Text

License Amendment Request Technical Specification Change Request for Modification of ANO-2 RCS
	ML041100661
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/15/2004
From:	Forbes J; Entergy Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	2CAN040405
	Download: ML041100661 (84)
	v • d • e

Entergy Operations, Inc.

1448SR 333 Entergy Russellville, AR 72802 Jeffery S. Forbes Vice President Operations ANO 2CAN040405 April 15, 2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

License Amendment Request Technical Specification Change Request for Modification of ANO-2 RCS Pressure/Temperature Cooldown Rates Arkansas Nuclear One, Unit 2 Docket No. 50-368 License No. NPF-6

REFERENCES:

1 Entergy letter dated October 30, 2001, Proposed Technical Specification Change Request Regarding Revised ANO-2 Pressure/Temperature and Low Temperature Overpressure Protection Limits for 32 Effective Full Power Years (2CAN1 00101) 2 NRC letter dated April 15, 2002, Arkansas Nuclear One, Unit No. 2 -

Issuance of Amendment Re: Reactor Vessel Pressure-Temperature Limits And Exemption From The Requirements Of 10 CFR Part 50, Section 50.60(a) (2CNA040205)

Dear Sir or Madam:

Pursuant to IOCFR50.90, Entergy Operations, Inc. (Entergy) hereby requests an amendment to the Arkansas Nuclear One, Unit 2 (ANO-2) Operating License No. NPF-6. This amendment request revises the ANO-2 Reactor Coolant System (RCS) Pressure/Temperature (P/T) cooldown limits contained in Technical Specification (TS) 3/4.4.9. The current ANO-2 P/T cooldown limits are contained in Limiting Condition for Operation (LCO) 3.4.9.1.b, which provides a variable cooldown rate based on RCS cold leg temperature conditions. The revision to LCO 3.4.9.1.b is proposed to be a single maximum cooldown rate of 1000 F per hour (constant) or 50° F in any half hour period (step) for RCS cold leg temperatures between 50° F and 5600 F. This change provides enhanced ANO-2 operational flexibility. The revised analysis for the P/T cooldown limits resulted in a slightly more restrictive change to TS Figure 3.4-2B, which is the Cooldown Curve for 32 Effective Full Power Years (EFPY). The reactor vessel AOl

2CAN040405 Page 2 of 3 material properties, fluence values and heatup rates are not being revised as part of this proposed change. Therefore, the heatup limits, LTOP and inservice hydrostatic test (ISLH) limits currently contained in the ANO-2 TSs are unaffected. The proposed changes are discussed in Attachment 1 and the proposed TS pages are contained in Attachment 2. A summary report has been developed by Framatome for the fracture toughness analyses of the ANO-2 reactor vessel. This report, BAW-2405, Revision 2, Appendix G Pressure-Temperature Limits for 32 EFPY, Using ASME Code Cases forANO-2 is contained in Attachment 4.

The proposed change has been evaluated in accordance with 10CFR50.91(a)(1) using criteria in 10CFR50.92(c) and it has been determined that this change involves no significant hazards consideration. Commitments being proposed by this amendment request are contained in. As discussed in Reference 1, Entergy had proposed a similar change for ANO-2 that established the current P/T limits for heatup, cooldown, and inservice hydrostatic testing for 32 EFPY in October of 2001 (Reference 1). The NRC approved this amendment request as discussed in the April 15, 2002 Safety Evaluation Report (Reference 2).

Entergy requests approval of the proposed amendment by January 28, 2005, in order to support the 2R17 refueling outage in the spring of 2005. Once approved, the amendment shall be implemented within 60 days. Although this request is neither exigent nor emergency, your prompt review is requested.

If you have any questions or require additional information, please contact Steve A Bennett at 479-858-4626.

I declare under penalty of perjury that the foregoing is true and correct. Executed on April 15, 2004.

