ML20137Y885
| ML20137Y885 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 09/30/1985 |
| From: | Miner S Office of Nuclear Reactor Regulation |
| To: | Reinaldo Rodriguez SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| Shared Package | |
| ML20133D117 | List: |
| References | |
| TAC-51007, NUDOCS 8510080112 | |
| Download: ML20137Y885 (2) | |
Text
m September 30, 1985 Docket No. 50-312 DISTRIBUTION, BGrimes Docket File WJones NRC PDR MVirgilio L P9R RDiggs Mr. Ronald J. Rodriguez ORBg4 Rdg RIngram Executive Director, Nuclear HThompson SMiner Sacramento Municipal Utility District 0 ELD Gray File +4 6201 S Street CMiles EBlackwood P. O. Box 15830 LHarmon H0rnstein Sacramento, California 95813 ACRS-10 SECY TRarnhart-4 EJordan
Dear Mr. Rodriguez:
SUBJECT:
AMENDMENT N0. 76 TO FACILITY OPERATING LICENSE NO. DPR-54, INSERVICE INSPECTION AND TESTING TECHNICAL SPECIFICATIONS The Commission has issued the enclosed Amendment No. 76 to Facility Operating License No. DPR-54 for the Rancho Seco Nuclear Generating Station. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated March 16, 1979, as supplemented by your letters of December 12, 1979, February 19, 1985, and April 24, 1985.
Your proposed request for a clarification of control rod testing requirements contained in the February 19, 1985, submittal will be addressed in a separate licensing action.
The amendment revises the Technical Specifications to provide conformance with the Consnission's regulations governing Inservice Inspection as set forth in 10 CFR 50.55a(g).
It also revises the Technical Specifications governing inspection of steam generator tubes.
A copy of the Safety Evaluation is also enclosed.
Notice of Issuance will be included in the Commission's Biweekly Notice.
Sincerely, ow
- W tri Sydney Miner, Project Manager Operating Reactors Branch #4 Division of Licensing
Enclosures:
1.
Amendment No.'76 to DPR-54 2.
Safety Evaluation cc w/ enclosures:
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Mr. R. J. Rodriguez Rancho Seco Nuclear Generating Sacramento Municipal Utility District Station cc:
Mr. David S. Kaplan, Secretary Sacramento County and General Counsel Board of Supervisors Sacramento Municipal Utility 827 7th Street, Room 424 District Sacramento, California 95814 6201 S Street i
P. O. Box 15830 Ms. Helen Hubbard Sacramento, California 95813 P. O. Box 63 Sunol, California 94586 Thomas Baxter, Esq.
Shaw, Pittman, Potts & Trowbridge 1800 M Street, N.W.
Washington, D.C.
20036 I
Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 220, 7910 Woodmont Avenue Bethesda, Maryland 20814 Resident Inspector / Rancho Seco c/o U. S. N. R. C.
14410 Twin Cities Road Herald, California 95638 Regional Administrator, Region V U.S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek, California 94596 Director Energy Facilities Siting Division Energy Resources Conservation &
Development Commission 1516 - 9th Street Sacramento, California 95814 Mr. Joseph 0. Ward, Chief Radiological Healtn Branch State Department of Health Services 714 P Street, Office Building #8 Sacramento, California 95814 i
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SACRAMENTO MUNICIPAL UTILITY DISTRICT DOCKET NO. 50-312 RANCHO SEC0 NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 76 License No. DPR-54 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by Sacramento Municipal Utility District (the licensee) dated March 16, 1979, as supplemented by letters of December 12, 1979, Februa ry 19, 1985, and April 24, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the
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the provisions of the Act, and the rules and regulations of the Comission; 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 comon 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 js amer.ded by changes to the Technical Specif' cations as indicateW'7n the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-54 is hereby amended to read as follows:
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Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
76, 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 its date of issuance.
FOR THE NUCLEAR REGU ATORY COMMISSION A
Klu /
Jo I'F. Stolz, Chief 0 rating Reactors Branch #4 ivision of Licensing l
Attachment:
j Changes to the Technical j
Specifications l
Date of Issuance: September 30, 1985 l
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ATTACHMENT TO LICENSE AMENDMENT NO. 75 FACILITY OPERATING LICENSE NO. DPR-54 DOCKET NO. 50-312 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert y
v IX iX 3-19 3-19 i
3-21 3-21 4-8 4-8 4-8a 4-10 4-10 4-11 4-11 4-12 4-12 4-12a 4-12a 4-12b 4-13 4-13 4-14 4-14 4-19 4-19 4-26 4-26 4-27 4-27 4-28 4-28 4-29 4-29 4-30 4-30 4-31 4-31 4-32 4-32*
4-39 4-39 4-39a 4-39a 4-45 4-45 4-46 4-46*
4-51 4-51 4-52 4-52 4-53 4-S3 4-54 4-54 4-55 4-55 4-55a 4-55a 4-57 4-57 4
4-57a 4-57a 4-57b 4-57b 4-57c 6-12f 6-12f
- 0verleaf page provided for document completeness.
