ML20205R138

Amend 104 to License DPR-40,changing Surveillance Criteria for Sample Selection & Insp Results Re Steam Generator Tube Insps from Fixed Numerical Criteria to Comparable Percentages

ML20205R138
Person / Time
Site:	Fort Calhoun
Issue date:	03/26/1986
From:	Thadani A Office of Nuclear Reactor Regulation
To:
Shared Package
ML20205R107	List: ML20205R101 ML20205R138 ML20205R143
References
TAC-63437, NUDOCS 8704060230
Download: ML20205R138 (13)

Text

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UNITED STATES 8 NUCLEAR REGULATORY COMMISSION t a WASHINGTON, D. C. 20588

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OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 104 License No. DPR-40 1

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

A. The application for amendment by the Omaha Public Power District (the licensee) dated November 5, 1986, complies with the standards and requirement? 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; ,

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

'l E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requi ements t have been satisfied.

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2. Accordingly, Facility Operating License No. DPR-40 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.8. of Facility Operating License No.

DPR-40 is hereby amended to read as follows:

B. Technical Specifications '

The Technical Specifications contained in Appendix A, as revised through Amendment No.104 , 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

, h A )O' Asho C. Thadani, Director PWR )roject Directorate #8 4

Division of PWR Licensing-B

Attachment:

Changes to the Technical Specifications Date of Issuance: March 26, 1987

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ATTACHMENT TO LICENSE AMENDMENT NO. Ina FACILITY OPERATING LICENSE NO. OPR-40 DOCKET NO. 50-285 ReviseAppendix"A"TechnicalSpecificationsasindicatedbelow. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

Remove Pages Insert Pages 11 11 3-21 through 3-29 3-21 3-22 3-27 3-86 through 3-91

TABLE'0FCONTENTS(Continued)

.P, age, 2.12 Control Room Systems...................................... 2-59 2.13 Nuclear Detector Cooling System........................... 2-60 2.14 Engineered Safety Features System Initiation Instrumentation Settings................................ 2-61 2.15 Instrumentation and Control Systems....................... 2-65 2.16 River Level............................................... 2-71 2.17 Miscellaneous Radioactive Material Sources................ 2-72 2.18 Shock Suppressors (Snubbers).............................. 2-73 2.19 Fire Protection System.................................... 2-89 2.20 Steam Generator Coolant Radioactivity..................... 2-96 2.21 Post-Accident Monitoring Instrumentation. . . . . . . . . . . . . . . . . . 2-97 2.22 Toxic Gas Monitors........................................ 2-99  ;

3.0 SURVEILLANCE REQUIREMENTS....................................... 3-1 3.1 Instrumentation and Control............................... 3-1 3.2 Equipment and Sampling Tests. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-17 3.3 Reactor Coolant System and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance......... 3-21 3.4 Reactor Coolant System Integrity Testing. . . . . . . . . . . . . . . . . . 3-36

3. 5 Containment Test.......................................... 3-37 3.6 Safety Injection and Containment Cooling Systems Tests........................................... 3-54 3.7 Emergency Power System Periodic Tests..................... 3-58 3.8 Main Steam Isolation Valves............................... 3-61 3.9 Auxil iary Feedwa ter Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-62 3.10 Reactor Core Parameters. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-63 3.11 Radiological Environmental Monitoring Programs............ 3-64 1 3.12 Radiological Waste Sampling and Monitoring. . . . . . . . . . . . . . . . 3-69 l 3.12.1 Liquid and Gaseous Ef fluents. . . . . . . . . . . . . . . . . . . . . . 3-69 3.12.2 Solid Radioactive Waste........................... 3-71a 3.13 Radioactive Material Sources Surveillance................ 3-76 3.14 Shock Suppressors (Snubbers).............................t. . 3-77 3.15 Fire Protection System.................................... 3-80 3.16 Recirculation Heat Removal System Inte 3-84 3.17 S team Genera tor Tubes. . . . . . . . . . . .................... . . . . . . gri ty Tes ting. 3-86 ......

