ML15112B016

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Revised Tech Specs Pages Per Order Modifying License, Requiring Periodic Surveillance Over Life of Plant & Specifying Limiting Conditions for Operation of Primary Coolant Sys Pressure Isolation Valves
ML15112B016
Person / Time
Site: Oconee  Duke Energy icon.png
Issue date: 04/20/1981
From:
Office of Nuclear Reactor Regulation
To:
Shared Package
ML15112B014 List:
References
NUDOCS 8104270264
Download: ML15112B016 (9)


Text

Attachment 3 ATTACHMENT TO ORDER FOR MODIFICATION OF FACILITY OPERATING LICENSES NOS. DPR-38, DPR-47 AND DPR-55 DOCKETS NOS. 50-269, 50-270 AND 50-287 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.

The revised pages contain vertical lines indicating the area of change.

3.1-14 3.1-14a (new) 3.1-15 3.1-16 4.5-1 4.5-2

4.

5-3 4.5-4 There are no changes to pages 3.1-16,,4.5-1 and 4.5-.4.

They are Provided to maintain document completeness.

83 1 5427

3.1.6 Leakage Specification 3.1.6.1 If the total reactor coolant leakage rate exceeds 10 gpm, the reactor shall be shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection.

3.1.6.2 If unidentified reactor coolant leakage (excluding normal evaporative losses) exceeds.1 gpm or if any reactor coolant leakage is evaluated as unsafe, the reactor shall be shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection.

3.1.6.3 If any reactor coolant leakage exists through a non-isolable fault in a RCS strength boundary (such as the reactor vessel, piping, valve body,. etc., except the steam generator tubes), the reactor shall be shutdown, and cooldown to the cold shutdown condition shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection.

3.1.6.4 If at any time, the leakage through the Unit 1 steam-generator tubes equals or exceeds 0.3 gpm, a reactor shutdown shall be initiated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and the reactor shall be in a cold condition within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. If the leakage is less than 0.3 gpm, an assessment shall be made whether operations may be continued safely or the plant should be shut down.

In either case, the NRC shall be notified in accordance with Section 6.6.2.1.

3.1.6.5 If reactor shutdown is required by Specification 3.1.6.1, 3.1.6.2 or 3.1.6.3, the rate of shutdown and the conditions of shutdown shall be determined by.the safety evaluation for each case and justified in writing as soon thereafter as practicable.

3.1.6.6 Action to evaluate the safety implication of reactor coolant leakage shall be initiated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of detection. The nature, as well as the magnitude, of the leak shall be considered in this evaluation. The safety evaluation shall assure that the exposure of offsite personnel to radia tion is within the guidelines of 10 CFR 20.

3.1.6.7 If reactor shutdown is required per Specification 3.1.6.1, 3.1.6.2, 3.1.6.3 or 3.1.6.4, the reactor shall not be restarted until the leak is repaired or until the problem is otherwise corrected.

3.1.6.8 When the reactor is critical and above 2% power, two reactor coolant leak detection systems of different operating principles shall be operable, with one.of the two systems sensitive to radioactivity. The systems sensitive to radioactivity may be out-of-service for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided two other means to detect leakage are operable.

3.1.6.9 Loss of reactor coolant through reactor coolant pump seals and system valves to connecting systems which vent to the gas yent header and from which coolant can be returned to the reactor coolant system shall not be considered as reactor coolant leakage and shall not be subject to the consideration of Specifications 3.1.6.1, 3.1.6,2, 3.1.6.3, 3.1.6.4, 3.1.6.5, 3.1.6.6 or 3.1.6.7 except that such losses when added to leakage shall not exceed 30 gpm.

3.1.6.10

a.

The maximum allowable leakage for valves CF-12, CF-14, LP-47 and LP-43 shall be as follows:

3.1-14 Order dtd. 4/20/81

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 per missible 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 differential test pressure sha1l not be less than 150 psid.
c. All pressure isolation valves listed in Section 4.5.1.2.3 shall be functional as a Dressure isolation device, except as specified in

.1.6.10.d. Valve leakage shall not exceed.the amounts in Section 3.1.6.10.a.

d. In the event that integrity of any pressure isolation valve in Section 4.5.1.2.3 cannot be demonstrated, reactor operation may continue, provided that at least two valves in each high pressure line having a non-functional valve are in, and remain in the mode corresponding to the isolated condition. (Motor operated vaives shall be placed in the closed.

position and power supplies deenergized).

e. If 3.1.6.10.c and d. cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Bases Every reasonable effort will be made to reduce reactor coolant leakage including evaporative losses (which may be on the order of,5 gpm) to the lowest possible rate and at least below 1 gpm in order to prevent a large leak from masking the presence of a smaller leak. Water inventory balances, radiation monitoring equip ment, boric acid crystalline deposits, and physical inspections can disclose reactor coolant leaks. Any leak of radioactive fluid, whether from the reactor coolant system primary boundary or not can be a serious problem with respect to in-plant radioactivity contamination and cleanup or it could develop into a still more serious problem; and therefore, first indications of such leakage will be followed up as soon as practicable.

