ML20199L709
| ML20199L709 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 06/25/1986 |
| From: | Rood H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20199L713 | List: |
| References | |
| GL-85-19, GL-85-19-39, TAC-60724, TAC-60725, NUDOCS 8607090569 | |
| Download: ML20199L709 (16) | |
Text
'f
'o UNITED STATES
+
E
'g NUCLEAR REGULATORY COMMISSION n
{'
WASHINGTON,0. C. 20555
%,...../
SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHEIM, CALIFORNIA 00CKET N0. 50-361 SAN ON0FRE NUCLEAR GENERATING STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 50 License No. NPF-10 1.
The huclear Regulatory Commission (the Commission) has found that:
A.
The applicaticn 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 February 7,1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations as set forth in 10 CFR Chapter I; 6.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Commission; C.
There is reasonable assurance:
(i) that the actisities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activ: ties 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; E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
8607090569 860625 PDR ADOCK 05000361 P
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this amendment and Paragraph 2.C(2) of Facility Operating License No. NPF-10 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 50, are hereby incorporated in the license.
SCE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The changes in the Technical Specifications are to become effective within 30 days of issuance of the amendment.
In the period between issuance of the amendment and the effective date of the new Technical Specifications, the licensees shall adhere to the Technical Specifications existing at the time. The period of time during changeover shall be minimized.
4.
This license amendment is effective as the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Harry Rood, Senior Project Manager PWR Project Directorate No. 7 Division of PWR Licensing-E
Attachment:
Changes to the Technical Specifications Date of Issuance: June 25, 1986
June 25, 1986 ATTACHMENT TO LICENSE AMENDMENT N0. 50 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.
Amendment Page Overleaf Page 3/4 4-23 3/4 4-24 B 3/4 4-5 B 3/4 4-6 6-16a
s REACTOR COOLANT SYSTEM 3/4.4.7 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.7 The specific activity of the primary coolant shall be limited to:
a.
Less than or equal to 1.0 microcurie / gram DOSE EQUIVALENT I-131, and b.
Less than or equal to 100/E microcuries/ gram.
APPLICABILITY:
MODES 1, 2, 3, 4 and 5.
ACTION:
a.
With the specific activity of the primary coolant greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T,yg less than 500 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
The provisions of Specification 3.0.4 are not applicable, b.
With the specific activity of the primary coolant greater than 100/E microcurie / gram, be in at least HOT STANDBY with T less avg than 500 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.7 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.
SAN ON0FRE - UNIT 2 3/4 4-23 AMENDMENT N0. 50
9 9*
e e
THIS PAGE INTENTIONALLY DELETED.
i I
l i
SAN ON0FRE - UNIT 2 3/4 4-24 AMENDMENT NO. 50 i
s REACTOR COOLANT SYSTEM BASES REACTOR COOLANT SYSTEM BASES CHEMISTRY (Continued) the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant.
The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent.
Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.
The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.
The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.
3/4.4.7 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appro-priately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM and a concurrent loss of offsite electrical power.
The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations.
These values are conservative in that specific site parameters of the San Onofre site, such as site boundary location and meteorological conditions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.
SAN ONOFRE-UNIT 2 B 3/4 4-5 AMEN 0 MENT NO. 50
- = _ -
BASES i
SPECIFIC ACTIVITY (Continued)
Reducing T to less than 500*F prevents the release of activity should asteamgenerat@9 tube rupture since the saturation pressure of the primary i
coolant is below the lift pressure of the atmospheric steam relief valves.
The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.
Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena.
A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
I 3/4.4.8 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand j
the effects of cyclic loads due to system temperature and pressure changes.
These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.
The various categories of load cycles l
used for design purposes are provided in Section 3.9.1.1 of the FSAR.
During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
l During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall.
These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.
Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.
The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location.
The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel.
These stresses are additive to the pressure induced tensile stresses which are already present.
The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.
Consequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.
1 I
l SAN ONOFRE-UNIT 2 B 3/4 4.6 1
ADMINISTRATIVE CONTROLS (6.9.1.5 Cont'd)
Reports required on an annual basis shall include the results of specific activity analysis in which the primary coolant exceeded the limits of Specifi-cation 3.4.7.
The following information shall be included in these reports:
(1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radio-iodine performed prior to exceeding the limit, results of analysis while limit was exceed and results of one analysis after the radioiodine activity was re-duced to less than limit.
Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting i
48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph i
of the I-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function time for the duration of the specific activity above the steady-state level; and (5) the time duration when the specific activity of the primary coolant exceeded the radioiodine limit.
