ML041880211

License Amendment Request, Miscellaneous Technical Specification Changes

ML041880211
Person / Time
Site:	San Onofre
Issue date:	06/29/2004
From:	Nunn D Southern California Edison Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML041880211 (105)

Text

.

.j SOUTHERN CALIFORNIA Dwight E. Nunn EDISON rVice President An EDISON INTERNATIONAAL-Company June 29, 2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Docket Nos. 50-361 and 60-362 Proposed Change Number (NPF-1 011 5) 533 License Amendment Request, Miscellaneous Technical Specification Changes San Onofre Nuclear Generating Station Units 2 and 3

Dear Sir or Madam:

Pursuant to 10CFR50.90, Southern California Edison (SCE) is submitting Enclosure 1 to request amendments to SCE Licenses NPF-10 and NPF-15 to change the Technical Specifications (TS) for San Onofre Units 2 and 3. The proposed changes revise the Technical Specifications to implement the following miscellaneous TS changes: revise TS 2.2.5 Safety Limit Violations Licensee Event Report (LER) reporting period from 30 days to 60 days, revise 3.4.3.1.2 Pressurizer Heatup/Cooldown Limits Surveillance Requirements frequency to reflect pressurizer spray cyclic limits being governed by the temperature differentials between the spray nozzle and the spray line, revise section 5.5.2.11.f.1 Steam Generator Tube Surveillance requirements to correct typographical errors, remove section 5.5.2.14 Configuration Risk Management Program (CRMP) in accordance with Federal Register Notice Vol. 64, No. 137 (July19, 1999), and revise 5.7.1.5 Core Operating Limits Report (COLR) to delete revision numbers and dates from the referenced documents in this section consistent with the NRC approved industry Technical Specifications Task Force (TSTF) Standard Technical Specifications Traveler number TSTF-363, "Revise Topical Report References in ITS (Improved Technical Specifications) 5.6.5 COLR" and incorporate editorial corrections.

Once approved, the amendment shall be implemented within 60 days.

P.O. Box 128 San Clemente, CA 92674-0128 949-368-1480 Fax 949-368-1490

Document Control Desk June 29, 2004 If you have any questions or require additional information, please contact Mr. Jack Rainsberry at (949) 368-7420.

Sincerely,

Enclosures:

1.

Notarized Affidavits

2.

Proposed Change Number (PCN)-533 cc:

B. S. Mallett, Regional Administrator, NRC Region IV B. M. Pham, NRC Project Manager, San Onofre Units 2, and 3 C. C. Osterholtz, NRC Senior Resident Inspector, San Onofre Units 2 and 3 S. Y. Hsu, Department of Health Services, Radiologic Health Branch UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA

)

EDISON COMPANY, ET AL. for a Class 103 )

Docket No. 50-361 License to Acquire, Possess, and Use

)

a Utilization Facility as Part of

)

Amendment Application Unit No. 2 of the San Onofre Nuclear

)

No. 225 Generating Station

)

SOUTHERN CALIFORNIA EDISON COMPANY, ET AL. pursuant to 10 CFR 50.90, hereby submit Amendment Application No. 225. This amendment application consists of Proposed Change No. NPF-10-533 to Facility Operating License NPF-10. Proposed Change No. NPF-10-533 is a request to revise the Technical Specification's (TS) to incorporate the following miscellaneous TS changes: revise TS 2.2.5 Safety Limit Violations Licensee Event Report (LER) reporting period from 30 days to 60 days, revise 3.4.3.1.2 Frequency to reflect pressurizer spray cyclic limits being governed by the temperature differentials between the spray nozzle and the spray line, revise section 5.5.2.11.f.1 to correct typographical errors, remove section 5.5.2.14 Configuration Risk Management Program (CRMP), and revise 5.7.1.5 Core Operating Limits Report (COLR) to delete revision numbers and dates from the referenced documents in this section and incorporate editorial corrections.

State of California County of San Diego Subscribed and sworn to (or affirmed) before me this a2 q day of

~

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,2004, by FRANCES M. THURBER Commisslon 1i 1295266 Notary Public - CollfomTa Son DTego County MyCa=Pn.

ress Mcr2325 Public

Enclosure I UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA

)

EDISON COMPANY, ET AL. for a Class 103 )

Docket No. 50-362 License to Acquire, Possess, and Use

)

a Utilization Facility as Part of

)

Amendment Application Unit No. 3 of the San Onofre Nuclear

)

No. 209 Generating Station

)

SOUTHERN CALIFORNIA EDISON COMPANY, ETAAL. pursuant to 10 CFR 50.90, hereby submit Amendment Application No. 209. This amendment application consists of Proposed Change No. NPF-15-533 to Facility Operating License NPF-15. Proposed Change No. NPF-1 5-533 is a request to revise the Technical Specification's (TS) to incorporate the following miscellaneous TS changes: revise TS 2.2.5 Safety Limit Violations Licensee Event Report (LER) reporting period from 30 days to 60 days, revise 3.4.3.1.2 Frequency to reflect pressurizer spray cyclic limits being governed by the temperature differentials between the spray nozzle and the spray line, revise section 5.5.2.11.f.1 to correct typographical errors, remove section 5.5.2.14 Configuration Risk Management Program (CRMP), and revise 5.7.1.5 Core Operating Limits Report (COLR) to delete revision numbers and dates from the referenced documents in this section and incorporate editorial corrections.

State of California County of San Diego Subscribed and sworn to (or affirmed) before me this Z f.

day of

, 2004, by Dwight E.

n Vice residenX FRANCESM.TURBER~

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ENCLOSURE 2 DESCRIPTION AND NO SIGNIFICANT HAZARDS ANALYSIS FOR PROPOSED CHANGE NPF-10115-533 Miscellaneous Technical Specification Changes San Onofre Nuclear Generating Station Units 2 and 3

DESCRIPTION AND NO SIGNIFICANT HAZARDS ANALYSIS FOR PROPOSED CHANGE NPF-10/15-533 Miscellaneous Technical Specification Changes San Onofre Nuclear Generating Station Units 2 and 3 EXISTING TECHNICAL SPECIFICATIONS Unit 2: See Attachment A Unit 3: See Attachment B PROPOSED TECHNICAL SPECIFICATIONS (Additions highlighted and deletions struck-out)

Unit 2: See Attachment C Unit 3: See Attachment D PROPOSED TECHNICAL SPECIFICATIONS (With changes)

Unit 2: See Attachment E Unit 3: See Attachment F PROPOSED TECHNICAL SPECIFICATION BASES CHANGES (Provided for information / additions highlighted and deletions struck-out)

Unit 2: See Attachment G (typical for both Units 2 and 3)

1.0 DESCRIPTION

Southern California Edison (SCE) proposes changes to the Units 2 and 3 Technical Specifications (TS) to accomplish miscellaneous TS changes. These proposed TS changes revise TS 2.2.5 Safety Limit Violations Licensee Event Report (LER) reporting period from 30 days to 60 days, revise 3.4.3.1.2 Pressurizer Heatup/Cooldown Limits Surveillance Requirements frequency to reflect pressurizer spray cyclic limits being governed by the temperature differentials between the spray nozzle and the spray line, revise section 5.5.2.11.f.1 Steam Generator Tube Surveillance requirements to correct typographical errors, remove section 5.5.2.14 Configuration Risk Management Program (CRMP) in accordance with Federal Register Notice Vol. 64, No. 137 (July19, 1999), and revise 5.7.1.5 Core Operating Limits Report (COLR) to delete revision numbers and dates from the referenced documents in this section consistent with the NRC approved Industry Technical Specifications Task Force (TSTF) Standard Technical Specifications Traveler number TSTF-363, "Revise Topical Report References in ITS (Improved Technical Specifications) 5.6.5 COLR" and incorporate editorial corrections.

2.0 PROPOSED CHANGE

S TS 2.2 SL (Safety Limit) Violations, Section 2.2.5 - It is proposed to change

'Within 30 days of the violation, a Licensee Event Report (LER) shall be prepared pursuant to 10 CFR 50.73" to 'Within 60 days of the violation, a Licensee Event Report (LER) shall be prepared pursuant to 10 CFR 50.73."

The 60 days LER reporting period corresponds with the current Code of Federal Regulations (CFR) requirements which was previously 30 days.

TS 3.4.3.1 Pressurizer Heatup/Cooldown Limits, SR (Surveillance Requirement) 3.4.3.1.2 - This proposed change replaces the statement "When less than 4 reactor coolant pumps are operating and for each cycle of auxiliary spray operation." with "For each cycle of auxiliary spray operation and for each cycle of main spray operation when the RCS cold leg temperature is <

5000F." This change is proposed to reflect that the pressurizer spray cyclic limits are governed by the temperature differentials between the spray nozzle and the spray line. This is consistent with the Updated Final Safety Analysis Report (UFSAR) Section 3.9.1.1.

5.5.2.11 Steam Generator (SG) Tube Surveillance Program, Section 5.5.2.11.f.1 - In 5.5.2.11.f.1.a), an editorial change is proposed to replace "Degradation - A service-induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube;" with "Degradation - A service-induced cracking, wastage, wear, or general corrosion occurring on either the inside or outside of a tube." In 5.5.2.11.f.1.b), it is proposed to correct a.

typographical error in "Degraded tube - A tube containing imperfections greater than or equal to 20% of the nominal wall thickness cause by degradation;" to "Degraded tube - A tube containing imperfections greater than or equal to 20%

of the nominal wall thickness caused by degradation;"

5.5.2.14 Configuration Risk Management Program (CRMP) - It is proposed to delete this section in its entirety and delete associated Bases sections (Attachment G) in accordance with Federal Register Notice Vol. 64, No. 137 (July19, 1999) which recognizes, in Section 11.5 "Regulatory Controls Overlapping Technical Specifications," that the final Maintenance Rule provides requirements that duplicate the CRMP requirements of the Technical Specifications. Federal Register Notice Vol. 64, No. 137 Section III 'The Final Rule" amends 10 CFR 50.65 "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," to incorporate paragraph (a)(4) which states: "Before performing maintenance activities (including but not limited to surveillance, post-maintenance testing and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities." Federal Register Notice Vol. 64, No. 137 Section 11.5 confirms that the revised maintenance rule enables the NRC to expeditiously support licensee requests to remove the CRMP requirements from plant Technical Specifications.

2

5.7.1.5 CORE OPERATING LIMITS REPORT (COLR) - It is proposed to eliminate revision numbers, supplements, dates, and associated correspondence references from the list of topical reports that contain the analytical methods used to determine the core operating limits. For documents other than topical reports that contain the analytical methods used to determine the core operating limits (such as licensing submittals), only the NRC approval letter will be identified in the Technical Specification. The NRC approval letter contains both the NRC safety evaluation and a listing of the documents that were submitted in support of the analytical method. This proposed change is consistent with the NRC approved Industry Technical Specifications Task Force (TSTF) Standard Technical Specifications Traveler number TSTF-363, "Revise Topical Report References in ITS (Improved Technical Specifications) 5.6.5 COLR." Due to the deletion of entries related to the correspondence, the numbering of the entries within Technical Specification 5.7.1.5 changes. This proposed change adds a brief parenthetical description (Cycle 3 SER) to 5.7.1.5.1.b.1 (revised to 5.7.1.5.b.6 in these amendment requests). This proposed change also adds a parenthetical description (Cycle 2 SER) to 5.7.1.5.2.b.1 (revised to 5.7.1.5.b.7 in these amendment requests). Correction of three typographical errors in TS 5.7.1.5 is also proposed in these amendment requests: addition of a quotation mark to 5.7.1.5.a.2, correction of 5.7.1.5.b.1.b.1 letter date from September 5, 1985 to May 16, 1986 (revised to 5.7.1.5.b.6 in these amendment requests), and addition of parentheses closure to 5.7.1.5.b.2.b.1 (revised to 5.7.1.5.b.7 in these amendment requests). No changes to the current analytical methods are being made. Implementation of these proposed changes will have no adverse impact on SCE's practices for controlling the methodologies used to develop the core operating limits for San Onofre Units 2 and 3. The complete citations (i.e., report number, title, revision number, report date, or NRC Safety Evaluation Report (SER) date, and any supplements) for each of the topical reports listed in TS 5.7.1.5 will be displayed as applicable in each station's COLR. NRC review and approval of new or revised topical reports will continue to be obtained in the same manner.

Changes to the COLRs will be controlled by 1 OCFR50.59. The last four pages of TS Section 5 are being re-numbered to accommodate the text reduction in Section 5.7.1.5 from this proposed change.

3.0 BACKGROUND

The proposed changes evolved from revision of the CFR LER requirements from 30 days to 60 days, need to clarify a pressurizer heatup/cooldown Surveillance Requirement, typographical errors, Maintenance Rule CRMP requirements superseding the TS requirements in accordance with Federal Register Notice Vol. 64, No. 137 (July19, 1999), and desire to eliminate TS amendment requests for cited COLR reference revisions consistent with the NRC approved Industry Technical Specifications Task Force (TSTF) Standard Technical Specifications Traveler number TSTF-363, uRevise Topical Report References in ITS (Improved Technical Specifications) 5.6.5 COLR."

3

4.0 TECHNICAL ANALYSIS

The proposed changes were developed to accomplish several San Onofre Units 2 and 3 TS improvements by providing consistency with current 10 CFR 50.73 LER reporting requirements of within 60 days of the violation, clarifying a pressurizer heatup/cooldown Surveillance Requirement to require that the spray water temperature differential is determined for each cycle of auxiliary or main spray operation when the cold leg temperature is below 5000F, incorporation of TS editorial corrections described above, removal of TS redundancy to the Maintenance Rule in accordance with Federal Register Notice Vol. 64, No. 137 (July19, 1999), and elimination of need for TS amendment requests for cited COLR reference revisions consistent with the NRC approved Industry Technical Specifications Task Force (TSTF) Standard Technical Specifications Traveler number TSTF-363, "Revise Topical Report References in ITS (Improved Technical Specifications) 5.6.5 COLR."

5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Southern California Edison (SCE) has concluded that operation of San Onofre Units 2 and 3, in accordance with the proposed changes to the Technical Specifications (TS), do not involve a significant hazards consideration. SCE's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(1), of the three standards set forth in 10 CFR 50.92(c).

1.

Does the proposed change involve a significant Increase in the probability or consequences of an accident previously evaluated?

