ML20212E407

Proposed Tech Specs 3.1.6, Leakage & 3.8, Fuel Loading & Refueling, Reflecting Replacement of Existing Radiation Monitors R-15001 a Through E W/Radiation Monitors R-15100 & R-15044

ML20212E407
Person / Time
Site:	Rancho Seco
Issue date:	02/20/1987
From:	SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
Shared Package
ML20212E368	List: ML20212E364 ML20212E407
References
TAC-64879, NUDOCS 8703040346
Download: ML20212E407 (3)

Text

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

When the source of leakage has been identified, .the situation can be evaluated to determine if operation can safely continue. This evaluation will be perfonned by the Operating Staff and will be documented in writing and approved by the Superintendent. Under these conditions, an allowable reactor coolant system leakage rate of 10 gpm has been established. This explained leakage rate of 10 gpm is also well within the capacity of one high pressure injection pump and makeup would be available even under the loss of off-site power condition.

If leakage is to the Reactor Building it may be identified by one or more of the following methods:

A. Sump Levels - All Reactor Building leakage is collected in the Reactor Building sumps. These sumps drain by gravity into a 120 gallon Reactor Building drain accumulation tank. The drain accumulation tank is used to measure the drain flow with level indicators at 20 gallons and 120 gallons. The tank is dumped into the East decay heat removal pump room sump. The frequency of dumping the accumulation tank and time interval between levels are recorded in the Control Room and are direct measures of the flow rate. Depending on the level at which the tank is dumped, the time to confirm a 1 gpm leak is between 40 minutes and 120 minutes.

Frequency of operation of the East DHR pump room sump pumps is r recorded in the control room to provide verification of proper operation of the Reactor Building drain accumulation tank.

Since the Reactor Building drain system collects drainage from all components in the Reactor Building, a change in drain flow does not necessarily indicate a reactor coolant system leak. One method available for determining if the additional drain flow is reactor coolant is to collect drainage in the drain accumulation tank, draw a sample from the tank, and analyze the sample for boric acid

' concentration and radioactivity.

I B. Radioactivity - Changes in the reactor coolant leakage rate in the I Reactor Building may cause changes in the control room indication of j 156> the Reactor Building atmosphere particulate radioactivity.

] '. The response time for the radiation monitor to detect a given leak rate is dependent on the coolant activity level, building

, equilibrium level, and the detector sensitivity. The radiation detection monitor's particulate channel has the ability to detect a small unidentified leak of approximately 1 gpm in approximately one hour. This is a qualitative leak detection method and the actual

< 1eakage rate can be determined from the other detection systems.

Proposed Amendment No.156 r 3-14 8703040346 870220 PDR ADOCK 03000312 P PDR

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RANCH 0'SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions' for Operation

' 156x C. . Reactor Coolant Inventory - Total reactor coolant' system leakage -

rate is periodically determined by comparing indications of reactor power, coolant temperature, pressurizer water level and makeup tank level over a time interval. All of these indications are recorded.

Since the pressurizer level is maintained essentially constant by the pressurizer level controller, any coolant leakage is replaced by coolant from the makeup tank resulting in a tank level decrease.

The makeup tank capacity is 31 gallons per inch of height and each

. graduation on the level recorder. represents 1 inch of tank height.

This inventory monitoring method is capable of detecting changes on the order of 31 gallons. A 1 gpm leak would therefore be detectable within approximately one-half hour.

As described above, in addition to direct observation, the means of detecting reactor coolant leakage are based on two different~

principles, i.e., activity and sump level and reactor coolant inventory measurements. Two systems of different principles provide, therefore, diversified ways of detecting leakage to the reactor building.

The upper limit of 30 gpm is based on the contingency of a complete loss of plant power. A 30 gpm loss of water in conjunction with a complete loss of plant power and subsequent cooldown of the reactor coolant system by the

, turbine bypass system (set at 1,040 psia) and steam driven emergency i feedwater pump would require more than 60 minutes to empty the pressurizer from the combined effect of system leakage and contraction. This will be ample time to restore electrical power to the plant and makeup flow to the reactor coolant system.

I j The plant is expected to be operated in a manner such that the secondary coolant will be nomally. maintained within those chemistry limits found to

result in negligible corrosion of the steam generator tubes. If the I secondary coolant chemistry is consistently not maintained within these chemistry limits, over some period of time localized corrosion could occur i and might result in stress corrosion cracking. The extent of cracking during 1

plant operation would be limited by the limitation of steam generator tube i leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 1 GPM). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of

!: safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 1 GPM can be detected by monitoring the secondary coolant. Leakage in excess of this limit will require plant 4 shutdown during which the leaking tubes will be located and plugged.

Proposed Amendment No. 156 '

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-RANCHO SECO UNIT ~1 s -

, TECHNICAL SPECIFICATIONS-

.; Limiting Conditions for Operation-f 3.8.8 ; ' When two irradiated fuel assemblies are being handled simul'taneou' sly '

within the fuel transfer canal, a minimum of 10 feet separation shall .

, - be maintained between the assemblies at all times. Irradiated fuel

, assemblies may be handled with the auxiliary bridge crane provided no-other irradiated fuel assembly is being handled in the fuel transfer canal.

3.8.9 If any of .the above specified limiting conditions for fuel loading and refueling are net met, movemerit of fuel _into the reactor core

- shall cease; action shall be initiated to correct the conditions so that the spectfied limits are met, and no operations which may increase the reactivity of the core shall be made. -

3.8.'10- The Reactor Building purge system, including the reactor building.

156x stack radiation monitor shall be tested and verified to be operable immediately prior to refueling operations. '

h 3.8.11 Irradiated fuel shall not be removed from the reactor until the unit has been subcritical for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

3.8.12 No loads will be handled over irradiated fuel stored in the spent- .

fuel pool, except the fuel assemblies themselves. - A dead weight load.

test at the rated load will be performed on the Fuel Storage Building handling bridge prior to each refueling.

Bases "

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Detailed written procedures will be available for use by refueling personnel. - These procedures, the above specifications, and the design of the fuel handling equipment, as described in subsection 9.7_of the FSAR i incorporating built-in interlocks and safety features, provide assurance that i no incident could occur during the refueling operations that would result in -

a hazard to public health and safety. If no change is being made _in core geometry, one flux monitor is sufficient. This permits maintenance on the ,

instrumentation. Continuous monitoring of radiation levels and neutron flux r provides immediate indication of an unsafe condition.

pump ia used to maintain a uniform boron concentration.{he The refueling decay heat rem boron concentration indicated in Specification 3.8.4 will be maintained to ensure that the more restrictive of the following reactivity conditions is met:

1. Either a keff of 0.95 or less with all control rods removed from the core.

2. A boron concentration of >1800 ppm.

The actual calculated boron concentration for item (1) above is 1974 ppm boron. Specification 3.8.5 allows the control room operator to inform the Reactor Building personnel of any impending unsafe condition detected from the main control board indicators during fuel movement.

The specification requiring testing Reacto/ Building purge termination is to i verify that these components will function as required should a fuel handling l accident occur that results in the release of significant fission products.  :

Proposed Amendment No. 156 "

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