ML20107M672
| ML20107M672 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 10/26/1984 |
| From: | Knighton G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20107M678 | List: |
| References | |
| TAC-55785, TAC-55786, TAC-56216, TAC-56217, TAC-56218, TAC-56221, TAC-56222, TAC-56223, NUDOCS 8411140225 | |
| Download: ML20107M672 (30) | |
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SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS'AND ELECTRIC COMPANY THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHEIM,' CALIFORNI'A.
DOCKET NO. 50-361 SAN ON0FRE NUCLEAR GENERATING STATION, UNITJ2:
AMENDMENT TO FACILITY OPERATING LICENSE-Amendment' No. 26 License No. NPF.
- 1..The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applications for amendment to the' license for San Onofre Nuclear Generating Station, Unit 2 (the facility) filed by the Southern-California Edison Company on behalf of itself and San Diego Gas and Electric Company, The City of Riverside and The City'of Anaheim, California (licensees) dated June 27, June 29, and July 18, 1984 comply with the standards and reauirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the applications, as amended, the provisions of the Act, and the regulations of the Commission; C.
There is reasonable assurance:
(i) that the activities authorized
'by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; e411140225 841026 PDR ADOCK 05000361 p
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The issuance of this amenc nent is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. NPF-In is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environ-mental Protection Plan contained in Appendix B, as revised through' Amendment No. 26, are hereby incorporated in the license. SCE'shall operate the facility in accordance with the Technical Specifications
-and the Environmental Protection Plan.
3.
This amendment is effective as of the date of 1ssuance.
FOR THE NUCLEAR REGULATORY COMMISSION canslM" George W. Knighton, Chief Licensing Branch No. 3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
October 26, 1984
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. ATTACHitENT TO LICENSE AMENDMENT NO. 26 FACILITY OPERATING LICENSE-NO. NPF-10 DOCKET NO. 50-361 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Also to be replaced are the following overleaf pages to the amended pages.
I Amendment Pages Overleaf Pages 3/4 4-12 3/4 4-11 3/4 9-6 3/4 9-5 3/4 9-11 3/4 9-12 3/4 9-14 3/4 9-13 B 3/4 4-3 B 3/4 4-4 B 3/4 9-2 B 3/4 9-1 Delete page 3/4 4-15a.
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-REACTOR COOLANT SYSTEMS
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SURVEILLANCE-REQUIREMENTS (Continued) 4.4.4.3 : Inspection Frequencies - The'above required: inservice inspections of steam generator tubes shall be performed at the~following frequencies:
- a.
The first inservice = inspection-shall be perfoceed after 6 Effective Full Power Months but within 24 calender months of initial crit-icality.
Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.
If-two consecutive inspections following service under AVT' conditions, not' including the preservice inspec-tion, result in both sets of inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously.
observed degradation has not continued and no additional 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 4.4-2 at 40 month intervals fall into Category C-3, the inspection frequency shall be increased to at least once per 20 months.
The increase in inspection frequency shall apply until the subsequent inspections satisfy the. criteria of
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Specification 4.4.4.3.a; the interval may then be extended to a maximum of once per 40 months.
c.
Additional', unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection i
specified in Table 4.4-2 during the shutdown subsequent to any of l
the following conditions:
i 1.
Primary-to-secondary tubes leaks (not including leaks originating.
from tube-to-tube sheet welds) in excess of the limits of 4
Specification 3.4.5.2.
1 2.
A seismic occurrence greater than the Operating Basis Earthquake.
3.
A loss-of-coolant accident requiring actuation of the engineered j'
safeguards.
4.
~A main steam line or feedwater line break.
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h SAN ONOFRE-UNIT 2 3/4 4-11
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.4.4 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 specifications.
Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as impgrfections.
2.
Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.
3.
Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation.
4.
% Degradation means the percentage of the tube wall thickness affected or removed by degradation.
5.
Defect means an imperfection of such severity that it exceeds the plugging limit.
A tube containing a defect is defective.
6.
