ML19256G480
| ML19256G480 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 12/12/1979 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML19256G476 | List: |
| References | |
| TAC-11288, TAC-51007, NUDOCS 7912310368 | |
| Download: ML19256G480 (26) | |
Text
e RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 4.
Two core flood tank pressure instrument channels shall be operable (one per tank minimum).
5.
The electrically operated vent valves (HV-26511 and HV-26512) f rom the core flood tanks shall be closed. The breakers shall be open and so tagged except during normal venting "perations.
C.
Reactor Building spray system and Reactor Building emergency cooling sys tem.
The following combination of system components must be operable:
1.
Two Reactor Building spray purrps and their associated spray headers with a minimum of 32 percent NaOH solution in the spray additive tanks and, 2.
A minimum level of 78 inches of solution shall be available in each spray addi tive tank.
3.
Four emergency cooling units, two with charcoal filter units.
D.
Nuclear service cooling and raw water cooling system.
1.
Two nuclear service cooling water (NSCW) pumps and raw water cooling pumps are operable.
2.
The manual valves in the NSCW suction and discharge of each operable Reacter Building emergency cooling unit and at the suction of each NSCW pump are locked open.
3 The manual valves in the suction and discharge lines of all operable heat exchangers served by the nuclear service raw water system are locked in their throttled or open position.
E.
Safety features valves and interlocks associated with each of the above sys tems are operable.
Inoperable valves shall be placed in the safety fea tures posi tion.
3.3 2 Maintenance shall be allowed during power operation on any component (s) in the high pressure, I;u pressure, nuclear. service cooling and raw water cooling, Reactor Building spray, or Reactor Building emergency cool' g sys tems, the core flooding system pressure instrunent channels or B',..T level channels, which wi l l. not degrade safety features system A or B below the level of performance with the single subsystem removed from service.
Ir i.he context of this speci.ficat ion, a safety features sys tem consis ts of the following subsystems:
high pressure injection, low pressure injection, Reactor Building emergency aC cooling, Reactor Building spray, diesel generator, nuclear servi c cooling water and nuclear service. raw water.
If the system being repaired is not restored to treet the requirements of speci fication 3 3.1 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the reactor shall be placed in a hot shutdown condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
i f the requi rements of speci fica tion 3.3.1 63 are not met within an additional 7 days, the reactor shall be placed i in e cold shutdown condition within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Amendmegh $3 1656 037'
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3-20 7910910
RANCHO SECO UNIT 1 TECHNICAL-SPECIFICATIONS Limit,ing Conditions for Operation
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3 3.3 Prior to initiating maintenance on any of. the component (s), the duplicate (redundant) component (s) shall be verified operable by checking that the surveillance test for the component (s) has been successfully completed and will remain in effect for the duration of the maintenance period.
63 Survelllance testing which removes a component from an automatic mode sF
- A be performed on any component (s) whose duplicate (redundant) cn>.~nen.
) has(have) been declared inoperable or is out of service for any rc. aon.
Bases The re
..ements of Speci fication 3.3.1 assure that, before the reactor can be made critical, adequate safety features are operable. Two high pressure injection pumps and two decay heat removal r mps 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 water for flooding the fuel transfer canal during refueling operation. (2) 390,000 gallons of borated water are supplied for emergency core cooling and Reactor Building spray :n the event of a loss-of-core coolant accident.
This amount fulfills requirements for emergency core cooling. The borated 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 I percent subcritical at 70*F without any control rods in the core. This concentration is 1585 ppm boron while the minimum value specified in the tanks is 1,800 ppm boron.
The requirement that one BMST isolation valve shall be open assures a static head to the injection pump not lined up to the makeup tank.
The post accident Reactor Building cooling may be accomplished by two spray units or by a combination of two emergency cooling unit:. and one spray unit.
The specified requirements assure that the required post accident components are available.
The spray system utilizes common suction lines with the decay heat removal system.
If a single train of equipment is removed from either system, the other train must be assured to be operable in each system.
Wnen the reactor is critical, maintenance is allowed per Specification 3.3.2 provided requirements in Specification 3.3.3 are met which assure operability of the duplicate components. Operability of the specified components shall be based on the results of testing as required by Technical Specification 4.5 The maintenance period of up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is acceptable if the equipment 63 redundant to that removed from service is veri fied operable. The basis of acceptabi1ity is a low 1ikelihood of failure within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> foilowing such confirmation.
