ML17313B049
| ML17313B049 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 08/05/1999 |
| From: | Stephen Dembek NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17313B050 | List: |
| References | |
| NPR-41-A-120, NUDOCS 9908120126 | |
| Download: ML17313B049 (92) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&4001 ARIZONAPUBLIC SERVICE COMPANY ET AL.
DOCKET NO. STN 50-528 PALO VERDE NUCLEAR GENERATING STATION UNIT 1 AMENDMENTTO FACILITYOPERATING LICENSE Amendment No. 120 License No. NPF-41 1.
The Nuclear Regulatory Commission (the Commission) has found that:
The application for amendment by the Arizona Public Service Company (APS or the licensee) on behalf of itself and the Salt River Project Agricultural Improvement and Power District, EI Paso Electric Company, Southern California Edison Company, Public Service Company of New Mexico, Los Angeles Department of Water and Power, and Southern California Public Power Authority dated May 23, 1997, as supplemented September 27, 1998, and May 26, '1999, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's'regulations set forth in 10 CFR Chapter I; B.
The facilitywilloperate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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 willbe conducted in compliance with the Commission's regulations; D.
The issuance of this amendment willnot be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
Accordingly, the. license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. NPF-41 is hereby amended to read. as follows:
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(2)
Technical S eciTications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 120, and the Environmental. Protection'Plan contained in Appendix 8, are hereby incorporated into this:license.
APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
This license amendment is effective as of the date of issuance, and shall be implemented within 45 days of the date of issuance.
FOR THE NUCLEAR REGULATORYCOMMISSION Step en Dembek, Chief, Section 2 Project Directorate IV 8 Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&4005 ARIZONAPUBLIC SERVICE COMPANY ET AL.
DOCKET NO. STN 50-529 PALO VERDE NUCLEAR GENERATING STATION UNIT2 AMENDMENTTO FACILITYOPERATING LICENSE Amendment No. 120 License No. NPF-51 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Arizona Public Service Company (APS or the licensee) on, behalf of itself and the. Salt River Project Agricultural'mprovement and Power District, El Paso Electric Company, Southern California Edison Company, Public Service Company of New Mexico, Los Angeles Department of Water and Power, and.Southern California Public Power, Authority dated May 23, 1997, as supplemented September 27, 1998, and May 26, 1999, complies with the. standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B.
The facilitywilloperate in conformity with the application, the provisions of the Act, and the. rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities willbe conducted in compliance with the Commission's regulations; D.
The issuance of this amendment willnot be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. NPF-51 is hereby amended to read as follows
0 0
1 (2)
Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 120, and the Environmental;Protection-Plan contained in Appendix B, are hereby incorporated into this, license.
APS,shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in.specific'license conditions.
This license amendment is effective as of the date of issuance, and shall be implemented within 45 days of the date of issuance.
FOR THE NUCLEAR REGULATORYCOMMISSION StepHen Dembek, Chief, Section 2 Project Directorate IV 8 Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
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UNITED STATES NUCL'EAR REGULATORY COMMISSION WASHINGTON, D.C.
2055&4001'RIZONA PUBLIC SERVICE COMPANY ET AL.
DOCKET NO. STN 50-530 PALO VERDE NUCLEAR GENERATING STATION'NIT3 AMENDMENTTO FACILITYOPERATING LICENSE Amendment No.120 License No. NP.F-74 1.
The Nuclear Regulatory Commission (the Commission) has found that:
The application for amendment by the Arizona'Public Service Company (APS.or the licensee) on behalf of itself and the Salt River Project Agricultural Improvement and Power District, EI Paso Electric Company,.Southern California Edison Company, Public Service Company of New Mexico, L'os Angeles Department of Water and Power, and Southern California Public Power Authority dated May 23, 1997, as supplemented September 27, 1998, and May 26, 1999, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B.
The facilitywill.operate in conformity with the application, the provisions of the Act, and-the rules and regulations of the Commission; 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 willbe conducted in compliance with the Commission's regulations; D.
The issuance of this amendment willnot be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this-amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. NPF-74 is hereby amended,to read as follows:
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(2)'echnical S ecifications and.Environmental*Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 120, and'the Environmental. Protection Plan contained in Appendix 8, are hereby incorporated into this license.
APS shall operate the facility in accordance with the Technical Specifications and.the Environmental Protection Plan, except where otherwise stated in specific'license conditions.
This license amendment is effective, as of the:date-of, issuance,.and shall be implemented within 45 days of the date of issuance.
FOR THE NUCLEAR,REGULATORYCOMMISSION Stephen Dembek, Chief, Section.2 Project Directorate IV 8 Decommissioning Division of Licensing Project Management Office of Nuclear'Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
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1999
ATTACHMENTTO'LICENSE AMENDMENTNO. 120 FACILITYOPERATING L'ICENSE NOS. NPF.-41 NPF-51: AND NPF-74 DOCKET NOS. STN50-528 STN50-529 AND STN50-530 Replace the'following pages of the Appendix A Technical Specifications with the attached revised pages.
The revised pages are identified by amendment number and contain marginal lines indicating the areas of.change.
- REMOVE, 3.4.14-1 INSERT.
'3.4.14-1 5.0-1 5.0-.2'.'0-3 5.0-4 5.0-5 5;0-6 5.0-7 5.0-8 5.0-9 5.0-10 5.0-11 5'.0-12 5.0-13 5.0-14 5.0-15 5.0-16 5.0-17 5.0-18 5.0-1 9 5;0-20 5.0-21 5.0-22 5.0-23 5.0-24 5;0-'25 5.0-26 5.0-'27 5;0-28 5.0-29 5.0-'30 5.0-31 5.0-32 5.0-33 5.1-1.'.2-1 5;2-2
,5.2-3 5.3-1
.5.4-1 5.5-1 5.5-2 5.5-3 5.'5-4 5.5-5 5.5-6
'5.5-7 5.5-8 5.5-9 5.5-10 5;5-11 5.5-12 5.5'-13 5.5-14'.5-15 5.5-16 5.5-17
,5.5-18
'5;5-19 5;5-20 5.5-21 5;5-'22 5.5-23 5;5-24 5.6-1 5.6-2 5.6-3 5.6-4 5.6-5 5.6-6 5.7-1 5.7-2
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t RCS Operational LEAKAGE 3.4.14 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.14 RCS Operational LEAKAGE LCO 3.4. 14 RCS operational LEAKAGE shall be limited to:
a.
No pressure boundary LEAKAGE:
b.
