leakage rate limit for the Decay Heat Removal System is a judgment value based on assuring that the components can be expected to operate without mechanical failure for a period on the order of 200 days after a loss of coolant accident.

The test pressures achieved either by normal system operation or by hydrostatically testing, give an adequate margin over the highest pressure within the system after a design basis accident.

Similarly, the pressure tests for the return lines from thc containment to the DecSy Heat Removal System are equivalent to the peak calculated pressure after a LOCA. A Decay Heat Removal System and Reactor Building Spray System sum 166-~

total leakage rate of 0.63 gph will limit offsite exposures due to leakage to insignificant levels relative to those calculated for leakage directly from the Reactor Building in the design basis accidcnt.

The dose to tne thyroid calculated as a result of this leakage is 7.21 rem for a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months exposure at the site boundary.

(1)

REFERENCES (1)

USAR, paragraph 14.3.9.3.

l l

i 1

l Proposed Amendment No. 166 4-33 L

RANCHO SECO UNIT 1 TE("'NICAL SPECIFICATIONS Surveillance Standards.

6.

Verifying the diesel generator operates for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Note 1).

During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generator shall be loaded between 2650 and 2850 kw for A and B and 3000 and 3300 kw for A2 and 82 and for the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be loaded between 2550 and 2750 kw for A and B and 3000 and 3300 kv for A2 and B2. Within five minutes after completing this 24-hour test, 166-~

perform surveillance requirement 4.6.3C.4.b).

(Note 2) 7.

Verifying that the auto-connected loads to each diesel 166-~

generator do not exceed the 2000-hour rating of 2750 kw for A and B and 3300 kw for A2 and B2.

D.

At least once per 10 years or after any modifications which could affect interdependence, by starting all four diesel generators simultaneously (Note 3) and verifying that they accelerate to a nominal 900 rpm for A and B and a nominal 450 rpm for A2 and B2 within 10 seconds after the start signal.

The generator voltage and frequency shall be 4160 (1420) volts and 60 (11.2) Hz within 10.0 seconds after the start signal.

Note 1 All planned engine starts may be preceded by an engine prelube period. With the exception of once per 184 days, all planned engine starts may be preceded by warmup procedures recommended by the manufacturer and may also include slow starting (greater than 10 seconds) and gradual loading (greater than 90 seconds) so that mechanical stress and wear on the diesel engine is minimized.

The testing performed once every 184 days shall include fast starting (less than or equal to 10 seconds) and fast loading (less than or equal to 90 seconds). Whenever : fast start is performed, the diesel generator shall start wit #1n 10 seconds.

166-Note 2 If Surveillance Requirement 4.6.3C.4.b is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test.

Instead, the diesel generator may be operated between 2550 and 2750 kw for A and B, or between 3000 and 3300 kw for A2 and B2 for one hour or until operating temperature has stabilized.

Note 3 All planned engine starts for the purpose of this surveillance testing may be preceded by an engine prelube period.

The testing shall include fast starting (less than or equal to 10 seconds).

Proposed Amendment No. 166 4-34g I

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.9 REACTIVITY ANOMALIES haplkabilitv Applies to potential reactivity anomalies.

Obiective To require the evaluation of reactivity anomalies of a specified magnitude occurring during the operation of the unit.

Soecification 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-s; ate concentrations reaches the equivalent of one percent in reactivity, an evaluation will be made to determine the cause of 16 6-~-

the discrepancy and reported to the Nuclear Regulatory Commission.

Bases 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 concentrattan, necessary to maintain adequate control characteristics, must be adjusted (normalized) to accurately reflect ctual core conditions. When full power is reached initially, and with the control rod groups in the desired positions, the boron concentration is ri.easured and the predicted curve is adjusted to this point. As power operation proceeds, the mear9 red 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 ccmpleted after about 10 percent 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 percent would be unexpected, and its occurrence would be thoroughly investigated and evaluated.

The value of 1 percent is considered a safe limit since a shutdown margin of at least 1 percent with the most reactive rod in the fully withdrawn position is always maintair.ed.

l Proposed Amendment No. 166 4-40

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Soecification (Continued) 4.11.1 B.

3.

(a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly and obtaining samples at least 2 inches in diameter with a length equal to the thickness of the bed, or (b) Emptying a longitudinal sample from.an adsorber tray, mixing the adsorber thoroughly, and obtaining samples at least 2 inches ~in diameter with a length equal to the thickness of the bed.

4.

Verifying system flow rate of 66,700 cfm 2 10% during system operation'when tested in accordance with ANSI N510.

C.

Verified by determining that the air distribution across the filter banks is uniform per ANSI N510 following original installation, modification or repair.

D.

Demonstrated operable after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by:

1.

Verifying within 31 days after removal that a laboratory analysis of a carbon sample from at least one test canister or carbon sample remwed from one of the charcoal adsorbers demonstrates a removal efficiency of 195% for radioactive methyl iodide when the sample is tested as used charcoal in accordance with ASTM D3803 (30*C, 95% R.H.).

The carbon samples not obtained from test canisters shall be prepared from either:

(a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least 2 inches in diameter and with a length equal to the thickness of the bed, or 166-(b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorber thoroughly, and obtaining samples at least 2 inches in diameter with a length equal to the thickness of the bed.

2.

If an adsorber tray is removed in obtaining the charcoal 166-~

sample per Specification 4.11.10.1 above:

(a) Verify that the charcoal adsorbers remove 199.5% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510 while operating the filter train at a flow rate of 66,700 cfm 10%.

j l

Proposed Amendment No. 166 4-42a

I RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.12 AUXILIARY AND SPENT FUEL BUILDING FILTER SYSTEMS Acolicability Applies to the Auxiliary Building exhaust filter system and to the Spent 166*

Fuel Pool Building when irradiated fuel which has decayed for less than 30 days is being moved or is stored in it, or when operating the crane with loads over the fuel storage pool.

~

Obiective To verify that the Auxiliary Building exhaust filter system and components will be able to perform their design functions.

Soecification 166--

4.12.1 When irradiated fuel which has decayed less than 30 days is in the spent fuel storage pool:

J A.

The spent fuel storage pool building exhaust ventilation system shall be verified to be operating with all spent fuel building doors closed (excepting intermittent personnel use) prior to fuel movement and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months during either fuel movement within the spent fuel storage pool or crane operation i

l 166-with loads over the spent fuel storage pool, or when stored, spent fuel has decayed for less than 30 days.

4.12.2 Proper operation of the ventilation system shall be:

A.

Verified at least once per 31 days by observing flow through the operating HEPA filter and charcoal adsorber train and verifying that the train operates with <6 inches water gauge pressure drop across the combined HEPA and charcoal filter banh and verifying system operattori for at least 15 minutes.

B.