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS (Continued)
Section Page 4
SURVEILLANCE STANDARDS 4-1 4.1 OPERATIONAL SAFETY REVIEW 4-1 4.2 SURVEILLANCE OF ASME CODE CLASS 1, 2, AND 3 SYSTEMS 4-10 4.2.1 Reactor Yessel Surveillance Specimens 4-10 4.2.2 Inservice Inspection 4-11 4.3 TESTING FOLLOWING OPENING OF SYSTEM 4-14 4.4 REACTOR BUILDING 4-15 4.4.1 Containt.wnt Leakage Tests 4-15 4.4.2 Structural Integrity 4-21 4.4.3 Hydrogen Purge System 4-25 4.5 EMERGENCY CORE COOLING AND REACTOR BUILDING 4-26 000LIlvG 5Y51EM PEitIUDIC TESTlhG 4.5.1 Emergency Core Cooling System 4-26 4.5.2 Reactor Building Cooling Systems 4-29 i
4.5.3 Decay Heat Aemoval System and Reactor Blailding Spray 4-32 System Leakage 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTING 4-34 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4-36 4.7.1 Control Rod Drive System Functional Tests 4-36 4.7.2 Control Rod Program Verification (Group vs. Core Positionst 4-37 4.8 AUXILIARY FEEDWATER PUMP PERIODIC TESTING 4-39 4
4.9 REACTIVITY ANOMALIES 4-40 4.10
- EMERGENCY CONTROL ROOM FILTERING SYSTEM 4-41 4.11 REACTOR BUILDING' PURGE EXHAUST FILTERING SYSTEM 4-42 4.12 AUXILIARY AND SPENT FUEL B'UILDING FILTER SYSTDtS 4-43 4.13 AUGMENTED INSERVICE INSPECTION PROGRAM FOR HIGH 4-44 ENERGY LINE5 UUI51DE OF Curdihllvnr.h4 Amendments Nos. /S, 5), J7,76 y
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS LIST OT TA3LIS Table Page 2.3-1 Reactor Protection System Trip Setting Limits 2-9 3.5.1-1 Instrments Operating Conditions 3-27
- 3. 6-1 Safety Features Containment Isolation Valves 3-40 3.14.1 Fire Detection Instruments for Safety Systems 3-55 3.14.2 Inside Building Fire Hose Stations 3-57a 3.15-1 Radioactive Liquid Effluent Monitoring Instrumentation 3-61 3.16-1 Radioactive Gases Effluent Monitoring Instrumentation 3-64 3.22-1 Radiological Environmental Monitoring Program 3-83 3.22-2 Reporting Levels for Radioactivity Concentrations
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in Environmental Samples 3-86 4.1-1 Instrument Surveillance Requirements 4-3 4.1-2 Minimum Equipment Test Frequency 4-8 4.1-3 Minimum Sampling Frequency 4-9 4.2-1 Capsule Assembly Withdrawal Schedule at Davis-Besse 1 4-12b 4.10-1 Environmental Radiation Monitoring Program 4-42 4.10-2 Operational Environmental Radiation Ponitoring Program 4-22a 4.14-1 Snubbers Accessible During Power Operations 4-47c 4.17-1 Minimum Number of Steam Generators to be Inspected During Inservice Inspection 4-56 4.17-2A Steam Generator Tube Inspection 4-57
~4.17-2B Steam Generator Tube Inspection (Specific Limited Area) 4-57a 4.17-3 OTSG Auxiliary Feedwater Header Surveillance 4-57b, 4-57c 4.19-1 Radioactive Liquid Effluent Monitoring Instrurentation Surveillance Requirements 4-64 4.20.-l Radioactive Caseous Effluent 7:onitoring Instrumentation Surveillan'ce Requirements 4-66 iX Amendment No. Js, %,- $$, $$, 76
RANCHO SECO UNIT 1 TECHN1 CAL SPEC 8FICATIONS Limiting Conditions for Operation e
3.3 EMERGENCY CORE COOLING, REACTOR BUILDING ENERGENCY COOLING AND REACTOR BUILDING SPRAY SYSTEMS Applicability Applies to the emergency core cooling, Reactor Building emergency cooling and Reactor Builcing spray systems.
Objective To define the conditions necessary to assure immediate availability of the emergency core cooling, Reactor Building emergency cooling and Reactor Building spray systems.
Specification 3.3.1 The reactor shall not remain cr'itical, unless the following conditions are met:
A.
Injection System 1.
The borated water storage tank shall contain a minimum of 390,000 gallons of water having a minimum concentration of 1,800 ppa boron at a temperature not less than 40*F.
The manual valves on the discharge line from the borated water starage tank shall be locked open.
2.
Two out of three high pressure injection pumps shall be operable.
l 3.
Two safety features actuated decay heat reuoval pumps shall be operable.
4.
Both decay beat removal ceslers shall be operable.
5.
Two BWST level instrument channels shall be operable.
6.
The Reactor Building emergency sump isolation valve shall De l
either manually or remote-cianually operable.
7.
One of the two BWST isolation valves shall be open (SFV 25003 or SFV 25004). This valve may be closed during the quarterly valve test specified in the Specifications 4.5.1.2A and 4.5.2.2A.
B.
Core Flooding System 1.
The two core flooding tanks shall each contain 1040
- 30 ft3 of borated water at 600
- 25 psIg.
2.
Core flooding tank boron concentration shall not be less than 1,800 ppm boron.
3.
The electrically operated discharge valves from the core flood tanks shall ne open. The breakers shall be open anc so tagged.
t Amendment No. f, 76 3 19
RANCHO SECO UNIT 1 TECHNICAL SPECIF2 CATIONS Lim Ming Conditions for Operation 3.3.3 Prior to initiating maintenance on any of the component (s), the duplicate (redundant) components shall be verified operable by checking that the surveillance test for the component (s) has been i
successfully completed and will remain in effect for the. duration of the maintenance period. Inservice testing per specification 4.2.2.1 shall not be performed on ar1y component (s) whose duplicate (redundant) component (s) has (have) been declared inoperable or is out of service for any reason.
3.3.4 During power operation, hot standby, hot shutdown or startup conditions, the primary coolant system pressure isolation valves shall be functional as follows:
1.
All pressure isolation valves listed in Table 3.3-1 shall De functional as a pressure isolation device, except as specified in 3.3.4.2.
Valve leakage shall not exceed the amounts indicated.
2.
In the event that integrity of ar1y pressure isolation valve specified in Table 3.31 cannot be demonstrated, reactor operation may c;,ntinue, 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.(a) 3.
If Specifications 3.3.4.1 and 3.3.4.2 cannot be met, a shutdown shall be initiated, the reactor shall not remain critical and shall be brought to a cold shutdown condition within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Bases The requirements of Spcification 3.3.1 assure that, before the reactor can be made critical, adequate :afety features are operable.
Two high pressure injection pumps and two deccy heat removal pumps are specified.
However, only one of each is necessary to supply emergency coolant to the reactor in the event of a loss-of-coolant accident. Both core flooding tanks are required as a single core flood tank has insufficient inventory to reflood the core.(1)
The borated water storage tank is used for two purposes:
A.
As a supply of borated water for accident conditions.
B.
As a supply of borated wate{2or flooaing the fuel transfer cans 1 during refueling operation.
390,000 gallons of borated water are supplied for emergency core cooling and Reactor Butiding spray in the event of a loss-of-core coolant accident. This sinount fulfills requirements for voergency core cooling. The boraped water storage tank minimum volume of 390,000 gallons is based on refueling volume requirements. Heaters maintain the borated water supply at a temperature to prevent freezing. The boron concentration is set at the amount of boron required to maintain the core 1 percent sIJbcritical at 70*F without any control rods in the core.
This concentration is 1585 ppm boron while tne minimum value specified in the tanks is 1,800 ppm boron.
(a)
Motor operatea valves shall be placed in the closed positten and power supplies deenergized.
Amendment No. A, 9f#ff WJ#/
3-21 f//$/M,67,76
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I RANCHO SECO UN3T 1 i
TECHN! CAL SPECIF! CATIONS Surveillance Standards TABLE 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY Item Test frequency
- 1. control roos Roa crop times of all Eacn refueling snutacan full length rods
- 2. Control rod Movement of each rod Every two weeks movement
- 3. Pressurizer code Setpoint Note 3 safety valves
- 4. Main steam safety Setpoint Note 3 valves
- 5. Refueling system Functional Each refueling interval interlocks prior to handling fuel.
