ML14175A834

Amend 83 to License DPR-23,revising Tech Specs to Incorporate Section 4.05 of Westinghouse STS

ML14175A834
Person / Time
Site:	Robinson
Issue date:	09/10/1984
From:	Varga S Office of Nuclear Reactor Regulation
To:
Shared Package
ML14175A835	List: ML14175A834
References
DPR-23-A-083 NUDOCS 8409260369
Download: ML14175A834 (30)

Text

pt REGUZ UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 CAROLINA POWER AND LIGHT COMPANY DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.8 3 License No. DPR-23

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Carolina Power and Light Company (the licensee) dated October 24, 1983 and December 12, 1983, as supplemented March 28, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the-Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-23 is hereby amended to read as follows:

6409260369 840910 PDR ADOCK 05000261 P

PDR

-2 (B) Technical Specifications The Technical Specifications contained in Appendices A and B. as revised through Amendment No.83, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 4 e n

.A ga, Ch ie f Operating Reactors Bra ch #1 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: September 10, 1984

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 83 FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Revise Appendix A as follows:

Remove Pages Insert Pages 3.5-9 3.5-9 4.1-1 thru 4.1.10a 4.1-1 thru 4.1-14 4.2.1 thru 4.2-25 4.2-1 thru 4.2-9 4.5.2 thru 4.5-6 4.5-2 thru 4.5-4

Tl T I 1,t1 1 1 RAI i

ING CO

(

11NIiTgI 113 I

2 3

MINLMUM OPERATOTR ACTION MINIMUM DEGREE IF CONDITIONS OF OPERABLE OF COLUMN I OR 2 No.

FUNCTIONAL UNIT CHANNELS REDUNDANCY CANNOT BE MET

12.

A) Lo Steam Generator 2

1 Maintain Hot Shutdown Water level

13.

Underfrequency 4 KV System 2

1 Maintain Hot Shutdown

14.

Undervoltage on 4 KV System 2

1 aintain Hot Shutdown

15.

Control Rod Misalignment Monitor****

a.

Rod Position Deviation 0

Log individual rod positions once/hour, and after a load change >10% or after >30 inches of control rod motion

b.

Quadrant Power Tilt 1

0 Log Individual upper and lower Monitor (upper and lower ion chamber currents once/hour ex-core neutron detectors) and after a load change >10% or after >30 inches of control rod motion (D

16.

Steam Flow/Feedwater Flow 1

0 Maintain Hot Shutdown ML match rD

17.

Low Steam Generator Water Level I0 intain Hot Shutdown 00 02*For zero power physics testi ng It Is permissible to take one channel out of service.

SWei~it two) of foir power chaitne s are greer thani 10%Z ft.I power hoot shutdown Is not requr red.

1410112 O

t two liteedIte range cha nnel s l.8 greater t hian 0

amps, hot sh catdown is not re(aIt r

red.

Ik*

f botht rod ml naigsiterit wtoni toirt (a anid b) are i noperable for two hours or more, theo nuiclear overpower trip s

ie.

reset to 93 percetii of rated power In addion to the inocreased surve ioance rtioted.

o.

i dpted Power

4.0 SURVEILLANCE REQUIREMENTS Specified intervals may be adjusted plus or minus 25% to accomodate normal test schedules. Performance of any surveillance test outlined in these specifications is not required when the system or nommnnet is onu tF ae, as permitted by the Limiting Conditions for Operation.

Prior to returning the system to service, the specified calibration and testing surveillance shall be performed.

4.0.1 Surveillance Requirements for inservice inspection.and testing of ASME Code Class 1, 2 and 3 components shall be applicable as follows:

a.

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

b.

Surveillance intervals specified in Section II of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

ASME Boiler and Pressure Vessel Required frequencies for Code and applicable Addenda perfotding inservice terminology for inservice inspection and testing inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days

c.

The combined time interval for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.

d.

Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.

e.

Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

4.1-1 Amendment No. 83

4.1 OPERATIONAL SAFETY REVIEW Applicability Applies to item directly relAte4 to safety li4ts and liiin.g c

i or operation.

Objective To specify the miningm frequency and type of surveillance to be applied to plant equipment and conditions.

Specification 4.1.1 Calibration, testing, and checking of instrumentation channels shall be performed as specified in Table 4.1-1.