Verified at least once per refueling interval, or once every 18 months, whichever occurs first, or after each partial or complete replacement of the HEPA filter bank or charcoal adsorber bank, or following painting, fire, or chemical release in the operating air makeup system, or after any structural maintenance on the HEPA filter or charcoal adsorber housings, by:

1.

Verifying that the charcoal adsorbers remove 199.5 percent of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510 while 166-~

operating the filter train at a flow rate of 43,400 cfm 10 percent.

2.

Verifying that the HEPA filter banks remove 199.91. of the DOP when they are tested in-place in accordance with ANSI 166-N510 while operating the filter train at a flow rate of 43,400 cfm i 10 percent.

Proposed Amendment No. 166 4-43

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Soecification (Continued) 4.12.2 B.

3.

Verifying that a negative pressure of 10.01 inches H.G. is maintained in the spent fuel building, with the Auxiliary Building exhaust system operating at a flow rate 166-~

1) of 43,400 cfm x 10 percent and exhausting through the HEPA filters and charcoal adsorbers to the facility vent, 166--

and 2) of 10,800 cfm 2 10 percent exhaust from the spent fuel pool area.

4.

Verifying within 31 days after removal that a laboratory analysis of a carbon sample from at least one test canister or carbon samp:e removed from one of the charcoal adsorbers demonstrates a.emoval efficiency of 195% for radioactive methyl iodide when the sample is tested in accordance with ASTM D3803 (30*C. 95% R.H.).

The carbon samples not obtained from test canisters shall be prepared from either:

(a)

Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or (b)

Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples of at least two inches in diameter and with a length equal to the thickness of the bed.

5.

Verifying a system flow rate of 43,400 cfm z 10 percent during system operation when tested in accordance with ANSI N510.

C.

Verified by determining that the air distribution across the adsorber section is uniform, per ANSI H510, following original installation, modification or major repair.

D.

Demonstrated operable after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by:

1.

Verifying within 30 days after removal that a laboratory analysis of a carbon sample from at least one test canister or carbon sample removed from one of the charcoal adsorbers demonstrates a removal efficiency of 195% for radioactive methyl iodide when the sample is tested in accordance with ASTH D3803 (30*C, 95% R.H.).

The carbon samples not obtained from test canisters shall be prepared by either:

(a)

Emptying one entire bed from a removed adsorber tray, 166-~

mixing the adsorbent thoroughly, and obtaining samples of at least two inches in diameter and with a length equal to the thickness of the bed.

(b)

Emptying a longitudinal sample from an adsorber tray, 166-~

mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal i

to the thickness o# the bed.

Proposed Amendment No. 166 4-43a

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards

)

Soecification (Continued) 4.12.2 D.

2.

After reinsta11ation of the sampled adsorber tray per 1664-Specification 4.12.20.1:

(a) Verify that the charcoal adsorbers remove 199.5% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510 while 1664-operating the filter train at a flow rate of 43,400 cfm 10 percent.

(b) Verify that the HEPA filter bank removes 199.9% of the D0P when tested in-place in accordance with ANSI N510 while operating the filter train at a flow rate of 43,400 cfm 10%.

E.

Started on a manual signal and operated for 15 minutes in each 31-day period, if not already operating.

Bases The Auxiliary Building exhaust system consists of two full capacity units arranged to take suction from a common plenum, draw the air through HEPA and charcoal filter banks and discharge it into the plant vent. Only one unit is operated at a time, allowing the other to be serviced or held in reserve.

This system draws all of the potentially radioactively contaminated air in the plant, external to the Reactor Building, through it.

The following major areas are served:

1) Spent Fuel Building

2) Radio-Chemical Lab Hoods and Service Area

3) Radwaste Area

4) Haste Gas Discharge

5) Condenser Air Ejector Exhaust

6) Various Instrumentation and Sampling Discharges While providing service to these areas the filters are credited with a minimum DF of 10 for radioactive iodine and a DF of 100 for particulate matter which may be released in the following:

1) Letdown Line rupture outside the Reactor Guilding

2) Post LOCA Decay Heat Removal Leakage

3) Dropped Fuel Assembly in Spent Fuel Pool

4) OTSG Tube Rupture

5) Makeup Tank Rupture Releases of radioactive materials, and the resulting dosage from these accidents, are based on the maximum flow rate from the plant vent.

Reduced flow rates are conservative as to the effect of plant releases.

Shutdown of the entire system in response to a specific occurrence is likewise allowable at the operator's discretion. The negative pressure requirement for the Spent Fuel Building is to ensure that all potential releases following a dropped fuel assembly accident are drawn into the exhaust system, filtered, and monitored prior to release.

Proposed Amendment No. 166 4-43b l

i

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Successive Insoection Intervals Every 10 years thereafter (or Volumetric inspection of 1/3 of nearest refueling outage) the welds at the expiration of each 1/3 of the inspection interval with a cumulative 100 percent coverage of all welds.

Hoh - The welds selected during each inspection period shall be 166-~

distributed among the total number to be examined to provide a representative sampling of the conditions of the welds.

3.

Examinations that reveal unacceptable structural defects in a weld during an inspection under 4.13 A.2 shall be extended to require an additional inspection of another 1/3 of the welds.

If further unacceptable defects are detected in the second sampling, the remainder of the welds shall be inspected.

4.

In the event repairs of any welds are required following any examination during successive inspection intervals, the inspection schedule for the repaired welds will revert back to the first 10 year inspection program.

B.

For all welds in critical areas other than those identified as postulated break location on Figures 4.13-1, 2 and 3:

1.

Inservice inspection shall be performed in accordance with the provisions of paragraph 4.2 of these Technical Specifications.

C.

For all welds in the critical areas as identified on Figures 4.13-1, 2 and 3:

1.

A visual inspection of the surface of the insulation at all weld locations shall be performed on a weekly basis for detection of leaks. Any detected leaks shall be investigated and evaluated.

If the leakage is caused by a through-wall flaw, either the plant shall be shutdown, or the leaking piping isolated.

Repairs shall be performed prior to return of this line to service.

2.

Repairs, re-examination and piping pressure tests shall be 166-~

conducted in accordance with the rules of ASHE Code,Section XI.

Proposed Amendment No. 166 4-45

r RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Stc.ndards 4.17 STEAM GENERATORS Acolicabilitv Applies to inservice inspection of the steam generator tubes.

Obj ec tive To verify the operability and structural integrity of the tubing as part of the reactor. coolant boundary.

Specification Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requireraents of Specification 1.3.

4.17.1 Steam Generator Samole Selection and Insoection Steam generator tubing shall be demonstrated OPERABLE by: selecting and inspecting steam generators as specified in Table 4.17-1.

4.17.2 Steam Generator Tube Samole Selection and Insoection The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.17-2A.

The inspection result classification and the corresponding action required for inspection of "specific j

limited areas" (see paragraph 4.17.2e) shall be as specified in Table 4.17-2B.