- 6. Turbine steam stop Movement of each valve Monthly valves
- 7. Reactor coolant Leakage Calculated inventory system weekly Leakage check daily.
- 8. Charcoal and high Charcoal and HEPA Each refueling in+.erni and efficiency filters filter for iodine at arty time work on filters and particulate could alter their integrity removal efficien-cies. DOP test on HEPA filters.
Freon test on charcoal filter units.
- 9. Fire pumps and Functional Monthly power supplies
- 10. Reactor Building Functional Each refueling isolation trip interval
- 11. Spent fuel Functional Each refueling cooling system interval prior to fuel handling
- 12. Turbine Overspeed Calibration Each refueling Trips interval 5
- 13. Internals Vent Manual actuation,1 Each refueling Valves remote visual inipection,2 interval verify valve not :; tuck open 1.
Verifying through manual actuation that the valve is fully open with a force of,400 lbs. (applied vertically upward).
5 2.
Check visually accessible surfaces to evaluate observed surface ir. qular* cies.
4.g Amendment No. 7, M, 76
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-2
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HINIMUM EQUIPMENT TEST FREQUENCY 3.
Tested in accordance with Section XI of the ASME Soiler and Pressure Yessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(1).
Amendment No. 76 4-8a l
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.2 SURVEILLANCE OF ASME CODE CLASS 1, 2 AND 3 SYSTEMS Applicability Applies to systems and components defined as ASHI Code Class 1, Class 2, and Class 3.
Objective To establish examinations whereby the integrity of ASME Code Class 1, 2 and 3 systems and components is monitored.
4.2.1 Reactor Vessel Surveillance Specimens 4.2.1.1 The reactor vessel material irradiation surveillance specimens removed from the reactor vessel at approximately 170 effective full power days shall be installed irradiated in and withdrawn from the Davis-Besse Unit No. I reactor vessel in accordance with the achedule shown in Table 4.2-1.
Following withdravi.1 of each capsule listed in Table 4.2-1, SMUD shall be responsible for testing the specimens and submitting a report of test results in accordance with 10 CFR 50, Appendix H.
4.2.1.2 A report or application for license amendment shall be submitted to the NRC within 90 days after the occurrence of any of the following:
l 1.
Failure of Davis-Besse Unit No. I to achieve commercial operation at 100% power by January 1,1978, or 2.
Beginning one year after attainment of commercial operation 100% power, any time that Davis-Besse Unit No. I fails at to maintain a cumulative reactor utilization factor of greater than 65%.
The report shall provide justification for continued operation of Rancho Seco with the reactor vessel surveillance program conducted i
t at Davis-Besse Unit No. 1 or the application for license amendment shall propose an alternative program for conduct of the Rancho Seco reactor vessel surveillance program.
I Amendment No. JE, 76 4-10 l
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Stancards 4.2.2 Inservice Inspection Specification 4.2.2.1 Inservice inspection of.ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed as closely as design permits in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50 Section 50.55alg), except where specific written relief has been granted b 50.55a(g)y the Commission pursuant to 10 CFR 50, Section (6)(1).
Performance of these inservice inspection and testing activities shall be in addition to other specified Surveillance Requi rements.
Nothing in the ASME Boiler and Pressure Yessel Code shall be construed to supersede the requirements of any Technical Specification.
In conducting inservice testing of pumps and valves, if the duplicate (redundant) component has been declared inoperable or is out of service for any reason, the component shall not be tested during power operation until the recundant component has been restored to operability or the operational actions required by the inoperable condition have been completed.
4.2.2.2 In addition to the requirements of Specification 4.2.2.1, each reactor coolant pump motor flywheel will be inspected volumetrically during the ten-year inspection interval. One hundred percent of the flywheel will be examined. All flywheels received a one hundred percent ultrasonic examination prior to installation on the motor.
4.2.2.3 If as a result of any of these inspections, defects are found, further examinations will be made as needed to assist evaluation. Evaluation of indications and repairs of defects shall be made in accordance with rules in Section XI of the ASME Soiler and Pressure Yessel Code.
4.2.2.4 Records of each inspection shall be kept to permit evaluation I
ar.d future comparison.
4.2.2.5 Periodic consideration will be given to incorpora} ion of new or I
improved inspection techniques into the surveillance program.
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Amendment No. 75, 76 4~11 e
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Bases Irradiation surveillance provides the capability of determining the radiation-induced changes in the mechanical and impact properties in the region of the reactor vessel surrounding the core. Test specimens of base metal, deposited weld metal and the heat-affected zone are installed in capsule assemblies placed inside the vessel.
In accordance with the schedules of Table 4.2-1 specimens will be removed; and a series of drop weight tests, Charpy impact tests and tension tests will be cunducted. Threshold neutron flux detectors and maximum temperature detectors will be installed with the specimens. Cnanges in nil-ductility transition temperature will tr determined, and appropriate alteration to plant operating parameters will be made.
To assure the availability of adequate surveillance data for the Rancho Seco No. I reactor vessel, a program has been developed to monitor the irradiation of the surveillance specimen capsules at the Davis Besse No. I reactor, and compare this to the irradiation of the Rancho Seco No. I reactor vessel.
Fluence estimates which are conservative in the appropriate direction are used for this comparison. The frequency of monitoring varies depending on the known neut ron fluence lead factor between the capsules and the reactor vessel. This provides ample time for ant icipat ing problems and initiating corrective action should operation of the host reactor be seriously delayed.
For the purpose of
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Technical Specificatien 4.2.8, the definition of Regulatory Guide 3.16 Revision 4 (August 1975) applies for the term "ce=nercial operation".
Cumulative reactor utilization factor is defined as:
[(Cumulative thernal zega att heurs since attainment of commercial operation at 100% pover) x 100) 4 1(licensed thertal pover) x (cumulat ive hours since at tain=ent of cc :ercial operation at 200*.' power)).
A preoperational examination was made which included all the items in ASME Code Class I systems that would normally be completed throughout the Inspection interval.
This survey established initial system integrity and provided a baseline for future testing.
Specification 4.2.2.1 ensures that inservice inspection of ASME Code Class 1 l
2 ano 3 components and inservice testing of ASME Code Class 1. 2 and 3 pumps and valves will be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Yessel Code and Aedenda as required by 10 CFR 50.55a.
Relief from any of the above requirements has been provided in writing by the Commission and is hot a part of these Technical Specifications.
t Amendment No. JJ, 76 4-12
RANCHO SECO UNfT 1 TECHNICAL SPECIFICATIONS Surveillance Stanuards Bases (continued)
Under the terus of specification 4.2.2.1, the more restrictive requirenents of the Technical Specifications take precedence over the ASME Boiler and Pressure Vessel Code and applicable Addenca. For example, the ASKE Boiler ana Pressure Vessel Code allows pumps to be tested up to one week af ter return to normal operation and allows a valve to be incapable of performing its specified function for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before being declared inoperable. Where these provisions are in conflict with the Technical Specifications, the Technical Specifications provisions will govern.