1 gpm unidentified LEAKAGE:
c.
10 gpm identified LEAKAGE; and d.
150 gallons per day primary to secondary LEAKAGE through any one SG.
APPLICABILITY:
MODES 1, 2, 3.
and 4.
ACTIONS CONDITION
'REQUIRED ACTION COMPLETION. TIME A.
RCS LEAKAGE not within A.l Reduce LEAKAGE to limits for reasons within limits.
other than pressure boundary LEAKAGE.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B.
Required Action and associated Completion Time of Condition A not met.
OR Pressure boundary LEAKAGE exists.
B.l Be in MODE 3.
AND B.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
.36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C.
One or more SGs inoperable.
C. 1 Enter LCO 3.0.3.
Iomediately PALO VERDE UNITS 1.2.3 3.4.14-1 AMENDMENT NO. 4P 120
IS
Responsibi 1 ity 5.1
'. 0 ADMINISTRATIVECONTROLS 5.1 Responsibi-lity 5.1.1 5.1.2 The Department
- Leader, Operations shall be responsible for overall unit.operation and shall delegate in writing the succession to this responsibility during his absence.
The Department Leader.
Operations or his'esignee shall
- approve, prior to implementation, each proposed test.
experiment or modification to systems or equipment that affect nuclear safety.
The Control Room Supervisor (CRS) shall be responsible for the control.room command function.
During any absence of the CRS from the control room while the unit is in 'MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator (SRO) license shall.be designated to assume the control room command function.
During any absence of the CRS from the control room while the unit is in MODE 5 or 6.
an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.
PALO VERDE UNITS 1.2,3
- 5. 1-1 AMENDMENT NO; ~120
Organization 5.2
'. 0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2 '
Onsite and Offsite Or anizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively.
The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.
a.
Lines of authority, responsibi lity, and communication shall be defined and established throughout highest management
- levels, intermediate levels, and all operating organization positions.
These relationships shall be documented and
- updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships.
and job descriptions for key personnel positions, or in equivalent forms of documentation.
These requirements shall be documented in the UFSAR; b.
The Vice President, Nuclear Production shall be responsible.
for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant; c.
The Senior Vice President, Nuclear shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety:
and d.
The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager;
- however, these individuals shall have sufficient organizational freedom.to ensure their independence from operating pressures.
5.2.2 Unit Staff The unit staff organization shall include the following:
a.
A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator (continued)
PALO VERDE UNITS 1',2,3 5.2-1 AMENDMENT NO. 447-120
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Organization 5.2
'.2 Organization 5.2.2 Unit Staff (continued) shall be assigned for each control room from which a reactor is operating in MODES 1, 2, 3. or 4.
b.
Shift crew composition shall meet the requirements stipulated herein and in 10 CFR 50.54(m). Shift crew composition may be.less than the minimum. requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.f for a peri.od of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members, provided immediate action is taken to restore the shi.ft crew composition to within the minimum requirements.
A Radiation Protection Technician shall be on site when fuel is in the reactor.'he position,may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
in order to provide for unexpected
- absence, provided immediate action is taken to fill the required position.
Administrative procedures shal.l be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g..
licensed SROs.
licensed ROs.
radiation protection technicians, auxiliary operators'nd key maintenance personnel).
The controls shall include guidelines on working hours that ensure adequate shift coverage shall be maintained without routine heavy use of overtime.
Any deviation from the working hour guidelines shall be authorized in advance by personnel at the Director level or designees, in accordance with approved administrative procedures and with documentation of the basis-for granting the deviation.
Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by these authorized. individuals or designees to ensure that excessive hours have not been assigned.
Routine deviation from the above guidelines is not authorized.
(continued)
PALO VERDE UNITS 1.2.3 5.2-2 AMENDMENT NO. 447.120
il ll
Organi zation 5.2
~ 5.2 Organization 5.2.2
.Unit Staff (continued) e.
The Operations Department Leader or Operations Supervisor shal.l:hold an SRO license.
f.
The Shift Technical Advisor (STA),shall provide advisory technical support to the Shift Manager in the areas of thermal hydraulics, reactor:engineering, and plant analysis with regard to the safe operation of the unit..
In addition.
the.STA shall, meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shi ft.
PALO VERDE UNITS 1.2.3 5.2-3
.AMENDMENT NO. ~120
ii Il
Unit Staff Qualifications 5.3
'.'0 ADMINISTRATIVECONTROLS 5.3 Unit Staff Qualifications 5.3.1 5.3.2'ach.
member of the unit staff shall meet or exceed the minimum qualifications of Regulatory Guide 1.8.
September 1975 and ANSI/ANS 3.1-1978.
except the Director, Site Radiation Protection shall meet or exceed the qualification of Regulatory Guide 1.8, September
- 1975, and the Shift Technical Advisor shall 'have a
bachelor's degree or equivalent in a scientific or engineering discipline with speci:fic training in plant design and plant operating characteristics.
including transients and accidents-.
For the purpose of 10 CFR 55;4, a,licensed senior, reactor operation (SRO) and a licensed reactor, operator (RO) are those individuals who, in addition to meeting the requirements of TS 5.3.1, perform the functions described in 10 CFR 50;54(m)..
PALO VERDE UNITS'.2.3 5.3-1'MENDMENT NO. ~120
0
Procedures
- 5. 4'
- 5. 0 ADMINISTRATIVECONTROLS 5.4 Procedures 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:
a.
The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978; b.
The emergency operating procedures required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1.
as stated in Generic Letter '82-33; c.
Quality assurance for effluent and environmental monitoring:
d.
Fire Protection Program implementation; and e.
All programs specified in Specification 5.5.
f.
Modification of core protection calculator (CPC) addressable constants.
These procedures shall include provisions to ensure that sufficient margin is maintained in CPC type I addressable constants to avoid excessive operator interaction with CPCs during reactor operation.
Modifications to the CPC software (including changes of algorithms and fuel cycle specific data) shall be performed in accordance with the most recent version of "CPC Protection Algorithm Software Change Procedure."
CEN-39(A)-P, which has been determined to be applicable to the facility.
Additions or deletions to CPC addressable constants or changes to addressable constant software limit values shall not be implemented without prior NRC approval.
PALO VERDE UNITS 1.2.3 5.4-1 AMENDMENT NO. ~120
0
Programs and Manuals 5.5
- 5. 0 ADMINISTRATIVECONTROLS 5.5 Programs and Manuals The following programs shall be established.
implemented, and maintained.