The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 166-4.17.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.17.4. 'The tubes selected for these inspections shall include at least 3% of the total number of tubes in both steam generators and be selected on a random basis except:

a.

If experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.

Proposed Amendment No. 166 4-51

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.17.2 (continued) b.

The first sample inspection during inservice inspection (subsequent to the first inservice inspection) of each steam generator shall include:

1.

All nonplugged tubes that previously had detectable wall penetrations (>20%), and 2.

Tubes in those areas where experience has indicated potential problems.

c.

The second and third sample inspections during each inservice inspection may be less than a full tube inspection by concentrating (selecting at least 50% of-the tubes to be inspected) the inspection on those areas of the tube sheet array and on those portions of the tubes where tubes with imperfections were previously found.

166-d.

A tube inspection (pursuant to Specification 4.17.4a.5) shall be performed on each selected tube.

If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

("Adjacent" is interpreted to mean the nearest tube capable of being inspected.)

Tubes which do not permit passage of the eddy current probe will be considered as degraded tubes when classifying inspection results.

e.

Tubes in specific limited areas which are distinguished by unique operating conditions and/or physical construction (for example, tubes adjacent to the open inspection lane or tubes whose 15th tube support plate hole is not broached but drilled) may be excluded from random samples if all such tubes in the specific area of a steam generator are inspected.

No credit will be taken for these tubes in meeting minimum sample size requirements.

The results of each sample inspection shall be classified into one of the following three categories:

Cateaorv Insoection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

Note:

In all inspections, previously degraded tubes must exhibit significant (>10%) further wall penetrations to l

be included in the above percentage calculations.

1 Proposed Amendment No. 166 4-52

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS

-Surveillance Standards TABLE 4.17-3 OTSG Auxiliary Feedwater Header Surveillance OTSG A Soecial Interest Tuhti Row Tube No 5

1, 46 6

2, 3, 49, 50, 51 7

1, 2, 53, 54 8

1, 2, 56, 57 44 1, 119 45 1, 120 166-~

46 1, 119 47 1,'122 48 1, 123 49 1, 124 103 1, 124 105 1

106 1, 119 107 1, 120 108 1, 119 144 1, 2, 56, 57 145 1, 2, 53, 54 146 2, 3, 49, 50, 51 147 1, 46 Proposed Amendment No. 166 4-57b

l RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.17-3 (Continued)

OTSG Auxiliary Feedwater Header Surveillance OTSG B Soecial Interest Tubes Row Tube No 5

1, 46 6

2, 3, 49, 50 7

1, 2, 53, 54 8

1, 2, 56, 57 44 1, 119 45 1, 120 1664-46 1, 119 47 122 48 1, 123 49 1, 124 103 1, 124 104 123 105 122 106 1, 119 107 1, 120 108 1,119 144 1, 2, 56, 57 145 1, 2, 53, 54 146 2, 3, 49, 50, 51 1

147 1, 46 i

l I

Proposed Amendment No. 166 4-57c

,..n.

.n,

RANCH 0-SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.18 FIRE SUPPRESSION SYSTEM SURVEILLANCE 4.18.1 Instrumentation 4.18.1.1 Except for fire detection instruments inaccessible during power operation, each of the fire detection instruments in Table 3.14-1 shall be demonstrated OPERABLE at least semi-annually by a CHANNEL FUNCTIONAL TEST.

Fire detectors which are not accessible during plant operation shall be demonstrated OPERABLE by the performance of a CHANNEL FUNCTIONAL TEST during each COLD SHUTDOWN exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless performed in the previous 6 months.

166-~

4.18.1.2 The NFPA Standard 720 supervised circuits associated with the detector alarms for each of the fire instruments in Table 3.14-1 shall be demonstrated OPERABLE at least semi-annually.

4.18.1.3 Non-supervised circuits associated with detector alarms, between the instrument and the control room shall be demonstrated OPERABLE at least once per 31 days.

4.18.2 Hater System 4.18.2.1 The fire suppression water system shall be demonstrated OPERABLE:

a.

At least once per 7 days by verifying the contained water supply volume, b.

At least once per 31 days on a STAGGERED TEST BASIS by starting each electric motor driven pump and operating it for at least 15 minutes on recirculation flow.

c.

At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path is in its correct position as indicated by position instrumentation.

d.

At least once per 6 months by performance of a system flush to each test fixture.

e.

At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full

travel, f.

At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:

166-~

1.

Verifying that each pump develops at least 2000 gpm at a minimum pressure of 125 psig.

j Proposed Amendment No. 166 4-58

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.18.2.1 (Continued) 166--

2.

Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel, and 166-~

3.

Verifying that each fire suppression pump starts.

(sequentially) to maintain the fire suppression water system pressure greater than or equal to 80 psig.

g.

At least once per 3 years by performing a flow test of the system in accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association.

Proposed Amendment No. 166 4-59

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.18.2.1 (Continued) h.

The fire pump diesel engine shall be demonstrated OPERABLE:

1.

At least once per 31 days by verifying:

a.

The fuel storage tank contains at least 250 gallons of fuel, and b.

The diesel starts from ambient conditions and operates for at least 30 minutes on recirculation flow.

2.

At least once per 92 days by verifying that a sample of diesel from the fuel storage tank, obtained in accordance with ASTH-D270-65, is within the acceptable limits l

166-specified in Table 1 of ASTH-D975-78 with respect to i

viscosity, water content and sediment.

3.

At least once per 18 months, by subjecting the diesel to an inspection in accordance with procedures prepared in l

conjunction with its manufacturer's recommendations for the class of service.

i.

The fire pump diesel starting 24-volt battery bank and charger shall be demonstrated OPERABLE:

1.

At least once per 7 days by verifying that; a)

The electrolyte level of each battery is above the

(

plates, and b)

The overall battery voltage is greater than or equal to 24 volts.

2.

At least once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery.

3.

At least once per 18 months by verifying that:

l l

a)

The batteries, cell plates and battery racks show no l

visual indication of physical damage or abnormal deterioration, and b)

The battery-to-battery and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.

Proposed Amendment No. 166 4-60

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.3 FACILITY STAFF 00ALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Manager, Radiation Protection who shall meet or exceed 166-~

the qualifications of Regulatory Guide 1.8, Revision 2, and (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline.

The STA shall receive specific training in plant design, and response and analysis of the plant for transients and accidents.

6.4 TRAINING 6.4.1 A retraining and replacement training program for the operating staff shall be maintained under the direction of the Manager, Training and shall meet or exceed the requirements and recommendations of Section 166-5.5 of ANSI N18.1-1971 or ANS 3.1-1981 as endorsed by Regulatory Guide 1.8, Revision 2, and 10 CFR, Part 55.

6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Manager, Training and shall meet or exceed the requirements of Section 27 of the NFPA Code - 1975, except refresher classroom training shall be on a quarterly schedule.