4.1.2 Sampling tests shall be conducted as specified in Table 4.1-2.

4.1.3 Equipment tests shall be conducted as specified in Table 4.1-3.

Basis Check Failures such as blown instrument fuses, defective indicators, faulted amplifiers which result in 'upscale' or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system. Furthermore, such failures are, in many cases, revealed by alarm or annunciator action, and a check supplements this type of built-in surveillance.

Based on experience in operation of both conventional and nuclear plant systems, when the plant is in operation, the mininum checking frequencies set forth are deemed adequate for reactor and steam system instrumentation.

4.1-2 Amendment No.

83

Calibration Calibration is performed to ensure the presentation and acquisition of accurate information.

The nuclear flux (linear level) channels daily calibration against a thermal power calculation will account for errors induced by changing rod patterns and core physics parameters.

Other channels are subject only to the "drift" errors induced within the instrumentation itself and, consequently, can tolerate longer intervals between calibration. Process system instranentation errors induced by drift can be expected to remain within acceptable tolerances if recalibration is performed at intervals of each refueling shutdown.

Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures.

Thus, the minimum calibration frequencies set forth are considered acceptable.

Testingi Minimum testing frequency is based on evaluation of asLsafe failure rate data and reliability analysis.

This is based on operating experience at conventional and nuclear plants. An "unsafe failure" is defined as one which negates channel operability and which, due to its nature, is revealed only when the channel is tested or attempts to respond to a bona fide signal. The minimum testing frequency for those instrument channels connected to the safety system is based on an average unsafe failure rate of 2.5 x 10-6 failure/hr per channel.

4.1-3 Amendment No. 83

For a specified test interval W and an H out of N redundant system with identical and independent channels having a constant failure rate X, the average availability A is given by:

W A

WQ N-+

2 I

N!

W (1-!+2)!

(M-1)!

where A is defined as the fraction of time during which the system is functional, and Q is the probability of failure of such a system during a time interval W.

For a 2-out-of-3 system A - 0.9999968, assuning a channel failure rate, X

equal to 2.5 x 10-6 hr-1 and a test interval, W, equal to 720 hrs.

This average availability of the 2-out-of-3 system is high, hence the test interval of one month is acceptable.

Because of their greater degree of redundancy, the 1/3 and 2/4 logic arrays provide an even greater measure of protection and are thereby acceptable for the same testing interval. Those items specified for monthly testing are associated with process components where other means of verification provide additional assurance that the channel is operable, thereby requiring less frequent testing.

Reference (1) FSAR Section 7.2.2 4.1-4 Amendment No.

83

TABLE 4.1-1 MINIMUM FREQIENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS Channel Description Check Calibrate Test Remarks

1. Nuclear Power Range S

D (1)

B/W (2)

(1)

Thermal Power calculations

• (3) during power operations R* (3)

(2)

Signal to 9r; bistable action (permissive, rod stop, trips)

(3)

Upper and lower chambers for symmetric offset:

monthly during power opera tions. When periods of re actor shutdown extend this interval beyond one month, the calibration shall be performed immediately fol lowing return to power.

2. Nuclear Intermediate Range S (1)

N.A S/U (2)

(1) Once/shift when in service (2)

Log level; bistable action (permissive, rod stop, trip)

3. Nuclear Source Range S (1)

N.A S/U (2)

(1) Once/shift when in service (2)

Btatable action (alarm, trip)

4. Reactor Coolant Temperature S

R B/W (1)

(1) Overtemperature -AT (2)

(2) Overpower -AT (D

5.

Reactor Coolant Flow S

R M

0

6.

Pressurizer Water Level S

R M

7.

Pressurizer Pressure S

M M

8. 4 Kv Voltage N.A R

M Reactor Protection circuits only

• By means of the movablt in-core detector system

TABLE 4.1-1 (Continued)

MINIMUM FREQUIJINC[ES FOR CIIIECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS Channel DescrLption Check Calibrate Test Remarks

9.

Analog Rod Position S (1,2)

R M

(1)

With step counters (2)

Following rod motion in excess of six inches when the computer is out of service

10. Rod Position Bank Counters S (1,2)

N.A.

N.A.

(1) Following rod motion in ex cess of aix inches when the computer to out of service (2)

With analog rod position

11. Steam Generator Level S

R H

12. Charging Flow N.A.

R N.A.

