Text

Attachment 1, Volume 2, San Onofre Nuclear Generating Station - Improved Technical Specifications Conversion - ITS Chapter 2.0, Safety Limits
	ML11251A095
Person / Time
Site:	San Onofre
Issue date:	07/29/2011
From:	; Edison International Co, Southern California Edison Co
To:	; Office of Nuclear Reactor Regulation
References
	NUREG-1432, Rev. 3.0
	Download: ML11251A095 (25)
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Attachment 1, Volume 2, Rev. 0, Page 1 of 25 ATTACHMENT 1 VOLUME 2 SAN ONOFRE NUCLEAR GENERATING STATION IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS CHAPTER 2.0 SAFETY LIMITS Attachment 1, Volume 2, Rev. 0, Page 1 of 25

Attachment 1, Volume 2, Rev. 0, Page 2 of 25 LIST OF ATTACHMENTS

1. ITS Chapter 2.0, SAFETY LIMITS Attachment 1, Volume 2, Rev. 0, Page 2 of 25

, Volume 2, Rev. 0, Page 3 of 25 ATTACHMENT 1 ITS Chapter 2.0, SAFETY LIMITS , Volume 2, Rev. 0, Page 3 of 25

Attachment 1, Volume 2, Rev. 0, Page 4 of 25 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

Attachment 1, Volume 2, Rev. 0, Page 4 of 25

Attachment 1, Volume 2, Rev. 0, Page 5 of 25 ITS A01 SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 2.1 SLs 2.1.1 2.1.1 Reactor Core SLs 2.1.1.1 2.1.1.1 In MODES 1 and 2, departure from nucleate boiling ratio (DNBR) shall be maintained at $ 1.31.

2.1.1.2 2.1.1.2 In MODES 1 and 2, peak fuel centerline temperature shall be maintained at < 5080EF, decreasing by 58EF per 10,000 CENPD-275-P, MWD/MTU and adjusted for burnable poison per A02 Revision 1-P-A CENPD-382-P-A.

or 2.1.2 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 2.2 SL Violations 2.2.1 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 2.2.2 If SL 2.1.2 is violated:

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

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

(continued)

SAN ONOFRE--UNIT 2 2.0-1 Amendment No. 207 Attachment 1, Volume 2, Rev. 0, Page 5 of 25

Attachment 1, Volume 2, Rev. 0, Page 6 of 25 SLs A01 2.0 2.0 SLs THIS PAGE INTENTIONALLY LEFT BLANK

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SAN ONOFRE--UNIT 2 2.0-2 Amendment No. 207 Attachment 1, Volume 2, Rev. 0, Page 6 of 25

Attachment 1, Volume 2, Rev. 0, Page 7 of 25 ITS A01 SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 2.1 SLs 2.1.1 2.1.1 Reactor Core SLs 2.1.1.1 2.1.1.1 In MODES 1 and 2, departure from nucleate boiling ratio (DNBR) shall be maintained at $ 1.31.

2.1.1.2 2.1.1.2 In MODES 1 and 2, peak fuel centerline temperature shall be maintained at < 5080EF, decreasing by 58EF per 10,000 CENPD-275-P, MWD/MTU and adjusted for burnable poison per A02 Revision 1-P-A CENPD-382-P-A.

or 2.1.2 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 2.2 SL Violations 2.2.1 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 2.2.2 If SL 2.1.2 is violated:

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

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

(continued)

SAN ONOFRE--UNIT 3 2.0-1 Amendment No. 199 Attachment 1, Volume 2, Rev. 0, Page 7 of 25

Attachment 1, Volume 2, Rev. 0, Page 8 of 25 SLs A01 2.0 2.0 SLs THIS PAGE INTENTIONALLY LEFT BLANK

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SAN ONOFRE--UNIT 3 2.0-2 Amendment No. 199 Attachment 1, Volume 2, Rev. 0, Page 8 of 25

Attachment 1, Volume 2, Rev. 0, Page 9 of 25 DISCUSSION OF CHANGES ITS 2.0, SAFETY LIMITS ADMINISTRATIVE CHANGES A01 In the conversion of the San Onofre Nuclear Generating Station (SONGS)

Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1432, Rev. 3.0, "Standard Technical Specifications Combustion Engineering Plants" (ISTS) and additional approved Technical Specification Task Force (TSTF) travelers included in this submittal.