6.5 REVIEH AND AUDIT 6.5.1 PLANT REVIEH COMMITTEE (PRC)

FUNCTION 6.5.1.1 The Plant Review Committee shall function to advise the AGM, Nuclear Power Production and Management Safety Review Committee on all matters related to nuclear safety.

COMPOSITION 6.5.1.2 The Plant Review Committee shall be composed of a chairman, and a minimum of six members.

i 1

I Proposed Amendment No. 166 6-3

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls Soecial Reoorts 166-6.9.5 Special reports shall be submitted to the Director of Nucicar Reactor Regulation, NRC within the time period specified for each report.

These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

A.

A one-time only, "Narrative Summary of Operating Experience" will be submitted to cover the transition period (calendar year 1977).

B.

A Reactor Building structural integrity report shall be submitted within ninety (90) days of completion of each of the following tests covered by Technical Specification 4.4.2 (the integrated leak rate test is covered in Technical Specification 4.4.1.1).

166-1.

Tendon Stress Surveillance 2.

End Anchorage Concrete Surveillance C.

In-Service Inspection Program D.

Reactor Vessel Haterial Surveillance Program E.

Status of Inoperable Fire Protection Equipment 30 days (3.14.1.2, 3.14.2.2, 3.14.3.2, 3.14.2.3, 3.14.5.2, 3.14.4.2, 3.14.6.2)

F.

Inoperable Emergency Control Room /TSC Ventilation Room Filter System G.

Radioactive Liquid Effluent Dose 30 days (3.17.2)

H.

Noble Gas Limits 30 days (3.18.2)

I.

Radio-Iodine anc' Particulates 30 days (3.18.3)

Proposed Amendment No. 166 6-12f i

4

. - i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls Soecial Reoorts (Continued)

J.

Gaseous Radwaste Treatment 30 days (3.19)

K.

Radiological Monitoring Program 30 days (3.22)

L.

Monitoring Point Substitution 30 days (3.22) 166-~

M.

Integrated Leak Rate Test 90 days (4.4.1.1.7)

N.

Fuel Cycle Dose 30 days (4.25) 0.

Deleted P.

Steam Generator Tube Inspection 30 days (4.17.5)

Proposed Amendment No. 166 6-129

RANCHO SECO UNIT 1.

TECHNICAL SPECIFICATIONS Administrative Controls 6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:

a.

Records and logs of facility operation covering time interval at each power level.

b.

Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety, c.

Licensee Event Reports, d.

Records of surveillance activities, inspections and calibrations required by these Technical Specifications, e.

Records of reactor tests and experiments, f.

Records of changes made to Operating Procedures.

g.

Records of radioactive shipments.

h.

Records of sealed source leak tests and results.

i.

Records of annual physical inventory of all sealed source material of record.

6.10.2 The following records shall be retained for the duration of the Facility Operating License:

a.

Record and drawing changes reflecting facility design modifications made to systems and equipment described in the 166-Updated Safety Analysis Report.

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Proposed Amendment No. 166 6-13

1

^

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ATTACHMENT III DESCRIPTION AND REASON FOR CHANGES Proposed Amendment No. 166 consists of various editorial and technical corrections to the Rancho Seco Technical Specifications. A description of and reason for each change is presented below:

1.

Change

Description:

List of Tables, page 1x:

Changed the page number for Table 3.3-1 from 3-22a to 3-22b.

Reason: The page number was changed to accommodate a text insertion on the previous page (see Item 12).

2.

Change Descriotion: Specification 1.4.2, page 1-2a: Changed "FSAR" to "USAR".

Reason:

Pursuant to 10 CFR 50.71, the Updated Safety Analysis Report (USAR) is the official updated version, issued annually, of the Final Safety Analysis Report (FSAR). Accordingly, the USAR is referenced in the Technical Specifications rather than the FSAR.

3.

Chanae Descriotion: Specifications 1.4.3, 1.4.4 and 1.4.5, page 1-3:

Changed "FSAR" to "USAR."

Reason: See the reason given for Change Description No. 2.

4.

Change Descriotion: Specification 1.9, page 1-5:

In first line, corrected spelling of word "accomodate" to "accommodate."

Reason: Typographical correction.

5.

Chanae Descriotion: Specification 2.3.1 Bases A, page 2-6:

In first paragraph changed:

"FSAR Section 14" to "USAR Section 14."

Reason: See the reason given for Change Description No. 2.

6.

Change Descriotion: Table 2.3-1, page 2-9:

a.

Item 9, added the words:

"(Turbine trip)."

b.

Added new Item 10:

"Anticipatory reactor trip (9) (Loss of both Main Feed Pumps)."

c.

Changed degree symbol to percent symbol in Note (2).

d.

Added Note (9):

"This trip is disabled below 20% power.

(power channel)."

Reason: Proposed Amendment No.140 (issued as Amendment No. 83 on 2/3/87) failed to provide the distinction that changing the arming threshold for anticipatory reactor trip from 20% to 45% of full power applied only to the Main Steam turbine, not to the Main Feed Pumps as well. This proposed change provides that distinction.

7.

Change Descriotion: Specification 3.1.2 References, page 3-4:

Corrected the ASHE Code title by adding the word:

"Vessel" in Reference (2).

Also deleted Reference (5), and renumbered Reference (6) and (7) as References (5) and (6).

Reason: The correction to the ASME title provides its proper designation.

The referenced FSAR paragraphs were changed to USAR in Proposed Amendment No. 139 (1/29/86).

Further review has determined that paragraph 4.4.4 as shown for Reference (5) is nonexistent in the USAR and is therefore deleted.

References (6) and (7) are renumbered accordingly.

8.

Change Descriotion: Specification 3.1.3 Bases, page 3-6:

In second and third paragraphs, deleted reference designator "(2)", and on page 3-7, deleted Reference (2).

Reason: The reference was changed to USAR by Amendment No. 87; however, further review has determined that paragraph 3.2.1.4 as shown for Reference (2) is nonexistent in the USAR and is therefore deleted.

9.

Change Descripl1GD: Specification 3.1.5, page 3-10:

In first line, changed "continous" to "continuous."

Reason: Typographical correction.

10. Change

Description:

Specification 3.1.8 Bases, page 3-15b: Added "of the USAR" after the number 7.2.2.1.3.

Reason: The proposed change provides more explicit identification.

11. Change Descriotion: Specification 3.3 Bases, page 3-22: At the top of the page, added the paragraph beginning:

"390,000 gallons of borated water..." Relocated the bottom paragraph and the References to page 3-22a, and moved Table 3.3-1 to a new page, 3-22b.

Page ix, List of Tables was changed to show the new page number for Table 3.3-1.

Reason: The above Bases paragroph was unintentionally deleted in Proposed Amendment No. 139, which was approved as Amendment No. 87.