Plugging Limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 44% of the nominal tube wall thickness.
7.
Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.4.3.c, above.
8.
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top suoport of the cold leg.
9.
Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing.
This inspection was performed prior to the field hydrostatic test and prior to initial POWEP. OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
SAN ONOFRE-UNIT 2 3/4 4-12 AMENDMENT NO. 26
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REFUELING OPERATIONS 3/4.9.5 COMMUNICATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct communications shall be maintained between the control room and personnel at the refueling station.
APPLICABILITY:
During CORE ALTERATIONS.
ACTION:
When direct communications between the control room and personnel'at the refueling sta, tion cannot be maintained, suspend all CORE ALTERATIONS.
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SURVEILLANCE REQUIREMENTS t
4.9.5 Direct communications between the control room and personnel at the refueling station shall be demonstrated within one hour prior to the start of and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS.
i SAN ON0FRE-UNIT 2 3/4 9-5
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REFUELING OPERATIONS 3/4.9.6 REFUELING MACHINE LIMITING CONDITION FOR OPERATION
-3.9.6 The refueling machine shall~be used for movement of CEAs* or fuel assemblies and shall be OPERABLE with:
a.
A minimum capacity of 3000 pounds, and b.
An overload cut off limit'of less than or equal to 3350 pounds.
APPLICABILITY:
During movement of CEAs* and/or fuel assemblies within the reactor pressure vessel.
ACTION:
With the requirements for the refueling machine OPERABILITY not satisfied, suspend all refueling machine operations involving the movement of CEAs* and fuel assemblies within the reactor pressure vessel.
SURVEILLANCE REQUIREMENTS 4.9.6 The refueling machine used for movement of CEAs* or fuel assemblies within the reactor pressure vessel shall be demonstrated OPERABLE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to the start of such operations by performing a load test of at least 3000 pounds and demonstrating an automatic load cut off when the refueling machine load exceeds 3350 pounds.
"Except for movement of four finger CEA's, coupling and uncoupling the CEA extension shafts or verifying the coupling and uncoupling.
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. SAN ON0FRE-UNIT 2 3/4 9-6 AMENDMENT NO. 26 f
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REFUELING' OPERATIONS 3/4.9.10 WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.10 At least 23 feet
- of water shall be maintained over the top of the l
reactor pressure vessel flange.
I APPLICABILITY:
During movement of fuel assemblies or CEAs within the reactor pressure vessel when either the fuel assemblies being moved or the fuel assemblies seated within the reactor pressure vessel are irradiated.
ACTION:
With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or CEAs within the pressure vessel.
SURVEILLANCE REQUIREMENTS 4.9.10 The water level shall be determined to be at least its minimum required depth within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to the start of and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter during movement of fuel assemblies or CEAs.
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" Water level may be lowered to a minimum of 23 feet above the top of the fuel for movement of four finger CEA's, coupling and uncoupling of CEA extension shafts or for verifying the coupling and uncoupling.
SAN ON0FRE-UNIT 2 3/4 9-11 AMENDMENT NO. 26 L_
REFUELING OPERATIONS
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3/4.9.11 WATER LEVEL-STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool.
ACTION:
With the requirement of the specification not satisfied, suspend all movement of fuel assemblies and restore the water level to within its limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS l
4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool.
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I SAN ON0i9E-UNIT 2 3/4 9-12 i
REFUELING OPERATIONS 3/4.9.12 FUEL HANDLING BUILDING POST-ACCIDENT CLEANUP FILTER SYSTEM LIMITING CONDITION FOR OPERATION 3.9.12 Two independent fuel handling building post-accident cleanup filter
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systems shall be OPERABLE.
APPLICABILITY: Whenever irradiated fuel is in the storage pool.
ACTION:
a.
With one fuel handling building post-accident cleanup filter system inoperable, fuel movement within the storage pool or operation of fuel handling machine over the storage pool may proceed provided the OPERABLE fuel handling building post-accident cleanup filter system is capable of being powered from an OPERABLE emergency power source.