Testing a component subject to the provisions of Section XI is to be oeferrec as stated if the redundant component is inoperable. This is because test requirements frequently make the test component temporarily inoperable.
In the event an accident should occur at this time, neither of the safety-related trains would be available to respond to the conoition. The oeferral of the test does not relieve the licensee of the responsibility for conducting the required testing in accordance with regulatory requirements governing the frequency of tests. The licensee shoulo establish a test schedule that can accommodate a reasonacle number of inoperability events without violating required test frequencies.
Amendment No. 76 4-12a i
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Table 4.2-1 Rancho Seco CAPSULE ASSEMBLY WITHDRAWAL SCHEDULE AT DAVIS-BESSE 1 CAPSULE INSERTION / WITHDRAWAL RSI-B Withdraw at end of first cycle.
RSI-E Insert at end of first cycle, withdraw at end of tenth cycle.
RSI-D Withdraw at end of second cycle.
RSI-A Insert at end of second cycle, withdraw at end of seventh cycle.
RSI-C Withdraw at end of twelfth cycle.
t RSI-F Withdraw at end of ninth cycle.
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- I Amendment No.j/,
)5',
76 4-12b'
RANCHO SECO UNIT 1 TECHNICAL $PECIFICATIONS Surveillance Standards i
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TABLE 4.2-2 DELETED l
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i 4-13 i
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Stancaros 4.3 TESTING FOLLOWING OPENING OF SYSTEM Applicability
, Applies to test requirenents for reactor coolant systeu integrity.
Objective To assure reacter coolant system integrity prior to return to criticality following normal opening, modification, or repair.
Specification 4.3.1 When reactor coolant system repairs or modifications have been made, these repairs or modifications shall be inspected and tested to meet all appitcable code requirements priur to tne reactor being uade critical.
4.3.2 Following any opening of the reactor coolant system, it shall be leak tested at not less than 2,255 psig prior to the reactor being made critical.
4.3.3 The limitations of Specification 3.1.2 shall apply.
Bases Repairs or modifications made to the reactor coolant system are inspectable and testable under Secti,on XI of the ASME Boiler and Pressure Vessel Code.
For normal opening, the integrity of the reactor coolant system, in terms of strength, is unchanged.
If the system does not leak at 2,255 psig (operating pressure +100 psi; e50 psi is n9rn}al pressure fluctuation), it will be leak tight during r.ormal operation.11 REFERENCES (1)
FSAR, section 4 Amendment No.
76 4,14
RANCHO SECO UN8T 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.4.1.3 Isolation Yalve Functional Tests Remotely operated Reactor Bu'11 ding isolation valves shall be stroked to the position required to fulfill their safety function in
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accordance with requirements of Section XI of the ASME Boiler ana Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55(a)(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55afg)(6)(i).
4.4.1.4 Annual Inspection A visual examination of the accessible interior and exterior surfaces of the containment structure and its components shall be perfomed annually anc prior to any integrateu leak test, to uncover any evidence of deterioration which may affect either the containment's structural integrity or leak-tightness. The discovery of any significant deterioration shall be accompanied by corrective actions in accord with acceptable procedures, nondestructive tests, and inspections, and local testing where practical, prior to the conduct of any integrated leak test.
Such repairs shall be reported as part of the test results.
4.4.1.5 Reactor Building Modifications Any major modification or replacement of components affecting the Reactor Building integrity shall be followed by either an integrated leak rate test or a local leak test, as appropriate, and shall meet the acceptance criteria of 4.4.1.1.5 and 4.4.1.2.3 respectively.
Bases The Reactor Building is designed for an internal pressure of Sg psig and a steam-air mixture temperature of 286 F.
Prior to initial operation, the containment will be strength tested at 115 percent of aesign pressure. The containment will also be leak tested prior to initial operation at Po and Pt (52 psig and 26 psig, respectively). These tests will verify that the leakage rate from Reactor Bu11d given in the specification. L1)1no pressurization satisfies the relationships L21 The performance of a periodic integrated leakage rate test during plant life provides a current assessment of potential leakage from the containment in case of an accident that would pressurize the interior of the containment.
In order to provide a realistic appraisal of the integrity of the containment under accident conditions, this periodic test is to be performea without preliminary leak detection surveys or leak repairs, and containment isolation valves are to be closed in the normal manner. The reduced test pressure of 26 psig fer the periodic integrated leakage rate test is sufficiently high to provide an accurate measurement of the leakage rate aaa it duplicates the pre-operational leakage rate test at 26 psig.
The specification provides a relationship for relating the measured ' leakage of air at 26 psig to the potential leakage at 52 psig. The minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was specified for the integrated leakage rate test to help stabilize conaitions and thus improve 4-19 Amendment No. 76 '
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.5 EHERGENCY CORE COOLING AND REACTOR BUILDING COOLING S STEM PERIODIC TESTING 4.5.1 EMERGEACY CORE C00LIhG SYSTEM
_ Applicability Applies to periodic testing requirement for emergency core cooling systems.
Objective To verify that the emergency core cooling systems are operable.
Specification 4.5.1.1 System Tests A.
High Pressure Injection 1.
During each refueling interval, a makeup and purification system test shall be conducted to uemonstrate the system is operable for high pressure injection. A manual initiate l
signal will be applied to demonstrate actuation of the makeup and purification for emergency core cooling operation.
2.
The test will be considered satisfactory if control board indication verifies that all components have responded to the actuation signal; all appropriate pump breakers shall have opened or closed and ali power actuated valves shall have completed their travel.
3.
The high pressure injection pump casings shall be vented monthly and prior to any ECCS flow tests.
8.
Low Pressure Injection 1.
During each refueling interval a decay heat renoval system test shail be conducted to demonstrate the system is operable for low pressure injection. A manual initiate signal will be applied to demonstrate actuation of the decay heat removal system for emergency core cooling operation.
l Amendment No. 9, 76 4-26
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 2.
The test will be considered satisfactory if control board indication verifies that all components have responded to the actuation signal and all appropriate pump breakers shall have opened or closed, and all power actuated valves have l
completed their travel.
3.
Decay heat pump casing shall be vented monthly and prior to any ECCS flow tests.
4.
Periodic leakage testing (a) on each valve listed in Table 3.3-1 shall be accomplished prior to plant operation at power after every time the plant is placeo 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 accomplisheo in the preceding 9 months, and prior to returning the valve to service after maintenance, repair or replacement work is perforneo.
5.