5.5. 1 Offsite Dose Calculation Manual (ODCM) a.
The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive, gaseous and liquid effluents, in the calculation of gaseous and liquid, effluent monitoring alarm and.trip setpoints
~
and in the conduct of the radiological environmental monitoring program; and b.
The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental, Operating, and Radioactive Effluent Release Reports required by Specification 5.6.2 and Specification 5.6.3..
Licensee initiated changes to the ODCM:
a.
Shall be documented and records of reviews performed shall be retained.
This documentation shall contain:
1.
Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s),
2.
A determination that the change(s) maintain the levels.
of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a.
and 10 CFR 50, Appendix I, and not adversely
.impact the accuracy or reliability of effluent, dose.
or setpoint calculations:
b.
Shall become effective after the approval of the Director, Site Chemistry:
and Shall be submitted to the NRC in the form of a complete.
legible copy of the entire ODCM as
- a. part of or concurrent with the Radioactive Effluent Rel'ease
'Report for the period of the report in which any change in the ODCM was made.
Each change shall be identified by markings in the margin of (continued).
PALO VERDE UNITS 1.2,3 5.5-1 AMENDMENT NO. ~120
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Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.1 Offsite Dose Calculation Manual (ODCH)
(continued) the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.
5.5.2 Primar Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable.
The systems include, recirculation portion of the high pressure injection system, the shutdown cooling portion of the low pressure safety injection system.
the post-accident sampling subsystem of the reactor coolant sampling
- system, the containment spray system, the post-accident sampling return piping of the radioactive waste gas system, the post-accident sampling return piping of the liquid radwaste
- system, and the post-accident containment atmosphere sampling piping of the hydrogen monitoring subsystem.
The program shall include the following:
a.
Preventive maintenance and periodic visual inspection requirements; and b.
Integrated leak test requirements for each system at refueling cycle intervals or less.
5.5.3 Post Accident Sam lin This program provides controls that ensure the capability to obtain and analyze reactor coolant.
radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions.
The program shall include the following:
a.
Training of.personnel:
b.
Procedures for sampling and analysis:
and c.
Provisions for maintenance of sampling and analysis equipment.
(continued)
PALO VERDE UNITS 1,2,3 5.5-2 NENDMENT NO. ~120
0 0
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Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Pro ram This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable.
The program shall be contained in the ODCM. shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded.
The program shall include the following elements:
a.
Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination. in accordance with the methodology in the ODCM; b.
Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas.
conforming to 10 times the concentration values in Appendix B. Table 2, Column 2 to 10 CFR 20. 1001-20.2402; c.
Monitoring, sampling, and analysis of radioactive liquid and
,gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM; d.
Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas.
conforming to 10 CFR 50, Appendix-I:
e.
Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days:
f.
Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected, doses in a period of 31 days would exceed 2X of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I:
(continued)
PALO VERDE UNITS 1.2.3 5.5-3 AMENDMENT NO. 447120
Programs and. Manuals 5.5
'.5 Programs and Manuals 5.5.4 5.5.5 5.5.6 Radioactive Effluent Controls Pro ram (continued) g.
Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary; 1'.
For noble gases:
less than or equal to a dose rate of 500 mrems/yr to the total body and less than or equal to a dose rate of 3000 mrems/yr to the skin, and 2.
For iodine-131.
iodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days:
less than or equal to a dose rate of 1500 mrems/yr to any organ:
h.
Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50.
Appendix I; i.
Limitations on the annual and. quarterly doses to a member of the public from iodine-131, iodine-133, tritium. and al1 radionuclides in particulate form with half lives ) 8 days in gaseous effluents released'from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;.and j.
Limitations on the annual dose or dose commitment to any member of the public beyond the site boundary due to releases of radioactivity and to radiation from uranium fuel cycle sources.
conforming to 40 CFR 190.
Com onent C clic or Transient Limit This program provides controls to track the UFSAR Section 3.9. 1. 1 cyclic and transient occurrences to ensure that components are maintained within the design limits.
Pre-Stressed Concrete Containment Tendon Surveillance Pro ram This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments,.
including effectiveness of its corrosion protection medium. to ensure containment structural integrity.
The program shall include baseline measurements prior to initial operations.
The Tendon Surveillance Program.
inspection frequencies, and acceptance criteria shall be in -accordance with Regulatory Guide 1.35, as described in Section 1.8 of the UFSAR.
The provisions of SR 3.0.2 and SR 3.'0.3 are applicable to the Tendon Surveillance Program inspection frequencies.
(continued)
PALO VERDE UNITS 1,2,3
- 5. 5-'4 AMENDMENT NO. ~120
il
Programs and Manuals 5.5
~ 5.5 Programs and Manuals 5.5.7 Reactor Coolant Pum Fl eel Ins ection. Pro ram This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendations of regulatory position c.4.b of Regulatory Guide 1.14, Revision 0, October 1971.
5.5.8 Inservice Testin Pro ram This program provides controls.for inservice testing of ASME I:ode Class 1
~ 2,. and 3 components including applicable supports.
The program, shall include the following:
a.
Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as follows:
ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for inservice testing activities Required Frequencies
- for performing inservice tes tin acti vities At least once per 7 days At least once per 31 days At least once per 92 days Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Biennially or every 2 years At least once per 184 days At least once per 276 days At least once per 366 days At least once per 731 days The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities; c.
The provisions of SR 3.0.3 are applicable to inservice
.testing activities:
and d.
Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.
(continued)
PALO VERDE UNITS 1.2.3 5.5-5 AMENDMENT NO. 447-120
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Programs and Manuals 5.5
'.5 Programs and Manuals (continued) 5.5.9 5.5.9.1 5.5.9.2a Steam Generator (SG) Tube Surveillance Pro ram This program provides controls for the Inservice Inspection of steam generator tubes and tube sleeves to ensure that structural integrity of this portion of the RCS is maintained.
The program shall include the following:
Steam Generator Sam le Selection and Ins ection
- Each steam genera or s a
e e ermine uring shutdown by selecting and inspecting tubes and tube sleeves in at least the minimum number of steam generators specified in Table 5.5.9-1.
Steam Generator Tube Sam le Selection and Ins ection
- The steam genera or u e minimum samp e size, inspec ion resu t classification, and the corresponding action required shall be as specified in Table 5.5.9-2.
The inservice inspection of steam generator tubes shall be performed at the frequencies specified in, 5.5.9.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of 5.5.9.4.
The tubes selected for each inservice inspection shall include at least 3X of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:
a.
Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50K of the tubes inspected shall be from these critical areas.
b.