Restore the inoperable fuel handling building post-accident cleanup filter system to OPERABLE status within 7 days or suspend all opera-tions involving movement of fuel within the storage pool or operation of the fuel handling machine over the storage pool.
b.
With no fuel handling building post-accident cleanup filter system OPERABLE, suspend all operations. involving movement of fuel within
-the storage pool or operation of fuel handling machine over the-storage pool until at least one fuel handling building post-accident cleanup filter system is restored to OPERABLE status.
c.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.12 The above required fuel handling building post-accident cleanup filter systems shall be demonstrated OPERABLE:
a.
At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at-least 10-hours with the heaters on.
b.
At least once per 18 months or (1) after any structural maintenance on the HEPA filter or' charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:
t SAN ONOFRE-UNIT 2 3/4 9-13
I REFUELING OPERATIONS
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SURVEILLANCE REQUIREMENTS (Continued)
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1.
Verifying that with the system operating at a flow rate of 12925 cfm i 10% and recirculating through the HEPA filters and E
charcoal adsorbers, the total bypass flow of the system through the system diverting valves, to the facility vent is less than or equal to 1% when the system is tested by admitting cold D0P 1
p at the system intake.
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2.
Verifying that the cleanup filter system satisfies the in place testing acceptance criteria and uses the test procedures of
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Regulatory Positions C.S.a, C.5.c and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is y
12925 cfm i 10%.
_7 3.
Verifying within 31 days after removal that a laboratory I
analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52, r
Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, b
Revision 2, March 1978.
4.
Verifying a system flow rate of 12925 cfm + 10% during systen operation when tested in accordance with ARSI N510-1975.
c.
E After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a k
representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, treets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.
d.
At least once per 18 months by:
i 1.
Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 7.3 inches
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Water Gauge while operating the system at a flow rate of i
12925 cfm i 10%.
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2.
Verifying that on a Fuel Handling Isolation (FHIS) test signal, g
the system automatically isolates normal ventilation and starts P
recirculation through the HEPA filters and charcoal adsorber L
banks.
3.
Verifying that the heaters dissipate 28.7 1 1.5 kw for E464, b
32.3 1 1.7 kw for E465, and 3.8 1 0.2 kw for E652 when tested in accordance with ANSI N510-1975 with the measured heater h
dissipation corrected to correspond to nominal voltage.
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SAN ON0FRE-UNIT 2 3/4 9-14 AMENDMENT NO. 26
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REACTOR COOLANT SYSTEM BASES
- STEAM GENERATORS (Continued)
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.
The plant'is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in.
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negligible corrosion of the steam generator tubes.
If the secondary coolant chemistry is not maintained within these limits, localized corrosion may-
.likely result in stress corrosion cracking.
The extent of cracking during l
plant operation would be limited by the limitation of steam generator tube.
leakage between the primary coolant' system and the secondary coolant system (primary-to secondary. leakage = 0.5 GPM per steam generator).
Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the' loads imposed during normal operation and by postulated accidents.
Operating plants have demonstrated that primary-to-secondary leakage of 0.5 GPM per steam generator can readily l-be detected by radiation monitors of steam generator blowdown.
Leakage in i
excess of this limit will require plant shutdown and an unscheduled inspection,-
j during which the leaking tubes will be located and plugged.
l' Wastage-type defects are unlikely with proper chemistry treatment of the i
secondary coolant.
However, even if a defect should~ develop in service, it J
will be found during scheduled inservice steam generator tube examinations.
Plugging will be required for all tubes with imperfections exceeding 44% of i
the nominal tube wall thickness.
This criteria was developed as a result of analyses performed for Primary Loop Pipe Break (PLPB) plus Design Basis Earth-quake (DBE) and Main Steam Line Break (MSLB) plus DBE.
These analyses examined families of tube rows as related to the number of vertical support grids. As horizontal tube spans become longer the loads become greater in spite of the increased number of tube supports.