Proposed Amendment No. 63 Rev. I 3-21 1656 038
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Ccnditions for Operation 3.4.2.2 Both auxiliary feed pumps are operable.
One auxiliary feed pump may be out of servi ce for maintenance for a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
if not repaired within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the reactor shall be placed in a hot shutdown condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Prior to initiating main-tenance on any of the component (s), the duplicate (redundant) com-ponent(s) shall be verified operable by chec'(ing that the surveil-63 lance test for the component (s) has been successfully completed and will remain in effect for the duration of the maintenance period.
Survei11ance testing which removes a component f rom an automatic mode sha!1 not be performed on any component (s) whose duplicate (redundant) component (s) has(have) been declared inoperable or is out of service for any reason.
Bases The feedwater system and the turbine bypass system are normally used for decay heat removal and cooldown above 280*F.
Feedwater makeup is supplied by operation of a condent. ate pump and main feedwater pump.
In the event of complete loss of electrical power, feedwater is supplied by a turbine driven auxiliary feedwater pump which takes suction from the condensate storage tank.
Steam relief would be through the system's atmospheric relief valves.
If neither main feed pump is available, feedwater can be supplied to the steam generators by an auxiliary feedwater pump and steam relief would be through the turbine bypass system to the condenser.
In order to heat the reactor coolant system above 280*F the maximum steam removal capabili ty required is 4-1/2 percent of rated power. This is the maximum decay heat rate at 30 seconds af ter a reactor trip. The require-ment for two steam system safety valves per steam generator provides a steam relief capability of over 10 percent per steam generator (1,341,938 lb/h).
In addition, two turbine bypass valves to the condenser or two atmospheric dump valves will provide the necessary capacity.
The 250,000 gallons of water in the condensate storage tank is the amount needed for cooling water to the steam generators for a period in excess of one day following a complete loss of all unit ac power.(1)
The minimum relief capacity of seventeen steam system safety valves is 13,329,163 lb/hr.(2) This is sufficient capacity to rotect the steam system under the design overpower condi tion' of 112 percent. (p)
~ 3 REFERENCES (1)
FSAR paragraph 14.1.2.8.4 (2)
FSAR paragraph 10.3.4 1656 039 (3)
FSAR Appendix 3A, Answer to question 3A.5 Proposed Amendment No. 63 Rev. 1 3-24
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.6 REACTOR BUILDING Applicability Applies to the containment when the reactor is subcritical by less than I percent ak/k.
Objective To assure containment integrity during startup and operation.
Speci fica tion
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3.6.1 Containment integrity shall be maintained whenever all three of the following conditions exist:
A.
Reactor coolant pressure is 300 psig or greater.
B.
Reactor coolant temperature is 2000F or greater.
C.
Nuclear fuel is in the core.
3.6.2~
Containment integrity shall be maintained when the reactor coolant system is open to lhe containment atmosphere and the requirements for a refueling shutdown are not met.
3.6.3 Positive reactivity insertions which would result in the reactor being subcritical by less than 1 percent Ak/k shall not be made by control rod motion or boron dilution whenever the containment integrity is not intact.
3.6.4 The reactor shall not remain critical if the Reactor Building internal pressure exceeds 1.5 psig or vacuum exceeds -1.5 psig.
3.6.5 Prior to criticality following refueling shutdown, a check shall be made to confirm that all manual containment isolation valves which should be closed are closed.
3.6.6 If, during critical operations, an automatic containment isolation valve is determined to be inoperable, the other containment isolation valve in the line shall be verified operable by checking that the 63 surveillance test has been successfully completed.
If the inoperable valve is not restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be brought to the 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 /> or the valve will be closed in the safety features position.
(63 1656 040 3-39 Proposed Amendment #63 Rev. I
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-2 MINIMUM EqulPMENT TEST FREQUENCY ltem Test 1.
Control rods Rod drop times of Each refueling shutdown all full length rods 2.
Control rod Movement of each Every two weeks movement rod 3
Pressurizer code Setpoint Note 3 safety valves 63 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 Note 3 l63 valves valve 7
Reactor Coolant Leakage Calculated inventory weekly system Leakage check daily.
8.
Charcoal and high Charcoal and HEPA Each refueling interval and efficiency filters filter for iodine at any time work on filters and particulate could alter their integrity.
removal efficien-cies.
D0P test on HEPA filters.