Subparagraph (a) beneath the reinserted paragraph was similarly deleted.

However, in reviewing the previous amendments affecting this Bases, there is no accounting for the appearance of this subparagraph sometime between Amendments No. 4 (5/19/76) and No. 61 (3/4/85).

Subparagraph (a) states:

"Hotor operated valves shall be placed in the closed position and power supplies deenergized." Since the wording is irrelevant to the borated water storage tank being discussed, the District proposes to not reinsert subparagraph (a).

12. Change Descriotion: Specification 3.5.1 Bases, page 3-26: Added Note (1), "Pressure setpoint tolerances are 600 1 25 psig and 100 50 psid." Also corrected the differential pressure unit from 100 psig to 100 psid.

Reason: Note (1) was added for clarification, and is in accord with the actual EFIC design as described in Proposed Amendment No.152, approved as Amendment No. 93 (1/5/88). Changing the differential pressure unit to psid is a technical correction. l

13. Chanae Descriotion: Table 3.5.1-1, pages 3-30a and 3-30b: Note 1, changed the maximum number of houre, that a channel can be out of service for testing, calibration, or maintenance from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 12.

Reason: Surveillance testing of the RPS requires placing an EFIC channel in maintenance bypass, thus entering the conditions of Note 1.

Past experience indicates a minimum of an eight-hour shift for this RPS testing. Allowing for possible test interruptions by operations evolutions or minor discrepancies, an additional 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is proposed for contingency.

14.

Chanae Descriptiort: Specification 3.5.2 Bases, page 3-33b: Corrected specification designation from:

"3.5.2.5.D.1" to "3.5.2.5D.1" and "3.5.2.5.0.2" to "3.5.2.5D.2."

Reason: The correction properly designates the specification.

15.

Change Descriotion: Specification 3.5.2 References, page 3-33b:

Changed "3.2.2.1.2" in Reference (1) to "3.2.2.1.3."

Reason: Review of the USAR has determined that Section 3.2.2.1.3 is a more appropriate reference to the subject of hot shutdown reactivity as addressed in the Bases.

16.

Chanae

Description:

Specification 3.5.6 Bases, page 3-38h: Added Note (1), "Pressure setpoint tolerances are 600 25 psig and 100 1 50 psid." Also corrected the differential pressure unit from "100 psig" to "100 psid."

Reason: Note (1) was added for clarification, and is in accord with the actual EFIC design as described in Proposed Amendment No. 152, approved by Amendment No. 93 (1/5/88). Changing the differential pressure unit from "100 psig" to "100 psid" is a technical correction.

17.

Change Descriotion: Specification 3.7.2J, page 3-42j:

Deleted the phrase "if the switchyard voltage goes below 219 KV," between Action statements (a.) and (b.).

Reason: The deleted phrase is redundant with a similar phrase in Action statement (a.).

18. Change Descriotion: Specification 3.9 Bases, page 3-46a:

Replaced the reference to a specific operating procedure with the words "as required by operating procedures."

Reason: Replacement of the wording "per Operating Procedure A.21, Section 7.3" with less specific wording allows future renumbering of the procedure or the section without necessitating a technical specification amendment.

19.

Change

Description:

Table 3.14-1, page 3-55:

Increased the quantity of smoke detectors in Detection Zone 13 from four to eight, and in Detection Zone 16 from four to six.

Reason:

The additional smoke detectors were installed to comply with 10 CFR 50, Appendix R, Section III.G.2, and with the District's commitment made in LER 87-29.

20. Chanae Descriotion:

Specification 3.18.3 Bases, page 3-75: Changed "Evaluting" to Evaluating."

Reason: Typographical correction.

21. Chanae

Description:

Specification 3.20 Bases, page 3-79:

In-second paragraph, changes:

"Vol. VI FSAR" to "Vol. VIII, USAR.' In the next paragraph changed the abbreviation "Ci" to "curies," and changed "FSAR Table..." to "USAR Table..."

In last paragraph changed the word "cooling" to "coolant."

ReA1Qn:

In :eviewing the USAR, it was determined that

...tr contents of page 140-25 of Volume VIII are equivalent to page 14D-25 of Volume VI of the FSAR.

The changes from "FSAR" to "USAR" are as explained for Change Description No. 2 of this attachment.

Spelling out "curies" is consistent with other uses of the unit in Specification 3.20, and changing the word "cooling" to "coolant" is a typographical correction.

22.

Change

Description:

Tabie 4.1-1, Items 20 and 21, page 4-6: Changed the test frequency from monthly (H) to refueling interval (R).

Reason: Complete testing, which includes the manual pushbutton, cannot be performed during operation without initiating a manual trip. Char.nel testing, excluding the manual pushbutton, is presently performed morithly in compliance with Item 16 of Table 4.1-1.

This surveillance frequency is in accord with the Standard Technical Specifications Table 4.3-2, Note (1).

23. Chanae

Description:

Table 4.1-1, Item 29, page 4-7: Added the wording "Equivalent to Item 49, but excludes SPDS" to the Remarks column.

Reason: The added wording makes clear that the daily check is not redundant with Item 49 which is checked monthly.

24.

Change Descriotion:

Table 4.1-1, Item 39, page 4-7a: Added calibration per refueling interval for control rod relative position.

Reason: Control rod absolute and relative positions are checked, one compared to the other, once per shift.

Since the control rod absolute l

position is calibrated each refueling interval, it is appropriate for control rod relative position to be calibrated at the same time..

25.

Change Descriotion: Table 4.1-1, page 4-7b: Deleted Item 45, Emergency plant radiation instruments and Item 51, Reactor Coolant System Subcooling Margin Monitor.

Reason:

Item 45, Emergency plant radiation instruments, was deleted because this equipment (hand-held radiation meters) does not fall under the applicability of Specification 4., which states:

"' $ lies to items directly related to safety limits and limiting cuditions for operation during power operation.

During cold shutdown, systems and components required to maintain safe shutdown will be tested."

Item 51 is redundant with Item 65.

26. Change Descriotion:

Table 4.1-1, pages 4-7d and 4-7e: Added remark for Items 68.c 68.h, 71.b, 71.c, and 71.d that pressure transmitters are to be calibrated annually.

Reason: The results of design engineering calculations indicate that these EFIC pressure transmitters, located outside of containment, require annual calibration rather than on a refueling interval as specified for the remainder of each of these channels.

27. Change Descriotion:

Table 4.1-1, page 4-7g: Changed the Channel Description of Item 85 from "Hide Range OSTG Pressure Pressurizer Level" to "Hide Range OTSG Pressure."

Inserted new Item 86 with the Channel Description "Pressurizer Level," and renumbered existing Items 87 through 90.

The Check, Test and Calibrate notations are the same for both Item 85 and 86.