Southern California Edison (SCE) proposes to modify the San Onofre Units 2 and 3 Technical Specifications (TS) to accomplish several improvements by providing consistency with current Code of Federal Regulations (CFR) Licensee Event Report (LER) reporting requirements, clarifying a pressurizer heatup/cooldown Surveillance Requirement, TS editorial corrections, removing TS redundancy to the Maintenance Rule in accordance with Federal Register Notice Vol. 64, No. 137 (July19, 1999), and eliminating need for TS amendment requests for cited Core Operating Limits Report (COLR) reference revisions consistent with the NRC approved Industry Technical Specifications Task Force (TSTF)

Standard Technical Specifications Traveler number TSTF-363, "Revise Topical Report References in ITS (Improved Technical Specifications) 5.6.5 COLR." These proposed changes do not involve any change in the design or operation of the plant. Therefore, the proposed change 4

does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Modifying the Technical Specifications to provide consistency with current CFR LER reporting requirements, clarify a pressurizer heatup/cooldown Surveillance Requirement, incorporate editorial corrections, remove TS redundancy to the Maintenance Rule in accordance with Federal Register Notice Vol. 64, No. 137 (July19, 1999), and to eliminate need for TS amendment requests for cited COLR reference revisions consistent with the NRC approved Industry Technical Specifications Task Force (TSTF) Standard Technical Specifications Traveler number TSTF-363, uRevise Topical Report References in ITS (Improved Technical Specifications) 5.6.5 COLR" does not involve any change in the design or operation of the plant. Therefore, a possibility of a new or different kind of accident from any accident previously evaluated is not created.

3.

Does the proposed change involve a significant reduction in a margin of safety?

Evaluation of these proposed modifications to the Technical Specifications to provide consistency with current CFR LER reporting requirements, clarify a pressurizer heatup/cooldown Surveillance Requirement, incorporate editorial corrections, remove TS redundancy to the Maintenance Rule in accordance with Federal Register Notice Vol.

64, No. 137 (July19, 1999), and to eliminate need for TS amendment requests for cited COLR reference revisions consistent with the NRC approved Industry Technical Specifications Task Force (TSTF) Standard Technical Specifications Traveler number TSTF-363, "Revise Topical Report References in ITS (Improved Technical Specifications) 5.6.5 COLR" does not involve any change in the design or operation of the plant and therefore does not create any reduction in a margin of safety.

Based on the above, SCE concludes that the proposed amendments present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

5

5.2 Applicable Regulatory Requirements/Criteria (ARR/C)

TS Change ARR/C Discussion 2.2 SL 10 CFR 50.73 This Licensee Event Report (LER) reporting period change from within 30 days of the violation to within 60 days of the violation corresponds with the current Code of Federal Regulations (CFR) requirements.

3.4.3.1 ASME Code.

This text clarification to require that the Section III spray water temperature differential for each cycle of auxiliary or main spray operation when the cold leg temperature is below 500 OF supports the ASME code requirements for the pressurizer.

5.5.2.11 n/a These are editorial changes only.

5.5.2.14 10 CFR 50.65 This deletion of the Configuration Risk Management Program (CRMP) from the TS removes duplication of the requirements of 10 CFR 50.65 which apply for licensee to assess and manage the increase in risk that may result from maintenance activities.

5.7.1.5 10 CFR 50.59 This change eliminates revision numbers, supplements, dates, and associated correspondence references from the list of topical reports in the Core Operating Limits Report (COLR) section of the TS that contain the analytical methods used to determine the core operating limits.

This information will be displayed as applicable in each station's COLR.

Changes to the COLR will be controlled by 10 CFR 50.59 In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

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6.0 Environmental Impact Consideration The proposed changes do not involve a significant hazards consideration, a significant change in the types of or significant increase in the amounts of any effluents that may be released offsite, or a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (c)(10)(ii). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed changes is not required.

PCN533R1 7

PCN 533 Attachment A (Existing Pages)

SONGS Unit 2

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, departure from nucleate boiling ratio (DNBR) shall be maintained at 2 1.31.

2.1.1.2 In MODES 1 and 2, the peak linear heat (adjusted for fuel rod dynamics) shall at

• 21.0 kW/ft.

rate (LHR) be maintained 2.1.2 Reactor Coolant System (RCS) Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained at s 2750 psia.

2.2 SL Violations 2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

2.2.3 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, notify the NRC Operations Center, in accordance with 10 CFR 50.72.

2.2.4 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, notify the Vice President - Nuclear Generation and the Nuclear Safety Group (NSG) Supervisor.

2.2.5 Within 30 days of the violation, a Licensee Event Report (LER) shall be prepared pursuant to 10 CFR 50.73.

The LER shall be submitted to the NRC, the NSG Supervisor, and the Vice President -

Nuclear Generation.

(continued)

SAN ONOFRE--UNIT 2 2.0-1 Amendment No. 127

Pressurizer Heatup/Cooldown Limits 3.4.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. ---------NOTE---------

C.1 Initiate action to Immediately Required Action C.2 restore parameter(s) shall be completed to within limits.

whenever this Condition is entered.

AND C.2 Determine Pressurizer Prior to Requirements of LCO is acceptable for entering MODE 4 not met any time in continued operation.

other than MODE 1, 2, 3, or 4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1.1 ------------------- NOTE--------------------

Only required to be performed during Pressurizer heatup and cooldown operations.

Verify Pressurizer heatup and cooldown rates within the following limits:

30 minutes

a. A maximum heatup of 200'F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period,

b. A maximum cooldown of 200'F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.

SR 3.4.3.1.2 The spray water temperature differential When less than shall be determined for use in the UFSAR.

4 reactor coolant pumps are operating and for each cycle of auxiliary spray operation.

SAN ONOFRE--UNIT 2 3.4-14 Amendment No. 127

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.11 Steam Generator (SG) Tube Surveillance Program (continued)

4.

The provisions of Technical Specifications Surveillance Requirement 3.0.2 are applicable to SG Tube Surveillance inspection frequencies except those established by Category C-3 inspection results.

The above required inservice inspections of SG tubes repaired by sleeving shall be performed at the following frequencies:

1.

Steam generator tube sleeves shall be inspected prior to initial operation and in service.

The initial operating period before the initial inservice sample inspection shall not be shorter than six months nor longer than 24 months.

The inspections of sleeves shall be configured to ensure that each individual sleeve is inspected at least once in 60 months.

2.

If the results of the inservice inspection of SG tube sleeves conducted in accordance with Table 5.5.2.11-2 fall in category C-3, the inspection frequency shall be increased to ensure that each remaining sleeve is inspected at least once in 30 months.

The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria for Category C-1.

f.

Acceptance Criteria

1.

Terms as used in this specification will be defined as follows:

a)

Degradation - A service-induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube; b)

Degraded tube - A tube containing imperfections greater than or equal to 20% of the nominal wall thickness cause by degradation; c)

% Degradation - The percentage of the tube wall thickness affected or removed by degradation; d)

Defect - An imperfection of such severity that it exceeds the repair limit. A tube containing a defect is defective.

(continued)

SAN ONOFRE--UNIT 2 5.0-17 Amendment No. i-72 140

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.12 Ventilation Filter Testing Program (VFTP)

(continued)

The provisions of Technical Specification Surveillance Requirement 3.0.2 and Technical Specification Surveillance Requirement 3.0.3 are applicable to the VFTP test frequencies.

5.5.2.13 Diesel Fuel Oil Testing Program This program implements required testing of both new fuel oil and stored fuel oil.

The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM standards.

The purpose of the program is to establish the following:

a.

Acceptability of new fuel oil 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.

a water and sediment content within limits.

b.

Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to the storage tanks, with exceptions noted in the Bases for Surveillance Requirement 3.8.3.3; and,

c.

Total particulate concentration of fuel oil is s 10 mg/l when tested every 92 days in accordance with ASTM D-2276, Method A.

5.5.2.14 Configuration Risk Management Program (CRMP)

The Confi uration Risk Management Program (CRMP) provides a proceduralized risk-informed assessment to manage the risk associated with equipment inoperability.

The program applies to technical specification structures, systems, or components for which a risk-informed Completion Time has been granted.

The program shall include the following elements:

a.

Provisions for the control and implementation of a Level 1 at power internal events PRA-informed methodology.

The assessment shall be capable of evaluating the applicable plant configuration.

b.

Provisions for performing an assessment prior to entering the LCO Condition for preplanned activities.

c.

Provisions for performing an assessment after entering the LCO Condition for unplanned entry into the LCO Condition.

SAN ONOFRE--UNIT 2 5.0-20 Amendment No. 439,167 l

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.14 Configuration Risk Management Program (CRMP)

(Continued)

d.

Provisions for assessing the need for additional actions after the discovery of additional equipment out of service conditions while in the LCO Condition.

e.

Provisions for considering other applicable risk significant contributors such as Level 2 issues, and external events, qualitatively or quantitatively.

5.5.2.15 Containment Leakage Rate Testing Program 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 modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.

The calculated peak containment internal pressure related to the design basis loss-of-coolant accident, Pa, is 45.9 psig (Pa will conservatively be assumed to be equal to the calculated peak containment internal pressure for the design basis Main Steam Line Break (56.5 psig) for the purpose of containment testing in accordance with this Technical Specification).

The maximum allowable containment leakage rate, La, at Pa, shall be 0.10% of containment air weight per day.

Leakage rate acceptance criteria are:

a.

The Containment overall leakage rate acceptance criterion is s 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are

• 0.60 La for the Type B and Type C tests and

• 0.75 La for the Type A tests;

b.

Air lock testing acceptance criteria are:

1)

Overall air lock leakage rate is

• 0.05 La when tested at 2 Pa-

2)

For each door, the leakage rate is

• 0.01 La when pressurized to 2 9.0 psig.

SAN ONOFRE--UNIT 2 5.0-20a Amendment No. 182 1

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.4 Monthly Operating Reports Routine reports of operating statistics and shutdown experience, including documentation of all challenges to pressurizer safety valves, shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, Attention: Document Control Desk, Washington, D.C., with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.

5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1.

Specification 3.1.4, "Moderator Temperature Coefficient;"

2.

Specification 3.1.5, Control Element Assembly (CEA)

A ignment;"

3.

Specification 3.1.7, "Regulating CEA Insertion Limits;"

4.

Specification 3.1.8, "Part Length Control Element Assembly Insertion Limits;"

5.

Specification 3.2.1, "Linear Heat Rate;"

6.

Specification 3.2.4, "Departure From Nucleate Boiling Ratio;"

7.

Specification 3.2.5, "Axial Shape Index;"

8.

Specification 3.9.1, "Boron Concentration."

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

1.a.1 Letter, dated September 5, 1985, M. 0. Medford (SCE) to G. W. Knighton (NRC), "Docket No. 50-361 and 50-362 Reload Analysis Report," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 3)

(continued)

SAN ONOFRE--UNIT 2 5.0-26 Amendment No. 127

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued) 1.b.1 Letter, dated September 5, 1985, G. W. Knighton (NRC) to K. P. Baskin (SCE), "Issuance of Amendment No. 47 to Facility Operating License NPF-10 and Amendment No. 36 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Methodology for Specifications 3.1.4 for Moderator Temperature Coefficient and 3.9.1 for Boron Concentration) 2.a.1 Letter, dated September 28, 1984, M. 0. Medford (SCE) to G. W. Knighton (NRC), "Reload Analysis Report," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 2) 2.b.1 Letter, dated January 9, 1985, G. W. Knighton NRC) to K. P. Baskin, "Issuance of Amendment No. 30 to Facility Operating License NPF-10 and Amendment No. 19 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Methodology for Specifications 3.1.5 for Control Element Assembly (CEA) Alignment, 3.1.7 for Regulating CEA Insertion Limits, and 3.1.8 for Part Length Control Element Assembly Insertion Limits) 3.a.1 "Calculative Methods for the C-E Large Break LOCA Evaluation Model," CENPD-132P, August 1974 3.a.2 "Calculational Methods for the C-E Large Break LOCA Evaluation Model," CENPD-132P, Supplement 1, February 1975 3.a.3 "Calculational Methods for the C-E Large Break LOCA Evaluation Model," CENPD-132P, Supplement 2-P, July 1975 3.a.4 "Calculative Methods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and W Designed NSSS," CEN-132, Supplement 3-P-A, June 1985 3.b.1 Letter, 0. D. Parr (NRC) to F. M. Stem (CE), dated June 13, 1975 (NRC Staff Review of the Combustion Engineering ECCS Evaluation Model)

(continued)

SAN ONOFRE--UNIT 2 5.0-27 Amendment No. 127

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued) 3.b.2 Letter, O. D. Parr (NRC) to A. E. Scherer (CE), dated December 9, 1975 (NRC Staff Review of the Proposed Combustion Engineering ECCS Evaluation Model Changes)

(Methodology for Specification 3.2.1 for Linear Heat Rate) 4.a.1 "Calculative Methods for the C-E Small Break LOCA Evaluation Model," CENPD-137P, August 1974 4.a.2 "Calculative Methods for the C-E Small Break LOCA Evaluation Model," CENPD-137, Supplement 1-P, January 1977 4.a.3 "Calculative Methods for the ABB C-E Small Break LOCA Evaluation Model," CENPD-137, Supplement 2-P-A, April 1998 4.b.1 Letter, K. Kniel (NRC) to A. E. Scherer (CE), dated September 27, 1977 (Evaluation of Topical Report CENPD-133, Supplement, 3-P and CENPD-137, Supplement 1-P)

(Methodology for Specification 3.2.1 for Linear Heat Rate) 4.b.2 Letter, T. H. Essig (NRC) to I. C. Rickord (ABB),

"Acceptance for Referencing of the Topical Report CENPD-137(P), Supplement, 2, 'Calculative Methods for the C-E Small Break LOCA Evaluation Model' (TAC M95687)," December 16, 1997.

5.

"Modified Statistical Combination of Uncertainties,"

CEN-356(V)-P-A, May 1988 (Methodology for Specifications 3.2.4 for Departure From Nucleate Boiling Ratio, and 3.2.5 for Axial Shape Index) 6.a "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3," SCE-9801-P, November 1998 (continued)

SAN ONOFRE--UNIT 2 5.0-28 Amendment No. +63--168 1

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued) 6.b "Identification of NRC Safety Evaluation Report Limitations and/or Constraints on Reload Analysis Methodology," CEN-635(S), Rev. 00, February 1999 6.c Letter, Stephen Dembek (NRC) to Harold B. Ray (SCE),

dated June 2, 1999, "San Onofre Nuclear Generating Station Units 2 and 3 - Evaluation of Reload Analysis Methodology Technology Transfer (TAC Nos. MA4289 and MA4290)"

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

Not Used Hazardous Cargo Traffic Report Hazardous cargo traffic on Interstate 5 (I-5) and the AT&SF railway shall be monitored and the results submitted to the NRC Regional Administrator once every three years.

5.7.1.6 5.7.1.7 (continued)

SAN ONOFRE--UNIT 2 5.0-29 Amendment No. 1-63, 168

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.2 Special Reports Special Reports may be required covering inspection, test, and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.