Freon test on charcoal filter units.
9 Fire pumps and Functional Monthly power supplles 10.
Reactor Building Functional Each refueling interval isolation trip 11.
Spent fuel Functional Each refueling interval prior cooling system to fuel handling 12.
Turbine Overspeed Calibration ach refueling interval Trips 13 Internals Vent Movement of each Note 3 63 valves valve 1656 041 1.
Deleted 63 2.
Deleted 3
Tested in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where 63 specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
4-8 Proposed Amendment #63 Rev.1
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.1-2 DELETED 63 1656 042 4-8a Proposed Amendment #63 Rev. I
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.2 SURVEILLANCE OF ASME CODE CLASS 1, 2 and 3 SYSTEMS Appilcability Applies to systems and components defined as ASME Code Classi, Class 2, and Class 3 Objective To establish examinations whereby the integrity of ASME Code Class 1, 2 and 3 sy' stems and components is monitored.
63 Speci fica tion 4.2.1 The reactor vessel material irradiation surveillance specimens removed from the reactor vessel at approximately 170 effective full-power days shall be installed, i rradiated in and withdrawn f rom the Davis-Besse Unit No. I reactor vessel in accordance wi th the schedule shown in Table 4.2-1.
Following withdrawal 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.
4.2.2 A report or application for license amendnent shall be submitted to the NRC within 90 days af ter the occurrence of any of the following:
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 at 100% power, any time that Davis-Besse Unit No. I fails 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 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 4.2.3 Inservice inspection of ASME Code Class 1, Class 2 and Class 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, 63 Section 50.55a(g), except where speci fic written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
Amendment No. 63 4-10 1656 043
RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Surveillance Stendards 4.2.4 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.
4.2.5 Each reactor coolant pump motor flywheel will be inspected volumetrically during the ten year inspection interval.
One hundred percent of the fly-63 wheel will be examined. All flywheels received a one hundred percent ultrasonic examination prior to installation on the notor.
4.2.6 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 Boiler and Pressure Vessel Code.
- 4. 2.' 7 Fbcords of each inspection shall be kept to permit evaluation and future comparison.
Bases Irradiation surveillance provides the capability of determining 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-af fected zone are installeJ in capsule assemblics placed inside the vessel.
In accordance with the schedules of Table 4.2-1 specimens will be re-moved; and a series of drop weight tests, Charpy impact tests and tension tests will be conducted. Threshold neutron flux detectors and maximum temperature detecters will be installed with the specimens. Changes in nil-ductility transition temperature will be 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 neutron fluence lead factor between the capsules and the reactor vessel. This provides 63 ample time for anticipating probiems and initiating corrective action should operation of the host reactor be seriously delayed.
For the purpose of Technical Speci ficat ion 4.2.8, the defini tion of Regulatory Guide 1.16, Revision 4 (August 1975) applies for the term " commercial operation".
Cumula-tive reactor utilization factor is defined as:
[(Cumulative thermal megawatt hours since attainment of commercial operation at 1003 power) x 100] : [(licensed thermal power) x (cumulative hours since attainment of commercial operation at 100% power)].
Amendment No. 63 4-11 1656 044 i
-RANCHO SECO UNIT l '
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TECHNICAL SPECIFICATIONS Surveillance Standards Preoperational and inservice inspections emphasize areas of highest stress con-centration and probability of failure.
The area predoninantly selected for these examinations are welds and the adjacent metal.
Examination of the welds 63 is of ten by a volumetric (ultrasonic or radiography) nethod which, when performed, examines surrounding base metal and the weld heat-af fected zone. Both testing methods will use present state-of-the-art equipment aperated by highly trained personnel quali fied within the requirements of the applicable codes.
l 1656 045 Amendnent No. 63 4-12
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards TABLE 4.2-2 63 DELETED O
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1656 046 Amendment No. 63 4-13 e
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.3 TESTING FOLLOWING OPENING OF SYSTEM Applicability Applies to test requirements for reactor coolant system integrity.
Objective T'o assure reactor coolant system Integrity prior to return to criticality fol-lowing 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 applicable code requirements prior to the reactor being made 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 Section XI of the ASME Boiler and Pressure Vessel Code.