13. Residual Heat Removal Pump Flow N.A.

R N.A.

14. Boric Acid Tank Level D (1)

R N.A.

(1) Bubbler tube rodded weekly

15. Refueling Water Storage W

R N.A.

'rank Level

16. Boron Injection Tank Level W

R N.A.

17.

Volume Control Tank Level N.A.

R N.A.

18. Containment Pressure D

R B/W (1)

(1) Containment isolation valve signal

19. Radiation Monitoring System*

1)

R M

20. Boric Acid Makeup Flow Channel N.A.

R N.A.

• Process

ysteiwi rdiation moonitoru aid area radiation monitors.

TABLE 4.1-1 (Continued)

MINIMUM1 FREQUENCIES FOR CHECKS, CALIBRATIONS, AND TEST OF INSTRUMENT CHANNELS Channel Descripti.on Check Calibrate Test Remarks

21.

Containment Sump Level N.A.

R N.A.

22.

Turbine Trip Setpoint**

N.A.

R R

23.

Accumulator Level and Pressure S

R N.A.

24.

Steam Generator Pressure S

R M

25.

Turbine First Stage Pressure S

R M

26.

DELETED

27.

Logic Channel Testing N.A.

N.A.

M(1)

(1)

During hot shutdown and power operations.

When periods of reactor cold shutdown and re fueling extend this interval beyond one month, the test shall be performed prior to start-up.

28.

Turbine Overspeed Protection N.A.

R M

TrLp Channel (Electrical)

(D

29.

4 kv Frequency N.A.

R R

(D,

30.

Control Rod Drive Trip Breakers N.A.

N.A.

M o

31.

Overpressure Protection System N.A.

R t

• • SLop valve chosure or low auto stop oil pressure.

TABLE 4.1-1 (Continued)

MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUHENf CHANNELS Channel Description Cheek Calibration Test Remarks

32.

Loss of Power

a.

480 Emerg. Bus N.A.

R R

Undervoltage (Loss of Voltage)

b.

480 Emerg.

Bus N.A.

R R

Undervoltage (Degraded Voltage)

33.

Auxiliary Feedwater Flow****

N N.A.

R Indication

34.

Reactor Coolant System**

H R

N.A.

Subcooling Monitor

35.

PORV Position Indicator***

N.A.

N.A.

R

36.

PORV Blocking Valve***

N.A.

N.A.

R Position indicator

37.

Safety Relief Valve Position***

Indicator N.A.

N.A.

R (D

38.

Pending -

Refer to CP&L Letter NA Dated February 22, 1982 cO

TABLE 4.1-1 (Continued)

MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CAINELS Channel Description Check Calibration Test Remarks

39.

Steam/Feedwater Flow Mismatch N.A.

R H

40.

Low Steam Generator Water Level N.A.

R H

Instrumentation for Detection of Inadequate Core Cooling -

NUREG 0578 Item 2.1.3.b.

Direct Indication of Power Operated Relief Valve and Safety Valve Position -

NUREG 0578 Item 2.1.3.a.

AA**

Auxiliary Feedwater Flow Indication to Steam Generator - NUREG 0578 Item 2.1.7.b.

S At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Q

At least once per 92 days D

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> S/U Prior to each reactor startup W

At least once per 7 days if not performed in the previous seven (7) days 11/W At least once per 14 days R

At least once per 18 months M

At least once per 31 days N.A. -

Not applicable O

CO

TABLE 4.1-2 FREQUENCIES FOR SAMPLING TESTS Maximum Time Between Check Frequency Tests

1.

Reactor Coolant Samples - Gross Activity (1) Minimum I Per 72 hrs.

3 days Radiochemical (2) Monthly 45 days Radiochemical for I per 6 mos. (6)(7) 6 months E Determination

- Isotopic Analysis I per 14 days (7) 14 days for Dose Equivalent 1-131 Concentration Isotopic Analysis a) Once per 4 for Iodine Inlud-hours (8) ing 1-131, 1-133 b) One sample (9) and 1-135 Tritium Activity Weekly 10 days Cl & 02 5 day/week 3 days

2.

Reactor Coolant Boron concentration Twice/week 5 days Boron

3.

Refueling Water Boron concentration Weekly 10 days Storage Tank Water Sample

4.