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4.0 DE S I GN F EATU RE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4. - 1 4.1 4.2 Site...................................................... 4-1 Containment Design Features............................... 4-1 1

4.2.1 Containment Structure.............................. 4-1 4.2.2 Penetrations....................................... 4-1 4.2.3 Containment Structure Cooling Systems.............. 4-2 11 Amendnent No. H .M .N . 104

3.0 SURVEILLANCE REQUIREMENTS 3.3 Reactor Coolant Systen and Other Components Subject to ASME XI l Boiler & Pressure Vessel Code Inspection and Testing Surveillance Applicability Applies to in-service surveillance of primary system components and other components subject to inspection and testing according to ASME XI Boiler & Pressure Vessel Code.

Objective e To ensure the integrity of the reactor coolant system and other l components subject to inspection and testing according to ASME XI Boiler & Pressure Vessel Code.

Specifications (1) Surveillance of the ASME Code Class 1, 2 and 3 systems, except the steam generator tubes inspection, should be covered by ASME XI Boiler & Pressure Vessel Code.

a. In-service inspection of ASME Code Class 1, Class 2, and Class 3 components and in-service testing of ASME Code Class 1, Class 2, and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, as required by 10 CFR Part 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part50,Section50.55a(g)(6)(1).

b. Surveillance of the reactor coolant pump flywheels shall be perfonned as indicated in Table 3-6.

c. A surveillance program to monitor radiation-induced changes in the mechanical and impact propert1 vessel materials shall be maintained.9g)of ti the specimen The reactor removal schedule shall be as indicated in Table 3-7. l (2) Surveillance of Reactor Coolant System Pressure Isolation Yhlves l

a. Periodic leakage testing on each valve listed in Table 2-9 ,

shall be accomplished prior to entering the power operation i mode every time the plant is placed in the cold shutdown '

• To satisfy ALARA requirements, leakage may be measured indirectly (as from the perfonnance of pressure indicators) If accomplished in accordance with a) proved procedures and supported by computa-tions showing that t1e method is capable of demonstrating valve compliance with the leakage criteria.

l 3-21 Amendment No. f6,7%,104

l 3.0 SURVEILLANCE REQUIREMENTS 3.3 Reactor Coolant System and Other Components Subject to ASME XI l 1

Boiler & Pressure Vessel Code Inspection and Testing Surveillance (Continued) condition for refueling, 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 return-ing the valve to service after maintenance, repair or replace-ment work is perfonned. ,

b. Whenever the integrity of a pressure isolation valve listed in Table 2-9 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 one other valve located in the high pressure line shall be recorded daily.

Basis Undetected prolonged leakage of borated reactor coolant onto carbon steel sets up an unusual corrosion mechanism. Detection of this leakage at an early stage can best be accommodated directly after an outage and before startup. The inspection program specified in Specification 3.3(1) places major emphasis on the areas of highest stress concentration as determined by general design evaluation and experience with similar systems. The inspections will rely on non-destructive analysis methods utilizing up-to-date analyzing equipment and trained personnel. Volumetric inspection of the reactor pressure vessel is to be performed completely from the outside diameter. The testing techniques and acceptance criteria of Section XI of the ASME B&PV Code will be utilized, except where specific relief is granted by the Comission.  ;

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i Reference (1) USAR, Section 4.5.3 1 l

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I 3- 22 Amendment No. pp.9f h t J/29/p7,

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I TABLE 3-6 '

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REACTOR COOLANT PUMP SURVEILLANCE i '

l REQUIREMENT METHOD FREQUENCY I l

1.1 Reactor Coolant Visual inspection of upper When motor is dis-Pump Flywheels surface of top disc and assembled for bottom surface of bottom maintenance disc; volum6tric inspection purposes.

from circumference of all disc segments.

i TABLE 3-7 4

CAPSULE REMOVAL SCHEDULE REMOVAL REFUELING SCHEDULE CAPSULE 7 SEQUENCE EFPY** REMOVED l-1 2.4 225' i j 2 5.9 -265* I

3 15 275*

l' 4 20 45* ,

5 21 85*  !

6 27 95* l 7 32 225** i l 8 Standby 265** l l

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• Replacement capsule assemblies were installed in the 225' and 265*

! locations following early removal of the 265* capsule. These cap- ,

sules benchmark the change in core loading design initiated at 5.9 1

! EFPY.