63 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; 50 psi is normal pressure fluctuation), it will be leak tight during normal operation.(I)
REFERENCES (1)
FSAR, section 4 O
Amendment No. 63 4-14 1656 047
RANCHO SECO UNIT l TECHNICAL SPEC 1FICAT10NS Surveillance Standards 4.4.1.3 Isolation Valve Functional Tests Except where valve failure during cycling would cause a loss of system function or where valve failure to close during cycling vould result in a loss of containment integrity, remotely-operated reactor building isolation valves shall be stroked to the position required 63 to fulfill their safety function in accordance with requirements of Section XI of the ASME Boiler and 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 50.55a(g)(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 performed annually and prior to any integrated leak test, to uncover any evidence of deterioration which may affect either the contain-ment'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 nodification 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 59 psig and a steam-air mixture temperature of 286 F.
Prior to initial operation, the con-tainment will be strength tested at 115 percent of design pressure. The containment will also be leak tested prior to initial operation at P and P (52 psig and 26 psig, respectively). These tests will verify that tee leakage g'ressurizationsatisfiestherelationshipsgiven rate from Reactor Buil in the speci fication.
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 performed without pre-liminary 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 for the periodic integrated leakage rate test is sufficiently high to provide an accurate measurement of the leakage rate and it duplicates the pre-ope ra tiona l 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 stabili2c conditions and thus improve Proposed Amendment No. 63 Rev. 1 4-19 1656 048
RANCHO SECO UtilT-1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.5 EMERGENCY CORE COOLING AND REACTOR DUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 EMERGENCY CORE COOLING SYSTEM ApplicabiIity Applies to periodic testing requirement for emergency core cooling sys tems.
Objective To veri fy that the ercrgency core cooling systems are operable.
Speci fica tion 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 demonstrate the system is operable for high pressure injection. A manual initiate signal will be applied to demonstrate actuation of high pressure 63 Injection 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 all power-63 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.
B.
Low Pressure injection 1.
During each refueling interval a decay heat removal system test shall be conducted to deronstrate the system is operable for low pressure injection.
The test shall be performed in accordance with the pro-cedure summarized below:
a.
A manual initiate signal will be applied to demonstrate l 63 actuation of the decay heat removal system for emergency core cooling operation.
b.
Verification of the safety features function of the nuclear service cooling water system and nuclear service raw water system which supplies cooling water to the decay heat removal cooiers shall be made to demonstrate operability of the coolers.
Proposed Amendment No. 63 Rev. 1 1656 049
RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Surveillance Standards Specification (Continued) i B.
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 63 have completed their travel.
3 Decay heat pump casings shall be vented monthly and prior to any ECCS flow tests.
C.
Core Flooding System 1.
During each refueling interval, a core floeding system test shall be conducted to demonstrate proper operation of the system.
During depressurization of the reactor coolant system, verification shall be made that core flood tank 63 fluid discharges to reactor coolant system.
2.
The test will be considered satisfactory if control board indication of core flood tank level verifies that core 63 flood tank level has decreased.
4.5.1.2 Components Tests A.
Except as outlined in C below, inservice testing of ECCS and Nuclear Service Cooling and Raw Water pumps and valves snail be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(9), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
B.
Flow Path Verification Following inservice testing of pumps and valves as required by paragraph 4.5.1.2A, required flow paths shall be demonstrated operable by veri fying that each valve (manual, power-operated 63 or automatic) in the flow path that is not locked in position is in its normal operating position.
Positions cf locked valves shall be verified during each refueling interval.
C.
Exceptions to Pump and Valve Testing Requirements 1.
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.
2.
Any valve,whose failure in a non-conservative position would cause total loss of system function,shall not be tested.
3 Any valve,whose failure to close during cycling would result in a loss of containment integrity shall not be tested during power operation.
t l
4.
Any valve, which when cycled could subject.a system to pressures in excess of its design pressure, shall not be tested.
I Proposed Amendment No. 63 Rev. 1 4-27 1656 050
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 63 Bases The 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 oy opening the borated water storage tank outlet valves and the test line valses 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 veri fy the check valves have opened.
REFERENCES FSAR subsection 6.2.
Amendment No. 63 4-28 16,56 051 9
i e
1 RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Surveillance Standards l
4.5 2 REACTOR BUILDING COOLING SYSTEMS Applicab!Iity Applies to testing of the Reactor Building cooling systems.
Obj ect ive 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 con-ducted to demonstrate proper operation of the system. A r.anual trip signal will be applied to demonstrate actuation of the Reactor Building spray system (except for Reactor Building motor-operated inlet valves which prevent water entering nozzles). Water will be circulated from the borated water storage tank through the Reactor Building spray pumps and returnad through the test line to the borated water storage tank.