Boric Acid Tank Boron concentration Twice/week 5 days

5.

Boron Injection Boron concentration Weekly (5) 10 days Tank

6.

Spray Additive NaOH concentration Monthly 45 days Tank

7.

Accumulator Boron concentration Mbnthly 45 days

8.

Spent Fuel Pit Boron concentration Prior to Refueling NA*

9.

Secondary Coolant Gross activity Minimum 1 Per 72 hrs.

3 days Isotopic Analysis a) 1 per 31 days (10) for Dose Equivalent b) I per 6 months (11) 1-131 Concentration

10.

Stack Gas Iodine 1-131 and particulate Weekly (3) 10 days

& Particulate radioactivity Samples releases

11.

Steam Generator Primary to secondary 5 days/week 3 days Samples tube leakage 4.1-10 Amendment No.

8 3

NOTES TO TABLE 4.1-2 (1)

A gross activity analysis shall consist of the quantitative measurement of the total radioactivity of the primary coolant in units of Xi/gram (2)

A radiochemical analysis shall consist of the quantitative measurement of each radionuclide with half life greater than 30 minutes making up at least 95% of the total activity of the primary coolant.

(3)

When iodine or particulate radioactivity levels excceed 10% of the limit in Specification 3.9.2.1, the sampling frequency shall be increased to a minimum of once each day.

(5)

The boron concentration in the boron injection tank shall be checked immediately after any actuation of the safety injection system that might result in dilution of the boron concentration in the boron injection tank.

(6)

Sample to be taken after a minimum of 2EFPD and 20 days of power operation have elapsed sInce the reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

(7)

Samples are to be taken in the power operating condition.

(8)

Sample taken at all operating conditions whenever the specific activity exceeds 1.0 uL/gram DOSE EQUIVALENT 1-131 or 100/ E Xi/gram.

These samples are to be taken until the specific activity of the reactor coolant system is restored within its limits.

(9)

One sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a thermal power change exceeding 15 percent of the rated thermal poder within a one hour period. Samples are required when in the hot shutdown or power operating modes.

(10)

Sample whenever that gross activity determination indicates iodine concentrations are greater than 10% of the allowable limit.

(11)

Sample whenever the gross activity determination indicates iodine concentrations are below 10% of the allowable limit.

NA* -

Not applicable.

4.1-11 Amendment No.83

TABLE 4.1-3 FREQUENCIES FOR EQUIPMENT TESTS Maximum Time Be tween Check Frequency Tests

1.

Control Rods Rod drop times of Each refueling shutdown NA*

all full length rods

2.

Control Rod Partial movement Every 2 weeks during 20 days of all full length reactor critical rods operations

3.

Pressurizer Set point Each refueling shutdown NA Safety Valves

4.

Main Steam Set point Each refueling shutdown NA Safety Valves

5.

Containment Iso-Functioning Each refueling shutdown NA lation Trip

6.

Refueling System Functioning Prior to each refueling NA Interlocks shutdown

7.

Service Water Functioning Each refueling shutdown NA System

8.

Fire Protection Functioning Monthl-45 days Pump and Power Supply

9.

Primary System Evaluate Daily when reactor NA Leakage coolant system is above cold shutdown condition

10.

Diesel Fuel Fuel Inventory Weekly 10 days Supply

11.

Critical Headers 100 Psig Hydro-Every five years 6 years of Auxiliary static Test Coolant System

12.

Turbine Steam Closure Monthly during power 45 days Stop, Control, operation and prior Reheat Stop, to startup and Interceptor Valves 4.1-12 Amendment No. 83

TABLE 4.1-3 (Continued)

FREOUENCIES FOR EOUIPMENT TESTS Maximum Time Check Freauency Eetween Tests

13.

Turbine Inspection Visual, Magnaflux Every five years 6 years and Die Penetrant i~

Fans ad tds Once per operating NA charcoal and Abso-Laboratory tests on cycle.

lute Filters for charcoal must show Residual Heat

> 99% iodine removal.

Removal Compartments In-place test must (HVE-5a and 5b) show > 99% removal of polydispersed DOP particles by the REPA filters and Freon by the charcoal filters.

15.

Isolation Seal Functioning Each refueling shutdown NA Water System

16.

Overpressure Functioning Each refueling shutdown NA Protection System

17.