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 2.1.1.2 states that peak fuel centerline temperature shall be maintained at

< 5080ºF, decreasing by 58ºF per 10,000 MWD/MTU and adjusted for burnable poison per CENPD-382-P-A. ITS 2.1.1.2 contains the same requirements, but also allows the burnable poison to be adjusted per CENPD-275-P, Revision 1-P-A. This changes the CTS by specifying an additional topical report that describes how to adjust for burnable poison.

CENPD-275-P, Revision 1-P-A, describes how to adjust for gadolinia burnable poison (Gd2O3). As stated in CTS 4.2.1, Gd2O3 is an NRC-approved burnable poison that is allowed to be used at SONGS Units 2 and 3. While not specifically stated in CTS 2.1.1.2, CENPD-275, Revision 1-P-A is already approved for use at SONGS Units 2 and 3. CTS 5.7.1.5 provides the CORE OPERATING LIMITS REPORT requirements. CTS 5.7.1.5.b provides a list of the analytical methods used to determine the core operating limits that have been previously approved by the NRC, and one of the identified documents in CTS 5.7.1.5.b is topical report SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3" (CTS 5.7.1.5.b.4). SCE-9801, Section 3.1.1, second paragraph, states that the Reactor Operation and Control Simulator and Mesh-Centered (References 1 and 41) physics models are constructed for the reload fuel cycle. Reference 41 is CENPD-275-P-A, May 1988 (which is Revision 1-P-A). Therefore, since the NRC has previously approved for use at SONGS Units 2 and 3 topical report CENPD-275, Revision 1-P-A (as documented in the NRC Safety Evaluation from S. Dembeck (NRC) to H. B. Ray (SCE) dated June 2, 1999), this change is considered acceptable and is designated as an administrative change.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None San Onofre Unit 2 and 3 Page 1 of 2 Attachment 1, Volume 2, Rev. 0, Page 9 of 25

Attachment 1, Volume 2, Rev. 0, Page 10 of 25 DISCUSSION OF CHANGES ITS 2.0, SAFETY LIMITS REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None San Onofre Unit 2 and 3 Page 2 of 2 Attachment 1, Volume 2, Rev. 0, Page 10 of 25

Attachment 1, Volume 2, Rev. 0, Page 11 of 25 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 2, Rev. 0, Page 11 of 25

Attachment 1, Volume 2, Rev. 0, Page 12 of 25 U2/U3 CTS SLs (Digital) 2 2.0 2.0 SAFETY LIMITS (SLs) (Digital) 2 2.1 2.1 SLs 2.1.1 2.1.1 Reactor Core SLs 2.1.1.1 2.1.1.1 In MODES 1 and 2, departure from nucleate boiling ratio (DNBR) shall be maintained at [1.19]. 1.31 3 2.1.1.2 2.1.1.2 In MODES 1 and 2, the peak fuel centerline temperature shall be maintained at < [5080]°F, decreasing by [58°F per 10,000 MWD/MTU]

3 and adjusted for burnable poison per [CENPD-275-P, Revision 1-P-A or CENPD-382-P-A].

(RCS) 2.1.2 2.1.2 Reactor Coolant System Pressure SL 4 In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained [2750] psia. 3 2.2 2.2 SAFETY LIMIT VIOLATIONS 2.2.1 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 2.2.2 If SL 2.1.2 is violated:

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

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

CEOG STS 2.0-1 Rev. 3.0, 03/31/04 1 San Onofre - Draft Amendment XXX Attachment 1, Volume 2, Rev. 0, Page 12 of 25

Attachment 1, Volume 2, Rev. 0, Page 13 of 25 JUSTIFICATION FOR DEVIATIONS ITS 2.0, SAFETY LIMITS

1. Changes are made (additions, deletions, and/or changes) to the ISTS which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The headings for ISTS 2.0 include the parenthetical expression "(Digital)." This identifying information is not included in the San Onofre Nuclear Generating Station (SONGS) ITS. This information is provided in the NUREG to assist in identifying the appropriate Specification to be used as a model for a plant specific ITS conversion, but serves no purpose in a plant specific ITS implementation. SONGS Units 2 and 3 are digital plants; therefore analog requirements and specific labels that identify a requirement is digital are not required.