Special Reports shall be submitted to the U. S. Nuclear Regulatory Commission, Attention: Document Control Desk, Washington, D. C.

20555, with a copy to the Regional Administrator of the Regional Office of the NRC, in accordance with 10 CFR 50.4 within the time period specified for each report.

The following Special Reports shall be submitted:

a.

When a pre-planned alternate method of monitoring post-accident instrumentation functions is required by Condition B or Condition H of LCO 3.3.11, a report shall be submitted within 30 days from the time the action is required. The report shall outline the action taken, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the function to OPERABLE status.

b.

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.

c.

Following each inservice inspection of steam generator (SG) tubes, in accordance with the SG Tube Surveillance Program, the number of tubes plugged and tubes sleeved in each SG shall be reported to the NRC within 15 days. The complete results of the SG tube inservice inspection shall be submitted to the NRC within 12 months following the completion of the inspection. The report shall include:

1.

Number and extent of tubes and sleeves inspected, and

2.

Location and percent of wall-thickness penetration for each indication of an imperfection, and

3.

Identification of tubes plugged and tubes sleeved.

(continued)

SAN ONOFRE--UNIT 2 5.0-30 Amendment No. +27140

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.2 Special Reports (continued)

Results of SG tube inspections which fall into Category C-3 shall be reported to the NRC prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

SAN ONOFRE--UNIT 2 5.0-31 Amendment No. 127

High Radiation Area 5.8 5.0 ADMINISTRATIVE CONTROLS 5.8 High Radiation Area Each high radiation area as defined in 10 CFR 20 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).

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

I I

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 rates in the area have been determined and personnel have been made knowledgeable of them,

c.

An individual qualified in radiation protection procedures with a radiation dose rate monitoring device.

This individual is responsible for providing positive radiation protection control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the radiation protection procedures or the applicable REP.

(continued)

SAN ONOFRE--UNIT 2 5.0-32 Amendment No. +27, 168 1

High Radiation Area 5.8 5.8.

High Radiation Area (continued) 5.8.2 In addition, areas that are accessible to personnel and that have radiation levels greater than 1.0 rem (but less than 500 rads at 1 meter) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm from the radiation source, or from any surface penetrated by the radiation, shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the shift supervisor on duty or health physics supervisor.

Doors shall remain locked except during periods of access by personnel under an approved REP that specifies the dose rates in the immediate work areas and the maximum allowable stay time for individuals in that area. In lieu of a stay time specification on the REP, direct or remote continuous surveillance (such as closed circuit TV cameras) may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

5.8.3 Individual high radiation areas that are accessible to personnel, that could result in radiation doses greater than 1.0 rem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and that are within large areas where no enclosure exists to enable locking and where no enclosure can be reasonably constructed around the individual area shall be barricaded and conspicuously posted. A flashing light shall be activated as a warning device whenever the dose rate in such an area exceeds or is expected to exceed 1.0 rem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm from the radiation source or from any surface penetrated by the radiation.

SAN ONOFRE--UNIT 2 5.0-33 Amendment No. 127, 168 1

PCN 533 Attachment B (Existing Pages)

SONGS Unit 3

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, departure from nucleate boiling ratio (DNBR) shall be maintained at 2 1.31.

2.1.1.2 In MODES 1 and 2, the peak linear heat (adjusted for fuel rod dynamics) shall at

• 21.0 kW/ft.

rate (LHR) be maintained 2.1.2 Reactor Coolant System (RCS) Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained at

• 2750 psia.

2.2 SL Violations 2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

2.2.3 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, notify the NRC Operations Center, in accordance with 10 CFR 50.72.

2.2.4 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, notify the Vice President - Nuclear Generation and the Nuclear Safety Group (NSG) Supervisor.

2.2.5 Within 30 days of the violation, a Licensee Event Report (LER) shall be prepared pursuant to 10 CFR 50.73. The LER shall be submitted to the NRC, the NSG Supervisor, and the Vice President -

Nuclear Generation.

(continued)

SAN ONOFRE--UNIT 3 2.0-1 Amendment No. 116

Pressurizer Heatup/Cooldown Limits 3.4.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. ---------NOTE---------

C.1 Initiate action to Immediately Required Action C.2 restore parameter(s) shall be completed to within limits.

whenever this Condition is entered.

AND C.2 Determine Pressurizer Prior to Requirements of LCO is acceptable for entering MODE 4 not met any time in continued operation.

other than MODE 1, 2, 3, or 4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1.1 -------------------NOTE--------------------

Only required to be performed during Pressurizer heatup and cooldown operations.

Verify Pressurizer heatup and cooldown rates within the following limits:

30 minutes

a. A maximum heatup of 2000F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period,

b. A maximum cooldown of 200'F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.

SR 3.4.3.1.2 The spray water temperature differential When less than shall be determined for use in the UFSAR.

4 reactor coolant pumps are operating and for each cycle of auxiliary spray operation.

SAN ONOFRE--UNIT 3 3.4-14 Amendment No. 116

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.11 Steam Generator (SG) Tube Surveillance Program (continued)

4.

The provisions of Technical Specifications Surveillance Requirement 3.0.2 are applicable to SG Tube Surveillance inspection frequencies except those established by Category C-3 inspection results.

The above required inservice inspections of SG tubes repaired by sleeving shall be performed at the following frequencies:

1.

Steam generator tube sleeves shall be inspected prior to initial operation and in service.

The initial operating period before the initial inservice sample inspection shall not be shorter than six months nor longer than 24 months. The inspections of sleeves shall be configured to ensure that each individual sleeve is inspected at least once in 60 months.

2.

If the results of the inservice inspection of SG tube sleeves conducted in accordance with Table 5.5.2.11-2 fall in category C-3, the inspection frequency shall be increased to ensure that each remaining sleeve is inspected at least once in 30 months.

The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria for Category C-1.

f.

Acceptance Criteria

1.

Terms as used in this specification will be defined as follows:

a)

Degradation - A service-induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube; b)

Degraded tube - A tube containing imperfections greater than or equal to 20% of the nominal wall thickness cause by degradation; c)

% Degradation - The percentage of the tube wall thickness affected or removed by degradation; d)

Defect - An imperfection of such severity that it exceeds the repair limit. A tube containing a defect is defective.

(continued)

SAN ONOFRE--UNIT 3 5.0-17 Amendment No.

i46 132

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.12 Ventilation Filter Testing Program (VFTP)

(continued)

The provi sions of Technical Specification Surveillance Requirement 3.0.2 and Technical Specification Surveillance Requirement 3.0.3 are applicable to the VFTP test frequencies.

5.5.2.13 Diesel Fuel Oil Testing Program This program implements required testing of both new fuel oil and stored fuel oil.

The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM standards.

The purpose of the program is to establish the following:

a.

Acceptability of new fuel oil 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.

a water and sediment content within limits.

b.

Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to the storage tanks, with exceptions noted in the Bases for Surveillance Requirement 3.8.3.3; and,

c.

Total particulate concentration of fuel oil is

• 10 mg/l when tested every 92 days in accordance with ASTM D-2276, Method A.

5.5.2.14 Configuration Risk Management Program (CRMP)

The Configuration Risk Management Program (CRMP) provides a proceduralized risk-informed assessment to manage the risk associated with equipment inoperability.

The program applies to technical specification structures, systems, or components for which a risk-informed Completion Time has been granted.

The program shall include the following elements:

a.

Provisions for the control and implementation of a Level 1 at power internal events PRA-informed methodology.

The assessment shall be capable of evaluating the applicable plant configuration.

b.

Provisions for performing an assessment prior to entering the LCO Condition for preplanned activities.

c.

Provisions for performing an assessment after entering the LCO Condition for unplanned entry into the LCO Condition.

(continued)

SAN ONOFRE--UNIT 3 5.0-20 Amendment No. +3-1,158 1

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.14 Configuration Risk Management Program (CRMP)

(Continued)

d.

Provisions for assessing the need for additional actions after the discovery of additional equipment out of service conditions while in the LCO Condition.

e.

Provisions for considering other applicable risk significant contributors such as Level 2 issues, and external events, qualitatively or quantitatively.

5.5.2.15 Containment Leakage Rate Testing Program 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 modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program", dated September 1995.

The calculated peak containment internal pressure related to the design basis loss-of-coolant accident, Pa, is 45.9 psig (Pa willl conservatively be assumed to be equal to the calculated peak containment internal pressure for the design basis Main Steam Line Break (56.5 psig) for the pur ose of containment testing in accordance with this Technical Specification).

The maximum allowable containment leakage rate, La, at Pa, shall be 0.10% of containment air weight per day.

Leakage rate acceptance criteria are:

a.

The Containment overall leakage rate acceptance criterion is

• 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are

• 0.60 La for the Type B and Type C tests and

• 0.75 La for the Type A tests;

b.

Air lock testing acceptance criteria are:

1)

Overall air lock leakage rate is

• 0.05 La when tested at 2 Pa-

2)

For each door, the leakage rate is

• 0.01 La when pressurized to 2 9.0 psig.

(continued)

SAN ONOFRE--UNIT 3 5.0-20a Amendment No. 173 1

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.4 Monthly Operating Reports Routine reports of operating statistics and shutdown experience, including documentation of all challenges to pressurizer safety valves, shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, Attention: Document Control Desk, Washington, D.C., with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.

5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1.

Specification 3.1.4, "Moderator Temperature Coefficient;"

2.

Specification 3.1.5, Control Element Assembly (CEA).

Alignment;"

3.

Specification 3.1.7, "Regulating CEA Insertion Limits;"

4.

Specification 3.1.8, "Part Length Control Element Assembly Insertion Limits;"

5.

Specification 3.2.1, "Linear Heat Rate;"

6.

Specification 3.2.4, "Departure From Nucleate Boiling Ratio;"

7.

Specification 3.2.5, "Axial Shape Index;"

8.

Specification 3.9.1, "Boron Concentration."

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

1.a.1 Letter, dated September 5, 1985, M. 0. Medford (SCE) to G. W. Knighton (NRC), "Docket No. 50-361 and 50-362 Reload Analysis Report," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 3)

(continued)

SAN ONOFRE--UNIT 3 5.0-26 Amendment No. 116

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued) 1.b.1 Letter, dated September 5, 1985, G. W. Knighton (NRC) to K. P. Baskin (SCE), "Issuance of Amendment No. 47 to Facility Operating License NPF-10 and Amendment No. 36 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Methodology for Specifications 3.1.4 for Moderator Temperature Coefficient and 3.9.1 for Boron Concentration) 2.a.1 Letter, dated September 28, 1984, M. 0. Medford (SCE) to G. W. Knighton (NRC), "Reload Analysis Report," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 2) 2.b.1 Letter, dated January 9, 1985, G. W. Knighton NRC) to K. P. Baskin, "Issuance of Amendment No. 30 to Facility Operating License NPF-10 and Amendment No. 19 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Methodology for Specifications 3.1.5 for Control Element Assembly (CEA) Alignment, 3.1.7 for Regulating CEA Insertion Limits, and 3.1.8 for Part Length Control Element Assembly Insertion Limits) 3.a.1 "Calculative Methods for the C-E Large Break LOCA Evaluation Model," CENPD-132P, August 1974 3.a.2 "Calculational Methods for the C-E Large Break LOCA Evaluation Model," CENPD-132P, Supplement 1, February 1975 3.a.3 "Calculational Methods for the C-E Large Break LOCA Evaluation Model," CENPD-132P, Supplement 2-P, July 1975 3.a.4 "Calculative Methods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and W Designed NSSS," CEN-132, Supplement 3-P-A, June 1985 3.b.1 Letter, 0. D. Parr (NRC) to F. M. Stem (CE),

dated June 13, 1975 (NRC Staff Review of the Combustion Engineering ECCS Evaluation Model)

(continued)

SAN ONOFRE--UNIT 3 5.0-27 Amendment No. 116

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued) 3.b.2 Letter, 0. D. Parr (NRC) to A. E. Scherer (CE), dated December 9, 1975 (NRC Staff Review of the Proposed Combustion Engineering ECCS Evaluation Model Changes)

(Methodology for Specification 3.2.1 for Linear Heat Rate) 4.a.1 "Calculative Methods for the C-E Small Break LOCA Evaluation Model," CENPD-137P, August 1974 4.a.2 "Calculative Methods for the C-E Small Break LOCA Evaluation Model," CENPD-137, Supplement 1-P, January 1977 4.a.3 "Calculative Methods for the ABB C-E Small Break LOCA Evaluation Model," CENPD-137,-Supplement 2-P-A, April 1998 4.b.1 Letter, K. Kniel (NRC) to A. E. Scherer (CE), dated September 27, 1977 (Evaluation of Topical Report CENPD-133, Supplement, 3-P and CENPD-137, Supplement 1-P)

(Methodology for Specification 3.2.1 for Linear Heat Rate) 4.b.2 Letter, T. H. Essig (NRC) to I. C. Rickord (ABB),

"Acceptance for Referencing of the Topical Report CENPD-137(P), Supplement, 2, 'Calculative Methods for the C-E Small Break LOCA Evaluation Model' (TAC M95687)," December 16, 1997.

5.

"Modified Statistical Combination of Uncertainties,"

CEN-356(V)-P-A, May 1988 (Methodology for Specifications 3.2.4 for Departure From Nucleate Boiling Ratio, and 3.2.5 for Axial Shape Index) 6.a "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3," SCE-9801-P, November 1998 (continued)

SAN ONOFRE--UNIT 3 5.0-28 Amendment No. +454-159 1

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued) 6.b "Identification of NRC Safety Evaluation Report Limitations and/or Constraints on Reload Analysis Methodology," CEN-635(S), Rev. 00, February 1999 6.c Letter, Stephen Dembek (NRC) to Harold B. Ray (SCE),

dated June 2, 1999, "San Onofre Nuclear Generating Station Units 2 and 3 - Evaluation of Reload Analysis Methodology Technology Transfer (TAC Nos. MA4289 and MA4290)"

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

Not Used Hazardous Cargo Traffic Report Hazardous cargo traffic on Interstate 5 (I-5) and the AT&SF railway shall be monitored and the results submitted to the NRC Regional Administrator once every three years.

5.7.1.6 5.7.1.7 (continued)

SAN ONOFRE--UNIT 3 5.0-29 Amendment No. 154-,159

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.2 Special Reports Special Reports may be required covering inspection, test, and maintenance activities.

These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.

Special Reports shall be submitted to the U. S. Nuclear Regulatory Commission, Attention: Document Control Desk, Washington, D. C.

20555, with a copy to the Regional Administrator of the Regional Office of the NRC, in accordance with 10 CFR 50.4 within the time period specified for each report.

The following Special Reports shall be submitted:

a.