Whenever integrity of a pressure isolation valve listed in Table 3.3-1 cannot be demonstrated, the integrity of the remaining valve in each high pressure line having a leaking valve shall be detemined and recordeo daily. In addition, the position of the other closed valve located in the high pitssure piping shall be recorded daily.
C.
Cort Flooding System 1.
During each refueling interval, a core flooding system test shall be conducted to demonstrate proper operation of the system. During depressurization of the reactor coolant system, verification shall be made that the check valves in the core flooding tank discharge lines operate.
2.
The test will be considered satisfactory if control board indication of core flood tank level verifies that all check valves have opened.
(a)To satisfy ALARA requirements, leakage may be measured indirectly (as j
from the perfomance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the metnod
' is capable of demonstrating valve compliance with the leakage criteria.
Amendment No. f, Of##/ #%#. f//$/ U.
4-27 76
e RANCHO SECO UNIT 1 TECHNICAL. SPECIFICATIONS Surveillance Standards D.
Nuclear Service Cooling and Raw Water Systems 1.
During each refueling interval, the safety features functiori of the nuclear service cooling water and raw water systems shall De tested. These tests may be in conjunction with other ECCS refueling interval tests which require automatic actuation of these systeus.
,i 2.
The test will be considered satisfactory if control board indication verifies all components have responcea to the actuation signal and all appropriate pump breakers shall have opened or closed, ano all power actuated valves have completed their travel.
4.5.1.2 Components Tests A.
Testing At least quarterly, Inservice testing of ECCS ano Nuclear Service Cooling and Raw Water Pumps and valves shall be performed in accordance with Section XI of the ASME Boiler anc Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50 Section 50.55a(g)(6)(1).
B.
Flow Path Verification Following Inservice testing of pumps and valves as required by paragraph 4.5.1.2A, required flow paths shall be demonstratec operable by verifying that each valve (manual, power-actuated or automatic) in the flow path that is not locked in position is in its normal operating position. Positions of locked valves shall be verified in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.
Bases Ine emergency core cooling systems are the principal reactor safeguards 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.
The decay heat removal pumps are tested singularly for operability by opening the borated water storage tank outlet valves and the test line valves to the borated water storage tank.
This allows water to be pumped from the borated water storage tank through each of the injection lines and back to the tank through a test line.
With the reactor shut down, the check 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 the check valves have opened.
REFERENCES FSAR subsection 6.2 Amendment No. 76 4-28
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Stancards 4.5.2 REACTOR BUILDING COOLING SYSTEMS Applicability Applies to testing of the Reactor Building cooling systems.
Objective To verify that the Reactor Building cooling systems are operable.
Specification 4.5.2.1 System Tests A.
Reactor Building Spray System 1.
During each refueling interval a system test shall be conducted to demonstrate proper operation of the system. A manual trip signal will be applied to demonstrate actuation of the Reactor Building spray system texcept for Reactor Building motor-operated inlet valves which prevent water entering nozzles). Water will be circulatea from the borated water storage tank through the Reactor Building spray pumps and returneo through the test line to the boratea water storage tank.
2.
The test will be considered satisfactory if visual observation and control board indication verifies that all components have responded to the actuation signal and the appropriate pump breakers shall have opened or closed, and all power actuated valves shall have completed their travel except the blocked Reactor Building inlet valve.
3.
Air will be introduced into the spray headers to verify the availability of the headers and spray nozzle at least every 10 years.
B.
Reactor Building Emergency Cooling System 1.
During each refueling interval, system tests shall be conducted to demonstrate proper operation of the system, including the upper dome air circulators. A manual initiate signal will be applied to actuate the Reactor Building emergency cooling system for Reactor Building cooling operation.
i Amendment No. 76 4-29
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 2.
The test will be considered satisfactory if ' control board indication verifies that all components have responded to the actuation signal and the appropriate breakers shall have completed their travel.
3.
Additionally it shall be verified that the NSCW flow through each operating cooler exceeds 1400 gpm and air flow througn the cooler exceeas 40,000 cfm.
4.5.2.2 Component Tests A.
Testing At least quarterly, Inservice testing of keactor Building Spray pumps and valves shall be perfomed in accordance with Section XI of the ASME Boiler and Pressure Yessel Code and applicable
- Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(1).
B.
Flow Path Verification Following Inservice testing of pumps and valves as required by paragraph 4.5.2.2A, required flow paths shall be demonstrated operable b automatic)y verifying that ' tach valve (manual, power-actuated or in the flow path that is not locked in position is in its nomal operating position. Positions of locked valves shall be verified in accordance with the provisions of Section XI of the ASME 8o'iler and Pressure Vessel Code.
Bases The Reactor Building emergency cooling systems anc Reactor Building spray system are designed to remove the heat in the containment atmosph prevent the building pressure from exceeding the design pressure.grg to LU The delivery capability of one Reactor Builoing spray pump at a time can be tested by opening the valve in the line from the borated water storage tank, opening the corresponding valve in the test line, and starting the corresponding pump. Pump discharge pressure and flow indication demonstrate perfomance.
With the pumps shut down and the borated water storage tank outlet valves closed, the Reactor Building spray injection valves can each be opened and closed by operator action.
With the Reactor Builoing spray inlet valves closed, air can be blown through the test connections of the Reactor Building spray nozzles to demonstrate that the flow paths are open.
l l
l Amendment No. 76 4,30
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Bases (continued)
The equipment, piping, valves, and instrumentation of the Weactor Building emergency cooling system are arranged so that they can be visually inspected.
The cooiing units ano associated piping are located outside the secondary concrete shield.
Personnel can enter the Reactor Building during power operations to inspect and maintain this equipment. The nuclear service cooling water piping and valves outside the Reactor Building are inspectable at all times.
REFERENCES (1)
FSAR, section 9.
Amendment No. 76 4-31
RANCHO SECO UN!T 1 TECHNICAL SRECI.ICATIONS Surveillance Standarcs 4.5.2 DECAY HEAT REMOVAL SYSTEM AND REACTOR EUILDING SPRAY SYSTEM LE Acclicability Applies to Decay Hea: Re= oval System and the Reactor Building Spray System leakage.
Obj ective To prevent significan offsite excesures by maintaining a creventive leakage rate for the Decay Hea Re= oval System anc :ne Reac:cr Builcing Spray Sys:s=.
Scecification 4.5.2.1 Accec:ance Li=it The maximum allowable leakage frca the ecmcenents (wnich incluce valve stems, flanges and pump seals) in tne Decay Heat Removal System and the Reac:cr Building Scray Syster snall nc: excesc a se:
ctal of 6.0 gallons per hcur for cc:n systems.
4.5.2.2.A Test - Decav Heat Re cyal System i
During each refueling interval, the following tests of :ne Decay Hea Removal System shall be concucted :: cetermine leakage:
1.