3.1-14a Order dtd. 4/20/81

Although some leak rates on the order of GPM may be tolerable from a dose point of view, especially if they are to closed systems, it must be recog-.

nized that leaks in the order of drops per minute through any of the walls of the primary system could be indicative of materials failure such as by stress corrosion cracking. If depressurization, isolation and/or other safety measures are not taken promptly, these small breaks could develop into much larger leaks, possibly into a gross pipe rupture. Therefore, the nature of the leak, as well as the magnitude of the leakage must be considered in the safety evaluation.

When the source of leakage has been identified, the situation can be evaluated to determine if operation can safely continue. This evaluation will be per formed 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 g;m has been established. This explained leakage rate of 10 gpm is also well within the capacity of one high pressure injection pump and cakeup 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. The reaccor building air particulate monitor is sensitive to low leak rates. The rates of reactor coolant leakage to which the instrument is sensitive are.10 gpm to greater than 30 gpm, assuming corrosion product activity and no fuel cladding leakage. Under these conditions, an increase in coolan: leakage of 1 gpm is detectable within 10 minutes after it occurs.
b. The iodine mcnitcr, gaseous monitor and area monitor are not as sensitive to corrosion produc activity. 1) It is calculated that the iodine monitor is sensitive to an 8 gpm leak and the gaseous monitor is sen sitive to a 230 gpm leak based on the presence of tramp uranium (no fission products from tramp uranium are assumed to be present). However, any fission products in the coolant will make these monitors more sensitive to coolant leakage.
c.

In addition to the radiation monitors, leakage is also monitored by a level indicator in the reactor building normal sump. Changes in normal sump level =ay be indicative of leakage from any of the systems located inside the reactor building such as reactor coolant system, low pressure service water system, component cooling system and steam and feedwater lines or condensation of humidity within the reactor building atmosphere.

The sump capacity is 15 gallons per inch of height and each graduation on the level indicates 1/2 inch of sump height. This indicator is capable of detecting changes on the order of 7.5 gallons of leakage into the su=p. A 1 gp= leak.would therefore be detectable within less than

-10 minutes.

3.1-15 Order dtd. 4/20/81

d. Total reactor coolant system leakage rate is periodically determined by comparing indications of reactor power, coolant temperature, pressurizer water level and letdown storage tank level over a time interval. All of these indications are recorded.

Since the pressurizer level is main tained essentially constant by the pressurizer level.controller, any coolant leakage is replaced by coolant from the letdown storage tank resulting in a tank level decrease. The letdown storage 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 2 different principles, i.e., activity, sump level and reactor constant 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 station power. A 30 gpm loss of water in conjunction with a complete loss of station power and subsequent cooldown of the reactor coolant system by the turbine bypass system (set at 1,040 psia) and steam driven emergency feedwater pump would require more than 60 minutes to empty the pressurizer from the com bined effect of system leakage and contraction. This will be ample time to restore electrical power to the station and. makeup flow to the reactor coolant system.

REFERENCES FSAR Section 11.1.2.4.1 3.1-16

4.5 EMERGENCY CORE COOLING SYSTEMS AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 Emergency Core Cooling Systems A.cDolicability Applies to periodic testing requirements for the Emergency Core Cooling Systems.

Objective To verify that the Emergency Core Cooling Systems are operable.

SDecification 4.5.1.1 System Tests 4.5.1.1.1 High Pressure Injection System

a. During each refueling outage, a system test shall be conducted to demonstrate that the system is operable. A test signal will be applied to demonstrate actuation of the High Pressure Injection System for emergency core cooling operation.
b. The test will be considered satisfactory if control board indication verifies that all components have responded to'the actuation signal properly; all appropriate pump breakers shall have opened or closed and all valves shall have completed their travel.

4.5.1.1.2 Low Pressure Injection.System

a. During each refueling outage, a system test shall be conducted to demonstrate that the system is operable. The test shall be performed in accordance with the procedure summarized below:

(1) A test signal will be applied to.demonstrate actuation of -the Low Pressure Injection System for emergency core cooling operation.