When a pre-planned alternate method of monitoring post-accident instrumentation functions is required by Condition B or Condition H of LCO 3.3.11, a report shall be submitted within 30 days from the time the action is required. The.

report shall outline the action taken, the cause of the -

inoperability, and the plans and schedule for restoring the instrumentation channels of the function to OPERABLE status.

b.

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.

c.

Following each inservice inspection of steam generator (SG) tubes, in accordance with the SG Tube Surveillance Program, the number of tubes plugged and tubes sleeved in each SG shall be reported to the NRC within 15 days. The complete results of the SG tube inservice inspection shall be submitted to the NRC within 12 months following the completion of the inspection. The report shall include:

1.

Number and extent of tubes and sleeves inspected, and

2.

Location and percent of wall-thickness penetration for each indication of an imperfection, and

3.

Identification of tubes plugged and tubes sleeved.

(continued)

SAN ONOFRE--UNIT 3 5.0-30 Amendment No. +-1 132

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.2 Special Reports (continued)

Results of SG tube inspections which fall into Category C-3 shall be reported to the NRC prior to resumption of plant operation.

This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

SAN ONOFRE--UNIT 3 5.0-31 Amendment No. 116

High Radiation Area 5.8 5.0 ADMINISTRATIVE CONTROLS 5.8 High Radiation Area 5.8.1 Each high radiation area as defined in 10 CFR 20 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)

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 rates in the area have been determined and personnel have been made knowledgeable of them,

c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device. This individual is responsible for providing positive radiation protection control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the radiation protection procedures or the applicable REP.

(continued)

SAN ONOFRE--UNIT 3 5.0-32 Amendment No. +-6, 159 1

High Radiation Area 5.8 5.8.

High Radiation Area (continued) 5.8.2 In addition, areas that are accessible to personnel and that have radiation levels greater than 1.0 rem (but less than 500 rads at 1 meter) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm from the radiation source, or from any surface penetrated by the radiation, shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the shift supervisor on duty or health physics supervisor. Doors shall remain locked except during periods of access by personnel under an approved REP that specifies the dose rates in the immediate work areas and the maximum allowable stay time for individuals in that area. In lieu of a stay time specification on the REP, direct or remote continuous surveillance (such as closed circuit TV cameras) may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

5.8.3 Individual high radiation areas that are accessible to personnel, that could result in radiation doses greater than 1.0 rem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and that are within large areas, where no enclosure exists to enable locking and where no enclosure can be reasonably constructed around the individual area shall be barricaded and conspicuously posted. A flashing light shall be activated as a warning device whenever the dose rate in such an area exceeds or is expected to exceed 1.0 rem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm from the radiation source or from any surface penetrated by the radiation.

SAN ONOFRE--UNIT 3 5.0-33 Amendment No. +1-6, 159 1

PCN 533 Attachment C (Proposed Pages)

(Redline and Strikeout)

SONGS Unit 2

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, departure from nucleate boiling ratio (DNBR) shall be maintained at 2 1.31.

2.1.1.2 In MODES 1 and 2, the peak linear heat (adjusted for fuel rod dynamics) shall at

• 21.0 kW/ft.

rate (LHR) be maintained 2.1.2 Reactor Coolant System (RCS) Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained at

• 2750 psia.

2.2 SL Violations 2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

2.2.3 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, notify the NRC Operations Center, in accordance with 10 CFR 50.72.

2.2.4 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, notify the Vice President - Nuclear Generation and the Nuclear Safety Group (NSG) Supervisor.

2.2.5 Within 360 days of the violation, a Licensee Event Report (LER) shall be prepared pursuant to 10 CFR 50.73.

The LER shall be submitted to the NRC, the NSG Supervisor, and the Vice President -

Nuclear Generation.

(continued)

SAN ONOFRE--UNIT 2 2.0-1 Amendment No. 14 Pressurizer Heatup/Cooldown Limits 3.4.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. --------- NOTE---------

C.1 Initiate action to Immediately Required Action C.2 restore parameter(s) shall be completed to within limits.

whenever this Condition is entered.

AND C.2 Determine Pressurizer Prior to Requirements of LCO is acceptable for entering MODE 4 not met any time in continued operation.

other than MODE 1, 2, 3, or 4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1.1 -------------------NOTE--------------------

Only required to be performed during Pressurizer heatup and cooldown operations.

Verify Pressurizer heatup and cooldown rates within the following limits:

30 minutes

a. A maximum heatup of 200'F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period,

b. A maximum cooldown of 2000F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.

SR 3.4.3.1.2 The spray water temperature differential When less than shall be determined for use in the UFSAR.

4 reaet&r are operatin and fFor each cycle"of auxiliary cnray pnPraqtjnnjUAndr fnr.porh rvrle nnf main qnrav' hnPrat inn when PIn.tomnprature is < 500AF.

SAN ONOFRE--UNIT 2 3.4-14 Amendment No. 4L7-

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.11 Steam Generator (SG) Tube Surveillance Program (continued)

4.

The provisions of Technical Specifications Surveillance Requirement 3.0.2 are applicable to SG Tube Surveillance inspection frequencies except those established by Category C-3 inspection results.

The above required inservice inspections of SG tubes repaired by sleeving shall be performed at the following frequencies:

1.

Steam generator tube sleeves shall be inspected prior to initial operation and in service. The initial operating period before the initial inservice sample inspection shall not be shorter than six months nor longer than 24 months.

The inspections of sleeves shall be configured to ensure that each individual sleeve is inspected at least once in 60 months.

2.

If the results of the inservice inspection of SG tube sleeves conducted in accordance with Table 5.5.2.11-2 fall in category C-3, the inspection frequency shall be increased to ensure that each remaining sleeve is inspected at least once in 30 months.

The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria for Category C-1.

f.

Acceptance Criteria

1.

Terms as used in this specification will be defined as follows:

a)

Degradation - A service-induced cracking, wastang wear, or general corrosion occurring on either the inside or outside of a tube; b)

Degraded tube - A tube containing imperfections greater than oreequal to 20% of the nominal wall thickness caused by degradation; c)

% Degradation - The percentage of the tube wall thickness affected or removed by degradation; d)

Defect - An imperfection of such severity that it exceeds the repair limit. A tube containing a defect is defective.

(continued)

SAN ONOFRE--UNIT 2 5.0-17 Amendment No. 127 140

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.12 Ventilation Filter Testing Program (VFTP)

(continued)

The provisions of Technical Specification Surveillance Requirement 3.0.2 and Technical Specification Surveillance Requirement 3.0.3 are applicable to the VFTP test frequencies.

5.5.2.13 Diesel Fuel Oil Testing Program This program implements required testing of both new fuel oil and stored fuel oil.

The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM standards.

The purpose of the program is to establish the following:

a.

Acceptability of new fuel oil 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.

a water and sediment content within limits.

b.

Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to the storage tanks, with exceptions noted in the Bases for Surveillance Requirement 3.8.3.3; and,

c.

Total particulate concentration of fuel oil is

• 10 mg/l when tested every 92 days in accordance with ASTM D-2276, Method A.

5.5.2.14 Deleted~enfiguration Risk Mlanagement Program (CRMP)P The Confi uraticn Risk Management Program (CRMP) prAvides a prcczdriT2 d

isk - inf a rmed assessment tW manag the rIsk asskciated with cquipment inoperability.

The program applis ta include the fellatin el ements s

a.

Provisions for the contral and implomentation af a Level 1 at power internal events PRA i-formed methodology.

The sAssessman shl bz capabll of evaluatn th)

.. 1

.ee t

_i: baplanrt configuration.

b.

Provision; for performing an assessment prior to entering the LC Canditgs:

fa prclann d act:+iiAs.

H-9

-lu5I I

-1 I ViX I

U ts ft J

UlY U

L I

r X

a o

r.

a n f a

nussessmnt after entering the Lfu Condition for unplanned entry into the LCO Condition.

(continued)

SAN ONOFRE--UNIT 2 5.0-20 Amendment No. 139,167

Procedures, Programs, and Manuals 5.5 (continued) 5.5 Procedures, Programs, and Manuals 5.5.2.14 Configuration Risk Management Program (CRMP)

(Continued)

d.

frovisians fr assess..

thce need for additional aetins after

+

tO~~~~~~~I A ;-i

_W ~

4

_A.__..s m 42+_1A A&

tLI di.

,avery of additi.,ana oqimn u f seryice canditions while in the LCO Condition.

e.

Prvisins-for considering other applicable risk significant contributors such as Level 2 issues, and external events, qualitatively or quantitatively.

5.5.2.15 Containment Leakage Rate Testing Program 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 modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.

The calculated peak containment internal pressure related to the design basis loss-of-coolant accident, Pas is 45.9 psig (Pa will conservatively be assumed to be equal to the calculated peak containment internal pressure for the design basis Main Steam Line Break (56.5 psig) for the purpose of containment testing in accordance with this Technical Specification).

The maximum allowable containment leakage rate, La, at Pa, shall be 0.10% of containment air weight per day.

Leakage rate acceptance criteria are:

a.

The Containment overall leakage rate acceptance criterion is

• 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are

• 0.60 La for the Type B and Type C tests and

• 0.75 La for the Type A tests;

b.

Air lock testing acceptance criteria are:

1)

Overall air lock leakage rate is

• at 2 Pa-

2)

For each door, the leakage rate is pressurized to 2 9.0 psig.

0.05 La when tested

• 0.01 La when (continued)

SAN ONOFRE--UNIT 2 5.0-20a Amendment No. +8R

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.4 Monthly Operating Reports Routine reports of operating statistics and shutdown experience, including documentation of all challenges to pressurizer safety valves, shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, Attention: Document Control Desk, Washington, D.C., with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.

5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1.

Specification 3.1.4, "Moderator Temperature Coefficient;"

2.

Specification 3.1.5,,FControl Element Assembly (CEA)

Alignment;"

3.

Specification 3.1.7, "Regulating CEA Insertion Limits;"

4.

Specification 3.1.8, "Part Length Control Element Assembly Insertion Limits;"

5.

Specification 3.2.1, "Linear Heat Rate;"

6.

Specification 3.2.4, "Departure From Nucleate Boiling Ratio;"

7.

Specification 3.2.5, "Axial Shape Index;"

8.

Specification 3.9.1, "Boron Concentration."

b.

The analytical methods used to determine the core operating 1 imitscch l,nl_.h,.tnhndnrpvin nle lv,,revei PWpH.,nH -a nnrnvpH-,hv, t he NRC-_specifically those described in1 the following-documents:

1.a.1 Letter, dated S7EptLImer 5, 1985, Pi. 0. MIedford (ICE) tc C. W. Knight-n (NRC), "Dcket No. 50 361 and 50 362 Reload Analysis-Report," San Onefre Nuclear Cenerating Station Units 2 and 3 (Cyele 3)

(continued)

SAN ONOFRE--UNIT 2 5.0-26 Amendment No. 1.7

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued) vl-b-I Letter, dated May,16, 1986aeptember 5, 1985, G. W.

Knighton (NRC)"to-K. P. Baskin (SCE), "Issuance of Amendment No. 47 to Facility Operating License NPF-10 and Amendment No. 36 to Facility Operating License NPF-15LSan-nnnfrp Nuclear Generating Station Units 2 and 3,(Cycle 3 SER)

(Methadology for Speaifications 3.1.4 for Moderater T-mporature Coefficient and 3.9.1 far Boren conee.lR.& t~enW6) 2.a.1 Letter, dated September 28. 1984, M. 0. Medford (SCE) to n -^ r-d cl_m^_no Mn It Pn 1^ldi:

{err L&LL U

U J&

U ftH.U.

1 uv~

C. W. Knighton (NRC), "Reload Analysis Report," San Onofre Nuclar Cenerating StatiAn Units 2 and 3 (Gyele-2+

72.b.1 Letter, dated January 9, 1985, G. W. Knighton rNRC) to K. P. Baskin, "Issuance of Amendment No. 30 to Facility Operating License NPF-10 and Amendment No. 19 to Facility Operating License NPF-15,...`5an-nnnfrP Nuclear Generating Station Units 2 and 3 (Cycle2 SER),

(Methodology for Specifications 3.1.5 for Control Element Assemmblly (CEA) Ali4gnment, 3.1.7 for Reultn CEA nsetion Limits, and 3.1.8 for Part LengthCnta 7-fr. CENPD-132P, "Calculative Methods for the C-E Large Break LOCA EV~lti6tion Model," GENPD 132PAugts-t-1-97-4 3.a.2 "Caleulational Mlethods for the C E Large Break LOCA Evaluation Made!," CENPB 132P, Supplement 1, Fbruary 3.a.3 "Caleulatinal Mlathds for the C E Large Break L rCA Evaluation Moddel,"

CENP 132P, SupplAment 2 P, July.74 3.a.4 "Calculative Methods for the C E Large Break LOCA Evaluation Model far the Analysis of C E and W Designed mss,"e GEN 132, Supplement 3 P A, June 1985 3.b.1 Letter, 0. D. Parr (NRC) to F. M. Stem (CE),

dated June (continued)

SAN ONOFRE--UNIT 2 5.0-27 Amendment No. +2*

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued) 3.b.2 Letter, 0. D. Parr (NRC) to A. C. Scherer (CE), dated Dccember 9, 1975 (NRC Staff Review of the Proposed Combustion Engineering ECCS Evaluation Model Changes)

(fl%,*adolagy for Speeifieatien 3.2.1 for Linear Heat 2.4.a.1E CNPD-137P,' "Calculative Methods for the C-E Small Break LOCA"Evalu-ation Model7 "GENPD-137P, August 1974 4.a.2 "Calculative Methods for the C-Small Break LOCA Evaluation Model," CENPP 137, Supplement 1 P. January

%I-rnI 4.a.3 "Calculative Methods for the AGB C-Small Dreak LOCA Evaluatien Madel," CENPD 137, Supplement 2 P A, April 4.b.1 Letter, K. Kniel (NIRC) to A. E. Seherer (CE), dated September 27, 1977 Ivauatr I an of T CENP9 13.3, Supplement, 3-P. and. CEN;D 17 ppTmntt (1tlthodolegy for Specification 3.2.1 for Linear lleat Rate) 4.b.2 Letter, T. H. Essig (NRC) to I. C. Riekard (ABD),

11 A^ -^ -r -~ -

l ^T n rmCnP-3()

Spplement,,

CluaieMtodfr the C-[ Small Break LOCA [valuation Madeal' (TAC M968 eember 16, 1997.

E5r.