The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
1.
All nonplugged tubes that previously had detectable wall penetrations (greater than 20K and not sleeved in that area).
2.
Tubes in those areas where experience has indicated potential problems.
3.
A tube inspection (pursuant to Specification 5.5.9.4.a. 10.) shall be performed 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.
C.
The tubes selected as the second and third samples (if required by Table 5.5.9-2) during each inservice inspection may be subjected to a partial tube inspection provided:
(continued)
PALO VERDE UNITS 1.2,3 5.5-6 AMENDMENT NO 447-120
Programs and Manuals 5.5
'.5 Programs and,Manuals (continued) 5.5.9.2a Steam Generator Tube Sam le Selection and Ins ection (continued) 1.
The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.
2.
The inspections include those portions of the tubes where imperfections. were previously found.
The results of each inspection sample shall be classified into one of the following three categories (this classification shall apply to the inspection of tubes and treated exclusive of the sleeve inspections in 5.5.9.2b):
~Cate or Ins ection Results C-1 Less than: 5K of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-'2 C-3 One or more tubes,.but not more than lX of the total tubes inspected are defective, or between 5C and. 10K of the total tubes.
inspected are degraded tubes.
More than lOX of the total tubes inspected are degraded tubes or more than 1X of the inspected tubes are defective.
-NOTE In all inspections, previously degraded tubes must exhibit.significant (greater than 10K) further wall penetrations to be included in the above percentage calculations.
(continued)
PALO VEROE UNITS 1'.2.3 5.5-7
,AMENDMENT NO. 447-120
l ik 4
I
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9.2b Steam Generator Tube Sleeve Sam le Selection and Ins ection - The steam genera or u e s eeve minimum samp e size.
inspec ion result classification, and the corresponding action required shall be as specified in Table 5.5.9-3.
The inservice inspection of steam generator tube sleeves shall be performed at the frequencies specified in 5.5.9.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of 5.5.9.4.
The tube sleeves selected for each inservice,inspection shall include at least 20K of the total number of tube sleeves in all steam generators:
the tube sleeves selected for these inspections shall be selected on a random basis except:
a.
Where experience in similar plants with similar water chemistry indicates critical areas to be inspected.
then at least 50K of.the tube sleeves inspected shall be from these critical areas.
Where the number of sleeves in the cri.tical areas.represents less than 50X of the initial sample, all sleeves in the critical areas shall be inspected.
b.
The first sample of tube sleeves selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
l.
All tube sleeves that previously had detectable wall
.penetrations (greater.
than 20X).
2.
Tube sleeves in those areas where experience has indicated potential problems.
3.
A tube sleeve inspection (pursuant to Specification 5.5.9.4.a.8.)
shall be performed on each selected tube sleeve.
If any selected tube sleeve does not permit the passage of the eddy current probe for a tube sleeve inspection, this shall be recorded and an adjacent tube sleeve shall be selected and subjected
.to a tube sleeve inspection.
The results of each inspection sample shall be classified into one of the following three categories (this classification shall apply to the inspection of sleeves and treated exclusive of the tube inspections in 5.5.9.2a).:
(continued)
PALO VERDE UNITS 1,2,3 5.5-8 AMENDMENT NO. &7120
I li
Programs and. Manuals 5.5
'.5 Programs and Manuals (continued) 5.5.9.2b Steam Generator Tube Sleeve Sam le Selection and Ins ection continue Cate<aory C-1 C-2 C-3 Ins ection Results Less than 5X of the total tube sleeves inspected are degraded tube sleeves and none of the inspected tube sleeves are defective.
One or more tube sleeves.
but not more than-1X of the total tube'sleeves inspected are defective.
or between 5K and 10K of the total tube sleeves inspected are degraded tube sleeves.
More than 10K of the total tube sleeves inspected are degraded tube sleeves or more than 1X of the inspected tube sleeves are defective.
NOTE In all inspections, previously degraded tube sleeves must exhibit significant (greater than 10K) further wall penetrations to be included in the above percentage calculations.
5.5.9.3 Ins ection Fre uencies
- The above required inservice inspections o
steam genera or u es and tube sleeves shall be performed at the following frequencies:
The first inservice,inspection of the steam generator tubes shall be.performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality.
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 inspection.
result in all inspection results falling into the
'-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.
(continued)
PALO VERDE UNITS 1,2,3 5.5-9 AMENDMENT NO. 447. 120
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9.3 Ins ection Fre uencies (continued) b.
The inservice inspection of steam generator tube sleeves shall be performed at the following-frequencies:
1.
Steam generator tube sleeves shall.be inspected prior to initial operation.
The. operating period before the initial inservice inspection shall not be shorter than six months nor longer than 24 months.
The inspections of tube sleeves shall be configured to ensure that eath individual tube sleeve is inspected at least once in 60 months.
2.
If the results of the inservi.ce inspection of steam, generator tube sleeves conducted in accordance with Table 5.5.9-3 fall in category C-3. the inspection frequency shall be increased to ensure that each remaining tube sleeve is inspected at least once in 30 months.
The increase in inspection frequency shall apply until the subsequent inspection satisfies the criteria for Category C-1.
If the results of the inservice inspection of a steam generator conducted in accordance with Tables 5.5.9-2 and 5.5.9-3 at 40 month intervals fall into Category C-3. the inspection frequency for the applicable tube or sleeve inspection 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 Specification 5.5.9.3.a (tubes) or 5.5.9.3.b.3 (sleeves):
the interval may then be extended to a maximum of once per 40 months (tubes) or
- 30. months (sleeves).
Additional. unscheduled inservice inspections shall be performed on each steam generator in accor'dance with the first.sample inspection specified in Tables 5.5.9-2 and 5.5.9-3 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. 14.
2.
A seismic occurrence greater than the Operating Basis Earthquake.
(continued)
PALO VEROE UNITS 1,2.3
- 5. 5-10 AMENDMENT NO. 447-120
ll 0
J l
(
Programs and Manuals 5.5
~ 5.5 Programs and Manuals (continued) 5.5.9.3 Ins ection Fre uencies (continued) 3.
A loss-of-coolant accident requiring actuation of the engineered safeguards.
4.
A main steam line or feedwater line break.
5.5.9.4 Acce tance Criteria a.
As used in this Specification 1.
Defect means an imperfection of such severity that it exceeefs the plugging or repair limit.
A tube or sleeve containing a defect is defective.
2.
g~dti 2 -td d
ttg.
tg wear, or general corrosion occurring on. either inside or outside of a tube.