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i ** Based on a rated power level of 1500 MWt. I

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, Amendment No. 44, M, $4/,104 3-27

(Nextpageis3-36) l 4 .

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3 3.:0 SURVEILLANCE-REQUIREMENTS f

3.17 Steam Generator Tubes l Applicability l

Applies to in-service surveillance of steam generator-tubes.

Objective To ensure the integrity of the steam generator tubes. l

! Specifications i

Each steam generator shall be demonstrated OPERABLE by performance of the following in-service inspection program.

I j (1) Steam Generator Sample Selection and Inspection Methods l i

i The in-service inspection shall be perfomed on each steam i generator on a rotating schedule. Under some circumstances. l

) the operating conditions in one steam generator may be found i to be more severe than those in the second steam generator.

! Under such circumstances, the sample sequence shall be

{ modified to inspect the steam generator with the most severe conditions.

(2) Steam Generator Tube Sample Selection and Inspection l ,

j The steam generator tube minimum sample size, inspection result

classification, and the corresponding. action required shall be as specified in Table 3-13. The in-service inspection of steam j generator tubes shall be perfomed according to Specification 3.17(4)(1), " Tube Inspection," and at the frequencies specified

in Specification 3.17(3). The inspected tubes shall be verified i acceptable per the acceptance criteria of Specification 3.17(4).

j The tubes selected for each in-service inspection shall include at least 3% of the total tubes in the steam generators and the tubes selected for these inspections shall be selected on a

! random basis, except: g l

(1) If the tube is recorded as a degraded tube, then an i

1 adjacent tube shall'be inspected.

(ii) The first sample inspection during each in-service inspec--

tion of each steam generator shall include all non-plugged tubes that previously had detectable wall penetrations

(>20%) and shall also include tubes in those areas where experience has indicated potential problems.

(iii) The second and third sample inspections, if required, may be less than an entire tube length inspection provided-

the inspection concentrates on those areas of the tube i

,' 3-86 Amendnent No. 104 i .

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3.0 SURVEILLANCE REQUIREMENTS , 3.17 Steam Generator Tubes (Continued) sheet array and on those portions of the tubes where defects were previously detected.

(iv) To the extent practical, where 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.

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

Category Inspection Results C-1 -Less than 5% of the tubes inspected are degraded and none of the inspected tubes are defective.

C-2 Less than 1% of the tubes inspected are defective, or between 5% and 10%

of the tubes inspected are degraded.

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

NOTE: In all inspections, previously degraded tubes must exhibit significant (>10%) further wall penetrations to be included in the above calculations.

(3) Inspection Frequencies l

The above required in-service inspections of steam generator tubes l shall be performed at the following frequencies-(inspections shall be performed, unless otherwise specified, coincident with refueling outages or any scheduled cold shutdown for plant repair and maintenance): ,

(i) In-service inspections shall be perfonned at intervals of not less than 12 nor more than 24 calendar months -

after the previous inspection, with one exception. If a plant operating cycle is less than 12 months, inspec-tions may be perfonned at the end of that ~ cycle. If two consecutive inspections following service under all volatile treatment conditions result in all inspection results falling into the C-l. category or if two consec-utive inspections demonstrate that previously observed degradation has not continued and no additional degrada-tion has occurred, the inspection interval may be extended to a maximum of once per 40 months.

3-87 Amendnent No.104

3.0 SURVEILLANCE REQUIREMENTS .

! 3.17 Steam Generator Tubes (Continued)

(ii)

If results of the in-service inspection of the steam generator tubes conducted in accordance with Table 3-13 I at 40-month intervals fall in Category C-3, the inspec-tion frequency shall be increased to at least once.per 20 months. The increase in inspection frequency shall apply until a subsequent inspection meets the conditions specified in (i) above, at which time the interval can .

be extended to a 40-month ' period.