3. The ISTS contains bracketed information and/or values that are generic to all Combustion Engineering vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the information/value is changed to reflect the current licensing basis.

4. The acronym "RCS" is used in ISTS 2.1.2 and needs to be defined prior to its use.

Therefore, the words "(RCS)" have been added to the header for ISTS 2.1.2. This is consistent with the Writers Guide for the Improved Standard Technical Specifications, TSTF GG-05-01.

San Onofre Unit 2 and 3 Page 1 of 1 Attachment 1, Volume 2, Rev. 0, Page 13 of 25

Attachment 1, Volume 2, Rev. 0, Page 14 of 25 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 2, Rev. 0, Page 14 of 25

Attachment 1, Volume 2, Rev. 0, Page 15 of 25 RCS Core SLs (Digital) 7 B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs (Digital) 7 BASES BACKGROUND GDC 10 (Ref. 1) requires and SLs ensure that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).

This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95% probability at a 95%

confidence level (95/95 DNB criterion) that DNB will not occur and by requiring that fuel centerline temperature stays below the melting temperature.

The restrictions of this SL prevent overheating of the fuel and cladding peak Centerline and possible cladding perforation that would result in the release of temperature 8 fission products to the reactor coolant. Overheating of the fuel is below the melting point prevented by maintaining the steady state, peak linear heat rate (LHR) below the level at which fuel centerline melting occurs. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

linear heat rate (LHR) 8 Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in the heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

The Reactor Protective System (RPS), in combination with the LCOs, is designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, pressure, and THERMAL POWER level that would result in a violation of the reactor core SLs.

CEOG STS San Onofre - Draft B 2.1.1-1 Revision XXX Rev. 3.0, 03/31/04 1 Attachment 1, Volume 2, Rev. 0, Page 15 of 25

Attachment 1, Volume 2, Rev. 0, Page 16 of 25 RCS Core SLs (Digital) 7 B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AOOs. The reactor core SLs are established to ANALYSES preclude violation of the following fuel design criteria:

a. There must be at least a 95% probability at a 95% confidence level (95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB and

b. The hot fuel pellet in the core must not experience centerline fuel melting.

The RPS setpoints, LCO 3.3.1, "Reactor Protective System (RPS)

Instrumentation," in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for RCS temperature, pressure, and THERMAL POWER level that would result in a departure from nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities.

Automatic enforcement of these reactor core SLs is provided by the following functions:

a. Pressurizer Pressure - High trip,

b. Pressurizer Pressure - Low trip,

c. Linear Power Level - High trip,

d. Steam Generator Pressure - Low trip,

e. Local Power Density - High trip,

f. DNBR - Low trip,

g. Steam Generator Level - Low trip,

h. Reactor Coolant Flow - Low trip, and ,

j. Logarithmic Power Level - High trip, and 2

i. Steam Generator Safety Valves.

k. Containment Pressure - High trip The limitation that the average enthalpy in the hot leg be less than or equal to the enthalpy of saturated liquid also ensures that the T measured by instrumentation used in the protection system design as a measure of the core power is proportional to core power.

CEOG STS San Onofre - Draft B 2.1.1-2 Revision XXX Rev. 3.0, 03/31/04 1 Attachment 1, Volume 2, Rev. 0, Page 16 of 25

Attachment 1, Volume 2, Rev. 0, Page 17 of 25 RCS Core SLs (Digital) 7 B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued)

The SL represents a design requirement for establishing the protection system trip setpoints identified previously. LCO 3.2.1, "Linear Heat Rate (LHR)," and LCO 3.2.4, "Departure From Nucleate Boiling Ratio (DNBR),"

U or the assumed initial conditions of the safety analyses (as indicated in the FSAR, Ref. 2) provide more restrictive limits to ensure that the SLs 1 are not exceeded.