CEN-356(V)-P-A,: "Modified Statistical Combination of Ucrainties," GEN 356(Y) P A,11ay 1988 (Methadalagy for Specifications 3.2.4 far Departure From Nucleata CAiling Ratio, and 3.2.5 far Axial Shape Index)

.4E~- SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuciear Generating Station Units 2 and 3,"

SCE 9801 P, Navember 1998 (continued)

SAN ONOFRE--UNIT 2 5.0-28 Amendment No. 163,168

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued) 5.-6v-b CEN-635(S), "Identification of NRC Safety Evaluation Report Limitations and or Constraints on Reload Analysis 6.e Lettar, Stephen Dembv k (NRC) to Harald B. Ray (SCE),

datM d JunA 2, 1999, "San Onafre Nuelear A

rneratinm Station Units 2 and 3 Evaluation of Reload Analysis Methodology Tchnology Transfer (TAC Nos. MA4289 and MA4290O)

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.7.1.6 Not Used 5.7.1.7 Hazardous Cargo Traffic Report Hazardous cargo traffic on Interstate 5 (1-5) and the AT&SF railway shall be monitored and the results submitted to the NRC Regional Administrator once every three years.

(continued)

SAN ONOFRE--UNIT 2 5.0-29 Amendment No. 163, 168

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.2 Special Reports Special Reports may be required covering inspection, test, and maintenance activities.

These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.

Special Reports shall be submitted to the U. S. Nuclear Regulatory Commission, Attention: Document Control Desk, Washington, D. C.

20555, with a copy to the Regional Administrator of the Regional Office of the NRC, in accordance with 10 CFR 50.4 within the time period specified for each report.

The following Special Reports shall be submitted:

a.

When a pre-planned alternate method of monitoring post-accident instrumentation functions is required by Condition B or Condition H of LCO 3.3.11, a report shall be submitted within 30 days from the time the action is required. The report shall outline the action taken, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the function to OPERABLE status.

b.

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.

c.

Following each inservice inspection of steam generator (SG) tubes, in accordance with the SG Tube Surveillance Program, the number of tubes plugged and tubes sleeved in each SG shall be reported to the NRC within 15 days. The complete results of the SG tube inservice inspection shall be submitted to the NRC within 12 months following the completion of the inspection. The report shall include:

1.

Number and extent of tubes and sleeves inspected, and

2.

Location and percent of wall-thickness penetration for each indication of an imperfection, and

3.

Identification of tubes plugged and tubes sleeved.

(continued)

SAN ONOFRE--UNIT 2 5.0-30 Amendment No. 172140

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.2 Special Reports (continued)

Results of SG tube inspections which fall into Category C-3 shall be reported to the NRC prior to resumption of plant operation.

This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

SAN ONOFRE--UNIT 2 5.0-31 Amendment No. 1.-7

High Radiation Area 5.8 5.0 ADMINISTRATIVE CONTROLS 5.8 High Radiation Area 5.8.1 Each high radiation area as defined in 10 CFR 20 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).

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 rates in the area have been determined and personnel have been made knowledgeable of them,

c.

An individual qualified in radiation protection procedures with a radiation dose rate monitoring device.

This individual is responsible for providing positive radiation protection control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the radiation protection procedures or the applicable REP.

(continued)

SAN ONOFRE--UNIT 2 5.0-32 Amendment No. 127, 168

High Radiation Area 5.8 5.8.

High Radiation Area (continued) 5.8.2 In addition, areas that are accessible to personnel and that have radiation levels greater than 1.0 rem (but less than 500 rads at 1 meter) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm from the radiation source, or from any surface penetrated by the radiation, shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the shift supervisor on duty or health physics supervisor.

Doors shall remain locked except during periods of access by personnel under an approved REP that specifies the dose rates in the immediate work areas and the maximum allowable stay time for individuals in that area. In lieu of a stay time specification on the REP, direct or remote continuous surveillance (such as closed circuit TV cameras) may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

5.8.3 Individual high radiation areas that are accessible to personnel, that could result in radiation doses greater than 1.0 rem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and that are within large areas where no enclosure exists to enable locking and where no enclosure can be reasonably constructed around the individual area shall be barricaded and conspicuously posted. A flashing light shall be activated as a warning device whenever the dose rate in such an area exceeds or is expected to exceed 1.0 rem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm from the radiation source or from any surface penetrated by the radiation.

SAN ONOFRE--UNIT 2 5.0-33 Amendment No. 127, 168

PCN 533 Attachment D (Proposed Pages)

(Redline and Strikeout)

SONGS Unit 3

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, departure from nucleate boiling ratio (DNBR) shall be maintained at 2 1.31.

2.1.1.2 In MODES 1 and 2, the peak linear heat rate (LHR)

(adjusted for fuel rod dynamics) shall be maintained at

• 21.0 kW/ft.

2.1.2 Reactor Coolant System (RCS) Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained at

• 2750 psia.

2.2 SL Violations 2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

2.2.3 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, notify the NRC Operations Center, in accordance with 10 CFR 50.72.

2.2.4 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, notify the Vice President - Nuclear Generation and the Nuclear Safety Group (NSG) Supervisor.

2.2.5 Within 360 days of the violation, a Licensee Event Report (LER) shall be prepared pursuant to 10 CFR 50.73. The LER shall be submitted to the NRC, the NSG Supervisor, and the Vice President -

Nuclear Generation.

(continued)

SAN ONOFRE--UNIT 3 2.0-1 Amendment No. 4+6

Pressurizer Heatup/Cooldown Limits 3.4.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. ---------NOTE---------

C.1 Initiate action to Immediately Required Action C.2 restore parameter(s) shall be completed to within limits.

whenever this Condition is entered.

AND C.2 Determine Pressurizer Prior to Requirements of LCO is acceptable for entering MODE 4 not met any time in continued operation.

other than MODE 1, 2, 3, or 4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY.

SR 3.4.3.1.1

---

NOTE--------------------

Only required to be performed during Pressurizer heatup and cooldown operations.

Verify Pressurizer heatup and cooldown rates within the following limits:

30 minutes

a. A maximum heatup of 2000F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period,

b. A maximum cooldown of 2000F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.

SR 3.4.3.1.2 The spray water temperature differential When less than shall be determined for use in the UFSAR.

4 reaetor eeeelant pumps are -&erating and fFor each cycle'-of auxiliary,nray nnprati nn nt, ifpr p-6arh rvrle

-nf main cnrav, innPrt inn when

,thp RrC nd..

Pn tpmnprature

_s < 5000F.

SAN ONOFRE--UNIT 3 3.4-14 Amendment No. 44-6

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.11 Steam Generator (SG) Tube Surveillance Program (continued)

4.

The provisions of Technical Specifications Surveillance Requirement 3.0.2 are applicable to SG Tube Surveillance inspection frequencies except those established by Category C-3 inspection results.

The above required inservice inspections of SG tubes repaired by sleeving shall be performed at the following frequencies:

1.

Steam generator tube sleeves shall be inspected prior to initial operation and in service.

The initial operating period before the initial inservice sample inspection shall not be shorter than six months nor longer than 24 months. The inspections of sleeves shall be configured to ensure that each individual sleeve is inspected at least once in 60 months.

2.

If the results of the inservice inspection of SG tube sleeves conducted in accordance with Table 5.5.2.11-2 fall in category C-3, the inspection frequency shall be increased to ensure that each remaining sleeve is inspected at least once in 30 months.

The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria for Category C-1.

f.

Acceptance Criteria

1.

Terms as used in this specification will be defined as follows:

a)

Degradation - A service-induced cracking, wastane.

wear, or general corrosion occurring on either, the inside or outside of a tube; b)

Degraded tube - A tube containing imperfections greater than orrequal to 20% of the nominal wall thickness caused by degradation; c)

% Degradation - The percentage of the tube wall thickness affected or removed by degradation; d)

Defect - An imperfection of such severity that it exceeds the repair limit. A tube containing a defect is defective.

(continued)

SAN ONOFRE--UNIT 3 5.0-17 Amendment No. 116 132

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.12 Ventilation Filter Testing Program (VFTP)

(continued)

The provisions of Technical Specification Surveillance Requirement 3.0.2 and Technical Specification Surveillance Requirement 3.0.3 are applicable to the VFTP test frequencies.

5.5.2.13 Diesel Fuel Oil Testing Program This program implements required testing of both new fuel oil and stored fuel oil.

The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM standards.

The purpose of the program is to establish the following:

a.

Acceptability of new fuel oil 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.

a water and sediment content within limits.

b.

Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to the storage tanks, with exceptions noted in the Bases for Surveillance Requirement 3.8.3.3; and,

c.

Total particulate concentration of fuel oil is

• 10 mg/l when tested every 92 days in accordance with ASTM D-2276, Method A.

5.5.2.14 DeletedGenfiguration Risk Management Program (CRGP)

ThzC-rf_6iguration Risk Managemaent Pr.gram (const pfid-c _

proeeduralized risk infermed asscssmcnt to manage th} risk associated with equipment inoperability.

The program applies to technical specificatian structures, systems, or componts for which a risk-infarmed Completion Time has been granted.

The program shall include the following elements:

a.

Provisions for the control and implemontation of a Level 1 at power internal events PRA-infrmed methodology.

The assessment shall be capable of evaluating the applicable plant

b.

P Conitions fa pprforming an assessmcnt prior to entering the LCCnita for preplanned activities.

c.Provisions fo pefrmn an%.

asseIIJ"

-ssment after entering the LGO Condition for unplanned entry into the LCD Condition.

(continued)

SAN ONOFRE--UNIT 3 5.0-20 Amendment No. 131,-158

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.14 Configuration Risk Mlanagement Program (CRPIP)

(Continued)

d.

Provisions for assessing the need for additional actions after thC discavery of additional eaui mprn cu Alit of crvIC canditians while in theLC Cadtin PrAvisins for considering other applicab.e risk significant cantributars sueh as Level 2 issues, and. external events, qualitatively or quantitatively-.

5.5.2.15 Containment Leakage Rate Testing Program 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 modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program", dated September 1995.

The calculated peak containment internal pressure related to the design basis loss-of-coolant accident, Pa, is 45.9 psig (Pa will conservatively be assumed to be equal to the calculated peak containment internal pressure for the design basis Main Steam Line Break (56.5 psig) for the purpose of containment testing in accordance with this Technical Specification).

The maximum allowable containment leakage rate, La, at P8nshall be 0.10% of containment air weight per day.

Leakage rate acceptance criteria are:

a.

The Containment overall leakage rate acceptance criterion is 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are

• 0.60 La for the Type B and Type C tests and

• 0.75 La for the Type A tests;

b.

Air lock testing acceptance criteria are:

1)

Overall air lock leakage rate is

• at 2 Pa-

2)

For each door, the leakage rate is pressurized to 2 9.0 psig.

0.05 La when tested

• 0.01 L. when (continued)

SAN ONOFRE--UNIT 3 5.0-20a Amendment No. 473

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.4 Monthly Operating Reports Routine reports of operating statistics and shutdown experience, including documentation of all challenges to pressurizer safety valves, shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, Attention: Document Control Desk, Washington, D.C., with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.

5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1.

Specification 3.1.4, "Moderator Temperature Coefficient;"

2.

Specification 3.1.5, F"Control Element Assembly (CEA):

Alignment;"

3.

Specification 3.1.7, "Regulating CEA Insertion Limits;"

4.

Specification 3.1.8, "Part Length Control Element Assembly Insertion Limits;"

5.

Specification 3.2.1, "Linear Heat Rate;"

6.

Specification 3.2.4, "Departure From Nucleate Boiling Ratio;"

7.

Specification 3.2.5, "Axial Shape Index;"

8.

Specification 3.9.1, "Boron Concentration."

b.

The analytical methods used to determine the core operating 1

AMitnd ha 1,.hpjhnrp_nrAPvim I vrevipwprl,;nr nnrnvpd.hv the NRC, specifically those described in Sthe following documents:

1.a.1 Letter, dated September 5, 1985, M. 0. Medford (SCE) to C. W. Knighton (N.RC), "Docket No. 50 361 and 50 362 Reload Analysis Report," San Onofre Nuclear Generating c_ :_ -L to r.,14.

o f ta ZT lakan UNit 2 alud 3 CYele 3)

(continued)

SAN ONOFRE--UNIT 3 5.0-26 Amendment No. i+16

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued) 6.1.b.1 Letter, dated May 16, 1986 eptemkr 5, 1985, G. W.

Knighton (NRC) to K. Pf.Baskin (SCE), "Issuance of Amendment No. 47 to Facility Operating License NPF-10 and Amendment No. 36 to Facility Operating License NPF-15 2'2Sn.;nnnfrp Nuclear Generating Station Units 2 and 3-(Cycle 3 SER)

(llethodology for Speeifieations 3.1.4 for Moderator Temperatur Coie fieiAnt and 3.9.1 for Boren GoneentatuU iUo n J)l vvu&

U Pvsa T: _: NJ "1

71 t

ORAp SAD!S Letto,8; dated Sapteff mbr 28, 194 MlUJ.v.

0.

I idfr (lCE) t GC. U4. Kn ailhtn 6wil C, §"Relad Analysis Reot"San Onrfr Nluclear Generating tai Ut2 and 01 (eye!1-e2) b u.-

Letter, dated January 9, 1985, G. W. Knighton,NRC) to K. P. Baskin, "Issuance of Amendment No. 30 to Facility Operating License NPF-10 and Amendment No. 19 to Facility Operating License NPF-15_1.__n-fnnnfre Nuclear Generating Station Units 2 and 3 [(Cycle,2 SER)

(Methodology for Specifications 3.1.5 for Control Element Assembly (CCA)

Alignment, 3.1.7 for Regulating CEA insertion LimitsLI, andu 1.8.U fuo art LPngth Control Element Assembly sT.lA Limits) 3-.a-f.

CENPD-132Pj, "Calculative Methods for the C-E Large Break LOCA Evaluation Mode1,"CENP9-132P, August 1974 3.a.2 "Calculational Mtethods for the C E Large Break LOCA Evaluation Mlodel," CENPD 132P, Supplement 1, February i75 3.a.3 "Calculational Methods for the C E Large Break LOCA Evaluation Model," CENPD 132P, Supplement 2 P, July 1975 3.a.4 "Caleulative Mcthods for the C E Large Break LOCIA Evaluation Model for the Analysis of C E and W Designed NSSS," CEN 132, Supplement 3 P A, June 1985 3.b.1 Letter, 0. D. Parr (NRC) to F. M. Stem (CE), dated June (continued)

SAN ONOFRE--UNIT 3 5.0-27 Amendment No. 6

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued) 3.b.2 Letter, 0. D. Parr (NRC) to A. E. Scherer (GE), dated December 9, 1975 (NRC Staff Review of the Proposed Combustion Engineering ECCS Evaluatian Model Changes)

(Methodology for Specification 3.2.1 for Linear Heat 2A.a.1 a

CENPD-137P, "Calculative Methods for the C-E Small Break

~

LOCA Evaluation ModelT"GENFID 137P, August 1974 4.a.2 "Caleulative Mlethads far Evaluation Moadel," CENPD

+9H the C E Small Break LOCA 137, Supplement 1 PL January 4.a.3 "Calculative Metheds for the ABB C-C Small Break LOCA Evaluation Model,." GENP9 13?, Supplement 2 P A, April

+998 4.b.1 Letter, K. Kniel (NRC) to A. E. Scherer (GE),

dated September 27, 1977 (Evaluation of Topical Report CENP9 133, Supplement, 3 P and CENIP 137, Supplement (Mflthadology for Specification 3.2.1 for Linear Heat Rate) 4.b.2 Letter, T. II.

