ML19312B855

Proposed Revision to Tech Specs 3.1.6 & 4.17 Re Provisions for Surveillance of Steam Generator Tubing

ML19312B855
Person / Time
Site:	Oconee
Issue date:	06/21/1977
From:	DUKE POWER CO.
To:
Shared Package
ML19312B850	List: ML19312B849 ML19312B855
References
NUDOCS 7911250060
Download: ML19312B855 (6)

Text

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• f- e 3.1.6 Lukaga Specification I

3.1.6.1 If the total reactor coolant leakage rate exceeds 10 gpm, the 1 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 evapora-tive 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 iritiated 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 reactor coolant system leakage exceeds 1 gpm through the steam generator tubes, 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 shutdown condition within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

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 =agnitude, of the leak shall be considered in this evaluation.

The safety evaluation shall assure that the exposure of offsite personnel to radiation is within the guidelines of 10CFR20.

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 t;he 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 syste=s 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 vent header and from which coolant can be returned to the reactor cool-ant 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.

Bases '

Every reasonable effort will be made to reduce reactor coolart leakage in-cluding evaporative losses (which may be on the order of .5 gpm) to the lowest possible rate and at least below 1 gpm in order.co prevent a large 3.1-14 7911250 0 6 O

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- 1erk'from masking th. presence of a smaller leak. Water inventory balances, radiation monitoring equipment, boric acid crystallina deposits, and physical

'insptctions can disclose reactor coolant leaks. Any leak of radioactive fluid, whether from the reactor coolant system pri=ary boundary or not can be a serious problam 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.

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. ,

1 When the source of leakage has been identified, the situation can be evaluated i 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 1 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 inj ection 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. The reactor 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 gym, assuming corrosion product activity and no fuel cladding leakage. Under these conditions, an increase in coolant' leakage of 1 gpm is detectable within 10 minutes after it occurs,

b. The iodine monitor, gasecus monitor and area monitor are not as sensitive to corrosion product activity.(1) It is calculated that the iodine monitor is sensitive to an 8 spm 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 may 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 sump. A 1 gym leak would therefore be detectable within less than 10 minutes.

3.1-15

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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 stecage 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 ecmplete loss of station power. A 30 gpm loss of water in conjunction with a complete loss of station pcwer and subsequent cooldown of the reactor coolant system by the turbine bypass system (set at 1,040 peia) and steam driven emergency feedwater pump would require more than 60 minutes to empty the pressurizer from the ecm-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.

REFEREh[CES

?SAR Section 11.1.2.4.1 3.1-ic 7/19f74

N 4.17 STEAM GENERATOR TUBING SURVEILLANCE Applicability Applies to the steam generator tubing surveillance.

Objective To define the in-service surveillance program for steam generator tubing.

Specification 4.17.1 Examination Methods In-service inspection of steam generator tubing shall include non-destruc-tive examination by eddy-current testing or other equivalent techniques.

The inspection equipment shall provide a sensitivity that will detect defects with a penetration of 20 percent or more of the minimum allowable as-manufactured tube wall thickness.

4.17.2 Selection and Testing The selection and testing of tubes shall be made on the basis of the following:

a. The examination =ay be from one steam generator and shall include 1 1/2%

of the total installed steam generator tubes for the unit.

b. Every inspection shall include all tubes which previously had detectable wall penetrations (greater than 20 percent and not including plugged tubes), and shall also consider tubes and those areas where design and experience have indicated potential problems.

If the inspection indicates more than 10 percent of the inspected tubes have detectable wall penetrations (greater than 20 percent), an additional inspection encompassing 3 percent of the tubes in both steam generators shall be examined, concentrating on areas of the tube array where the tubes with defects were found. In the event the inspection of these additional tubes indicates that more than 10 percent of the tubes examined in any single steam generator have detectable wall penetrations (greater than 20 percent), an additional inspection encompassing 3 percent of the tubes in both steam generators shall be examined.

4.17.3 Insoection Intervals

a. The inservice examinations of steam generator tubing shall be performed during 40 month periods except as indicated in Specification 4.17.3.b and 4.17.3.c.

b. If in any examination of steam generator tubing an excess of 10 percent of the tubes exhibit indications in excess of 20 percent of the wall thickness, the next two inspections shall be performed at 12 to 24 month intervals. If, in these examinations, no more than 10 percent of the tubes examined exhibit either additional degradation (greater than 10 percent of wall thickness) of previously degraded tubes, tubes

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with new dsiccts ... exessa of 20 percent of w211 thickness, or a combin -

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tion of both, thn inspection interval may continus at 40 months.

c. Additional inservice examinations shall be performed during the shutdown subsequent to any of the following conditions:

1) a seismic occurrence greater than the operating basis earthquake,

2) a loss-of-coolant accident requiring actuation of the engineered safeguards system,

3) a main steam line or feedwater line break.

4.17.4 Acceotance Criteria

a. If less than 10 percent of the total tubes inspected have detectable wall penetration (greater than 20 percent), operation may resume af ter required corrective measures have been taken.

b. If more than 10 percent of the total tubes inspected have detectable wall penetrations (greater than 20 percent), operation =ay resume after required corrective measures have been taken, and the situation and remedial action shall be reported to the NRC.

4.17.5 Corrective Measures All tubes with unacceptable bufects (greater than 40 percent wall thinning) shall be plugged.

4.17.6 Reports

a. Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the NRC within 30 days.

b. The results of the steam generator tube inservice inspection shall be included in the Annual Operating Report for the period in which this inspection was completed. This report shall include:

1) N:.mber and extent of tubes inspected,

2) Locatien and percent of wall-thickness penetration for each indication of an imperfection.

3) Identification of tubes plugged.

Bases The program of periodic in-service inspection of steam generators provides the means of monitoring the integrity of the tubing and to =aintain surveil-lance in the event there is evidence of mechanical damage or progressive deterioration due to design, manufacturing errors, or operating conditions.

In-service inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that correc-tive measures =ay be taken. In-service inspection includes non-destructive examination using a suitable eddy-current inspection system (or other equivalent techniques), capable of locating and identifying defects due

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, . .to , stress corrosion cracking, mtchnnical damage, chemical wistage, or othsr causes.

An unacceptable defect is defined as one which would result in not satisfying the calculated acceptable minimum tube wall thickness that can sustain a loss-of-coolant accident in combination with a saf e shutdown earthquake.

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