Primary Coolant Functioning

1. Periodic Jeqkage testing(a)('

System check on each c valve listed valves in Table 3.1-1 shall be accomplished prior to entering reactor operation condition (1) after every time the plant is placed in the cold shutdown condition for refueling, (2) 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 pre ceding 9 months, (3) after maintenance, repair or replacement work is performed.

(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.

(b )Iinimum test differential pressure shall not be less than 150 psid.

(c)More than one valve may be tested in parallel.

The combined leakage shall not exceed 5.0 gpm.

Redundant valves in each line shall not be tested in series.

4.1-13 Amendment No.83

TABLE 4.1-3 (Cuntinued)

Maximum Time Check Fre uency Between Test

2. Whenever integrity of a pressure isolation valve listed in Table 3.1-1 cannot be demonstrated the integrity of the remaining valve "n each high pressure line having a lcaking 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.

4.1-14 Amendment No.83

4.2 PRIMARY SYSTEM SURVEILLANCE Applicability Applies to in-service structural surveillance of the reactor vessel and primary system boundary.

Objective To assure the continued integrity of the primary system boundary.

Specification 4.2.1.1 Inservice Inspection of Steam Generator Tubes 4.2.1.1.1 Tube Inspection Entry from the hot-leg side with examination from the point of entry completely around the U-bend to the top support of the cold leg is considered a tube inspection.

4.2.1.1.2 Sample Selection and Testing Selection and testing of steam generator tubes shall be made on the following basis:

(a) One steam generator shall be inspected during inservice inspection in accordance with the following requirements:

1. The inservice inspection may be limited to one steam generator on a rotating sequence basis.

This examination shall include at least 9% of the tubes if the results of the first or a prior inspection indicate that all three generators are performing in a comparable nanner.

4.2-1 Amendment No. 83

2.

When other steam generators are required to be examined by Table 4.2-2 and if the condition of the tubes in one or more generators is found to be more severe than in the other steam generators, the steam generator sampling sequence at the subsequent inservice inspection shall be modified to examine the steam generator or generators with the more severe condition.

(b) The minimum sample size, inspection result classification and the associated required action shall be im conformance with the requirements specified in Table 4.2-2.

The results of each sampling examination of a steam generator shall be classified into the following three categories:

Category C-1:

less than 5% of the total amber of tubes examined are degraded bu none are defective.

Category C-2:

Between 5% and 10% of the total number of tubes examined are degraded, but none are defective or one tube to not more than 1% of the sample is defective.

Category C-3:

More than 10% of the total number of tubes examined are degraded, but none are defective or more than 1% of the sample is defective.

In the first sample of a given steam generator during any inservice inspection, degraded tubes not beyond the plugging limit detected by the prior examinations in that steam generator shall be included in the above percentage calculations, only if these tubes are demonstrated to have a further wall penetration of greater than 10% of the nominal tube wall thickness.

(c) Tubes shall be selected for examination primarily from those areas of the tube bundle where service experience has shown the most sdere tube degradation.

4.2-2 Amendment No. 83

(d) The tubes examined in a given steam generator during the first examination of any inservice inspection shall include all non plugged tubes in that steli generator that from prior examination were degradpd, plus additional tubes are required to satisfy the minimum sample size specified in Table 4.2-2.

If any selected tube does not permit passage of the eddy current probe for a tube inspection, this shall be recorded and in adjacent tube shall be selected and subjected to a tube inspection. This information shall be included in the report required by Specification 4.2.1.3.2.

(e) During the second and third sample examinations of any inservice inspection, the tube inspection may be limited to those sections of the tube lengths where imperfections were detected during the prior examination.

(f)

During subsequent Lnservice inspections, the tube inspection may be limited to certain areas of the tube sheet array and those sections of the tube lengths where imperfections were detected during previous inservice inspections.

4.2.1.1.3 Examination Method and Requirements Steam generator tubes shall be examined in accordance with the method prescribed in Appendix IV, "Eddy Current Examination of Non-Ferromagnetic Steam Generator Heat Exchanger Tubes," as contained in ASME Boiler and Pressure Vessel Code -

Section XI "Inservice Inspection of Nuclear Power Plant Components."

4.2.1.1.4 Inspection Intervals (a) Inservice inspections shall not be more than 24 calendar months apart, except that reduced,or tightened inspection intervals shall be governed as specified in 4.2.1.4(c) and (d).