SAN ON0FRE-UNIT 2 6-16a AMENDMENT NO. 50
8
'o UNITED STATES
-[ '
N NUCLE AR REGULATORY COMMISSION 3
- p W ASHINGTON, D. C. 20555
\\..../
SOUTHERN CALIFORNIA EDIS0N COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHEIM, CALIFORNIA DOCKET NO. 50-362 SAN ON0FRE NUCLEAR GENERATING STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 39 License No. NPF-15 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application 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 Company, The City of Riverside and The City of Anaheim, California (licensees) dated February 7, 1986, complies with the standards and requirements 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 application, 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; E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this amendment and Paragraph 2.C(2) of Facility Operating License No. NPF-15 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 39, are hereby incorporated in the license.
SCE shall operate the facility in r.ccordance with the Technical Specifications and the Environmental Protection Plan.
3.
The changes in the Technical Specifications are to become effective within 30 days of issuance of the amendment.
In the period between issuance of the amendment and the effective date of the new Technical Specifications, the licensees shall adhere to the Technical Specifications existing at the time. The period of time during changeover shall be minimized.
4.
This license amendment is effective as the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Harry Road, Senior Project Manager PWR Project Directorate No. 7 Division of PWR Licensing-B
Attachment:
Changes to the Technical Specifications Date of Issuance:
June 25, 1986
June 25, 1986
. ATTACHMENT TO LICENSE AMENDMENT NO. 39 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 ta the amended pages.
Amendment Page Overleaf Page 3/4 4-24 3/4 4-23 3/4 4-25 3/4 4-26 B 3/4 4-5 B 3/4'4-6 6-17a
't
^
3/4.4.7 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.7 The specific activity of the primary coolant shall be limited to:
a.
Less than or equal to 1.0 microcurie / gram DOSE EQUIVALENT I-131, and b.
Less than or equal to 100/E microcuries/ gram.
APPLICABILITY:
MODES 1, 2, 3, 4 and 5.
ACTION:
a.
With the specific activity of the primary coolant greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T less avg than 500 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
The provisicas of Specification 3.0.4 are not applicable.
b.
With the specific activity of the primary coolant greater than 100/E microcuries/ gram, be in at least HOT STANDBY with T less than 500 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
avg SURVEILLANCE REQUIREMENTS 4.4.7 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.
SAN ONOFRE - UNIT 3 3/4 4-24 AMENDMENT N0. 39
4 0
9 9
l 1
THIS PAGE INTENTIONALLY DELETED SAN ONOFRE - UNIT 3 3/4 4-25 AMEN 0 MENT NO. 39
~
REACTOR COOLANT SYSTEM BASES CHEMISTRY (Continued) the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant.
The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent.
Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.
The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.
The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.
3/4.4.7 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2-hour doses at the site boundary will not exceed an appro-priately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM and a concurrent loss of offsite electrical power.
The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the San Onofre site, such as site boundary location and meteorological conditions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.
i SAN ON0FRE-UNIT 3 B 3/4 4-5 AMENDMENT NO. 39
. _ = _ _ -
BASES SPECIFIC ACTIVITY (Continued)
Reducing T to less than 500*F prevents the release of activity should asteamgenerat8f9 tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.
The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.
Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena.
A j
reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
1 3/4.4.8 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand trie eff ects of cyclic loads due to system temperature and pressure changes.
These cyclic loads are introduced by normal load transients,' reactor trips, and stattup and shutdown operations.
The various categories of load cycles l
used for design purposes are provided in Section 3.9.1.1 of the FSAR.
During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
l l During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall.
These thermally induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.
Therefore, a pressure-temperature curve based on steady-state conditions (i.e., no thermal stresses) i represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.
The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location.
The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive l ~
to the pressure induced tensile stresses which are already present.
The thermally induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.
Consequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.
k SAN ONOFRE-UNIT 3 8 3/4 4-6
ADMINISTRATIVE CONTROLS (6.9.1.5 Cont'd)
Reports required on an annual basis shall include the results of specific activity analysis in which the primary coolant exceeded the limits of Speci-fication 3.4.7.
The following information shall be included in these reports:
(1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first r :ple in which 1
the limit was exceeded; (2) Results of the last isotopic analysis for radio-iodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than limit.
Each result should include date and time of sampl-ing and the radioiodine concentrations; (3) Cleanup system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the spe-cific activity of the primary coolant exceeded the radioiodine limit.
4 i
l SAN ONOFRE-UNIT 3 6-17a AMENDMENT NO. 39
-r-
.