The partien of the Decay Heat Removal System, excect as specified in (2), that is cu:sice the containment snall be i
tested either by use in ncrmal opera:icn or by hycrestatically testing at 450 psig.
2.
Piping from :ne containmen: energency su=c to :ne cecay nea:
removal pu=p suction iscla:icn valve snall be pressure ested at no less than 52 psig as a centainmen: local leak ra:e tes:
uncer paragraph 4.4.1.2.
3.
Visual inscection shall be mace for excessive leakage frem cc=conents of tne system.
Any excessive leakage snall ce measured by ccliection and weighing er by anctner ectivaler.-
me tned.
4.5.2.2.B Test - Reactor Builcine Scrav Svs:s=
During each refueling interval, the following tests cf the Reactor Building Spray System shall be concutted in order to ceter=ine leakage:
Amendment No. 57 a-32
RANCHO SECO UN1T 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.8 AUXILIARY FEEDWATER PUMP PERIODIC TESTING Applicability Applies to the periodic testing of the turbine and motor driven auxiliary feedwater pumps.
Obj ective To verify that the auxiliary feedwater pump and associated valves are operaule.
Specification 4.8.1 System Tests A.
At least once per 18 months during a shutdown:
1.
Verify that each automatic valve in the flow path actuates to its correct position upon receipt of each auxiliary feedwater actuation test signal.
2.
Verify that each auxiliary feedwater pump starts as cesignec automatically upon receipt of each auxiliary feedwater actuation test signal.
B.
The test will be considered satisfactory if visual observation and control board indication veriffes that all components have responded to the actuation signal and the appropriate pump breakers shall have opened or closed, and all power actuated valves shall have completed their travel.
C.
Prior to startup following a refueling shutdown or any cold shutdown of longr:r than 30 cays duration, conduct a test to demonstrate that the motor-driven AFW pumps can pump water from the CST to the steam generator.
D.
All valves including those that are locked, sealed, or otherwise secured in position, are to ue inspectea monthly to verify they are in the proper position.
Amendment No. 31, 76 4-39
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards t
4.8.2 Comoonent Tests A.
Testing At least quarterly, when tne average reactor coolant system temperature is > 305*F, Inservice testing of Auxiliary feedwater System pumps anli valves shall be perfomed in accordance with Section XI of the ASME Boiler and Pressure Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the f2C pursuant to 10 CFR 50 Section SU.55a(g)(6)(i).
The quarterly test requirement shall be brought current within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> af ter the average Reactor Coolant System temperature is 1
5"F.
30 B.
Flow Path Verification Following Inservice testing of pumps and valves as required by paragraphs 4.8.1 ano 4.8.2, requireo flow paths shall be demonstrated operable by verifying that each valve (manual, i
power-actuated or automatic) in the flow path that is not locked in position is in its normal operating position.
Bases The quarterly test frequency will be sufficient to verify that the turbine / motor driven and motor driven auxiliary feedwater pumps are operable.
Verification of correct operation will be made both from the control room instrumentation and direct visual observation of the pumps.
The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 305*F from normal operating conditions in the event of a total loss of off-site power.
Each electric driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 780 gpm at a pressure of 1050 psig to the entrance of the steam generators.
The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 780 gpm at a pressure of 1050 psig to the s
entrance of the steam generators.
This capacity is sufficient to ensure that adequate feedwater flow is available to remove deca heat and reduce the Reactor Coolant System temperature to less than 300{F when the Decay Heat Aemoval System may be placed into operation.
Amendment No. H, 76 4 39, i
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Successive Inspection Intervals i
Every 10 years thereafter (or Volumetric inspection of 1/3 of l
nearest refueling outage) the welds at the expiration of f
each 1/3 of the inspection interval with a cumulative 100 percent coverage of all welds.
Note - The welds selected during each inspection period shall be distributed among the total nuuber to be exmainea to provide a representative sampling of the conoitions of the welds.
3.
Examinations that reveal unacceptable structural defects in a weld during an inspection under 4.13 A 2 shall be extended to require an additional inspection of another 1/3 of the welus.
If further unacceptable defects are detected in the second sampling, the remainder of the welds shall be inspected.
4.
In the event repairs of any welds are required following any examination during successive inspection intervals, the inspection schedule for the repaired welds will revert back to the first 10 year inspection program.
B.
For-all welds in critical areas other than those identified as postulated break location on figure 4.13-1, 2 and 3:
1.
Inservice inspection shall be perfomed in accordance with the provisions'of paragraph 4.2 of these Technical Specifications.
C.
For all welds in the critical areas as identified on figure 4.13-1, 2 and 3:
1.
A visual inspection of the surface of the insulation at all weld locations shall be perfomed on a weekly basis for detection of leaks.
Any detected leaks shall be investigated ano evaluated.
If the leakage is caused by a through-wall flaw, either the plant shall be shutdown, or the leaking piping isolated.
Repairs shall be perfomed prior to return of this line to service.
2.
Repairs, re-examination and piping pressure tests shall be conducted in accordance with the rules of ASME Section XI Code.
Amendment No. 76 4-45
..-. x
I RANCHO SECO UNIT 1 i
TECHNICAL SPECIFICATIONS Basis Under normal plant operating conditions, the piping materials operate under ductile conditions and within the stress limits considerably below the ultimate s trength properties of the materials. Flaws which could grow under such conditio ns are generally associated with cyclic loads that f a tigue the metal, and lead to leakage cracks. The inservice examination and the f re-quency of inspection will provide a means for timely detection even before the flaw penetrates the wall of the piping.
e i
i 1
4-46
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Stancards 4.17 STEAM GENERATORS Applicability Applies to inservice inspection of the steam generator tubes.
' Objective To verify the operability and structural integrity of the tubing as part of the reactor coolant boundary.
Specification Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 1.3.
4.17.1 Steam Generator Sample Selection and Inspection Steam generator tubing shall be demonstrated OPERABLE by selecting and inspecting steara generators as specified in Table 4.17-1.
4.17.2 Steam Generator Tube Sample Selection and Inspection The steam generator tube minimum sample size, inspection result classification, and the corresponoing action required shall be as specified in Table 4.17-2A.
Tne inspection result classification anc the corresponding action required for inspection of " specific limited areas" (see paragraph 4.17.2e) shall be as specified in Table 4.1 7-28. The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specif'ication 4.17-3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.17.4.
The tubes selectea for these inspections shall include at least 3% of the total number of tubes in both steam generators and be selected on a ranoom Dasis except:
a.
If 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.
l l
l Amendment No. 7), 76 4-51
RANCHO SECO UNTT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.17.2 (continued) b.
The first sample in'spection during inservice inspection (subsequent to the first inservice inspection) of each steam generator shall include:
1 1.