3.
X De radation means the percentage of the tube wall ic ness a
ected or removed by degradation.
4.
De raded Tube or Sleeve means a tube or sleeve containing imper ections greater than or equal to 20K of the nominal wall.thickness caused by degradation.
5.
Im erfection means an exception to the dimensions, inis
, or contour of a tube from that required by fabrication drawings or specifications.
Eddy-current testing indications below 20K of the nominal tube wall thickness, if detectable, may be considered as imperfections.
6.
Plu in or Re air Limit means the imperfection depth at or eyon w ic e
ube shall be removed from service by plugging or repaired by sleeving in the affected. area.
The plugging or repair imperfection depths specified below are in percentage of nominal wall thickness:
a.
Original tube wall b.
ABB-CE leak tight sleeve wall 40K 35K (continued)
PALO VERDE UNITS 1.2,3 5.5-11 AMENDMENT NO. 447120
0 I
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9.4 Acce tance Criteria (continued) 7.
Preservice Ins ection in the context of new steam genera ors 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.
Preservice Ins ection for steam generator tubes repaire y tu e s eeving means an inspection of the full length of the pressure boundary portion of the sleeved area performed by eddy current techniques prior to service to establish a baseline condition of the sleeved area.
The sleeved area includes the pressure retaining portions of the parent tube in contact with the sleeve, the sleeve-to-tube weld and the pressure retaining portion of the sleeve.
8.
Sleeve Ins ection for sleeves selected in accordance wit ta e
. -3 means an inspection of the sleeved area.
including the pressure retaining portions of the parent tube in contact with the sleeve.
the sleeve-to-tube weld and the pressure retaining portion of the sleeve.
9.
Tube or Tubin means that portion of the tube that orms e primary system to secondary system pressure boundary.
10.
Tube Ins ection for tubes selected in accordance with a
e 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, excluding sleeved areas.
(continued)
PALO VERDE UNITS 1.2.3 5.'5-12 AMENDMENT NO, 447-120
0 I
0 Programs and Manuals 5.5 5.5 Programs and Manuals (continued)
.5.5.9.4 Acce tance Criteria (continued) 11.
Tube Re air or Sleeve refers to welded sleeving, as escri e
in om us ion Engineering, Inc.
(CE or ABB-CE) report CEN-630-P, "Repair of 3/4" O.D. Steam Generator Tubes Using Leak Tight Sleeves,"
Revision 02, June 1997, which is used to maintain a tube in service or to return a previously plugged tube to service.
Returning a previously plugged tube to service includes the removal of the tube plugs that were installed as a preventive or corrective measure and performing a tube inspection of the tube in accordance with Technical Specification 5.5.9.4.a.8.
12.
Unserviceable describes the condition of a tube if it df t1 g
ght ff tit 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 5.5.9.3.d.,
above.
The steam generator shall be determined OPERABLE after completing the corresponding actions required by Tables 5.5.9-2 and 5.5.9-3, including the following:
1.
Plug or repair all defective tubes and all tubes containing through-wall cracks.
2.
Plug all tubes containing any defective sleeves and all.
tubes containing any sleeves with through-wall cracks.
C (continued)
PALO VERDE UNITS 1.2,3 5.5-13 NENDMENT NO. ~120
IQ 0
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I
TABLE 5.5.9-1 HINIHUH NUHBER OF STEAH GENERATORS TO BE INSPECTEO DURING INSERVICE INSPECTION Preservice Inspection No Yes No. of Steam Generators per Unit Two Two First Inservice Inspection Al1 One Second 8 Subsequent Inservice Inspection One*
One*
TABLE NOTATION "The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3
N X of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a like manner.
Note that under so'me circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators.
Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.
PALO VERDE UNITS 1.2,3 5.5-14 AHENDHENT NO. 447-120
I 4
IST SAMPLE INSPECTION TABLE 5.5.9-2 STEN GENERATOR TUBE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sa le Size A minimum of S Tubes per S.G Result C-I C-2 C-3 Action Re ired Hone Plug or repair defective tubes and inspect additional 2S tubes in this S.G.
Inspect all tubes in this S.G.. plug or repair defective tubes and inspect 25 tubes in each other S.G.
Notification to NRC pursuant to 10 CFR 50.72 (b)(2)
Result C-I C-2 C-3 All other S.G.s are C-1 Some S.G.s C-2 but no additional S.G.
are C-3 Additional S.G.
is C-3 Action Re fred N.A.
Hone Plug or repair defective tubes and inspect additional 45 tubes in this S.G.
Perform action for C-3 result of Hone Perform action for C-2 result of second sample Inspect all tubes in each S.G.
and plug or repair defective tubes.
Hotification to NRC pursuant to 10 CFR 50.72 (b)(2)
Result N.A.
C-I C-2 C-3 N.A.
N.A.
N.A.
Action Re ired Hone Plug or repair defective tubes Perform action for C-3 result of first sam le N.A.
N.A.
3 Q Where N is the number of steam generators in the unit, and n is the number of steam generators N
inspected during an inspection.
PALO VERDE UNITS 1.2,3 5.5-15 AMENDMENT NO. ~120
I
TAHLE 5.5.9-3 STEN GENERATOR SLEEVE INSPECTION IST SAHPLE INSPECTION 2ND SNPLE INSPECTION Sa le Size A miniwm of 20K of the sleeves per S.G.
Result C-1 C-2 Action Re ired None Plug tubes containing defective sleeves and inspect all remaining installed sleeves in this S.G.
Result N.A.
C-1 C-2 Action Re ired Hone Plug tubes containing defective sleeves C-3 Perform action for C-3 result of first sample C-3 Inspect all installed sleeves in this S.G.. plug tubes containing defective sleeves and inspect IDOL of the installed sleeves in the other S.G.
Notification to NRC pursuant to 10 CFR 50.72 (b)(2)
Other S.G. is C-1 Other S.G. is C-2 Other S.G. is C-3 Hone Plug tubes containing defective sleeves Inspect all sleeves in each S.G.
and plug tubes containing defective sleeves.
Notification to NRC pursuant to 10 CFR 50.72(b)(2)
PALO VERDE UNITS 1.2.3 5.5-16 NENDMENT N0.1 20
1 Programs and Manuals 5.5
~ 5.5 Programs and Manuals (continued) 5.5.10 Secondar Water Chemistr Pro ram This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion cracking.
The program shall include:
a.
Identif'ication of a sampling schedule for the critical variables and control points for these variables; b.