Steam' generator tube inspections of operating i
plants have demonstrated the capability to reliably detect degradation that has penetrated ~20% of the original tube wall thickness.
I Whenever the results of any stean generator tubing inservice inspection i
fall into Category C-3, these results M il be promptly reported to the Commission pursuant to Specification 6.9.1 prior the resumption of plant operation.
Such cases will be considered by the Commission ~on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
SAN ONOFRE-UNIT 2 B 3/4 4-3 AMENDMENT NO. 26
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BASES 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.5.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Ldakage Detection Systems," May 1973.
3/4.4.5.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM.
This threshold value is sufficiently low to ensure early detection of additional leakage.
The 10 GPM IDENTIFIED LEAKAGE limitation provides allowances for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.
The surveillance requirements for RCS Pressure Isolation Valves provide
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added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.
Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowable limit.
The total steam generator tube leakage limit of 1 GPM for all steam generators ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break.
The 1 GPM limit is consistent with the assumptions used in the analysis of these accidents.
The 0.5 GPM leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.
PRESSURE B0UNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.
Therefore, the presence of any PRESSURE B0UNDARY LEAKAGE requ*res the unit to be promptly placed in COLD SHUTDOWN.
3/4.4.6 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion.
Maintaining k
SAN ONOFRE-UNIT 2 8 3/4 4-4 i
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3/4.9 REFUELING OPERATIONS
-BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:
- 1) the' reactor will remain subcritical during CORE ALTERATIONS, and
- 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel.
These limitations are consistent with the initial conditions assumed for the boron dilution incident in the accident analyses.
The value of 0.95 or less for K includes a 1% delta K/K conservative allowance for uncertainties.
Similarly*f[heboron concentration value of 1720 ppm or greater also includes a conservative uncertainty allowance of 50 ppm boron.
3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.
3/4.9.3 DECAY' TIME Tha minimum requirement for reactor subcriticality prior to movement of irradiated fuel. assemblies.in the reactor pressure vessel ensures that suf"icient time has elapsed to allow the radioactive decay of the short lived fis icn products.
This decay time is consistent with the assumptions used in the accident analyses.
3/4.9.4 CONTAINMENT PENETRATIONS The requirements on containment penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment.
The OPERABILITY and closure restrictions are sufficient to restrict. radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.
i 3/4.9.5 COMMUNICATIONS r
The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during CORE ALTERATIONS.
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REFUELING OPERATIONS BASES 3/4.9.6 REFUELING MACHINE The OPERABILITY requirements for the refueling machine ensure that:
(1) the refueling machine will be used for movement of all fuel assemblies including those with a CEA inserted, (2) each machine has sufficient load capacity to lift a fuel assembly including those with a CEA, and (3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
With the exception of the four finger CEA's, CEA's are removed from the reactor vessel along with the fuel bundle in which they are inserted utilizing the refueling machine.
The four finger CEA's are inserted through the upper guide structure with two fingers in each of the two adjacent fuel bundles in the periphery of the core. The four finger CEA's are either removed with the upper guide structure and lift rig or can be removed with separate tooling prior to upper guide structure removal utilizing the auxiliary hoist of the polar crane.
Coupling and uncoupling of the CEA's and the CEE, drive shaft extensions is accomplished using the gripper operating tool.
The coupling and uncoupling is verified by weighing the drive shaft extensions.
3/4.9.7 FUEL HANDLING MACHINE - SPENT FUEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel assembly, CEA and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly and (2) any post,wle distortion of fuel in the storage racks wili not result in a critical array.
This assumption is consistent with the activity release assumed in the accident analyses.
3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION The requirement that at least one shutdown cooling train be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140*F as required during the REFUELING MODE, and (2) sufficient coolant circulation is main-tained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification.
The requirement to have two shutdown cooling trains OPERABLE when there is less thc 23 feet of water above the reactor pressure vessel flange, ensures that a single failure of the operating shutdown cooling loop will not result i
in a complete loss of decay heat removal capability.