2.
Air will be introduced into the spray headers to verify the availibility of the headers and spray nozzle at least every 10 years.
3 The test will be considered satisfactory if visual obser-vation and control board indication verifies that all components have responded to the actuation signal and the appropriate pump breakers shall have opened and closed, and all power-operated valves shall have completed their 63 travel except the blocked Reactor Building inlet valve.
B.
Reactor Building Emergency Cooling System 1.
During each refueling interval, a system test shall be conducted to demonstrate prcper operation of the system, including the upper dome air circulators.
The test shall be performed in accordance with the procedure summarized below:
A manual trip signal will be applied to actuate the a.
Reactor Building emergency cooling system for Reactor Building cooling operation.
Proposed Amendment 63 Rev. I 1
4-29 1656 052
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards b.
Verification of the safety features function of the nuclear service cooling water system which supplies coolant water to the Reactor Building coolers shall be made to demonstrate operability of the coolers.
2.
The test will be considered satisfactory if control board indication verifies that all components have responded to the actuation signal and the appropriate pump breakers shall have completed their travel.
NSCW flow through each oper: ting cooler exceeds 1400 gpm, fans are running a,d air flow through each cooler exceeds 40,000 cfn.
4.5.2.2 Component Tests A.
Except as outlined in C below; inservice testing of Reactor
!63 Building Spray pumps and valves shall be performed in accord-ance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Q, Section 50.55a(g), except where specific written relief has 'oeen granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
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 by verifying that each valve (manual, power-operated or automatic) in the flow path that 1s not locked in position is in its rormal operating position.
Positions of locked valves shall be verified during each refueling interval.
C.
Exceptions to Pump and Valve Testing Requirements 1.
If the duplicate (redundant) component has been declared inoperable, or is out of service for any reason, the com-ponent shall not be tesied during power operation.
2.
Any valve whose failure in a nonconservative position would cause total loss of system function shall not be tested during power operation.
63 3
Any valve whose failure to close during cycling would re-sult in a loss of containment integrity shall not be tested during power operation.
4.
Any valve, which when cycled, could subject a system to pressures in excess of its design pressure shall not be tested.
Bases The Reactor Building emergency cooling systems and Reactor Building spray system are designed to remove the heat in the containment tmosphere to prevent the build-
\\I Ing pressure from exceeding the design presrure.
The delivery capability of one Reactor Building 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 performance.
4-30 Proposed Amendment #63 Rev. 1 1656 09
RANCHO SECO UNI.T l-TECHNICAL SPECIFICATIONS Surveillance Standards 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 Building spray inlet valves 63 closed, air can be blown through the test conncctions of the Reactor Building spray nozzles to demonstrate that the flow paths are open.
The equipment, piping, valves, and instrumentation of the Reactor Building emergency cooling system are arranged so that they can be visually inspected.
The cooling enits and associated piping are located outside the secondary concrete shield.
Personnel can enter the Reactor Building during power opera-tions to inspect and maintain this equipment. The nuclear service cooling water piping and valves outside the Reactor Building are inspectable at all times. Operational tests shall be performed prior to initial startup.
REFERENCES (1)
FSAR, section 9
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o Armndment No. 63 4-31 1656 054
RANCHO SECO UNIT 1 i
TECHNICAL SPECIFICATIONS Surveillance Standards 4.8 AUXILIARY FEEDVATER PUMP PERIODIC TESTING Applicability Applies to the periodic testing of the turbine and motor-driven auxillary feedwater pumps when the average RCS temperature is >305 F.
I Obj ective i
To verify that the auxiliary feedwater pump and associated valves are operable.
Specification 4.8.1 Except as outlined in 4.8.3 below, Inservice testing of Auxiliary Feedwater System pumps and valves shall be performed in accordance 6
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 f
specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50 55a(g)(6)(i).
The ASME Section XI 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 /> after the a arage Reactor Coolant System temperature is >305 F 4.8.2 Following inservice testing of pumps and valves as required by paragraph 4.8.1, required flow paths shall be demonstrated operable by verifying that each valve (manual, power-operated 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 during each 63 refueling interval.
4.8.3 Exceptions to Pump and Valve Testing Requirements A.
If the duplicate (redundant) component has been declared inoper-able, or is out of service for any reason, the component shall not be tested during power operation.