Identification of the procedures used to measure the va1ues of the critical variables; d.
Identification of process sampling points which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage; Procedures for the recording and management of data; e.
Procedures defining corrective actions for all off control point chemistry conditions; and f.
A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.
5.5.11 Ventilation Filter Testin Pro ram (VFTP)
A program shall be established to,implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52.
Revision 2.
and in accordance with Regulatory Guide 1.52.
Revision 2 and ANSI N510-1980 at the system flowrate specified below + 10K.
a.
Demonstrate for 'each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass 6 1.0 X when tested in accordance with Regulatory Guide 1.52, Revision 2.
and ANSI N510-1980. at the system flowrate specified as follows
+ lOX:
(continued)
PALO VERDE UNITS 1.2.3 5.5-17 AMENDMENT NO. ~120.
il 0
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.11 Ventilation Filter Testin Pro ram (VFTP)
(continued)
ESF Ventilation S stem Control Room Essential Filtration System (CREFS)
Engineered Safety Feature (ESF)
Pump Room Exhaust Air Cleanup System (PREACS)
Flowrate 28,600 CFM 6.000 CFM b.
Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration
.and system bypass
~ 1.0 X when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980 at the system flowrate specified as follows x 10K:
Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber.
when obtained as described in Regulatory Guide 1.52. Revision 2.
and ANSI N510-1980 shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1979 at a temperature of 80'C a 0.5'C and greater than or equal to the relative humidity specified as follows:
ESF Ventilation S stem CREFS ESF PREACS Penetration s 1.0X RH 70K 70K (continued)
PALO VERDE UNITS 1.2.3 5:5-18 AMENDMENT NO. ~120
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l Programs and Manuals 5.5
, 5.5 Programs and Manuals (continued) 5.5.11 Ventilation Filter Testin Pro ram (VFTP)
(continued) d.
For each of the ESF systems, demonstrate the pressure drop across the combined HEPA filters, the prefilters,,and the charcoal adsorbers is less than the value specified below when tested in accordance with Regulatory Guide 1;52, Revision 2, and ANSI N510-1980 at the system flowrate specified as follows + 10X:
ESF Ventilation S stem Delta P
Flowrate CREFS ESF PREACS 8.4 inches water gauge 28,600 CFM 8.4 inches water gauge 6,000 CFM e.
Demonstrate that the heaters for each of the ESF systems dissipate the following specified value when tested in accordance with ANSI N510-1980:
ESF Ventilation S stem ESF PREACS
~Watta e
) 19 kW 5.5.12 The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.
Ex losive Gas and Stora e'ank Radioactivit Monitorin Pro ram This program provides control for,potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.
The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP)
ETSB 11-5.
"Postulated Radioactive Release due to Waste Gas System Leak or Failure".
The liquid radwaste quantities shall be determined in accordance with Standard Review Plan. Section 15.7.3,
- "Postulated Radioacti.ve Release due to Tank Failures".
(continued)
PALO VERDE UNITS 1.2.3 5.5-19 AMENDMENT NO 4P-120
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I I-
Programs and Manuals 5.5
, 5.5 Programs and Manuals (continued) 5.5.12 Ex losive Gas and Stora e Tank Radioactivit Monitorin Pro ram continue The program shall include:
a.
The 1'imits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a surveillance program to ensure the limits are maintained.
Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion):
b.
A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whol'e body exposure of t 0.5 rem to any individual in an unrestricted area.
in the event of an uncontrolled release of the tanks'ontents; and C.
A surveillance program to ensure that the quantity of.:
radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks'ontents and that do not have'tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR Part 20, Appendix 8, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted
- area, in the event of an uncontrolled release of the tanks'ontents.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program survei 1 1 ance frequenci es.
5.5.13 Diesel Fuel Oil Testin Pro ram A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established.
The program shall include sampling and testing requirements.
and acceptance criteria, all in accordance with applicable ASTM Standards as referenced in the UFSAR.
The purpose of the program is to establish the following:
(continued)
PALO VERDE UNITS 1.2.3 5.5-20 AMENDMENT NO. 447-120
0 I
~,
Programs and Manuals 5.5
, 5.5 Programs and Manuals (continued) 5.5.13 Diesel Fuel Oil Testin Pro ram (continued) a.
Acceptability of new fuel oil for, use prior to addition to storage tanks by determining that the fuel oil has:
1.
.An API gravity or an absolute specific gravity within
- limits, 2.
A flash point and kinematic viscosity, within limits for ASTM 2D fuel oil, and 3.
Water and sediment are within the limits of ASTM D1796:
b.
Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to storage tanks:
and c.
Total particulate concentration of the stored fuel oil is 5 10 mg/1 when tested every 92 days in accordance with ASTM D-2276.
Method A-,2 or A-3.
5.5.14 Technical S ecifications (TS) Bases Control Pro ram This program provides a means for processing changes to the Bases of these Technical Specifications.
a.
Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
b.
Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:
A change in the TS incorporated in the license; or A change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.
C.
The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.
(continued)
PALO VERDE UNITS 1.2.3 5.5-21 AMENDMENT NO. 447-120
~"
~ a rw
0
Programs and Manuals 5.5
, 5.5 Programs and Manuals (continued) 5.5.14 Technical S ecifications (TS) Bases Control Pro ram (continued) d.
Proposed changes, that meet the criteria of Specification 5.5.14b above shall be reviewed and approved by the NRC prior to implementation.
Changes to.the Bases implemented without prior NRC approval shall be provided to the NRC on a
frequency consistent with 10 CFR 50.71(e).
5.5.15 Safet Functions Determination Pro ram (SFDP)
This program ensures loss of safety function is detected:and appropriate actions taken.
Upon entry into LCO 3.0.6, an evaluation shall, be made to determine if loss of safety function exists.
Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of.the support system inoperability and'orresponding exception to entering supported system Condition and Required Actions.
This program implements the requirements of'CO 3.0.6.
The SFDP shall contain the following:
a.
Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected:
b.
Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists; c'.
Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and d.
Other appropriate limitations and remedial or compensatory actions.
A loss of safety function exists
- when, assuming no concurrent single failure.
a safety function assumed in the accident analysis cannot be performed.
For the purpose of this program, a loss of safety function may exist when a support system is inoperable.
and:
a.
A required system redundant to system(s) supported by the inoperable support system is also inoperable:
or PALO VERDE UNITS 1.2,3 5.5-22 (continued)
AMENDMENT NO. 447120
0 4
I
)
~,
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.15 5.5.16 Safet Functions Determination Pro ram (continued) b.