Essig (NRC) to I. C. Rickard (ABD),

"Aeeeptanee for Refereneing of the Topic-al Deo ND 13(P), Supplement,,, Calculative Methods fo the C C Small Break LOCA Evaluatian l odel' (TAC 195687)," December 16, 1997.

B6.

,CEN-356(V)-P-Ai "Modified Statistical Combination of Unceftaintie6s"CEN 356(V) P A, May 1988 (Methodology for Spceifications 3.2.4 fAr Departure From Nulceate Boiling Ratio, and 3.2.5 for Axial Shape Index) 6--.6 SCE-9801-P7A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3-,"

SCE-9801eP, November 1998 (continued)

SAN ONOFRE--UNIT 3 5.0-28 Amendment No. 154,1'9

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued) 5.6-b CEN-635(S), "Identification of NRC Safety Evaluation Report Limitations and/or Constraints on Reload Analysis Methodology,"' C N-63(S), Rev. 00, February 1999 6.e Letter, Stephen Dembek (NRC) to Hlareld B. Ray (SCE),

dated June 2, 1999, "San Onfre Nula Generatiny Station Units 2 and 3 _ Evaluation of Reload Analysis Methodolo-TacnoogyTrnsfer (TAG Nos. MA4289 and MA429W*

C.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.7.1.6 Not Used 5.7.1.7 Hazardous Cargo Traffic Report Hazardous cargo traffic on Interstate 5 (1-5) and the AT&SF railway shall be monitored and the results submitted to the NRC Regional Administrator once every three years.

(continued)

SAN ONOFRE--UNIT 3 5.0-29 Amendment No. 154,159

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.2 Special Reports Special Reports may be required covering inspection, test, and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.

Special Reports shall be submitted to the U. S. Nuclear Regulatory Commission, Attention: Document Control Desk, Washington, D. C.

20555, with a copy to the Regional Administrator of the Regional Office of the NRC, in accordance with 10 CFR 50.4 within the time period specified for each report.

The following Special Reports shall be submitted:

a.

When a pre-planned alternate method of monitoring post-accident instrumentation functions is required by Condition B or Condition H of LCO 3.3.11, a report shall be submitted within 30 days from the time the action is required.

The, report shall outline the action taken, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the function to OPERABLE status.

b.

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.

c.

Following each inservice inspection of steam generator (SG) tubes, in accordance with the SG Tube Surveillance Program, the number of tubes plugged and tubes sleeved in each SG shall be reported to the NRC within 15 days. The complete results of the SG tube inservice inspection shall be submitted to the NRC within 12 months following the completion of the inspection.

The report shall include:

1.

Number and extent of tubes and sleeves inspected, and

2.

Location and percent of wall-thickness penetration for each indication of an imperfection, and

3.

Identification of tubes plugged and tubes sleeved.

(continued)

SAN ONOFRE--UNIT 3 5.0-30 Amendment No. 116 132

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.2 Special Reports (continued)

Results of SG tube inspections which fall into Category C-3 shall be reported to the NRC prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

SAN ONOFRE--UNIT 3 5.0-31 Amendment No. +1-6

High Radiation Area 5.8 5.0 ADMINISTRATIVE CONTROLS 5.8 High Radiation Area 5.8.1 Each high radiation area as defined in 10 CFR 20 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)

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 rates in the area have been determined and personnel have been made knowledgeable of them,

c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device. This individual is responsible for providing positive radiation protection control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the radiation protection procedures or the applicable REP.

(continued)

SAN ONOFRE--UNIT 3 5.0-32 Amendment No. 116, 159

High Radiation Area 5.8 5.8.

High Radiation Area (continued) 5.8.2 In addition, areas that are accessible to personnel and that have radiation levels greater than 1.0 rem (but less than 500 rads at 1 meter) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm from the radiation source, or from any surface penetrated by the radiation, shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the shift supervisor on duty or health physics supervisor. Doors shall remain locked except during periods of access by personnel under an approved REP that specifies the dose rates in the immediate work areas and the maximum allowable stay time for individuals in that area. In lieu of a stay time specification on the REP, direct or remote continuous surveillance (such as closed circuit TV cameras) may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

5.8.3 Individual high radiation areas that are accessible to personnel, that could result in radiation doses greater than 1.0 rem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and that are within large areas, where no enclosure exists to enable locking and where no enclosure can be reasonably constructed around the individual area shall be barricaded and conspicuously posted. A flashing light.

shall be activated as a warning device whenever the dose rate in such an area exceeds or is expected to exceed 1.0 rem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm from the radiation source or from any surface penetrated by the radiation.

SAN ONOFRE--UNIT 3 5.0-33 Amendment No. 116, 159

PCN 533 Attachment E (Proposed Pages)

SONGS Unit 2

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, departure from nucleate boiling ratio (DNBR) shall be maintained at Ž 1.31.

2.1.1.2 In MODES 1 and 2, the peak linear heat (adjusted for fuel rod dynamics) shall at

• 21.0 kW/ft.

rate (LHR) be maintained 2.1.2 Reactor Coolant System (RCS) Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained at

• 2750 psia.

2.2 SL Violations 2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

2.2.3 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, notify the NRC Operations Center, in accordance with 10 CFR 50.72.

2.2.4 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, notify the Vice President - Nuclear Generation and the Nuclear Safety Group (NSG) Supervisor.

2.2.5 Within 60 days of the violation, a Licensee Event Report (LER) shall be prepared pursuant to 10 CFR 50.73. The LER shall be submitted to the NRC, the NSG Supervisor, and the Vice President -

Nuclear Generation.

I (continued)

SAN ONOFRE--UNIT 2 2.0-1 Amendment No.

Pressurizer Heatup/Cooldown Limits 3.4.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. ----

NOTE---------

C.1 Initiate action to Immediately Required Action C.2 restore parameter(s) shall be completed to within limits.

whenever this Condition is entered.

AND C.2 Determine Pressurizer Prior to Requirements of LCO is acceptable for entering MODE 4 not met any time in continued operation.

other than MODE 1, 2, 3, or 4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1.1 -------------------NOTE--------------------

Only required to be performed during Pressurizer heatup and cooldown operations.

Verify Pressurizer heatup and cooldown rates within the following limits:

30 minutes

a. A maximum heatup of 2000F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period,

b. A maximum cooldown of 200'F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.

SR 3.4.3.1.2 The spray water temperature differential For each cycle shall be determined for use in the UFSAR.

of auxiliary spray operation and for each cycle of main spray operation when the RCS cold leg temperature is

< 5000F.

I SAN ONOFRE--UNIT 2 3.4-14 Amendment No.

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.11 Steam Generator (SG) Tube Surveillance Program (continued)

4.

The provisions of Technical Specifications Surveillance Requirement 3.0.2 are applicable to SG Tube Surveillance inspection frequencies except those established by Category C-3 inspection results.

The above required inservice inspections of SG tubes repaired by sleeving shall be performed at the following frequencies:

1.

Steam generator tube sleeves shall be inspected prior to initial operation and in service.

The initial operating period before the initial inservice sample inspection shall not be shorter than six months nor longer than 24 months. The inspections of sleeves shall be configured to ensure that each individual sleeve is inspected at least once in 60 months.

2.

If the results of the inservice inspection of SG tube sleeves conducted in accordance with Table 5.5.2.11-2 fall in category C-3, the inspection frequency shall be increased to ensure that each remaining sleeve is inspected at least once in 30 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria for Category C-1.

f.

Acceptance Criteria

1.

Terms as used in this specification will be defined as follows:

a)

Degradation - A service-induced cracking, wastage, wear, or general corrosion occurring on either the inside or outside of a tube; b)

Degraded tube - A tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation; c)

% Degradation - The percentage of the tube wall thickness affected or removed by degradation; d)

Defect - An imperfection of such severity that it exceeds the repair limit. A tube containing a defect is defective.

(continued)

SAN ONOFRE--UNIT 2 5.0-17 Amendment No.

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.12 Ventilation Filter Testing Program (VFTP)

(continued)

The provisions of Technical Specification Surveillance Requirement 3.0.2 and Technical Specification Surveillance Requirement 3.0.3 are applicable to the VFTP test frequencies.

5.5.2.13 Diesel Fuel Oil Testing Program This program implements required testing of both new fuel oil and stored fuel oil.

The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM standards.

The purpose of the program is to establish the following:

a.

Acceptability of new fuel oil 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.

a water and sediment content within limits.

b.

Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to the storage tanks, with exceptions noted in the Bases for Surveillance Requirement 3.8.3.3; and,

c.

Total particulate concentration of fuel oil is

• 10 mg/l when tested every 92 days in accordance with ASTM D-2276, Method A.

5.5.2.14 Deleted I

(continued)

SAN ONOFRE--UNIT 2 5.0-20 Amendment No.

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.15 Containment Leakage Rate Testing Program 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 modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.

The calculated peak containment internal pressure related to the design basis loss-of-coolant accident, Pas is 45.9 psig (Pa will conservatively be assumed to be equal to the calculated peak containment internal pressure for the design basis Main Steam Line Break (56.5 psig) for the purpose of containment testing in accordance with this Technical Specification).

The maximum allowable containment leakage rate, La, at Pas shall be 0.10% of containment air weight per day.

Leakage rate acceptance criteria are:

a.

The Containment overall leakage rate acceptance criterion is 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are

• 0.60 La for the Type B and Type C tests and

• 0.75 La for the Type A tests;

b.

Air lock testing acceptance criteria are:

1)

Overall air lock leakage rate is

• at 2 Pd-0.05 La when tested

2)

For each door, pressurized to the leakage rate is 2 9.0 psig.

• 0.01 La when (continued)

SAN ONOFRE--UNIT 2 5.0-20a Amendment No.

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.4 Monthly Operating Reports Routine reports of operating statistics and shutdown experience, including documentation of all challenges to pressurizer safety valves, shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, Attention: Document Control Desk, Washington, D.C., with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.

5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1.

Specification 3.1.4, "Moderator Temperature Coefficient;"

2.

Specification 3.1.5, "Control Element Assembly (CEA)

Alignment;"

3.

Specification 3.1.7, "Regulating CEA Insertion Limits;"

4.

Specification 3.1.8, "Part Length Control Element Assembly Insertion Limits;"

5.

Specification 3.2.1, "Linear Heat Rate;"

6.

Specification 3.2.4, "Departure From Nucleate Boiling Ratio;"

7.

Specification 3.2.5, "Axial Shape Index;"

8.

Specification 3.9.1, "Boron Concentration."

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:

(continued)

SAN ONOFRE--UNIT 2 5.0-26 Amendment No.

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued)

1.

CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model"

2.

CENPD-137P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model"

3.

CEN-356(V)-P-A, "Modified Statistical Combination of Uncertainties"

4.

SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3"

5.

CEN-635(S), "Identification of NRC Safety Evaluation l

Report Limitations and/or Constraints on Reload Analysis Methodology"

6.

Letter, dated May 16, 1986, G. W. Knighton (NRC) to K.

P. Baskin (SCE), "Issuance of Amendment No. 47 to.

Facility Operating License NPF-10 and Amendment No'.36 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 3 SER)

7.

Letter, dated January 9, 1985, G. W. Knighton (NRC) to K. P. Baskin, "Issuance of Amendment No. 30 to Facility Operating License NPF-10 and Amendment No. 19 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 2 SER)

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.7.1.6 Not Used 5.7.1.7 Hazardous Cargo Traffic Report Hazardous cargo traffic on Interstate 5 (I-5) and the AT&SF railway shall be monitored and the results submitted to the NRC Regional Administrator once every three years.

(continued)

SAN ONOFRE--UNIT 2 5.0-27 Amendment No.

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.2 Special Reports Special Reports may be required covering inspection, test, and maintenance activities.

These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.

Special Reports shall be submitted to the U. S. Nuclear Regulatory Commission, Attention: Document Control Desk, Washington, D. C.

20555, with a copy to the Regional Administrator of the Regional Office of the NRC, in accordance with 10 CFR 50.4 within the time period specified for each report.

The following Special Reports shall be submitted:

a.

When a pre-planned alternate method of monitoring post-accident instrumentation functions is required by Condition B or Condition H of LCO 3.3.11, a report shall be submitted within 30 days from the time the action is required. The report shall outline the action taken, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the function to OPERABLE status.

b.

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.

c.

Following each inservice inspection of steam generator (SG) tubes, in accordance with the SG Tube Surveillance Program, the number of tubes plugged and tubes sleeved in each SG shall be reported to the NRC within 15 days. The complete results of the SG tube inservice inspection shall be submitted to the NRC within 12 months following the completion of the inspection.

The report shall include:

1.

Number and extent of tubes and sleeves inspected, and

2.

Location and percent of wall-thickness penetration for each indication of an imperfection, and

3.

Identification of tubes plugged and tubes sleeved.

(continued)

SAN ONOFRE--UNIT 2 5.0-28 Amendment No.

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.2 Special Reports (continued)

Results of SG tube inspections which fall into Category C-3 shall be reported to the NRC prior to resumption of plant operation.

This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

SAN ONOFRE--UNIT 2 5.0-29 Amendment No.

High Radiation Area 5.8 5.0 ADMINISTRATIVE CONTROLS 5.8 High Radiation Area 5.8.1 Each high radiation area as defined in 10 CFR 20 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).

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 rates in the area have been determined and personnel have been made knowledgeable of them,

c.

An individual qualified in radiation protection procedures with a radiation dose rate monitoring device.

This individual is responsible for providing positive radiation protection control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the radiation protection procedures or the applicable REP.

(continued)

SAN ONOFRE--UNIT 2 5.0-30 Amendment No.

High Radiation Area 5.8 5.8.

High Radiation Area (continued) 5.8.2 In addition, areas that are accessible to personnel and that have radiation levels greater than 1.0 rem (but less than 500 rads at 1 meter) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm from the radiation source, or from any surface penetrated by the radiation, shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the shift supervisor on duty or health physics supervisor.