(2) Verification of the engineered safety features function of the Low Pressure Service Water System which supplies cooling water to the low pressure coolers shall be made to demonstrate operability of the coolers.

The test will be considered satisfactory if control board indication verifies that all components have responded to the actuation signal properly; all appropriate pump breakers shall have opened or closed, a'd all valves shall have completed their travel.

4.5.1.1.3 Core Flooding System

a. During each refueling outage, a system test shall be conducted to demon strate proper operation of the system. During pressurization of the Amendments Nos.

96, 96 & 93 4.3-1

Reactor Coolant System, verification shall be made that the check and isolation valves in the core flooding tank discharge lines operate properly.

b. The test will be considered satisfactory if control board indication of core flood tank level verifies that all valves have opened.

4.5.1.2 Component Tests 4.5.1.2.1 Pumps Quarterly, the high pressure and low pressure injection pumps shall be started and operated to verify proper operation. Acceptable performance will be indicated if the pump starts, operates for 15 minutes, and the discharge pressure and flow are within + 10 percent of a point on the pump head curve.

(Figures 4.5.1-1 and 4.5.1-21.

4.5.1.2.2 Valves - Power Operated

a. Quarterly, each-Engineered Safety Features valve in the Emergency Core Cooling Systems, except LP-17, -18, and each Engineered Safety Features valve associated with emergency core cooling in the Low Pressure Service Hater System shall be tested to verify operability. LP-17, -18 shall only be tested every cold shutdown unless previously tested during the current quarter.
b. The acceptable performance of each power-operated valve will be that motion is indicated upon actuation by appropriate signals.
c. During each refueling outage, low pressure injection pump discharge (engineered safety features) valves, low pressure injection discharge throttling valves, and low pressure injection discharge header crossover valves shall be cycled manually to verify the manual operability of these power-operated val ves.

4.5.1.2.3 Check Valves Periodic individual leakage testing (a) of valves CF-12, CF-14, LP-47 and LP-48 shall be accomplished prior to power operation after every time the plant is.

placed in the cold shutdown condition for refueling, after each time the plant is placed in a cold shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomplished in the preceding 9 months, and prior to returning the valve to service after maintenance, repair or replacement work is performed. Whenever integrity of these valves cannot be demonstrated, the integrity of the remaining valve in each high pressure line having a leaking valve shall be determined and recorded daily. In addition, the position of the other closed valve located in the high pressure piping shall be recorded.daily. For the allowable leakage rates and limiting conditions for operation, see Technical Specification 3.1.6.10.

Bases The Emergency Core Cooling Systems are the principle reactor safety features in the event of a loss of coolant accident. The removal of heat from the core provided by these systems is designed to limit core damage.

(a)

To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if.

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

4.5-2 Order dtd. 4/20/8i

The High Pressure Injection System under normal operating conditions has one pump operating.

At least once per month, operation is rotated to another high pressure injection pump.

This verifies that the high pressure 'injection pumps are operable.

The requirements of the Low Pressure Service Water System for cooling water are more severe during normal operation than under accident conditions.

Rotation of the pump in operation on a monthly basis verifies that twb pumps are operable.

The low pressure injection pumps are tested singularly for operability by opening the borated water storage tank outlet valves and the bypass valves in the borated water storage tank fill line. This allows water to be pumped from the borated water storage tank through each of the injection lines and back to the tank.

Testing the manual. operability of power-operated valves in the Low Pressure Injection System gives assurance that flow can be established in a timely manner even if the capability to operate a valve from the control room is lost.

With the reactor shut down, the valves in each core flooding line are checked for operability by reducing the Reactor Coolant.System Pressure until the indicated level in the core flood tanks verify that the check and isolation valves have opened.

Power Operated Valves LP-17, -18 and Check Valves CF-12, CF-14., LP-47and LP-48, are boundary valves between high pressure and low pressure design piping. As such, functionally testing of these valves is per formed during cold shutdown conditions when the Reactor Coolant System pressure is below the design pressure of the.Low Pressure Injection System piping and the potential for over-pressurization of the low pressure system is eliminated.

REFERENCE (1) FSAR, Section 6 4.5-3 Order dtd. 4/20/81

1~

40 NPSN 7r-30 7000~ ~

~ ~~r TONl 2

25o 5000 20 15 3000 2000 0

100 200 300 400 500 500 Capacity, Rp HIGH PRESSURE INJECTION PUMP CHARACTERISTICS E

UEPa OCONEE NUCLEAR STATION Figure 4.5.1-1 4.5-4