A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable:
or c.
A required system redundant to support system(s) for the supported systems (a) and (b) above is also inoperable.
The SFDP identifies where a loss of safety function exists.
If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.
Containment Leaka e Rate Testin Pro ram A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50. Appendix J, Option B, as modi.fied by approved exemptions'his program shall be in accordance with the guidelines contained in Regulatory Guide 1.163.
"Performance-Based Containment Leak-Test =Program," dated September, 1995.
as modified by the following exceptions:
The peak calculated containment internal pressure for the design basis loss of coolant accident.
P,. is 52.0 psig.
The containment design pressure is 60 psig.
The maximum allowable containment leakage rate.
L,. at P,. shall be 0. 1 X of containment air weight per day.
Leakage Rate acceptance criteria are:
a.
Containment leakage rate acceptance criterion is c 1.0 L,.
During the first unit startup following testing in accordance with this program, the leakage rate acceptance are
< 0.60 L, for the Type B and C tests and s 0.75 L, for Type A tests.
b.
Air lock testing acceptance criteria are:
1.
Overall air lock leakage rate is s 0.05 L, when tested at zP,.
2.
For each door.
leakage rate is s 0.01 L, when pressurized to h 14.5 psig.
PALO VERDE UNITS 1,2,3 5.5-23 (continued)
AMENDMENT NO. ~ 120 r
0 1
I
,C
Programs and Manuals 5.5
, 5.5 Programs and Manuals (continued) 5.5.16 Containment Leaka e Rate Testin Pro ram (continued)
The provisions of SR 3.0.2 do not apply to the test frequencies in the Containment Leakage Rate Testing Program.
The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
PALO YERDE UNITS 1,2,3 5.5-24 AMENDMENT NO. 447-120.
Reporting Requirements 5.6
. 5.0 ADMINISTRATIVECONTROLS
- 5. 6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.
5.6.1 Occu ational Radiation Ex osure Re ort NOTE-A single submittal may be made for a multiple unit station.
The submittal should combine sections common to all units at the station.
A tabulation on an annual basis of the number of station, utility.
and other personnel (including contractors) receiving exposures
) 100 mrem/yr and their associated man rem exposure according to work and job functions (e.g.',
reactor operations and surveillance.
inservice inspection, routine maintenance.
special maintenance.
waste processing.
and refueling).
This tabulation supplements the requirements of 10 CFR 20.2206.
The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge measurements.
Small exposures totalling ( 20K of the individual total dose need not be accounted for.
In the aggregate.
at least 80X of the total whole body dose received from external sources should be assigned to specific major work functions.
The report shall be submitted by April 30 of each year.
5.6.2 Annual Radiolo ical Environmental 0 eratin Re ort NOTE A single submittal may be made for a multiple unit station.
The submittal should combine sections codon to all units at the station.
The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year.
The report shall include summaries, interpretations.
and analyses of trends of the results of the radiological environmental monitoring program for the reporting period.
The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual
,(ODCM). and in 10 CFR 50. Appendix I. Sections IV.B.2, IV.B.3, and IV.C.
(continued)
PALO VERDE UNITS 1,2.3 5.6-1 AMENDMENT NO. 4%120
0
~ ~
C'
Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5'.2 Annual Radiolo ical Environmental 0 eratin Re ort (continued)
The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979.
The report shall identify the TLB results that represent collocated dosimeters in relation to the NRC TLD program and the exposure period associated with each result.
In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.
The missing data shall be submitted in a supplementary report as soon as possible.
5.6.3 Radioactive Effluent Release Re ort OTE---------------------'--------
N A single submittal may be made for a multiple unit station.
The submittal should combine sections comoon to all units at the station; however, for units with separate radwaste
- system, the submittal shall specify the releases of radioactive material from each unit.
The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a.
The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.
The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.
5.6.4 Monthl 0 eratin Re orts Routine reports of operating statistics and shutdown experience.
including documentation of all challenges to the shutdown cooling system suction line relief valves or pressurizer safety valves, shall be submitted on a monthly basis no later than the. 15th of each month following the calendar month covered by the report.
(continued)
PALO VERDE UNITS 1,2.3 5.6-2 AMENDMENT NO. 4%120
il I
4
Reporting Requirements5.6
~ 5.6 Reporting Requirements (continued) 5.6.5 CORE 'OPERATING LIMITS REPORT (COLR) a.
Core operating limits shall be established prior to each reload cycleor prior to any remaining portion of a reload cycle.
and shall be documented in the COLR for the following:
1.
Shutdown Margin - Reactor Trip Breakers Open for Specification 3.1.1.
2.
Shutdown Margin - Reactor Trip Breakers Closed for Specification 3.1.2.
3.
Moderator Temperature Coefficient BOL and EOL limits for Specification 3.1.4.
4.
Boron Dilution Alarm System for Specification 3.3.12.
5.
CEA Alignment for Specification 3.1.5.
6.
Regulating CEA Insertion Limits for Specification 3.1.7.
7.
Part Length CEA Insertion Limits for Specification 3.1.8.
8.
Linear Heat Rate for Specification 3.2.1.
9.
Azimuthal Power Tilt - T, for Specification 3.2.3.
10.
DNBR for Specification 3.2.4.
11.
Axial Shape Index for Specification 3.2.5.
12.
Boron Concentration (Mode 6) f'r Specification 3.9.1.
b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the. NRC, specifically those'described in the following documents:
1.
"CE Method for Control Element Assembly Ejection
- Analysis, "CENPD-0190-A, January 1976 (Methodology for Specification 3.1.7.
Regulating=CEA Insertion Limits).
(continued)
PALO VERDE UNITS 1,2.3 5.6-3 AMENDMENT NO. 447-120
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Reporting Requirements
'5. 6
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5.6 Reporting Requirements (continued) 5.6.5 Core 0 eratin Limits Re ort (COLR) (continued) 2.
6.
"The ROCS and DIT Computer Codes for Nuclear Design,"
CENPD-266-P-A, April 1983 [Methodology for Specifications 3.1.1.
Shutdown Margin - Reactor Trip Breakers Open; 3.1.2, Shutdown Margin - Reactor Trip Breakers Closed; 3.1.4, Moderator Temperature Coefficient BOL and EOL limits: 3.1.7, Regulating CEA Insertion Limits and 3.9.1.
Boron Concentration (Mode 6)j.
"Safety Evaluation Report related to the Final Design of the Standard Nuclear Steam Supply Reference Systems CESSAR System 80, Docket No.