(b) The inservice inspections may be scheduled to be coincident with refueling outages or any plant shutdown, provided the inspection intervals of 4.2.1.1.4(a), (c) or (d), as applicable, are not exceeded.

(c) If two consecutive inservice inspections covering a time span of at least 12 months yield results that fall in C-1 category, the inspection frequency may be extended to 40 month intervals between inspections.

(d)

If the results of the inservice inspection of steam generator tubing conducted in accordance with Table 4.2-2 at 40 month intervals fall in category C-3, the inspection frequency shall be reduced to at least once per 20 months.

The increase in inspection frequency shall apply until a subsequent inspection meets the conditions specified in 4.2.1.1.4(c) and the interval can be extended to a 40 month period.

4.2-3 Amendment.No.83

(e) Unscheduled inspections shall be conducted in accordance with Specification 4.2.1.1.2 on any steam generator with primary-to-secondary tube leakage (not including leaks originating from tube-to-tube sheet welds) exceeding Specification 3.1.5.3.

All steam generators shall be inspected before returning to power in the event of a seismic occurrence greater than an operatina basis earthquake, a LOCA requiring actuation of engineered safeguards, or a main steam line or feedwater line break.

4.2.1.1.5 Acceptance Limits Defini tions:

Imperfection is an exception to the dimension, finish, or contour of a tube from that required by fabrication drawings or specifications.

Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

Degradation means a service induced cracking, wastage, wear or general corrosion occurrfng on either inside or outside of a tube.

Degraded Tube is a tube that contains imperfections caused by degradation equal to or greater than 20% of the nominal tube wall thickness.

Defect is an imperfection of such severity that it exceeds the plugging limit.

A tube containing a dAfect is defective.

Plugging Limit is the imperfection depth beyond which a degraded tube must be removed from service by plugging, because the tube may become defective prior to the next scheduled inspection of that tube.

The plugging limit is 47% of the nominal tube wall thickness if the next inspection interval of that tube is 12 months, and a 2% reduction in the plugging limit for each 12 month period until the next inspection of the inspected steam generator.

4.2.1.2 Corrective Measures All tubes that leak or are determined to have degradation exceeding the plugging limit shall be plugged prior to return to power.

4.2.1.3 Reports

1.

After each inservice examination, the number of tubes plugged in each steam generator shall be reported to the Commission in accordance with Specification 6.9.2.a(3).

4.2-4 Amendment No.

83

2.

The complete results of the steam generator tube inservice inspection shall be inclu4d in the Operating Report for the period in which the inspection was completed.

Reports shall include:

(a) Number and extent of tubes inspected (b) Location and percent of wall thickness penetration for each eddy current indication and any leaks.

(c) Identification of tubes plugged.

3.

All results in Category C-3 of Table 4.2.2 shall be reported to the Commission as a prompt notification of Specification 6.9.2.a prior to resumption of plant operation.

The written follow-up shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

Basis:

The inspection program is in compliance with Section XI of the ASME Rules for In-service Inspection of Nuclear Power Plant Components. It should be recognized that examinations in certain areas are desirable but impractical due to the state-of-the-art.

The areas indicated for inspection represent those of representative stress levels and therefore will serve to indicate potential problems before significant flaws develop there or in other areas.

As more experience is gained in operation of pressurized water reactors, the time schedule and location of inspection may be altered or, should new eq4ipment and/or techniques be developed, consideration may be given to incorporate these into this inspection program.

The use of conventional nondestructive, diredt visual and remote visual test techniques can be applied to the inspection of most primary loop components except the reactor vessel.

The reactor vessel presents special problems because of the radiation levels and the requirement for remote underwater accessibility to this component.

Because of,hese limitations on access to the reactor vessel, several steps have been incorporated into the design and manufacturing procedures in preparation for nondestructive test techniques which may be available in the future.

The techniques used for in-service inspection include visual inspections, ultrasonic, radiographic, magnetic particle and dye penetrant testing of selected parts during refueling periods.

The primary pressure boundary class I components covered by this inspection will include the primary reactor coolant system and branch lines greater than I" from the reactor coolant system to the second design isolation valve. Credit is taken in the design of this plant for check valves.