Doors shall remain locked except during periods of access by personnel under an approved REP that specifies the dose rates in the immediate work areas and the maximum allowable stay time for individuals in that area. In lieu of a stay time specification on the REP, direct or remote continuous surveillance (such as closed circuit TV cameras) may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

5.8.3 Individual high radiation areas that are accessible to personnel, that could result in radiation doses greater than 1.0 rem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and that are within large areas where no enclosure exists to enable locking and where no enclosure can be reasonably constructed around the individual area shall be barricaded and conspicuously posted. A flashing light.shall be activated as a warning device whenever the dose rate in such an area exceeds or is expected to exceed 1.0 rem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm from the radiation source or from any surface penetrated by the radiation.

SAN ONOFRE--UNIT 2 5.0-31 Amendment No.

PCN 533 Attachment F (Proposed Pages)

SONGS Unit 3

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, departure from nucleate boiling ratio (DNBR) shall be maintained at 2 1.31.

2.1.1.2 In MODES 1 and 2, the peak linear heat rate (LHR)

(adjusted for fuel rod dynamics) shall be maintained at s 21.0 kW/ft.

2.1.2 Reactor Coolant System (RCS) Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained at

• 2750 psia.

2.2 SL Violations 2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

2.2.3 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, notify the NRC Operations Center, in accordance with 10 CFR 50.72.

2.2.4 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, notify the Vice President - Nuclear Generation and the Nuclear Safety Group (NSG) Supervisor.

2.2.5 Within 60 days of the violation, a Licensee Event Report (LER) shall be prepared pursuant to 10 CFR 50.73.

The LER shall be submitted to the NRC, the NSG Supervisor, and the Vice President -

Nuclear Generation.

(continued)

SAN ONOFRE--UNIT 3 2.0-1 Amendment No.

Pressurizer Heatup/Cooldown Limits 3.4.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. ---------NOTE---------

C.1 Initiate action to Immediately Required Action C.2 restore parameter(s) shall be completed to within limits.

whenever this Condition is entered.

AND C.2 Determine Pressurizer Prior to Requirements of LCO is acceptable for entering MODE 4 not met any time in continued operation.

other than MODE 1, 2, 3, or 4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1.1 ------------------- NOTE--------------------

Only required to be performed during Pressurizer heatup and cooldown operations.

Verify Pressurizer heatup and cooldown rates within the following limits:

30 minutes

a. A maximum heatup of 2000F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period,

b. A maximum cooldown of 2000F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.

For each cycle SR 3.4.3.1.2 The spray water temperature differential of auxiliary shall be determined for use in the UFSAR.

spray operation and for each cycle of main spray operation when the RCS cold leg temperature is

< 500 0F.

I SAN ONOFRE--UNIT 3 3.4-14 Amendment No.

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.11 Steam Generator (SG) Tube Surveillance Program (continued)

4.

The provisions of Technical Specifications Surveillance Requirement 3.0.2 are applicable to SG Tube Surveillance inspection frequencies except those established by Category C-3 inspection results.

The above required inservice inspections of SG tubes repaired by sleeving shall be performed at the following frequencies:

1.

Steam generator tube sleeves shall be inspected prior to initial operation and in service.

The initial operating period before the initial inservice sample inspection shall not be shorter than six months nor longer than 24 months. The inspections of sleeves shall be configured to ensure that each individual sleeve is inspected at least once in 60 months.

2.

If the results of the inservice inspection of SG tube sleeves conducted in accordance with Table 5.5.2.11-2 fall in category C-3, the inspection frequency shall'be increased to ensure that each remaining sleeve is inspected at least once in 30 months.

The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria for Category C-1.

f.

Acceptance Criteria

1.

Terms as used in this specification will be defined as follows:

a)

Degradation - A service-induced cracking, wastage, wear, or general corrosion occurring on either the inside or outside of a tube; b)

Degraded tube - A tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation; c)

% Degradation - The percentage of the tube wall thickness affected or removed by degradation; d)

Defect - An imperfection of such severity that it exceeds the repair limit. A tube containing a defect is defective.

(continued)

SAN ONOFRE--UNIT 3 5.0-17 Amendment No.

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.12 Ventilation Filter Testing Program (VFTP)

(continued)

The provisions of Technical Specification Surveillance Requirement 3.0.2 and Technical Specification Surveillance Requirement 3.0.3 are applicable to the VFTP test frequencies.

5.5.2.13 Diesel Fuel Oil Testing Program This program implements required testing of both new fuel oil and stored fuel oil.

The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM standards. The purpose of the program is to establish the following:

a.

Acceptability of new fuel oil 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.

a water and sediment content within limits.

b.

Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to the storage tanks, with exceptions noted in the Bases for Surveillance Requirement 3.8.3.3; and,

c.

Total particulate concentration of fuel oil is

• 10 mg/l when tested every 92 days in accordance with ASTM D-2276, Method A.

5.5.2.14 Deleted I

(continued)

SAN ONOFRE--UNIT 3 5.0-20 Amendment No.

Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.15 Containment Leakage Rate Testing Program 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 modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program", dated September 1995.

The calculated peak containment internal pressure related to the design basis loss-of-coolant accident, Pa, is 45.9 psig (P. will conservatively be assumed to be equal to the calculated peak containment internal pressure for the design basis Main Steam Line Break (56.5 psig) for the purpose of containment testing in accordance with this Technical Specification).

The maximum allowable containment leakage rate, La, at Pa, shall be 0.10% of containment air weight per day.

Leakage rate acceptance criteria are:

a.

The Containment overall leakage rate acceptance criterion is

• 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are

• 0.60 La for the Type B and Type C tests and

• 0.75 La for the Type A tests;

b.

Air lock testing acceptance criteria are:

I

1)

Overall air lock leakage rate is -

at 2 Pa.

2)

For each door, the leakage rate is pressurized to 2 9.0 psig.

0.05 La when tested

• 0.01 La when (continued)

SAN ONOFRE--UNIT 3 5.0-20a Amendment No.

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.4 Monthly Operating Reports Routine reports of operating statistics and shutdown experience, including documentation of all challenges to pressurizer safety valves, shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, Attention: Document Control Desk, Washington, D.C., with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.

5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1.

Specification 3.1.4, "Moderator Temperature Coefficient;"

2.

Specification 3.1.5, "Control Element Assembly (CEA)

Alignment;"

3.

Specification 3.1.7, "Regulating CEA Insertion Limits;"

4.

Specification 3.1.8, "Part Length Control Element Assembly Insertion Limits;"

5.

Specification 3.2.1, "Linear Heat Rate;"

6.

Specification 3.2.4, "Departure From Nucleate Boiling Ratio;"

7.

Specification 3.2.5, "Axial Shape Index;"

8.

Specification 3.9.1, "Boron Concentration."

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:

(continued)

SAN ONOFRE--UNIT 3 5.0-26 Amendment No.

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR)

(continued)

1.

CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model"

2.

CENPD-137P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model"

3.

CEN-356(V)-P-A, "Modified Statistical Combination of Uncertainties"

4.

SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3"

5.

CEN-635(S), "Identification of NRC Safety Evaluation Report Limitations and/or Constraints on Reload Analysis Methodology"

6.

Letter, dated May 16, 1986, G. W. Knighton (NRC) to K.

P. Baskin (SCE), "Issuance of Amendment No. 47 to.

Facility Operating License NPF-10 and Amendment No. 36 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 3 SER)

7.

Letter, dated January 9, 1985, G. W. Knighton (NRC) to K. P. Baskin, "Issuance of Amendment No. 30 to Facility Operating License NPF-10 and Amendment No. 19 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 2 SER)

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.7.1.6 Not Used 5.7.1.7 Hazardous Cargo Traffic Report Hazardous cargo traffic on Interstate 5 (1-5) and the AT&SF railway shall be monitored and the results submitted to the NRC Regional Administrator once every three years.

(continued)

SAN ONOFRE--UNIT 3 5.0-27 Amendment No.

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.2 Special Reports Special Reports may be required covering inspection, test, and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.

Special Reports shall be submitted to the U. S. Nuclear Regulatory Commission, Attention: Document Control Desk, Washington, D. C.

20555, with a copy to the Regional Administrator of the Regional Office of the NRC, in accordance with 10 CFR 50.4 within the time period specified for each report.

The following Special Reports shall be submitted:

a.

When a pre-planned alternate method of monitoring post-accident instrumentation functions is required by Condition B or Condition H of LCO 3.3.11, a report shall be submitted within 30 days from the time the action is required.

The report shall outline the action taken, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the function to OPERABLE status.

b.

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.

c.

Following each inservice inspection of steam generator (SG) tubes, in accordance with the SG Tube Surveillance Program, the number of tubes plugged and tubes sleeved in each SG shall be reported to the NRC within 15 days. The complete results of the SG tube inservice inspection shall be submitted to the NRC within 12 months following the completion of the inspection. The report shall include:

1.

Number and extent of tubes and sleeves inspected, and

2.

Location and percent of wall-thickness penetration for each indication of an imperfection, and

3.

Identification of tubes plugged and tubes sleeved.

(continued)

SAN ONOFRE--UNIT 3 5.0-28 Amendment No.

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.2 Special Reports (continued)

Results of SG tube inspections which fall into Category C-3 shall be reported to the NRC prior to resumption of plant operation.

This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

SAN ONOFRE--UNIT 3 5.0-29 Amendment No.

High Radiation Area 5.8 5.0 ADMINISTRATIVE CONTROLS 5.8 High Radiation Area 5.8.1 Each high radiation area as defined in 10 CFR 20 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)

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 rates in the area have been determined and personnel have been made knowledgeable of them,

c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device.

This individual is responsible for providing positive radiation protection control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the radiation protection procedures or the applicable REP.

(continued)

SAN ONOFRE--UNIT 3 5.0-30 Amendment No.

High Radiation Area 5.8 5.8.

High Radiation Area (continued) 5.8.2 In addition, areas that are accessible to personnel and that have radiation levels greater than 1.0 rem (but less than 500 rads at 1 meter) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm from the radiation source, or from any surface penetrated by the radiation, shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the shift supervisor on duty or health physics supervisor. Doors shall remain locked except during periods of access by personnel under an approved REP that specifies the dose rates in the immediate work areas and the maximum allowable stay time for individuals in that area.

In lieu of a stay time specification on the REP, direct or remote continuous surveillance (such as closed circuit TV cameras) may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

5.8.3 Individual high radiation areas that are accessible to personnel, that could result in radiation doses greater than 1.0 rem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and that are within large areas, where no enclosure exists to enable locking and where no enclosure can be reasonably constructed around the individual area shall be barricaded and conspicuously posted. A flashing light shall be activated as a warning device whenever the dose rate in such an area exceeds or is expected to exceed 1.0 rem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm from the radiation source or from any surface penetrated by the radiation.

SAN ONOFRE--UNIT 3 5.0-31 Amendment No.

Bases03-009 Attachment G Proposed Technical Specification Bases Changes (Redline and Strikeout)

San Onofre Unit 2 (typical for both Units)

Reactor Core SLs B 2.1.1 BASES (continued)

SAFETY LIMIT VIOLATIONS The following violation responses are applicable to the reactor core SLs.

2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, the requirement to go to MODE 3 places the unit in a MODE in which this SL is not applicable.

The allowed Completion importance of bringing applicable and reduces Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the the unit to a MODE where this SL is not the probability of fuel damage.

2.2.3 If SL 2.1.1.1 or SL 2.1.1.2 is violated, the NRC Operations Center must be notified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in accordance with 10 CFR 50.72 (Ref. 3).

2.2.4 If SL 2.1.1.1 or SL 2.1.1.2 is violated, the appropriate senior management of the nuclear plant and the utility shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

This 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period provides time for the plant operators and staff to take the appropriate immediate action and assess the condition of the unit before reporting to the senior management.

2.2.5 If SL 2.1.1.1 or SL 2.1.1.2 is violated, a Licensee Event Report shall be prepared and submitted within 360 days to the NRC, Vice President - Nuclear Generation, and the NSG Supervisor. This requirement is in accordance with 10 CFR 50.73 (Ref. 4).

2.2.6 If SL 2.1.1.1 or SL 2.1.1.2 is violated, restart of the unit shall not commence until authorized by the NRC.

This requirement ensures the NRC that all necessary reviews, (continued) l1 SAN ONOFRE--UNIT 2 B 2.0-4 Amendment No. +-F I

RCS Pressure SL B 2.1.2 BASES SAFETY LIMIT VIOLATIONS (continued) 2.2.4 If the RCS pressure SL is violated, the appropriate senior management of the nuclear plant and the utility shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

This 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period provides time for the plant operators and staff to take the appropriate immediate action and to assess the condition of the unit before reporting to the senior management.

2.2.5 If the RCS pressure SL is violated, a Lirensee Event Report shall be prepared and submitted within 4bO days to the NRC, Vice President -

Nuclear Generation, and the NSG Supervisor.

This requirement is in accordance with 10 CFR 50.73 (Ref. 7).

2.2.6 I

If the RCS pressure SL is violated, restart of the unit shall not commence until authorized by the NRC.

This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to normal operation.

REFERENCES

1.

10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.

2. ASME, Boiler and Pressure Vessel Code,Section III, Article NB-7000.

3. ASME, Boiler and Pressure Vessel Code,Section XI, Article IWX-5000.

4.

10 CFR 100.

5.

UFSAR, Section 7.2, "Reactor Protective Systems"

6.

10 CFR 50.72.

7.

10 CFR 50.73.

SAN ONOFRE--UNIT 2 B 2.0-9 Amendment No. g-2-e I

ESFAS Instrumentation B 3.3.5 BASES (continued)

ACTIONS D.1 and D.2 (continued)

Function is in two-out-of-three logic in the bypassed input parameter, but with another channel failed, the ESFAS may be operating with a two-out-of-two logic.

This is outside the assumptions made in the analyses and should be corrected.

To correct the problem, the second channel is placed in trip.

This places the ESFAS Function in a one-out-of-two logic.

If any of the other OPERABLE channels receives a trip signal, ESFAS actuation will occur.

Action D.2 provides a limit of 7 days for operation with 2 inoperable channels. In the one-out-of-two configuration, a single channel failure can cause a spurious trip.

For RAS and EFAS functions, a spurious trip can lead to undesireable consequences during certain Design Basis Events.

The 7 day time limit provides operational flexibility to perform a required CHANNEL FUNCTIONAL TEST on one channel (which is bypassed) while a second channel is inoperable (and is tripped).

The 7 day time limit also maintains acceptable core damage frequency as discussed in NSG 98-007, Time Limit for RAS or EFAS Channel in Trip (Reference 11).

As required by Section 5.5.2.14, a Configuration Risk Management Program is imple:mnted in te event of ConditA-on-D 7.

E.1. E.2.1. and E.2.2 Condition C applies to one automatic bypass removal channel inoperable.

The only automatic bypass removal on an ESFAS is on the Pressurizer Pressure -Low signal.

This bypass removal is shared with the RPS Pressurizer Pressure-Low bypass removal.