B.
Any valve whose failure in a non-conservative position would cause total loss of system function shall not be tested during power operation.
C.
Any valve whose failure to close during cycling would result in a loss of containment integrity shall not be tested during power operation.
D.
any valve, which when cycled could subject a system to pressures in excess of its design pressure, shall not be tested.
Bases The test frequency required by Section XI will be sufficient to verify that'the l63 turbine / motor driven and motor d-iven auxiliary feedwater pumps are operable.
Verification of correct operation will be made both f rom the control room instrumentation and direct visual observation af thc pumps.
Proposed Amendme % Q 6h h. I 4-39
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Star.dards Successive inspection Intervals Every 10 years thereaf ter (or Volumetric Inspection of 1/3 of nearest refueling outage) the welds at the expiration of each 1/3 of the inspection interval witt a cumulative 100 percent coverage of all welds.
Note - The welds selected during each inspection period shall be distributed among the total number to be examined to provide a representative sampiing of the conditions of the welds.
3.
Examinations that reveal unacceptable structural defec'ts 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 welds.
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 ucids 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 or. figure 4.13-1, 2 and 3:
1.
Inservice inspection shall be performed in accordance with the 63 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 performed on a weekly basis for detection of leaks. Any detected leaks shall be investigated and evaluated.
If the leakage is caused by a through-wall flaw, ei ther the plant shall be shutdown, or the leaking piping isolated. Repairs shall be perforred 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. 63 4-45 1656 056
RkNCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.17 STEAM GENERATORS kpplicability Applies to inservice inspection of the steam generator tubes.
Objective To verify the operability and structural integrity of tubing as part of the 63 reactor coolant bounda ry.
Speci fica tion f:
7, Each steam generator shall be demonstrated'pPERABLE by pdformance of the following augmented inservice inspection program and the requirements of Specification 1.3 4.17.1 Steam Generator Sampie Selection and Inspection Steam generator tubing shall be demonstrated OPERABLE by 63 by selecting and inspecting steam generators as specified in Table 4.17-1.
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4.17.2 Steam Generator Tube Samole Selection and inspection The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.17-2A.
The inspection result classi fication and the corre.sponding action required for inspection of " specific 63 limited areas" (see paragraph 4.17.2e) shall be as specified in Table 4.17-2B. The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.17-3 and the inspected tubes shall be veri fled acceptable per the acceptancc cri teria of Speci fication 4.17.4.
The tubes selected for these inspections shall include at least 3% of the total number of tubes in both stean generators and be selected on a random basis except:
If experience in similar plants with similar water chemistry a.
indicates critical areas to be inspected, then at least 50%
of the tubes from these c ri tical areas shall be inspected.
l 63 b.'
The first sample inspection during inservice inspection (sub-sequent to the fi rs t inservice ins'pection) of each steam generator shall include:
1.
All nonplugged tubes that previously had detectable wali penetrations (>20%), and 2.
Tubes in those areas where experience has indicated poten-tial problems.
Amendment No. 63 4-51 1656 057
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RANCHO SECO UNIT I
, TECHNICAL SPECIFICATIONS Surveillance Standards c.
The second and third sample inspections during each inservice inspection may be less than a full tube inspection by concentrating
(=clecting at 1 cast 50% of the tubes to be inspected) the inspec-tion on those areas of the tube sheet array and on those portions of the tubes where tubes with imperfections were previously found.
d.
A tube inspection (pursuant to Specification 4.17.4.5) shall be per-forned on each selected tube.
If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
("Adj acen t" is interpreted to mean the nearest tube capable of being inspected.)
Tubes which do not permit passage of the eddy current probe will be considered as degraded tubes when classifying inspection results.
e.
Tubes in specific limited areas which are distinguished by unique operating conditions and/or physical construction (for example, tubes adjacent to the open inspection lane or tubes whose 15th tube support plate hole is not broached but drilled) 63 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 requi rements.
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The results of each sample inspection shall be classified into one of the following three categories:
Category inspection Resul ts C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2 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 de-graded tubes.
C-3 Hore than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
Note:
In all inspections, previously degraded tubes must exhibit signi ficant (>10%) further wall penetrations to be included in the above percentage calculations.
4.17.3 Inspection Frequencies.
The above required inservice inspections ~ of steam generator tubes shall be perforred at the following frequencies:
a.