Reason: Combining the two Channel Description titles was a typographical error.

28.

Change

Description:

T ele. 4.1-3, page 4-9: Added note designator

"(3)" to Item 6.

Deleted Items 7, 8 and 10, and renumbered remaining items. Also deleted Note (4) and renumbered Note (5) as Note (4).

Reason:

The addition of Note (3), "Not performed during cold shutdown," permits maintenance on the Concentrated Boric Acid Storage Tank during shutdown without violating Specification 3.2.1.

Items 7, 8 and 10, and Note (4) are moved to a new table by Proposed Amendment No. 155 (RETS).

29.

Change Descriotion:

Specification 4.2.2 Bases, page 4-12:

In second paragraph, changed "4.2.8" to "4.2.1.2."

Reason:

For editorial purposes, Specification 4.2.8 was moved to Specification 4.2.1 by Amendment No. 76 (9/30/85).

Through oversight, the corresponding number change was not mad 0 in the Bases, and is herein proposed for correction.

30.

Change

Description:

Specification 4.4.1.1.7, page 4-17: Added specific requirement for reporting any deterioration noted in the containment surfaces inspection performed before each integrated leak rate test.

Reason: Clarification is provided that such reporting is required by 10 CFR 50, Appendix J, Section V.B, and "Special Report" Specification 6.9.5H.

31.

Change

Description:

Specification 4.4.1.4, page 4-19: Changed title of specification from "Annual Inspection" to "Containment Surfaces Inspection." Deleted the requirement for an annual inspection.

Also changed the term "integrated leak test" to "integrated leak rate test,"

and referenced "Special Report" Specification 6.9.5M.

In addition, deleted all but the first sentence from the second paragraph of the Bases, and moved the top three lines of Bases, page 4-20 to page 4-19.

Reason: An annual inspection of the containment surfaces is not required by 10 CFR 50, App. J,Section V.A only an inspection prior to an integrated leak rate test.

The change in test terminology is editorial, and addition of the special report reference is for clarification.

The Bases statement that periodic tests are to be performed without preliminary leak detection surveys or leak repaire, etc., is deleted as an option allowed by I & E Notice 85-71, and 10 CFR 50, Appendix J.

The remaining deleted sentences from the Bases concerns reduced pressure testing, an option which is not performed at Rancho Seco.

32.

Change Descriotion: Specification 4.4.2.5, page 4-23: Deleted that part of the first sentence calling for report preparation and review by the District's Generation Engineering Department, and changed "Specification 6.9-1" to "Specification 6.9.5.B."

Also deleted Specifications 4.4.2.6, 4.4.2.6.2, 4.4.2.6.3 and 4.4.2.6.4.

Reason: Reporting responsibilities are established by administrative procedures, and thus need not be made a Technical Specification requirement.

The number 6.9-1 is an incorrect designation for a specification. Specification 6.9.1 discusses general reporting requirements.

Specification 6.9.5B is specific to the type of report being discussed in Specification 4.4.2.5.

The Liner Plate Surveillance specified in 4.4.2.6, 4.4.2.6.2, 4.4.2.6.3 and 4.4.2.6.4, was only required after the first year of initial operation and is no longer applicable.

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33. Chanae Descriotion: Specification 4.5.3.1, page 4-32, and Specification 4.5.3 Bases, page 4-33:

Changed the combined maximum allowable leakage from components ir. the Decay Heat Removal System and Reactor Building Spray System from 6.0 gallons per hour (gph) to 0.63 gph.

Reason: Amendment No. 57 dated October 30, 1984, approved increasing the combined maximum allowable leakage from components in the Decay Heat Removal System and the Reactor Building Spray System from the original Technical Specifications value of 0.63 gph to 6.0 gph.. Current radiation dose calculations for the Control Room /TSC give acceptable results using 0.63 gph leakage, but give unacceptable results using 6.0 gph.

Because su veillance experience has found this combined maximum leakage to be well below the 0.63 gph value, and in order to achieve acceptable dose calculations for the Control Room /TSC, the District proposes this conservative reduction to 0.63 gph for the combined maximum allowable leakage.

34. Change Descriotion: Specification 4.5.3.2.B2, page 4-33: Changed the term "pump isolation" to "pump suction isolation."

Reason: The proposed wording change corrects the term, and makes it consistent with the wording in Specification 4.5.3.2.Bl.

35. Chanae Descriotion: Specifications 4.6.3C.6 and 4.6.3C.7, page 4-34g:

Added Note 2 for Specification 4.6.3C.6 as follows:

"If Surveillance Requirement 4.6.3.C.4.b is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test.

Instead, the diesel generator may be operated between 2550 and 2750 kw for A and B, or between 3000 and 3300 kw for A2 and 82 for one hour or until operating temperature has stabilized."

Also in Specification 4.6.3C.7, changed the 2000-hour rating shown for diesel generators A and B from 2850 kw to 2750 kw.

Reason: The addition of Note 2 for Specification 4.6.3C.6 eliminates an unnecessary 24-hour diesel run if the test is unsatisfactory.

The purpose of the 24-hour test is to demonstrate functional capability at full-load temperature conditions.

This is achieved by Note 2, and is consistent with the technical specifications at other nuclear plants.

The 2850 kw previously shown in Specification 4.6.3C.7 is the actual 2-hour rating, not the 2000-hour rating which is 2750 kw, as changed.

36.

Change Descriotion: Specification 4.9, page 4-40: Changed the wording "Atomic Energy Commission" to "Nuclear Regulatory Commission."

ReAion: Title correction.

37. Change Descriotion: Specification 4.11.10.l(b), page 4-42a: Added a comma after "adsorber tray" and "thoroughly."

Reason:

Editorial 38.

Chanae Descriotion:

Specification 4.11.10.2, page 4-42a: Changed the designation "D.1" to "Specification 4.11.1D.1," and made a similar change to Specification 4.12.20.2 on page 4-43b.

Reason:

This change is proposed for clarification.

I

39. Change

Description:

Specifications 4.12 and 4.12.1 A, page 4-43:

Expanded the wording in the Applicability statement for Specification 4.12 to include the condition when the crane is being operated with loads over the fuel storage pool.

In Specification 4.12.1A, changed 90 days fuel decay to 30.

Reason: Surveillance Requirement (SR) 4.12 and its corresponding LC0 j

Specification 3.13 were approved by the NRC in Amendment No. 39 dated September 13, 1982.

The accompanying NRC Safety Evaluation Report (SER) stated for the LC0 specification that the auxiliary and spent fuel building filter system must be operating whenever spent fuel movement occurs unless the spent fuel has decayed for a contirsous 30-day period.

The SER goes on to say that after such a decay pericJ (39 days) the doses would be less than those calculated for the fuel har#',g accident with the auxiliary and spent fuel building filter system operating.