With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a i
l large heat sink is available for core cooling, thus in the event of a failure j
i of the operating shutdown cooling train, adequate time is provided to initiate emergency procedures to cool the core.
i SAN ONOFRE-UNIT 2 B 3/4 9-2 AMENDMENT NO. 26 L
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SOUTHERN CALIFORNIA EDISON COMPANY n._
. SAN DIEGO GAS AND ELECTRIC COMPANY
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.THE CITY 0F RIVERSIDE,: CALIFORNIA.
cTHE CITY OF ANAHEIM, CALIFORNIA a.
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2 DOCKET NO. 50-362?
SAN ON0FRE NUCLEAR GENERATING STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 15 License No.'NPF-15 1.
The Nuclear Regulatory Comission'(the.Comission) has~ found that:
AL The' applications for amendment to the license for San-Onofre Nuclear Generating Station, Unit 3 (the facility) filed by the Southern California Edison Company on behalf'of itself and: San Diego Gas and Electric Com California (pany, The City of Riverside and The City of Anaheim, licensees) dated June 27, June 29, and July 18..-1984 comply with the standards and requirements of the Atomic Energy Act:
of 1954, as amended (the Act) and the Comission's regulations as set forth in 10 CFR Chapter.I; B.
The facility will operate in conformity with the applications, as amended, the provisions of the'Act,'and the 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 Comission's regulations set forth in 10 CFR Chapter I; D;
The issuance of this license amendment will not be inimical to the common. defense and security or to the health and safety of the public; t
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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.
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2.
'Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this' license amendment, and paragraph 2.C(2) of Facility Operating License No. NPF-15 is hereby amended to read as follows:
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(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environ-mental Protection Plan contained in Appendix B, as revised through Amendment No. 15, are hereby incorporated in the license.
SCE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3. -
This amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION George W. Knighton, Chief Licensing Branch No. 3 Division of. Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: October 26, 1984 7
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... ATTACHMENT TO LICENSE' AMENDMENT NO. 15' FACILITY OPERATING LICENSE NO. NPF-15 DOCKET NO. 50-362 Replace the.following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Also to be replaced are
_the following overleaf pages to the amended pages.
Amendment Pages Overleaf Pages 3/4 4-12 3/4 4-11 3/4 9-6 3/4 9-5 3/4 9-11 3/4 9-12 3/4 9-14 3/4 9-13 8 3/4 4-3 B 3/4 4-4 B 3/4 9-2 B 3/4 9-1 Delete page 3/4 4-16.
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REACTOR COOLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.4.4.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
a.
The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calender months of initial crit-icality.
Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.
If two consecutive inspections following service under AVT conditions, not including the preservice inspec-tion, result in both sets of inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once par 40 months.
b.
If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40 month intervals fall into Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specificatio,4.4.4.3.a; the interval may then be extended to a maximum of once per 40 months.
c.
Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
1.
Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.5.2.
2.
A seismic occurrence greater than the Operating Basis Earthquake.
3.
A loss-of-coolant accident requiring actuation of the engineered safeguards.
4.
A main steam line or feedwater line break.
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l SAN ONOFRE-UNIT 3 3/4 4-11 t
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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.4.4 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 specifications.
Eddy-current testing indications below 20% of the nominal tube 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.
3.
Degraded Tube means a tube containing imperfections greater than or equai to 20% of the nominal wall thickness caused by degradation.
4.
% Degradation means the percentage of the tube wall thickness affected or removed by degradation.
5.
Defect means an imperfection of such severity that it exceeds the plugging limit.
A tube containing a defect is defective.
6.
Plugging Limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 44% of the nominal tube wall thickness.
7.
Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.4.3.c, above.
8.
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.
9.
Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing.
This inspection was performed prior to the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
SAN ONOFRE-UNIT 3 3/4 4-12 AMENDMENT NO. 15
cf REFUELING OPERATIONS
\\
3/4.9.5 COMMUNICATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct communications shall be maintained between the control room and personnel at the refueling station.