(iii) Unscheduled in-service inspections shall be perfomed on each steam generator-in accordance with the first sample inspection specified in Table 3-13 during the i shutdown subsequent to any of the following conditions:

1. Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess-of the limits of Section 2.1.4 of the Technical Specifications,

2. A seismic occurrence greater than the Operating Basis Earthquake,

3. A loss-of-coolant accident requiring actuation of the engineered safeguards, or

4. A main steam line or main feedwater line break.

(4) Acceptance Criteria

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(i) As used in this specification:

Imperfection means an exception to'the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indica- i tions 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 occurring on either inside or outside of a tube.

Degraded Tube means a tube containing imperfections >20%

of the nominal wall thickness caused by degradation. Any tube which does not permit the passage of the eddy-current inspection probe through its entire length and U-bend shall be deemed a degraded tube.

1 Degradation means'the percentage of the tube wall thick-ness affected or removed by degradation.

3-88 Amendment No. 104

4 3.0 SURVEILLANCE REQUIREMENTS 3.17 Steam Generator Tubes (Continued) '

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Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective.

! Plugging Limit means the imperfection depth at or beyond which the tube shall be removed from service because it may become unserviceable prior to the next inspection and is equal to 40% of the nominal tube wall thickness.  ;

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break.

TubeInspect[onmeansaninspectionofthesteamgenerator tube from the point of entry (hot leg side) completely <

around the U-bend to the. top support of the cold leg.

(ii) The steam generator shall be determined OPERABLE after complet.ing the corresponding actions (plug all tubes exceeding the plugging limit a'nd all tubes containing through-wall cracks) required by Table 3-13. l (5) Reporting Requirements l ,

i (1) Following each in-service inspection of steam generator i

tubes, the number of tubes plugged in each steam generator shall be reported to the Comission within 30 days.

(ii) The complete results of the steam generator tube in-i

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service inspection shall be reported to the Comission within 6 months following completion of the inspection.

This report shall include:

1. Number and extent of tubes-inspected.

2. Location and , ? cent of wall thickness penetration for each imperfection.

, 3. Identification of tubes plugged. I (iii) Results of steam generator tube inspections which fall into Category C-3 and require prompt notification of the.

Commission shall be reported pursuant to Section 5.6 of the Technical Specifications prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted -

to determine cause of the tube degradation and correc-tive measures taken to prevent recurrence.

l 3-89 Amendment No. 104 t

TABLE 3-13 l STEAM GENERATOR TUBE INSPECTION IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Samp1.e Size Result Action Required Result Action Required Result Action Required Am m C-1 None N/A N/A N/A N/A f bes per S.G.

C-2 Plug defective tubes C-1 None N/A N/A and inspect additional Plug defective tubes 600 tubes in this S.G. C-2 C-1 None and inspect additional 1200 tubes in this C-2 Plug defective tubes S*G* C-3 Perfonn action for C-3 result of first sample w

C-3 Perfonn action for C-3 result of first sample E

C-3 Inspect all tubes in The this S.G., plug defec- jnd tive tubes and inspect 600 tubes in other is C-1 '

k The Perform action for Prompt notification second C-2 result of second N/A N/A k' to NRC pursuant to S.G. sample a specification 5.6 is C-2 '-

2 P The Inspect all tubes in w

second the second S.G. and

, S.G. plug defective tubes. N/A N/A

,q is C-3 Prompt notification

. to NRC pursuant to 2; -

specification 5.6 N/A = Not Applicable.

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3.0 SURVEILLANCE REQUIREMENTS , 3.17 Steam Generator Tubes (Continued) i Basis The surveillance requirements for inspection of the steam generator

, tubes ensure that the structural integrity of this portion of the RCS i will be maintained. The program for in-service inspection of the steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1 dated July 1975. In-service inspection of steam generator tubing is essential in order to main'tain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage

, or progressive degradation due to design, manufacturing errors, or.

in-service conditions that lead to corrosion.

In-service inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so t. hat corrective measures can be taken.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled in-service steam generator tube examinations. Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants 4

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.in-service inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Section 5.6 of the Technical Specifications prior to the resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement 4 for analysis, laboratory examinations, tests, additional eddy-current

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inspection, and revision of the Technical Specifications, if necessary. .

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.r 3-91 Amendment No.104

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