SAFETY LIMITS SL 2.1.1.1 and SL 2.1.1.2 ensure that the minimum DNBR is not less than the safety analyses limit and that fuel centerline temperature remains below melting.

1.31 The minimum value of the DNBR during normal operation and design basis AOOs is limited to [1.19], based on a statistical combination of 6 CE-1 CHF correlation and engineering factor uncertainties, and is established as an SL. Additional factors such as rod bow and spacer grid size and placement will determine the limiting safety system settings required to ensure that the SL is maintained. Maintaining the dynamically 8

adjusted peak LHR to 21 kW/ft ensures that fuel centerline melt will not occur during normal operating conditions or design AOOs.

SL 2.1.1.2 ensures that fuel centerline temperature remains below the fuel melt temperature of [5080]F during normal operating conditions or design AOOs with adjustments for burnup and burnable poison. An adjustment of [58F per 10,000 MWD/MTU] has been established in 6

[Topical Report CEN-386-P-A] (Ref. 3) and adjustments for burnable poisons are established based on [Topical Reports CENPD-275-P]

(Ref. 4) and [CENPD-382-P-A] (Ref.5).

APPLICABILITY SL 2.1.1.1 and SL 2.1.1.2 only apply in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are 4

specified in LCO 3.3.1.

In MODES 3, 4, 5, and 6, Applicability is not required, since the reactor is not generating significant THERMAL POWER.

CEOG STS San Onofre - Draft B 2.1.1-3 Revision XXX Rev. 3.0, 03/31/04 1 Attachment 1, Volume 2, Rev. 0, Page 17 of 25

Attachment 1, Volume 2, Rev. 0, Page 18 of 25 RCS Core SLs (Digital) 7 B 2.1.1 BASES SAFETY LIMIT The following SL violation responses are applicable to the reactor core VIOLATIONS SLs. 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 Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of bringing the unit to a MODE where this SL is not applicable and reduces the probability of fuel damage.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10, 1988.

U 15.0.3.2

2. FSAR, Section [ ]. 1 6

3. [Topical Report CEN-386-P-A, Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/kgU for Combustion Engineering 6 16x16 PWR Fuel, August 1992.]

4. [Topical Report CENPD-275-P, Revision 1-P-A, CE Methodology for 6 Core Designs Containing Gadolini-Urania Burnable Absorbers, 1 May 1988.] a 1

5. [Topical Report CENPD-382-P-A, Methodology for Core Designs 6

Containing Erbium Burnable Absorbers, August 1993.]

CEOG STS San Onofre - Draft B 2.1.1-4 Rev. 3.0, 03/31/04 1 Revision XXX Attachment 1, Volume 2, Rev. 0, Page 18 of 25

Attachment 1, Volume 2, Rev. 0, Page 19 of 25 RCS Pressure SLs (Digital) 7 B 2.1.2 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL (Digital) 7 BASES BACKGROUND The SL on RCS pressure protects the integrity of the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. By establishing an upper limit on RCS pressure, continued RCS integrity is ensured. According to 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and GDC 15, "Reactor Coolant System Design" (Ref. 1), the reactor coolant pressure boundary (RCPB) design conditions are not to be exceeded during normal operation and anticipated operational occurrences (AOOs). Also, according to GDC 28 (Ref. 1), "Reactivity Limits," reactivity accidents, including rod ejection, do not result in damage to the RCPB greater than limited local yielding.

The design pressure of the RCS is 2500 psia. During normal operation and AOOs, the RCS pressure is kept from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, according to the ASME Code requirements prior to initial operation, when there is no fuel in the core. Following inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of ASME Code,Section XI (Ref. 3).

Overpressurization of the RCS could result in a breach of the RCPB. If this occurs in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases specified in 10 CFR 100, "Reactor Site 1 Criteria" (Ref. 4). 50.67, "Accident Source Term" APPLICABLE The RCS pressurizer safety valves, the main steam safety valves SAFETY (MSSVs), and the Reactor Pressure - High trip have settings established ANALYSES to ensure that the RCS pressure SL will not be exceeded.