APPLICABILITY:
During CORE ALTERATIONS.
ACTION:
When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS.
SURVEILLANCE REQUIREMENTS 4.9.5 Direct communications between the control room and personnel at the refueling station shall be demonstrated within one hour prior to the start of and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS.
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SAN ONOFRE-UNIT 3 3/4 9-5
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REFUELING OPERATIONS:
3/4.9.6' REFUELING MACHINE
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t LIMITING CONDITION FOR OPERATION J
3.9.6'.The refueling machine shall.be used for' movement of CEAs* or fuel assemblies and shall be OPERABLE with:
a.
A minimum capacity of 3000 pounds, and b.
An overload cut off limit of less than or equal to 3350 pounds.
APPLICABILITY:- During movement of CEAs* and/or fuel assemblies within the reactor pressure vessel.
ACTION:
With the requirements for the refueling machine OPERABILITY not satisfied, suspend all refueling machine operations involving the movement of CEAs* and fuel assemblies within the reactor pressure vessel.
SURVEILLANCE REQUIREMENTS 4.9.6 The refueling machine used for movement of CEAs* or fuel assemblies within the reactor pressure vessel shall be demonstrated OPERABLE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to the start of such operations by performing a load test of at least 3000 pounds and demonstrating an automatic load cut off when the refueling machine load exceeds 3350 pounds.
"Except for movement of four finger CEA's, coupling and uncoupling the CEA extension shafts or verifying the coupling and uncoupling.
SAN ONOFRE-UNIT 3-3/4 9-6 AMENDMENT NO.15
REFUELING OPERATIONS' 3/4.9.10 ~ WATER LEVEL - REACTOR VESSEL ~
LIMITING CONDITION FOR OPERATION-3.9.10k.Atleast'23 feet *ofwater shall be maintained over the top of.the
-l reactor pressure vessel flange.-
APPLICABILITY:
During movement of fuel assemblies or CEAs within the reactor pressure vessel when either the fuel assemblies being moved or the fuel assemblies seated within the reactor pressure vessel are. irradiated.
' ACTION:
With the requirements of theLabove specification not satisfied, suspend all
^
operations involving movement of fuel assemblies or CEAs within the pressure i
vessel.
4 4
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SURVEILLANCE REQUIREMENTS 4.9.10 The water leue'. shall~ be determined to be at -least its minimum required depth within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to the start of and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter during movement of fuel assembliesior CEAs.
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" Water level may be lowered to a minimum of 23 feet above the' top of the fuel d
t for movement of four finger CEA's,. coupling and uncoupling of CEA extension
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shafts or for verifying the coupling or uncoupling.
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' SAN ONOFRE-UNIT 3 3/4 9-11
' AMENDMENT N0.15 e
y
. REFUELING OPERATIONS 3/4.9.11 WATER LEVEL - STORAGE POOL
(
9 LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in thc storage racks.
APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool.
ACTION:
With the requirement of the specification not satisfied, suspend all movement of fuel assemblies and restore the water level to within its limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS f
4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool.
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SAN ONOFRE-UNIT 3 3/4 9-12 L
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' REFUELING OPERATIONS-
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' 3/4.9'.12 FUEL ~ FANDLING BUILDING POST-ACCIDENT CLL%NUP FILTER SYSTEM LIMITING CONDli10N FOR OPERATION w
-3.9.12 Two independent. fuel handling building post-accident cleanup filter
. systems shall be OPERABLE.,
APPLICABILITY: Whenever irradiated fuel is in the storage pool.
~ ACTION:
With one fuel handling building post-accident cleanup filter system a.
inoperable, fuel movement within the storage pool or operation of 1
fuel handling machine over the storage pool may proceed provided the OPERABLE fuel handling building post-accident cleanup filter system is~ capable of being powered from an OPERABLE ~ emergency power source.
Restore the inoperable fuel-handling building post-accident cleanup filter system to~0PERABLE status within 7 days or suspend all opera-tions involving movement of fuel within the storage pool-or operation of the fuel handling machine over the storage pool.