4.2-5 Amendment No. 83

Class 2 and Class 3 components were chosen based on Regulatory Guide 1.26 and A.NSI N18.2 and N18.2a "Nuclear Safety Criteria for the Design of stationary Pressurized Water Reactor Plants."

4.2.2 Materials Irradiation Surveillance Specimens The reactor vessel surveillance program includes eight specimen capsules to evaluate radiation damage based on pre~irradiation and post-irradiation testing of specimens.

The specimens are located about three inches from the vessel wall at the (il midplace and are spaced radially at 0*,

10*,

20*,

30*, and 40.

Capsule No.

I is scheduled to be removed at the first region replacement. The exposure of this capsule leads the vessel maximum exposure by a factor of 2.1.

Thus, this capsule provides information for approximately a four-year exposure to the vessel.

Capsule No. 2 is scheduled to be removed at the fourth region replacement. This capsule leads the vessel maximum exposure by a factor of 0.8 and thus will provide data for a four-year exposure to the vessel. This sample also contains weld metal which is not present in Capsule No. 1.

Capsule No. 3 leads the vessel maximum exposure by a factor of 2.2 and is scheduled to be removed after twenty years.

Thus, sample No.

3 will provide data for an exposure to the vessel of approximately forty years.

Capsule Nos.

4 and 5 lag the maximum vessel exposure by factors of 0.7 and 0.5, respectively. Thus, Capsule No. 4, which is scheduled to be removed after thirty years, provides d.ata for a vessel exposure of twenty-one years and Capsule No. 5, which is scheduled to be removed at forty years, provides data for a vessel exposure of twenty years.

In addition to the capsules discussed above, there are three spares.

Two are located at the same location as Capsule No. 5 and one is located at the same location as Capsule No. 4.

The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes for evidence of mechanical damage or progressive degradation.

Lnservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

4.2-6 Amendment No. 83

Wastage-type defects will be minimized with proper chemistry treatment of the secondary coolant.

If defects or significant degradations should develop in service, this condition is expected to be detected during inservice steam generator tube examinations.

Plugging will be required for all tubes with imperfections exceeding the plugging limit.

Steam generator tube inspections by means of eddy current testing have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-,3, these results will be promptly reported to the Commission pursuant to Specification 6.9.2.a prior to resumption of plant operation.

Such cases will be considered by the Commission on a case-by-case basis and may result in a -requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications.

4.2.3 Primary Pump Flywheels The flywheels shall be visually examined at the first refueling after each ten year inspection.

At the fourth refueling after each ten tear inspection and at each fourth refueling thereafter, the outside surfaces shall be examined by ultrasonic methods.

References (1)

FSAR, Section 4.4 (2)

FSAR, Volume 4, Tab VII, Question VI.C 4.2-7 Amendment No. 83

Table 4.2-1 DELETED 4.2-8 Amendment No. 83

TABIE 4.2-2 S'EAM GENERATOR TUBE INSPECTION II. B. ROBINSON UNIT NO.

2 IST SAMPLE EXAMINATION 2ND SAMPLE EXAMINATION 3RD SAMPLE EXAMINATION Sample Size Result Action Required Result Action Required Result Action Required A minimum of C-1 Acceptable for N/A N/A N/A N/A S

tubes per Continued Service Steam C-2 Plug tubes exceeding the 'C-I Acceptable for N/A N/A Generator plugging limit and pro-continued Service (S.G.)

ceed with 2nd sample examination of 2S tubes IC-2 Plug tubes exceeding C-1 Acceptable for S=3(N/n)%

in same steam generator the plugging limit Continued Service and proceed with 3rd Plug tubes exc. plug sample examination of C-2 limit.

Acceptable for j

where:

4S tubes in same continued service steam gpnerator Perform action required N is the C-3 under C-3 of 1st number of sample examination

> steam genera-Perform action required D tors in the C-3 tinder C-3 of ist sample N/A N/A plant =3 examination C-3 Inspect all All other Acceptable for N/A N/A n is the tubes in this S.G, plug S.G s are Continued Service z number of tubes exceeding the C-1 steam genera-plugging limit and Some S.G.s L tors inspect-proceed with 2nd sample C-2 but no Perform action required N/A N/A ed during an examination of 2S tubes additional under C-2 of 2nd sample examination in each other steam S.G are examination above generator not Lnclided in C-3 the InservLce inspection Additional program.