The RCS pressurizer safety valves are sized to prevent system pressure from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code for Nuclear Power Plant Components (Ref. 2). The transient that establishes the required relief capacity, and 6

hence the valve size requirements and lift settings, is a [complete loss of CEOG STS B 2.1.2-5 Revision XXX Rev. 3.0, 03/31/04 1 San Onofre - Draft Attachment 1, Volume 2, Rev. 0, Page 19 of 25

Attachment 1, Volume 2, Rev. 0, Page 20 of 25 RCS Pressure SLs (Digital) 7 B 2.1.2 BASES APPLICABLE SAFETY ANALYSES (continued) 6 external load without a direct reactor trip]. During the transient, no control actions are assumed except that the safety valves on the secondary plant are assumed to open when the steam pressure reaches the secondary plant safety valve settings, and nominal feedwater supply is maintained.

The Reactor Protective System (RPS) trip setpoints (LCO 3.3.1, "Reactor Protective System (RPS) Instrumentation"), together with the settings of the MSSVs (LCO 3.7.1, "Main Steam Safety Valves (MSSVs)") and the pressurizer safety valves, provide pressure protection for normal operation and AOOs. In particular, the Pressurizer Pressure - High Trip setpoint is specifically set to provide protection against overpressurization (Ref. 5). Safety analyses for both the Pressure - High Trip and the RCS pressurizer safety valves are performed, using conservative assumptions relative to pressure control devices.

More specifically, no credit is taken for operation of any of the following:

a. Pressurizer power operated relief valves (PORVs),

a b. Steam Bypass Control System, 3

b c. Pressurizer Level Control System, or c d. Pressurizer Pressure Control System.

SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure. The 5 maximum transient pressure allowable in the RCS piping, valves, and fittings under [USAS, Section B31.1 (Ref. 6)], is 120% of design pressure.

The most limiting of these two allowances is the 110% of design pressure; therefore, the SL on maximum allowable RCS pressure is established at 2750 psia.

APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4, and 5 because this SL could be approached or exceeded in these MODES due to overpressurization events. The SL is not applicable in MODE 6 because the reactor vessel head closure bolts are not fully tightened, making it unlikely that the RCS can be pressurized.

CEOG STS B 2.1.2-6 Revision XXX Rev. 3.0, 03/31/04 1 San Onofre - Draft Attachment 1, Volume 2, Rev. 0, Page 20 of 25

Attachment 1, Volume 2, Rev. 0, Page 21 of 25 RCS Pressure SLs (Digital) 7 B 2.1.2 BASES SAFETY LIMIT The following SL violation responses are applicable to the RCS pressure VIOLATIONS SL.

2.2.2.1 If the RCS pressure SL is violated when the reactor is in MODE 1 or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

With RCS pressure greater than the value specified in SL 2.1.2 in MODE 1 or 2, the pressure must be reduced to below this value. A pressure greater that the value specified in SL 2.1.2 exceeds 110% of the RCS design pressure and may challenge system integrity.

The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provides the operator time to complete the necessary actions to reduce RCS pressure by terminating the cause of the pressure increase, removing mass or energy from the RCS, or a combination of these actions, and to establish MODE 3 conditions.

2.2.2.2 If the RCS pressure SL is exceeded in MODE 3, 4, or 5, RCS pressure must be restored to within the SL value within 5 minutes.

Exceeding the RCS pressure SL in MODE 3, 4, or 5 is potentially more severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. This action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.

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.

50.67

4. 10 CFR 100. 1 U

1 6

5. FSAR, Section [ ]. 7.2

[ 6. ASME, USAS B31.1, Standard Code for Pressure Piping, 1967. ] 5 CEOG STS B 2.1.2-7 Revision XXX Rev. 3.0, 03/31/04 1 San Onofre - Draft Attachment 1, Volume 2, Rev. 0, Page 21 of 25

Attachment 1, Volume 2, Rev. 0, Page 22 of 25 JUSTIFICATION FOR DEVIATIONS ITS 2.0 BASES, SAFETY LIMITS

1. Changes are made (additions, deletions, and/or changes) to the Improved Standard Technical Specification (ISTS) Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. The addition of "Logarithmic Power Level - High trip" and Containment Pressure -

High trip" to the San Onofre Nuclear Generating Station (SONGS) Units 2 and 3 Specific ITS Bases 2.1.1, ASA, is consistent with the CTS Bases 2.1.1, ASA. This deviation from the ISTS reflects additional methods of automatic enforcement of the reactor core SLs utilized by SONGS, and is therefore added to the list.