All nonplugged tubes that previously had detectable wall 2
J penetrations (>20*.), and 2.
Tubes in those areas where experience has inoicated Potential problems.
The second and third sample inspections during each inservice c.
inspection may be less than a full tube inspection by concentrating (selecting at least 50% of the tubes to be inspected) the !nspection on those areas of the tube sheet erray and on those portions of the tubes where tubes with imperfections were previously founc.
d.
A tube inspection (pursuant to Specification 4.17.4.5) shall be perfomed on each selected tube.
If arty selectea tube does not pemit the passage of the eddy current probe for a tube inspection, this shall be recorded and an aaf acent tube sna11 be selected and subjected to a tube inspection.
(" Adjacent" is interpreted to mean the nearest tube capable of being inspected.)
Tubes which do not pemit passage of the eddy current probe will be considered as cegraded tubes when 4
classifying inspection results, Tubes in specific limited areas which are distinguished by e.
unique operating conditions and/or pflysical construction (for example, tubes adjacent to the open inspection lane or tubes whose 15th' tube support plate hole is not broached but crilled) may be excluded from random samples if all such tubes in the specific area of a steam generator are inspected. No credit will be taken for these tubes in meeting minimum sample size i
requirements.
i The results of each scmple inspection shall be classified into one of the following three categories:
Category Inspection Results C-1 Less than 5% of the total tubes inspected are desraced tubes and none of the inspected tubes are uefective.
C-2 i
One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
C-3 Hore than 10% of the total tubes inspected er'e uegracee tubes or more thari 1% of the inspected tubes are defective.
l Note:
In all inspections, previously uegrated tubes must exhibit significant (>10% ) further wall penetrations to be incluoed in the above percentage calculations.
Amendment No. D, 76 4-52
RANCHO SECO UNZT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.17.3 Inspection Frequencies The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
The first inserMed tr. spec'tf on shall be performed during the a.
first refueling outage. Subsequent inservice inspections shall be perfomed at intervals of not less than 12 nor more than 24 i
calendar months after the previous inspection.
If two consecutive inspections following service result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no significant additional i
degracation has occurred, the inspection interval may be extended to a maximum of once per 40 months.
b.
If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.17-2A and/or Table 4.17-2B l
at 40-month intervals falls in Category C-3, the inspection frequency shall be increased to at least once per 20 months.
The increase in inspection frequency shall apply until a 1
subsequent inspection meets the conditions specified in 4.17.3a and the interval can be extended to a 40-month period.
Additional, unscheduled inservice inspections shall be perfomec c.
on each steam generator in accordance with the first sample inspection specified in Table 4.17-2A during the shutcown subsequent to any of the following conditions:
1.
l Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the j
limits of Specification 3.10, 2.
{
A seismic occurrence greater than the Operating Basis Earthquake, 3.
A loss-of-coQant accicent requiring automatic actuation of 1
the engineered safeguards, or i
4.
A main steam line or feedwater line break as defined in the USAR.
l 4.17.4 Acceptance Criteria As used in this Specification:
a.
1.
Imperfection means an exception to the dimensions, finish or contour of a tube from tnat required by fabrication drawings or specifications. Eddy-current testing indications of less than 20% of the nominal tune wall thickness, if detectable, may be considered as imperfections.
2.
Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.
Amendment No. JA, 76 4 53
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.17.4 A.ce ptance Criteria 3.
Degraded Tube means a tube containing imperfections >20 % of the nominal wall thickness caused by degradation.
4.
Defective Tube means a tube containing an imperfection >40 %
of the nominal tube wall thickness unless higher limits are shown acceptable by analysis. Defective tubes shall be plugged.
5.
Tube Inspection means an inspection of the steam generator tube f rom the point of entry completely to the point of exit (except as noted in 4.17.2c).
b.
The steam generator shall be determined OPERABLE after completing the corresponding actions required by Table 4.17-2A (and Table 4.17-28 if provisions of paragraph 4.17.2e are utilized).
4.17.5 Reports Following each inservice inspection of steam generator tubes, a.
the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.
b.
The results of the steam gene.ator tube inservice inspection shall be included in a Monthly Operating Report for the period in which this inspection was completed. This report shall include:
1.
Number and extent of tubes inspected.
2.
Location and percent of wall-thickness penetration for each indication of an imperfection.
3.
Identification of tubes plugged.
c.
Results of staat generator tube inspections which fall into Category C-3 and require notification of the Concission shall be reported pursuant to Specification 6.9.5.P prior to resumption of plant operation. The written followup of this report shall provioe a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
4.17.6 OTSG Auxiliary Feedwater Header Surveillance On the first refueling outage following the 1983 refueling outage, and at the 10-year ISI, the following inspections will take place:
Amendment No. 73, $$, 76 4-54 a
,-._-,,,,,-e- -., -, - - - -,,, _ --.,
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standarcs 4.17.6 a.
Visual inspections of the secured Internal Heacer, attachment welds ano external headers thermal sleeves will be made thrcugh selected openings, and will be performed in such a manner that the known cracks in the Internal Header will be inspected.
b.
Selected special interest peripheral tubes, designated in Table 4.17-3. will be Eddy Current inspected but shall not be considered a part of the Eddy Current inspection that is conductec pursuant to Technical Specification 4.17.1 through 4.17.5.
4.17.7 Insoection Acceptance Criteria and_Currective Actions Video taped inspections of the known cracks performed during the a.
initial discovery will be compared with the crack configuration found during this surveillance. This comparison will allow a determination to be made as to whetner or not the crack has propagated.
If any inspected special interest peripneral tube indicates D.
clearance (less than 1/4") or greater than 40% througn wall indications, it will be plugged.
4.17.8 Reoorts A report of these inspections will be prepared and incorporated in tne subsequent Monthly Report to the NRC.
Bases The Surveillance Requirements for inspection of the steam generator tuoes ensure that the structural' integrity of this portion of the RCS will be maintained. The surveillance requirements of steam generator tuoes are based on a mccification of B&W - Standard Technical Specifications dated June 1, 1976.
Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event ~ that t j
evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can ce taken.
Operational experience has shown that tube oefects can be the result of unique operating cor.ditions or physical arrangeuents in certain areas of the steam generators. A full inspection of all of the tubes in such limited areas will provide complete assurance that degraded or defective tubes in these areas are detected.
Because no credit is taken for these distinctive tubes in the constitution of the first sample or its results, the requirements for the first sample are unchanged. This requirement is essentially equiavlent to and meets the intent of the requirements set forth in NRC P.agulatory Guide 1.83, Supplement I and does not reduce the margin of safety provided by those requirements.
i l
Amendment No. 7), JHf, 76 4.55
1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Stancarcs j
Bases (Continued) i
~
Wastage-type defects are unlikely with AVT chemistry treatment of the seconcary coolant. However, even if a defect should develop in arvice, it will be found during scheduled inservice steam generator tube examinations.