1 s
b.
With no fuel. handling building post-accident cleanup filter system OPERABLE, suspend all 07erations involving movement of fuel within-the storage pool or operation of-fuel handling machine over the storage pool untii at least one fuel handling-building post-accident cleanup filter system is restored to OPERABLE-status.
'l c.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE-REQUIREMENTS i
4.9.12 The above required fuel handling building post-accident cleanup filter systems shall be demonstrated OPERABLE:
At least once per 31 days on a STAGGERED TEST < BASIS by initiating, a.
l from the control room, flow through the HEPA filters and charcoal l
adsorbers and verifying.that the system operates for at least~10 hours with the heaters on.
b.
At least once per 18 months or (1) af ter any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone 7~
J communicating with-the syste,n by:
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l-SAN ON0FRE-UNIT 3 3/4 9-13 l
= -
s
. REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) 1.
. Verifying.that with the system operating at a flow rate of
~12925 cfm + 10% and recirculating through the HEPA filters and charcoal adsorbers, the total bypass flow of the system through
.the system diverting valves,:to the facility vent is less than or equal to 1% when the system is tested by admitting cold DOP at the system intake.
2.
Verifying that the cleanup filter system _ satisfies the in place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a,. C.5.c and C.S.d of Regulatory Guide 1.52,;Revit, ion 2, March 1978, and the system flow rate is 12925 cfm + 10%.
3.
Verifying within 31 days after removal that a laboratory
+
analysis of a representative. carbon sample obtained in accor.
dance with Regulatory Position' C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria:
of Regulatory Position C.6.a of Regulatory Guide 1.52, i
Revision 2, March 1978.
4.
Verifying a system flow rate of 12925 cfm + 10% during system operation when tested in accordance with ANST N510-1975.
After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying 4
c.
within 31 days after removal that a laboratory analysis.of a representative carbon sample.obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a l
of Regulatory Guide 1.52, Revision 2, March 1978.
d.
At least once per 18 months by:
1.
Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 7.3 inches Water Gauge while operating the system at a flow rate of 12925 cfm + 10%.
2.
Verifying that on a Fuel Handling Isolation (FHIS) test signal, the system automatically isolates normal ventilation and starts recirculation through the HEPA filters and charcoal adsorber banks.
3.
Verifying that the heaters dissipate 28.7-t 1.5 kw for E464, 32.3 i 1.7 kw.for E465, and 3.8 i 0.2_kw for E652 when tested i
in a':cordance with ANSI N510-1975 with the measured heater dissipation corrected to correspond to nominal voltage.
SAN ON0FRE-UNIT 3 3/4 9-14 AMENDMENT NO. 15 i
REACTOR COOLANT-SYSTEM BASES STEAM GENERATORS (Continued)
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.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in i
negligible corrosion of the steam generator tubes.
If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.
The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to secondary leakage = 0.5 GPM per steam generator).
Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.
Operating plants have demonstrated.
that primary-to secondary leakage of 0.5 GPM per steam generator can readily be detected by radiation monitors of stear generator blowdown.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.
However, even if a defect should develop in service,-it will be found during scheduled inservice steam generator tube examinations.
Plugging will be required for all tubes with imperfections exceeding 44% of the nominal tube wall thickness. This criteria was developed as a result of analyses performed for Primary Loop Pipe Break (PLPB) plus Design Basis Earth-quake (DBE) and Main Steam Line Break (MSLB) plus DBE.
These analyses examined families of tubt rows as related to the number of vertical support grids.
As horizontal tube spans become longer the loads become greater in spite of the increased number of tube supports.
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 into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.9.1 prior the resumption of plant operation.
Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
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SAN ONOFRE-UNIT 3 B 3/4 4-3 AMENDMENT N0. 15
REACTOR COOLANT SYSTEM BASES 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.5.1 LEAKAGE DETECTION SYSTEMS i
The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45. " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
3/4.4.5.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of aMitional leakage.