If the bypass removal channel for any operating bypass cannot be restored to OPERABLE status, the associated ESFAS channel may be considered OPERABLE only if the bypass is not in effect.

Otherwise, the affected ESFAS channel must be declared inoperable, as in Conditions A and B, and the bypass either removed or the bypass removal channel repaired.

The Bases for the Required Actions and required Completion Times are consistent with Conditions A and B.

(continued)

SAN ONOFRE--UNIT 2 B 3.3-100 Amendment No. 127 10/06/99

Pressurizer Heatup and Cooldown Limits B 3.4.3.1 BASES (continued)

ACTIONS C.1 and C.2 (continued)

If requirements of the LCO are not met at any time in other than MODES 1, 2, 3, or 4, the Required Action C.1 requires to immediately initiate action to restore parameter(s) to within limits specified in the Pressure/Temperature Limits.

Also, Required Action C.2 requires to perform engineering evaluation prior to entering MODE 4 to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer.

CONDITION C is modified by a Note which requires to determine the pressurizer is acceptable for continued operation whenever the requirements of the LCO not met any time in other than MODES 1, 2, 3, or 4.

SURVEILLANCE SR 3.4.3.1.1 REQUIREMENTS To minimize the potential thermal stresses of the pressurizer during startup and shutdown, the rate of temperature changes should be monitored during startup and shutdown.

The verification these rates are within limits specified in the Pressure/Temperature Limits should be made every 30 minutes.

This FREQUENCY is based on operating experience and reflects the importance of the possible effect of temperature changes rate during such Unit evolutions as startup and shutdown on pressurizer and its components integrity.

This SURVEILLANCE REQUIREMENT is modified by a Note which requires to perform this SR during pressurizer heatup and cooldown operations only.

SR 3.4.3.1.2 SR 3.4.3.1.2 requires to determine for use in the Pressure/Temperature Limits the spray water temperature differential for each cycle of main spray when less than 4 reactar colant pumps are

'r.r-i" t

-nP.,,hr.A

_f-

IaIei~ji~ry. cnrav..nnPratinn ;Anrl fnr patrh r'velP nf main-spray pperationowhen -the RCS cold leg temperature is;< 500 0F.

The spray nozz ezle e th'~r_1 a

ensiets f6r noirmal4-RC P) and auxiliary spray operations are developed in the calculation package S-PEC-368, and are used as design input for the Pressurizer Class-1 stress report. A maximum temperature differential of 200'F is assumed for normal spray operations. Of particular (continued)

SAN-ONOFRE--UNIT 2 B 3.4-20 Amendment No. Age 1

SITs B 3.5.1 BASES (continued)

ACTIONS B.1 (continued) condition be established for the specific case, where "One accumulator [SIT] is inoperable due to the inoperability of water level and pressure channels," in which the completion time to restore the accumulator to operable status will be 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

While technically inoperable, the accumulator would be available to fulfill its safety function during this time and, thus, this change would have a negligible increase in risk."

Although Action B.1 has a risk informed Completion Time, iL e i i

Admitrative Controells Section 5.'.2.4, i: _at reguired as stat.: in Rfer:ne C.1 If one SIT is inoperable, for a reason other than boron concentration or the inability to verify level or pressure, the SIT must be returned to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

In this Condition, the required contents of three SITs cannot be assumed to reach the core during a LOCA as is assumed in Appendix K to 10CFR50.

Reference 7 provides series of deterministic and probabilistic findings that support 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as being either "risk beneficial" or "risk neutral" in comparison to shorter periods for restoring the SIT to OPERABLE status.

Reference 7 discusses a best-estimate analysis that confirmed that, during large-break LOCA scenarios, core melt can be prevented by either operation of one Low Pressure Safety Injection (LPSI) pump or the operation of one High Pressure Safety Injection (HPSI) pump and a single SIT.

Reference 7 also discusses a plant-specific probabilistic analysis that evaluated the risk-impact of the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> recovery period in comparison to shorter recovery periods.

Although Action C.1 has a risk informed Completion Time, imleenaton of the Cniuaio~n isk~ IMangmetPrga (CMP dsciedinAdinsrative.1 Cotrl Seto 5rs.5.2.14 is not required as stated in Reference 8.

D.1 and D.2 If the SIT cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply.

To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and pressurizer pressure reduced to < 715 psia (continued)

SAN ONOFRE--UNIT 2 B 3.5-8 Amendment No. 127 O7J!5J98 l

I SITs B 3.5.1 BASES (continued)

REFERENCES

1.

IEEE Standard 279-1971.

2.

UFSAR, Section 6.3.

3.

10 CFR 50.46.

4.

UFSAR, Chapter 15.

5.

NUREG-1366, December 1992.

6.

NRC Generic Letter 93-05, "Line-Item Technical Specification Improvements to Reduce Surveillance Requirements for Testing During Power Operations,"

September 27, 1993.

7. CE NPSD-994, "CEOG Joint Application Report for Safety Injection Tank AOT/STI Extension," April 1995.

8.

NRC Safety, Evaluation Report, Junc 19, 1998.

SAN ONOFRE--UNIT 2 B 3.5-10a Amendment No. 127 07/1'7/98 l

ECCS - Operating B 3.5.2 BASES (continued)

ACTIONS A.1 and B.1 (continued)

The LCO requires the OPERABILITY of a number of independent subsystems. Due to the redundancy of trains and the diversity of subsystems, the inoperability of one component in a train does not render the ECCS incapable of performing its function.

Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS.

The intent of each of Condition A and Condition B is to maintain a combination of OPERABLE equipment such that 100% of the ECCS flow equivalent to 100% of a single OPERABLE train remains available. This allows increased flexibility in plant operations when components in opposite trains are inoperable.

Each of Condition A and Condition B includes a combination of OPERABLE equipment such that at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train remains available.

Condition A addresses the specific condition where the only affected ECCS subsystem is a single LPSI subtrain.

The availability of a least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train is implicit in the definition of Condition A.

If LCO 3.5.2 requirements are not met due only to the existence of Condition A, then the inoperable LPSI subtrain components must be returned to OPERABLE status within 7 days of discovery of Condition A. A Configuration Risk Ma~nagement PARPgArn (CRIP) defined in Administrative Control ee.

5524is imlmnte irln the_ evn rof adt::

A:.. r This 7-day Completion Time is based on the findings of the deterministic and probabilistic analysis that are discussed in Reference 6. Seven days is a reasonable amount of time to perform many corrective and preventative maintenance items on the affected LPSI subtrain.

Reference 6 concluded that the overall risk impact of this Completion Time was either risk-beneficial or risk-neutral.

(continued)

SAN ONOFRE--UNIT 2 B 3.5-16 Amendment No.

127 03/03/I00

Containment Spray and Cooling Systems B 3.6.6.1 BASES (continued)

LCO During a DBA, a minimum of two containment cooling trains or two containment spray trains, or one of each, is required to maintain the containment peak pressure and temperature below the design limits (Ref. 2.

Additionally, one containment spray train is also required to remove iodine from the containment atmosphere and maintain concentrations below those assumed in the safety analysis. To ensure that these requirements are met, two containment spray trains and two containment cooling units must be OPERABLE.

Therefore, in the event of an accident, the minimum requirements are met, assuming that the worst case single active failure occurs.

Each Containment Spray System includes a spray pump, spray headers, nozzles, valves, piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST upon an ESF actuation signal and automatically transferring suction to the containment sump.

Each Containment Cooling System includes demisters, cooling coils, dampers, fans, instruments, and controls to ensure an OPERABLE flow path.

APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to containment and an increase in containment pressure and temperature, requiring the operation of the containment spray trains and containment cooling trains.

In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES.

Thus, the Containment Spray and Containment Cooling systems are not required to be OPERABLE in MODES 5 and 6.

ACTIONS A.1 With one containment spray train inoperable, the inoperable containment spray train must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE spray and cooling trains are adequate to perform the iodine removal and containment cooling functions. A Configuration Risk Management (CRMP) deofined in the Administrative Controls Section 5.5...14 is implemented in the event of (continued)

SAN ONOFRE--UNIT 2 B 3.6-37 Amendment No. 127 O4/297O3

Containment Spray and Cooling Systems B 3.6.6.1 BASES (continued)

ACTIONS A.1 (continued)

Condition A. The 7-day Completion Time is based on the findings of the deterministic and probabilistic analysis that was reviewed and approved in Reference 3. Seven days is a reasonable amount of time to perform many corrective and preventive maintenance items on the affected Containment Spray Train.

The 14 day portion of the Completion Time is based upon engineering judgement.

It takes into account the low probability of coincident entry into two conditions in this Specification coupled with the low probability of an accident occurring during this time.

Refer to Section 1.3, "Completion Times," for a more detailed discussion of the purpose of the "from discovery of failure to meet the LCO" portion of the Completion Time.

B.1 and B.2 If the inoperable containment spray train cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply.

To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 4 within 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />.

The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

The extended interval to reach MODE 4 allows additional time for the restoration of the containment spray train and is reasonable when considering that the driving force for a release of radioactive material from the Reactor Coolant System is reduced in MODE 3.

C.1 With one required containment cooling train inoperable, the inoperable containment cooling train must be restored to OPERABLE status within 7 days. The components in this degraded condition provide iodine removal capabilities and are capable of providing at least 100% of the heat removal needs after an accident. The 7 day Completion Time was developed taking into account the redundant heat removal capabilities afforded by combinations of the Containment Spray System and Containment Cooling System and the low probability of a DBA occurring during this period.

(continued)

SAN ONOFRE--UNIT 2 B 3.6-38 Amendment No. 127 04/29/43 l

ECW System B 3.7.10 BASES (continued)

BACKGROUND If while implementing LCO 3.7.10 Action A for an inoperable (continued)

ECW train, the opposite ECW train for the affected Unit(s) becomes inoperable, enter LCO 3.0.3 on the applicable Unit(s).

TS 3.7.10 allows 14 days for restoring operability of one ECWS train.

The 14 day AOT is based on a probabilistic risk assessment that was done in accordance with the guidance of Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk Informed Decisionmaking: Technical Specifications."

The 14 day AOT is implemented in the three-tiered approach.

First, the risk of the 14 day AOT is acceptable based on the single AOT risk.

Second, administrative controls must be established to ensure that planned maintenance on the normal chilled water system does not coincide-with planned maintenance on the ECW system.

Third T Ihe I=ON Configuratian Risk -Management Program (CRnr) program is employed to ensure that risk-significant configurations are identified and managed appropriately per the Maintenance Rule (a)(4).

Allowing only one 14 day clock even in the case of multiple single train component failures is conservative.

This approach prohibits exceeding the intent of the LCO, which is to ensure an ECWS train remains out of service for no more than 14 days, regardless of circumstances.

LCO 3.7.10 allows only one ECW train to be inoperable.

Therefore, with both trains inoperable, a LCO 3.0.3 entry is required.

An emergency chiller is considered OPERABLE when it is or can be aligned to either Unit's operating or standby OPERABLE Component Cooling Water (CCW) critical loop, provided that the OPERABLE CCW critical loop can be placed in operation within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after a design basis event is detected in the Control Room.

(Reference 2) Thus, an emergency chiller, under normal circumstances, remains OPERABLE during a transfer operation between OPERABLE CCW critical loops completed in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Likewise, an emergency chiller is considered OPERABLE when it is aligned to either Unit's energized 4 kV bus.

Under normal circumstances, the emergency chiller remains OPERABLE during a transfer operation between 4 kV buses, provided the transfer operation is completed in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Room Coolers OPERABILITY. General If one or more required individual room coolers for a Unit are inoperable and the backup cooling listed in Table 1 for the affected room(s) is also inoperable, OR if the temperature in the affected room(s) increases above its design temperature, declare the safety related equipment in the cooled room(s) inoperable and enter the LCO action (continued)

SAN ONOFRE--UNIT 2 B 3.7-50 Amendment No. 127 10/17/1

-W

AC Sources -Operating B 3.8.1 BASES (continued)

ACTIONS A.2 (continued)

As in Required Action A.2, the Completion Time allows for an exception to the normal "time zero' for beginning the allowed outage time "clock."

This will result in establishing the "time zero" at the time that the LCO was initially not met, instead of at the time Condition A was entered.

As required by Scetien 5.5.2.14, a Configuration Risk Management Program is implemented in the event of Ccnfditicn A.

B.1 To ensure a highly reliable power source remains when one of the required DGs is inoperable, it is necessary to verify the availability of the offsite circuits on a more frequent basis.

Since the Required Action only specifies "perform,"

a failure of SR 3.8.1.1 acceptance criteria does not result in a Required Action being not met.

However, if a circuit fails to pass SR 3.8.1.1, it is inoperable.

Upon offsite circuit inoperability, additional Conditions and Required Actions must then be entered.

B.2 Required Action B.2 is intended to provide assurance that a loss of offsite power, during the period that a DG is inoperable, does not result in a complete loss of safety function of critical systems. These features are designed with redundant safety related trains. This includes motor driven auxiliary feedwater umps. Single train systems, such as turbine driven auxi iary feedwater pumps, are not included. Redundant required feature failures consist of inoperable features associated with a train, redundant to the train that has an inoperable DG.

(continued)

SAN ONOFRE--UNIT 2 B 3.8-7 Amendment No. 127 12/10/99 I

1-h AC Sources -Operating B 3.8.1 BASES (continued)

ACTIONS B.4 (continued)

"time zero" at the time that the LCO was initially not met, instead of at the time Condition B was entered.

As required by Section 5.5.2.14, a Configuration Risk Management Program is implemented in the event of retaion B7 C.1 and C.2 Required Action C.1, which applies when two offsite circuits are inoperable, is intended to provide assurance that an event with a coincident single failure will not result in a complete loss of redundant required safety functions.

The Completion Time for this failure of redundant required features is reduced to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowed by Regulatory Guide 1.93 (Ref. 6) for two inoperable required offsite circuits. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance is based upon the assumption that two complete safety trains are OPERABLE.

When a concurrent redundant required feature failure exists, this assumption is not the case and a shorter Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is appropriate. These features are powered from redundant AC safety trains. This includes motor driven auxiliary feedwater pumps. Single train turbine driven auxiliary pumps, are not included in the list.

The Completion Time for Required Action C.1 is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."

In this Required Action, the Completion Time only begins on discovery that both:

a.

All required offsite circuits are inoperable; and

b. A required feature is inoperable.

If at any time during the existence of Condition C (two offsite circuits inoperable) and a required feature becomes inoperable, this Completion Time begins to be tracked.

According to Regulatory Guide 1.93 (Ref. 6), operation may continue in Condition C for a period that should not exceed (continued)

SAN ONOFRE--UNIT 2 B 3.8-10 Amendment No. 127 10/06/99