STN 50-470, "NUREG-0852 (November 1981).
Supplements No.
1 (March 1983),
No.
2 (September 1983),
No.
3 (December 1987) [Methodology for Specifications 3.1.2, Shutdown Margin - Reactor Trip Breakers Closed; 3.1.4
~ Moderator Temperature Coefficient BOL and EOL limits; 3.3.12.
Boron Dilution Alarm System; '3.1.5, CEA 'Alignment; 3.1.7, Regulating CEA. Insertion Limits: 3.1.8.
Part Length CEA Insertion Limits and 3.2.3, Azimuthal Power Tilt - T,].
"Hodified Statistical Combination of Uncertainties."
CEN-356(V)-P-A Revision 01-P-A.
May 1988 and "System 80' Inlet.Flow Distribution." Supplement 1-P,to Enclosure 1-P to LD-82-054, February 1993 (Methodology for Specification 3.2.4, DNBR and 3.2.5 Axial Shape Index).
"Calculative Methods for the CE Large Break LOCA Evaluation Model for the Analysis of CE and W Designed NSSS."
CENPD-132.
Supplement 3-P-A. June 1985 (Methodology for Specification 3.2.1, Linear Heat Rate).
"Calculative Methods for the CE Small Break LOCA Evaluation Model." CENPD-137-P.
August 1974 (Methodology for Specification 3.2.1, Linear Heat Rate).
"Calculative Methods for the CE Small Break LOCA Evaluation Model." CENPD-137-P.
Supplement 1P, January 1977 (Methodology for Specification 3.2.1, Linear Heat Rate).
(continued)
PALO VERDE UNITS 1,2,3 5;6-4 AMENDMENT NO. ~120.
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,Reporting Requi rements 5.6 5.6 Reporting Requirements (continued) 5.6.5 Core 0 eratin Limits Re ort (COLR) (continued) 8.
Letter:
0.0. Parr (NRC) to F.
M. Stern (CE), dated June 13.
1975 (NRC Staff Review of the Combustion Engineering ECCS Evaluation Model).
NRC approval for:
5.6.5.b.6.
9.
Letter:
K. Kniel (NRC) to A. E. Scherer (CE), dated September 27, 1977 (Evaluation of Topical Reports CENPD-133, Supplement 3-P and CENPD-137, Supplement 1-P).
NRC approval for 5'.5.b.7.
10.
"Fuel Rod Maximum Allowable Pressure,"
CEN-372-P-A, May 1990 (Methodology for Specification 3.2.1. Linear Heat Rate).
ll.
Letter: A. C. Thadani (NRC) to A. E..Scherer (CE),
dated April 10.
1990, ("Acceptance for Reference CE Topical 'Report. CEN-372-P").
NRC approval for 5.6.5.b.lo.
C.
The core operating limits shall be determined such 4hat all applicable limits (e.g.. fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM. transient analysis limits. and accident analysis limits) of the safety analysis are met.
5.6.6 d.
The COLR, including any mid cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
~PAM Re art When a report is required by Condition B or G of LCO 3.3.10, "Post Accident Monitoring (PAM) Instrumentation."
a report shall be submitted within the following 14 days.
The report shall outline the preplanned alternate method of monitoring. the cause of the inoperability, and'he plans and schedule for restoring the instrumentation channels of the Function to.OPERABLE status.
(continued)
PALO VERDE UNITS 1.2.3 5.6-5 AMENDMENT NO. ~ i20.
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Reporting Requirements 5.6
, 5.6 Reporting Requirements (continued) 5.6.7 5.6.8 Tendon Surveillance Re ort Any abnormal degradation of the containment structure detected during the tests required by the Pre-Stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days.
The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages).
the inspection procedures, the tolerances on cracking.
and the corrective action taken.
Steam Generator Tube Ins ection Re ort Within 15 days following the completion of each inservice inspection of steam generator
- tubes, the number of tubes plugged and/or repaired in each steam generator shall be reported to the Commission in a Special Report.
The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report within 12 months following completion of the inspecti.on.
This Special Report shall include:
b.
Number and extent of tubes inspected.
Location and percent of wall-thickness penetration for each indication of an imperfection.
c.
Identification of tubes plugged and/or repai'red.,
Results of steam generator tube and sleeve inspections which fall into Category C-3 shall be reported in a Special Report to the Commission within 30 days and prior to resumption of plant operation and shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
PALO VERDE UNITS 1.2.3 5.6-6 AMENDMENT NO. ~ 120
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High'Radiation Area 5.7
- 5. 0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area 5.7.1 In addition to the provisions of 10 CFR 20.1601.
the following controls provide an alternate method for controlling access to high radiation areas as provided by paragraph 20.1601(c) of 10 CFR part 20.
as defined in 10 CFR 20, in which the intensity of radiation is > 100 mrem/hr but 5 1000 mrem/hr, shall be barricaded and conspicuously posted as a
high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Exposure Permit (REP).
Individuals qualified in radiation protection procedures (e.g.,
Radiation Protection Technicians) or personnel continuously escorted by such individuals may be exempt from the REP issuance requirement during the performance of thei r assigned duties in
'high radiation areas with exposure rates
~ 1000 mrem/hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.
Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
a.
A radiation monitoring device that continuously indicates the radiation dose rate in the area.
b.
A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.
Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.
C.
An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Radiation Protection Section Leader or designated alternate in the REP.
(continued)
PALO VERDE UNITS 1,2.3 5.7-1 AMENDMENT NO. ~120.
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5.7 High Radiation Area 5.7.2 In addition to the requirements of Specification 5.7.1, areas accessible to personnel with radiation levels such that an individual could receive in I hour a dose greater than 1000 mrem shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Manager on duty or Radiation Protection supervision.
Doors shall remain locked except during periods of access by personnel under an approved REP that shal.l specify the dose rate levels in the irrrnediate work areas and the maximum allowable stay times for individuals in those areas.
In lieu of the stay time specification of the
- REP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel. qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the. area.
5.7.3 For individual high radiation areas accessible to personnel with radiation levels such that an individual could receive in I hour a
dose in excess of 1000 mrem (measurement made at 30 cm from.source of radioactivity), that are located within large areas such as reactor containment, where no enclosure exists for purposes of
- locking, and where no enclosure can be reasonably constructed around the individual arear'hat individual area shall be barricaded and conspicuously, posted, and a flashirig light shall be activated as a warning device.
PALO VERDE UNITS 1,2,3 A
5.7-2 AMENDMENT NO. ~120
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