S.G.

is Inspect all N/A N/A Report results to NRC C-3 tubes In S.G. and plug within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> In tubes exceeding plug accordalnce with Techlical ging limit.

Report to Specif icatioi 6.9.2.a(3).

NRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with Technical Specification 6.9.2.a(3).

Containment Soray System 4.5.1.3 System tests shall be performed at each refueling interval. The test shall be performed with the isolation valves in the spray supply lines at the containment and spray additive tank blocked closed. Operation of the system is initiated by tripping the normal actuation instrumentation.

4.5.1.4 The spray nozzles shall be checked for proper functioning at least every five years.

4.5.1.5 The tests discussed in-4.5.1.3 and 4.5.1.4 will be considered satisfactory if visual observations indicate all components have operated satsifactorily.

Containment Fan Coolers 4.5.1.6 Each fan cooler unit shall be tested at intervals not to exceed one month to verify proper operation of all essential features including valves, dampers and piping.

4.5.2 Component Verification 4.5.2.1 When the reactor coolant pressure is in excess of 1,000 psi, it shall be verified at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (from the RTGB indicators/controls) that the following valves are in their proper position with control power to the valve operators removed.

Valve Number Valve Position 1-MOV 862 A&B Open 2-MOV 863 A&B Closed 3-MOV 864 A&B Open 4-MOV 866 A&B Closed 4.5-2 Amendment No. 83

4.5.2.2 At monthly intervals during power operations each valve (manual, power operated, or automatic) in the safety injection (low and high pressure) and containment spray system flow paths that is not locked, sealed or otherwise secured in position shall be verified as correctly positioned.

Basis The Safety Injection System and the Containment Spray System are principal plant safeguards that are normally inoperative during reactor operation.

Complete systems tests cannot be performed when the reactor is operating because a safety injection signal causes reactor trip, main feedwater isolation and containment isolation, and a Containment Spray System test requires the system to be temporarily disabled. The method of assuring operability of these systems is therefore to combine systems tests to be performed during annual plant shutdowns, with more frequent component tests, which can be performed during reactor operation.

The systems tests demonstrate proper automatic operation of the Safety Injection and Containment Spray Systems.

A test signal is applied to initiate automatic action and verification made that the coqgonents receive the safety injection in the proper sequence. The test demonstrates the operation of the valves, pump circuit breakers, and automatic circuitry.(1)(2)(4)

During reactor operation, the instrunentation which is depended on :o initiate safety injection and containment spray is generally checked each shift and the initiating circuits are tested monthly (in accordance with Specification 4.1).

The testing of the analog channel inputs is accomplished in the same manner as for the reactor protection system. The engineered safety features logic system is tested by means of test switches to simulate inputs from the analog channels. The test switches interrupt the logic matrix output to the master relay to prevent actuation. Verification that the logic is accomplished is indicated by the matrix test light.

Upon completion of the logic checks, verification that the circuit from the logic matrices to the master relay is complete is accomplished by use of an ohmmeter to check continuity.

In 4.5-3 Amendment No. 8 3

addition, the active components (pumps and signal valves) are to be tested quarterly to check the operation of the starting circuits and to verify that the pumps are in satisfactory running order. The quarterly test interval is based on the judgment that more frequent testing would not significantly increase the reliability (i.e., the probability that the component would operate when required), and that more frequent testing would result in increased wear over a long period of time.

Quarterly testing of valves is consistent with the requirements of ASME Section XI.

Quarterly testing of the safety injection pumps, residual heat removal pumps, containment spray pumps and the boron injection tank isolation valves is not required when in the cold shutdown condition. These components are not required for plant safety when the reactor is in cold shutdown and testing during this condition will result in unnecessary wear on the equipment.

Other systems that are also important to the emergency cooling function are the accumulators, the Component Cooling System, the Service Water System and the containment fan coolers.

The accumulators are a passive safeguard. In accordance with Specification 4.1, the water volume and pressure in the accumulators are checked periodically. The other systems mentioned operate when the reactor is in operation and by these means are continuously monitored for satisfactory performance.

References (1) FSAR Section 6.2 (2) FSAR Section 6.4 (3) FSAR Section 6.1 (4) CP&L report and supplemental letters of September 29, November 5, December 8, 1971, and March 20, 1972.

4.5-4 Amendment No. 83