Accordingly, for consistency with the LC0 Specification 3.13, the required period of fuel decay during which the spent fuel storage pool building exhaust ventilation system must be operating is changed from 90 days to 30 days.

The Applicability statement in Specification 4.12 is expanded to include crane operation to make it consistent with the wording in Specification 4.12.1 A.

40. Change Descr1Dtion: Specifications 4.12.28.1, 4.12.2B.2, 4.12.2b.3, and 4.12.20.2, pages 4-43, 4-43a, and 4-43b: Changed the qualifier for the flow rates shown as "not exceeding cfm i 10 percent" to "flow rate of cfm 10 percent.'

Reason: The use of the words "not exceeding" when referring to a numerical parameter with a specified tolerance is technically incorrect.

41. Change Descriollon: Specifications 4.12.2D.1(a) and (b), page 4-43a:

Added comma after "thoroughly."

Reason:

Editorial, and for consistency with Change Description No. 38.

42. Change

Description:

Specification 4.13A.2, Note page 4-45: Changed "exmained" to "examined," and in Specification 4.13C.2, changed "Section XI Code" to "ASPE Code,Section XI."

Reason: Typographical and editorial corrections.

43.

Change Descriotion: Specification 4.17.2, page 4-51:

Changed "Specification 4.17-3" to "Specification 4.17.3."

Reason: The numerical designation with a hyphen as used in Specification 4.17.2 is proper for identifying a figure or a table in the format of the technical specifications.

The corrected designation is proper for referring to the specification.

44. Change Descriotion: Specification 4.17.2.d, page 4-52:

Changed "Specification 4.17.4.5" to "Specification 4.17.4a.5."

RE11DD: Specification 4.17.4a.5 is the proper designation.

45. Change

Description:

Table 4.17-3, pages 4-57b and 4-57c: Changed the tube number in Row 46 from "121" to "119 "

l Reason: Tube No. 119 is the correct number of the Special Interest Tube in Row 46.

46. Chanae

Description:

Specification 4.18.1.2, 4-58:

Changed "NFPA Standard 720" to "NFPA Standard 720."

Reason: Typographical correction.

47.

Change Descriotion: Specification 4.18.2.1, pages 4-58 and 4-59:

Deleted subparagraph f.1 which states "Verifying that each automatic valve in the flow path actuates to its correct position." Also renumbered subparagraphs f.2, f.3 and f.4 accordingly.

Reason: The only automatic valves in the fire suppression water system are surveillance tested by Specification 4.18.3.lc.l.

Thus Specification 4.18.2.lf.1 is deleted because of redundancy.

48.

Change Descriotion: Specification 4.18.2. lh.2, page 4-60:

Changed "ASTH-D975-74" to "ASTH-0975-78."

Reason: The change represents the updated version of the ASTM method used at Rancho Seco for diesel oil analysis.

49. Chanae

Description:

Specification 6.4.1, page 6-3: Deleted reference to Appendix A of 10 CFR 55. Also changed qualification references in Specifications 6.3.1 and 6.4.1.

Reason: Appendix A of 10 CFR 55 has been incorporated directly into 10 CFR 55.

Changes in qualification references ware erlitorial updates.

50.

Chanae Descriotion: Specification 6.9.5, page 6-12f: Changed "the Regulatory Operations Regional Office" to "Nuclear Reactor Regulation, NRC."

Reason: Editorial update.

51.

Chanae Descriotion: Specification 6.9.5B, page 6-12f:

Deleted Specifications 6.9.58.1 and 6.9.58.4, and renumbered Specifications 6.9.5B.2 and.3 as Specifications 6.9.58.1 and.2.

Added 6.9.5H, "Integrated Leak Rate Test" on page 6-12g.

Reason: An annual structural integrity inspection is not required by 10 CFR 50, Appendix J, on'y an inspection prior to an integrated leak rate test. A report of the structural integrity inspection is required to be included in the Integrated Leak Rate Test Special Report, but only if the inspection reveals abnormalities.

The Liner Plate Surveillance was required before and after the initial pressure test, and within approximately one year after that.

This surveillance was discontinued after the first year per Specification 4.4.2.6.4 and therefore is deleted.

A special report is required of the Integrated Leak Rate Test results per 10 CFR 50, Appendix J, and by Specification 4.4.1.1.7 within 90 days of the test completion.

_g_

52. Change

Description:

Specification 6.10.2a, page 6-13:

Changed "Final Safety Analysis Report" to "Updated Safety Analysis Report."

Reaion: The reason for the change is as' explained in Change.

.i Description No. 2.

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ATTACHMENT IV SAFETY ANALYSIS FOR NO SIGNIFICANT HAZARDS CONSIDERATIONS The proposed changes, described in Attachment III, have been evaluated by the District using the standards of 10 CFR 50.92.

The conclusion has been reached that the changes constitute no significant hazards to the public.

Following are the bases for the District's conclusion:

1.

Except for Proposed Change Nos. 13, 22, 35 and 39, all the rest of the changes are purely administrative, consisting of typographical and grammatical corrections, renumbering and relocating of specifications, updating references to codes and standards, and specification clarifications.

The proposed administrative changes have been judged to not constitute a significant hazards consideratic,n because the changes:

a.

Would not involve a significant increase in the probability or consequences of an accident previously evaluated since the proposed, purely administrative changes do not affect existing plant designs or cause changes to existing plant operations, b.

Would not create the possibility of a new or different kind of accident from any accident previously evaluated since the proposed changes, purely administrative, do not introduce any new operational requirement that could affect plant safety.

c.

Hould not involve a significant reduction in the margin of safety since the purely administrative changes, by their very character, could not cause a reduction in the conservative nature of the Technical Specifications.

2.

Proposed Change Nos. 13, 22, 35, and 39 which are not purely administrative, are discussed individually below:

a.

Proposed Change No. 13 increases to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months the existing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months during which one EFIC channel at a time can be out-of-service for surveillance testing, calibration and maintenance.

The 6-hour time in Note 1 of Table 3.5.1-1, Instrument Operating Conditions, was based on the estimated time required to perform surveillance testing solely of an EFIC channel. A condition that was not considered in estimating the EFIC channel out-of-service duration was the circumstance of RPS channel surveillance testing during which the associated EFIC channel is disabled.

RPS surveillance testing requires at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and allowing for minor calibrations or maintenance, the District considers a 12-hour period to be a more realistic duration for the EFIC channel to be disabled for RPS testing.

The District considers the 12-hour tiEe duration to be acceptable because the remaining EFIC channels must be OPERABLE, and in all cases the required LCO operator ACTION if one channel is inoperable is greater than the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months proposed in this change.

The 12-hour out-of-service allowance would permit surveillance of the RPS channel to be completed without entering the associated ACTION.

The District has reviewed the above proposed change against the standards of 10 CFR 50.92, and concludes, for the fcilowing reasons, that no significant hazards would result.