3. The deletion of the Pressurizer Power Operated Relief Valves (PORVs) from the SONGS ITS Bases Section 2.1.2, ASA, is consistent with the SONGS CTS and proposed ITS which do not contain a PORV TS. The CE 3410 vintage plants (which, SONGS Units 2 and 3 are) do not credit a PORV in any safety analysis nor do they have a PORV as outlined in NUREG-1432. Due to this deletion the subsequent items in the list have been renumbered.

4. The SONGS ITS Bases Section 2.1.1, Applicability, will not include the last sentence in the paragraph, "Setpoints for the reactor trip functions are specified in LCO 3.3.1."

This deviation is consistent with the SONGS CTS Bases and with the intent of TSTF-493 Option B which relocates the "Allowable Values" to Licensee control; therefore reference to the setpoints being in LCO 3.3.1 is not needed.

5. The following statement was not included in the San Onofre ITS 2.1.2 Bases.

"The maximum transient pressure allowable in the RCS piping, valves, and fittings under [USAS Section B31.1 (Ref. 6)] is 120% of design pressure. The most limiting of these two allowances is the 110% of design pressure."

This deviation from the ISTS is consistent with the SONGS CTS. This statement is not applicable to SONGS (SONGS is designed to ASME Section III Code) and therefore should not be included in the TS Bases.

6. The ISTS Bases contains bracketed information and/or values that are generic to all Combustion Engineering vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the information/value is changed to reflect the current licensing basis.

7. The headings for ISTS 2.0 Bases include the parenthetical expression "(Digital)."

This identifying information is not included in the SONGS ITS. This information is provided in the NUREG to assist in identifying the appropriate Specification to be used as a model for a plant specific ITS conversion, but serves no purpose in a plant specific ITS implementation. SONGS Units 2 and 3 are digital plants; therefore analog requirements and specific labels that identify a requirement is digital are not required.

8. The Safety Limit is based on the peak centerline fuel temperature, not the linear heat rate. The linear heat rate limit specified in LCO 3.2.1 is a steady state limit, not an accident limit. Therefore, the words in the second paragraph of the Background section of the Bases have been changed to be consistent with the actual Safety Limit requirement. The linear heat rate limit was changed to fuel centerline limit in TSTF-San Onofre Unit 2 and 3 Page 1 of 2 Attachment 1, Volume 2, Rev. 0, Page 22 of 25

Attachment 1, Volume 2, Rev. 0, Page 23 of 25 JUSTIFICATION FOR DEVIATIONS ITS 2.0 BASES, SAFETY LIMITS 445, but these words were apparently not modified. Furthermore, since the term LHR is not being defined in this paragraph, the subsequent use of LHR in the next paragraph is being defined, consistent with the ISTS format.

In addition, the last sentence in the second paragraph of the Safety Limit section of the Bases is also being deleted, consistent with TSTF-445. This sentence was deleted by this TSTF, but when Rev 3.0 was issued, the words were inadvertently left in.

San Onofre Unit 2 and 3 Page 2 of 2 Attachment 1, Volume 2, Rev. 0, Page 23 of 25

Attachment 1, Volume 2, Rev. 0, Page 24 of 25 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 2, Rev. 0, Page 24 of 25

Attachment 1, Volume 2, Rev. 0, Page 25 of 25 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 2.0, SAFETY LIMITS There are no specific No Significant Hazards Considerations for this Specification.

San Onofre Unit 2 and 3 Page 1 of 1 Attachment 1, Volume 2, Rev. 0, Page 25 of 25

ML11251A095

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