Plugging will be required for defective tubes. Steam generator tube inspections of operating plants 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 l
into Category C-3, these results will be reported to the Commission pursuant i
to Specification 6.9 prior to resuuption 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 inspectien and evision of the Technical Specifications, if necessary.
The visual and eddy current inspections provide the capability of ueterminine the success of the internal aux feedwater header stabilization by monitoring the long term effects to tube integrity and crack propagation. The inspections will focus on known crack growth, new crack identification (if any), and tube ef fects in localized areas near the internal header DracAe i.
Adcitionally, visual inspection of the external header thermal sleeves will provide assurance that the new design header will not introduce adoitional problems by demonstrating sleeve integrity.
i 4
I
?
Amendment No. 7), $$, 76 4-55a i
i i
l j
. - _ - _. _ - _. _. _,. _ _ _ _. _, _. _. _. _ _. _. _ ~ -. _. _, _ _ _ _ _
AANCHD SECO Intf l IttiselCAL SPECIFICAll045 g
5erveillance Standards TAett 4.17-24 i
STEAM GlutRATOR Test INSPECTICII a
15i SAMPtE INSPtC1104 240 SAMPLt INSPECTlose Jne SAMPtE INSPtCIfou Sample 51se nesult Actien negsIred mesett Actien secIred mesult Action neguired 5
A ofalous of C-1 lbne N/A N/A N/A N/A 5 et the Iobes per c-z PIvy detective c-I Ilone 5.G.
tubes and f aspect N/A N/A addittenal 25 et E-z FIng eerective tubes one t-s the tubes in this inspect additional 45 of none 5.G.
the tubes In tlsts 5.G.
E-z r wg serective tubes 0-J Perform action for U.J 1
result of first sasyle C-3 Ferrern action for [-3 N/A nin result of first sample CJ Inspec t aII tubes The et ~
None N/A n/A in thfs 5.G.. plug 5.G. Is defectlee tubes C-1 and inspect 25 of time tubes in the The other Perform action Ier L-2 N/A N/A ether 5.G.
5.G. Is result ef secend sample C-2 hetIfIcetfen to anc perseant to specificatfon 6.9 The other Inspect aII tubes in N/A N/A 5.G. Is each 5.G. and pIwg C-3 defective tubes.
Ilottf fcatfen to NRC pursuant to specifica-i Lion 6.9 6
5. g % eere a f s the neder of steam generators f aspected during en f aspection l
4-57 e
i
RANClio 5(CD UNii I o
TECHNICAL SPECIFICAll0NS 1
B Survefilence Standards IAtt[ 4.17-20 5
STEAM GENERATOR Tvet INSPttil0lf E
151 591PLE INSPECIl0N OF A '5PECIFIC Linlitu Amt4*
2ND 5 AMPLE INSPfCiloll 0F A *5PECIFIC LIMITES AmtA*
Sample Stre Result Action Regufred Result Action Required 1001 of Area C-1 None N/A N/A la both 015Gs C-2 Plug defective tubes N/A N/A C-3 Plug defective tubes.
N/A N/A Notification to INIC pursuant to specification 6.9.
3001 of Area C-1 Ilone N/A N/A la one 015G C-2 Play defective tubes and f aspect C-!
Ilone 1001 of corresponding area in other 015G.
C-2 Plug defective tubes C-3 Plug defective tubes C-3 Plug defective tubes and f aspect C-1 Ilene 1005 of corresponding area la ether OI5G. Ilotification to Inst. pursuant to specification C-2 Plug defective tubes 6.9.
C-3 Plug defective tubes 4 51a l
i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS I
Surveillance Stancarus TABLE 4.17-3 OTSG Auxiliary Feedwater Header Surveillance OTSG A Special Interest Tubes l
Row Tube No 5
1, 46 6
2, 3, 49, 50, 51 7
1, 2, 53, 54 8
1, 2, 56, 57 44 1, 119 4
45 1, 120 46 1, 121 47 1, 122 48 1, 123 49 1, 124 103 1, 124 105 1
106 1, 119 1
107 1, 120 l
l 108 1, 119 144 1, 2, 56, 57 145 1, 2, 53, 54 3
l 146 2, 3, 49, 50, 51 147 1, 46 i
~
Amendment tio. %; 76 4-57 b l
s j
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standarcs TABLE 4.17-3 (Continued)
OTSG Auxiliary Feedwater Heauer Surveillance i
OTSG B Scecial Interest Tubes Row Tube No 5
1, 46 6
2, 3, 49, 50 7
1, 2, 53, 54 8
1, 2, 56, 57 44 1, 119 d
45 1, 120 46 1, 121 l
47 1
122 48 1, 123 49 1, 124 103 1, 124 104 123 105 122 106 1, 119 -
107 1, 120 106 1, 119 144 1, 2, 56, 57 145 1, 2, 53, 54 146 2, 3, 49, 50, 51 147 1, 46 1
l 9
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Amendment flo. 56, 76 4-57c l
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RANCHO SECO UNIT 1 IECHNICAL SPECIFICATIONS Acministrative Controls Scecial Reoorts 6.9.5 Special reports shall be submitted to the Regional Acministrator Region Y Office within the time period specified for each report.
These reports shall be submitted covering the act?vities icentified below pursuant to the requirements of the applicable reference specification:
A.
A one-time only, " Narrative Sumiary of Operating Experience" will be submitted to cover the transition period (calencar year 1977).
E.
A Reactor Builcing structural integrity report shall be submittec within ninety (90) days of comoletion of each of the following tests coverec cy Tecnnical Specification 4.4.2 (the integratec leak rate test is coverec in Tecnnical Specification 4.4.1.1).
1.
Annual Inspection 2.
Tencon Stress Surveillance 3.
Enc Ancnorage Concrete Surveillance 4.
Liner Plate Survefila.c4 Inservice Inspection Program D.
Reservec for Proposed Amencment No. 43 E.
Status of Ino;eracle Fire Protection Equipment F.
Inoperacle Emergency Control Room /TS Yentilation Room Filter System 3.
Radioactive Liquid Effluent Dose 30 days (3.17.2',
H.
Noble Gas Limits 30 days (3.18.2) 1.
Radioicdine and Particulates 30 days (3.18.3)
J.
Gaseous Radwaste Treatment 30 days (3.19)
K.
Radiological Monitoring Program 30 days (3.22)
L.
Monitoring Point Substitutions 30 days (3.22) n.
Deleted N.
Fuel Cycle Dose 30 days (3.25) 0.
Deleted p.
Steam Generator Tube Inspection 30 days (4.17.5)
Amendment No.
. 72, 52, X, 75, 6-12f
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