1 The 10 GPM IDENTIFIED LEAKAGE limitation provides allowances for a limited amount of leakage from known sources whose presence will not interfere with'the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.
The surveillance requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross
(
valve failure and consequent intersystem LOCA.
Leakage from the RCS Pressure Isolation Valves is IDENTIEIED LEAKAGE and will be considered as a portion of the allowable limit.
The total steam generator tube leakage limit of 1 GPM for all steam generators ensures that the dosage contribution from the tube leakage vill be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break.
The 1 GPM limit, is consistent with the assumptions used in the analysis of these accidents.
The 0.5 GPM leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.
PRESSURE BOUNDARY' LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.
Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly i
placed in COLD SHUTDOWN.
l 3/4.4.6 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion.
Maintaining k
SAN ONOFRE-UNIT 3 B 3/4 4-4
3/4.9 REFUELING OPERATIONS e
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BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure
- 1) the reactor will remain subcritical during CORE ALTERATIONS, and that:
- 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel.
These limitations are consistent with the initial conditions assumed for the boron' dilution includes incident in the accident analyses. The value of 0.95 or less for K*f[he boron a 1% delta K/K conservative allowance for uncertainties.
Similarly concentration value of 1720 ppm or greater also includes a conservative uncertainty allowance of 50 ppm boron.
3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.
3/4.9.3 DECAY TIME i
The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived This decay time is consistent with the assumptions used in fission products.
the 9ccident analyses.
3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment.
The OPERABILITY and closure-restrictions are sufficient to restrict radioactive material r.elease from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.
f 3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during CORE ALTERATIONS.
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4 SAN ONOFRE-UNIT 3 8 3/4 9-1
er
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REFUELING OPERATIONS -
8ASES-
'3/4.9.6'REFUELINGMACHINE The OPERABILITY requirements for the refueling machine ensure that:
(1) the. refueling machine will be used for movement of all fuel assemblies including those with a CEA inserted, (2) each machine has ~ sufficient load 4 '
capacity to lift a. fuel assembly including those with a CEA, and (3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
With the exception'of the four finger CEA's,- CEA's are removed from the reactor vessel along with the fuel bundle in which they are. inserted utilizing 4
the refueling ~ machine.
The four finger CEA's are inserted through the upperc guide structure with two fingers in each of.two adjacent fuel bundles in the-periphery of the core.
The four finger CEA's are either removed with the upper j -
guide structure and lift rig or can be removed with separate tooling prior to
~
upper guide structure removal utilizing the auxiliary hoist of the polar crane.
Coupling and uncoupling of the CEA's and the CEDM drive shaft extensions is accomplished using the gripper operating tool.
The coupling and uncoupling is verifled by weighing the drive shaft extensions.
i 3/4.9.7 FUEL HANDLING MACHINE - SPENT FUEL STORAGE BUILDING j
i The restriction on movement of loads in excess of the nominal weight of a i
fuel assembly, CEA and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped (1) the 4
activity release will be limited to that contained in a single fuel assembly and (2) any possible distortion of fuel in the storage racks will not result in a critical array.
This assumption is consistent with the activity release assumed in the accident analyses.
3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION The requirement that at least one shutdown cooling train be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water.in the reactor pressure vessel below 140*F as required during the REFUELING MODE, and (2) sufficient coolant circulation is main-4 tained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification.-
i The requirement to have two shutdown cooling trains OPERABLE when there is less than 23 feet of water above the reactor pressure vessel flange ensures that a single failure of the operating shutdown cooling loop will not result in a complete loss of decay heat removal capability. With the reactor vessel i
head removed and 23 feet of water above the reactor pressure vessel flange, a L
large heat sink is available for core cooling, thus in the event of a failure of the operating shutdown cooling train, adequate time is provided to initiate emergency procedures to cool the core.
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SAN ONOFRE-UNIT 3 B 3/4 9-2 AMEN 0 MENT NO. 15 LE
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