Extension of the EFIC channel out-of-service time duration from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months would not:

1.

Involve a significant increase in the probability or consequence of an accident previously evaluated, because redundancy maintains sufficient instrumt it channels in operation to protect the plant from any previously evaluated accident.

11.

Create the possibility of a new or different kind of accident from any accident previously evaluated, because the proposed change does not lessen the operator's ability to maintain the plant in a safe condition under all circumstances.

iii.

Involve a significant reduction in the margin of safety, because the proposed time extension is well within the allowed instrument inoperability duration before operator ACTION to preserve the safety margin would become necessary.

b.

Proposed Change No. 22 involves the surveillance test frequency of High Pressure Injection, Reactor Building isolation, and Reactor Building emergency cooling manual trip (Items 20 and 21 of Table 4.1-1, Instrument Surveillance Requirements) from monthly (H) to refueling intervals (R). Complete testing of these instrument channels includes the manual pushbutton which, if actuated during operation, initiates a manual trip of the associated channel.

Since ths channel testing, excluding the manual pushbutton, is performed monthly in accordance with Item 16 of Table 4.1-1, the District considers complete testing on a refueling interval frequency to be an acceptable time span for ccafirming the operability of the manual pushbutton.

The Dictrir.t has reviewed the above proposed change against the standards of 10 CFR 50.92, and concludes, for the reasons presented below that no significant hazards would result.

Extension of the test frequency of High Pressure Injection, Reactor Building isolation, and Reactor Building emergency cooling manual trip from monthly to refueling intervals would not: l

i.

Involve a significant increase in the probability or consequences of an accident previously evaluated, because the manual pushbuttons for which the time extension of surveillance testing is requested are very reliable devices, and testing on a refueling interval frequency is consistent with the trip test frequency for the other SFAS channels (Items 22 through 27 of Table 4.1-1)-

On this precedent, it is concluded that there is no increase in the probability or consequences of an accident previously evaluated.

ii.

Create the possibility of a new or different kina of accident from any accident previously evaluated, because the proposed change does not intrcJuce new equipment or design, nor lessen the ability to maintain the plant in a safe condition.

iii.

Involve a significant reduction in the margin of safety, because the proposed extension in test frequency is consistent with the test frequency for the other SFAS channels for which their margin of safety has been found accsotable, c.

Proposed Change No. 35 involves Specification 4.6.3.C.6, which presently requires an auto-start of the emergency o3esel generator within 5 minutes of completion of the surveillance 24-hour run.

The specification does not address what step to take in the event that the restart does not occur within the prescribed 5-minute period, or is otherwise not satisfactorily initiated or completed.

This lack of alternate instructions implies that another 24-hour run is necessary before the restart can again be attempted.

It should be explained that the purpose of the diesel generator rest 6ft within 5 minutes of the 24-hour run is to demonstrate the functional capability of the diesel to start and pick up load while still in the full-load temperature condition.

Note 2, as follows, was added to Specification 4.6.3C.6:

"If Surveillance Requirement 4.6.3C.4.b is not satisfactorily completed, it is not necessary to repeat the 24-hour test.

Instead, the diesel genarator may be orerated between 2550 and 2750 kw for A and B, or between 3000 and 3300 kw for A2 and 82 for one hour or until operating temperab re has stabilized."

t The addition of Note 2 eliminates an unnecessary 24-hour run while still accomplishing the surveillance required restart of the diesel while at ful'3-load operating temperature.

This alternate surviillance instruction is consistent with the Technical Specification provision at other nuclear plants.

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1 Because the alternate surveillance step presented in Note 2 of Specification 4.6.3C.6 results in the required surveillance demonstration of the diesel's functional capability, the District concludes that the proposed change would not:

i.

Involve-a significant increase in the probability or consequences of an accident previously evaluated. The proposed alternate surveillance step still accomplishes the required surveillance objective while eliminating an unnecessar.y 24-hour run of the diesel. Since the proposed change involves equipment undergoing surveillance testing, and not in actual operational use, an increased probability or consequences of an accident already evaluated is not foreseen.

ii.

Create the possibility of a new or different kind of accident from any accident previously evaluated, because the simple elimination of an unnecessary run of the diesel generator cannot be foreseen as contributing to the creation of the possibility of a new or different kind of accident from any accident previously cvaluated.

iii.

Hould not involve a significant reduction in the margin of safety, because the proposed change, the addition of an alternate step in surveillance testing, will have no effect on the design or operation of the plant.

d.

Proposed Change No. 39 consists of a revision to the wording of Surveillance Requirement 4.12.lA that presently requires verifying the operation of the Spent Fuel Storage Pool Building exhaust ventilation system when the irradiated fuel has decayed for less than 90 days.

The proposed change requires that the system be verified to be operating for the first 30 days of fuel decay for consistency with LC0 Specification 3.13.

The NRC's Safety Evaluation Report for Amendment No. 39 (September 13, 1982) which introduced both the Limiting Condition for Operation (LCO) and the Survelliance Requirement (SR) stipulating the fuel decay period for the system's operation stattd:

..., it is our position, based upon the doses calculated for a fuel handling accident that the auxiliary and spent fuel building filter system must be coerating whenever spent fuel movement occurs unless the spent fuel has decayed for a continuous 30-day period. After such a decay period, the doses would be less than those calculated for the fuel handling accident with the auxiliary and spent fuel building filter system operating."

The LC0 (Specification 3.13.2) is in accord with the above NRC statement.

Specification 3.13.2 states:

"One auxiliary and Spent Fuel Buildirig filter system must be operating whenever spent fuel moveinent is being made unless the spent fuel has decayed fer a continuous 30-day period."

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1 Since there is no further LCO requirement that the systen remsin OPERABLE for an additional 60 days of fuel decay, the District proposes to change Surveillance Requirement 4.12.1A to be consistent with the LCO.

It should be noted that verification of system operability every 31 days, and the method for doing so, is addressed in SR Specification 4.12.2A.

The District has reviewed the above proposed change against the standards of 10 CFR 50.92, and concludes, for the following reasons, that no significant hazards would result.

Reducing the verification of system operation from 90 days to 30 days would not:

1.

Involve a significant increase in the probability or consequence of an accident previously evaluated, because, as stated in the NRC's SER after a 30-day decay period, the doses would be less than those calculated for the fuel handling accident with the auxiliary and spent fuel building filter system operating.

ii.

Create the possibility of a new or different kind of accident from any previously evaluated, because the requirement for system operation until fuel decay has reached 30 days remains unchanged.

iii.

Involve a significant reduction in the margin of safety, because the proposed change does not affect the limiting conditions of plant operation.

Based on the evaluation of the proposed changes against the standards of 10 l

CFR 50.92, the District concludes that Proposed Amendment No. 166 does not involve a significant hazards consideration.

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ML20150D186

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