The first inservice inspection shall be performed during the first refueling outage. Subsequent inse rvi ce inspections shall be performed at intervals of not less than 12 nor more than,24 Amendment No. 63 4-52 1656 058
{ll lll RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards calendar months after the previous inspection.
If two con-secutive inspections following service resul t in all inspec-tion results falling into the C-l category or i f two consecu-tive inspections demonstrate that previously observed degradation has not continued and no signi ficant addi tional degradation 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 l63 at 40-month intervals falls in Category C-3, the inspection f requency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until a subsequent inspection meets the condi tions speci fied in 4.17.3a and the interval can be extended to a 40-mcnth period.
c.
Additional, unscheduled inservice inspections shall be per-formed on each steam generator in accordance with the first sample inspection speci fied in Table 4.17-2A during the shut-l 63 down 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 limi ts of Speci fication 3.10, 2.
A seismic occurrence greater than the Operating Basis Earthquake, 3
A loss-of-coolant accident requi ring automatic actuation of the engineered safeguards, or 4.
A main steam line or feedwater line break as defined in the FSAR.
l 4.17.4 Acceptance Criteria a.
As used in this Specification:
1.
Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or speci fications. Eddy-current testing indi-cations of less than 20% of the nominal tube wall thick-ness, if detectable, may be considered as imperfections.
2.
Degradation neans a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.
Ancndrent No. 63 4-53 1656 059
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O RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 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 140% 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 tubing shall be determined OPERABLE af ter completing the corresponding actions required by Table 4.17-2A 63 (and Table 4.17-2B if provisions of paragraph 4.17.2e are utilized.)
4.17.5 Reports Following each inservice inspection of steam generator tubes, the a.
number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.
b.
The results of the steam generator tube inservice inspection shall be included in a Monthly Operating Report. This repo rt 63 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.
Results of steam generator tube inspections which fall into c.
Category C-3 and requi re noti fication of the Commission shall be reported pursuant to Specification 6.9 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 corrective measures taken to prevent recurrence.
Bases 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 surveillance requirements of steam generator tubes are based on a modi fication of B&W - Standard Technical Speci fications dated June 1,1976.
Inservice inspection Amendment No. 63 4-54 1656 060
RANCHO SECO UNIT.1 TECHNICAL SPECIFICATIONS Surveillance Standards of steam generator tubing is essential in order to maintain 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 inser-vice 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 be taken.
Operational experience has shown that tube defects can be the result of unique operating conditions or physical arrangements in certain areas of the steam gen e ra tors. 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 63 constitution of the first sample or its results, the requirements for the first sample a re unchanged. This requirement is essentially equivalent to and meets the intent of the requi rements set forth in NRC Regulatory Guide 1.83, Supplement I and does not reduce the margin of safety provided by those requi remen ts.
Vastage-type defects are unlikely with AVT chemistry treatment of the secondary coolant.
Howeve r, even i f a defect should develop in service, 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 thickncss.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to Speci-fication 6.9 prior to resumption of plant operation.
Such cases will be con-sidered by the Commission on a case-by-case basis and may result in a requi rement for analysis, laboratory examinations, tests, additional eddy-current inspection and revision of the Technical Speci fications, i f necessary.
s Amendment No. 63 4-55 1656 061
TABLE 4.17-2A STEAM GENERATOR TUBE INSPECTION IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required A minimum of C-1 None N/A N/A N/A N/A S of the i
Tubes per C-2 Plug defective C-1 None N/A N/A l
5.G.
tubes and inspect 8-additional 2S of C-2 Plug defective tubes and C-1 None W
the tubes in this inspect additional 4S of S.G.
the tubes in this S.G.
C-2 Plug defective tubes C-3 Perform action for C-31 result of first sample.
C-3 Perform action for C-3 N/A N/A result of this sample 2-t un C-3 inspect all tubes The other None N/A N/A in this S.G., plug S.G.
Is defective tubes C-1 l
and inspect 25 of the tubes in the The other Perform action for C-2 N/A N/A Cys other S.G.
S.G.
Is result of second sample (J7 C-2 cd Notification to NRC pursuant The other inspect all tubes in N/A N/A c) f('
to specification S.G.
Is each S.G. and plug
)
6.9 C-3 defective tubes.
Motification to NRC pursuant to specifica-tion 6.9 S=6 %
Where n is the number of steam generators inspected during an inspection
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