Sincerely, JSF/sab Attachments:

1. Analysis of Proposed Technical Specification Change

2. Proposed Technical Specification Changes (mark-up)

3. List of Regulatory Commitments

4. BAW-2405, Revision 2, Appendix G Pressure-Temperature Limits for 32 EFPY, Using ASME Code Cases

2CAN040405 Page 3 of 3 cc:

Dr. Bruce S. Mallett Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Thomas W. Alexion MS O-7D1 Washington, DC 20555-0001 Mr. Bernard R. Bevill Director Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 West Markham Street Little Rock, AR 72205 to 2CAN040405 Page 1 of 6

1.0 DESCRIPTION

This letter requests an amendment to Operating License No. NPF-6 for Arkansas Nuclear One, Unit 2 (ANO-2). The proposed change revises the ANO-2 Operating License to modify the Reactor Coolant System (RCS) Pressure/Temperature (P/T) cooldown limits contained in Technical Specification (TS) 3/4.4.9.

2.0 PROPOSED CHANGE

The current ANO-2 P/T cooldown limits are contained in Limiting Condition for Operation (LCO) 3.4.9.1.b, which provides a variable cooldown rate based on RCS cold leg temperature (Tc) conditions as follows:

A maximum cooldown rate based on:

RCS Temperature (To)

Maximum Cooldown Rate Tc > 200OF 1200F *TY *200OF Tc < 120OF 100 cF per hour (constant) or 50 0 F in any half hour period (step) 60 OF per hour (constant) or 30 IF in any half hour period (step) 25OF per hour (constant) or 12.5'F in any half hour period (step)

The revision to LCO 3.4.9.1.b is proposed to be:

A maximum cooldown rate of 1000Fperhour(constant) or 500 F in anyhalf hourperiod (step) for RCS cold leg temperatures between 500F and 5600F.

The revised analysis for the new P/T cooldown limits also resulted in a change to TS Figure 3.4-2B, Cooldown Curve - 32 EFPY. This curve is slightly more restrictive than the current curve for normal cooling down of the unit.

The existing TS Bases discuss how the TS limits were developed, but are not prescriptive in the actual limits. Therefore, no changes to the TS Bases are necessary.

The reactor vessel material properties, fluence values and heatup rates are not being revised as part of this proposed change. Therefore, the heatup limits, and inservice hydrostatic test (ISLH) limits in TS 3.4.9 do not require revision.

to 2CAN040405 Page 2 of 6

3.0 BACKGROUND

As discussed in Reference 1, Entergy had previously proposed changes that established the current pressure/temperature (PIT) limits for 32 effective full power years (EFPY) in October of 2001. In preparation for the proposed change, a second reactor vessel specimen had been removed and analyzed for better determination of fluence and for establishing reactor vessel fracture toughness. As a result of revised RTNDT analyses and fluence values, new heatup, cooldown, and inservice hydrostatic test (ISLH) curves were developed. The low temperature overpressure protection (LTOP) conditions were also reanalyzed. A summary report developed by Framatome for the revised fracture toughness analyses of the Arkansas Nuclear One, Unit 2 (ANO-2) reactor vessel was provided in the initial issuance of BAW-2405, Appendix G Pressure-Temperature Limits for 32 EFPY, Using ASME Code Cases for ANO-2, September 2001: The NRC approved these changes on April 15, 2002 as discussed in the Safety Evaluation Report in Reference 2.

New analysis was performed with more flexible cooldown limitations using a maximum cooldown rate of 1000F per hour (constant) or 50° F in any half hour period (step) for RCS cold leg temperatures between 50° F and 5600 F. It was determined that plant cooldown can be accomplished under appropriately restrictive conditions to protect against reactor vessel brittleness for 32 EFPY using these limits. Therefore, the proposed license amendment establishes revised P/T limits for cooling down the RCS. The revised analysis was also performed by Framatome ANP. Revision 2 to BAW-24051 is contained in Attachment 2 and provides the basis for the acceptability of this proposed change.

4.0 TECHNICAL ANALYSIS

10CFR50, Appendix G, Fracture Toughness Requirements, specifies minimum fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of light water-cooled power reactors and provides specific guidelines for determining the pressure-temperature limitations. The fracture toughness and operational requirements are specified to provide adequate safety margins during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime.

The purpose of the fracture toughness analysis is to provide location adjusted (corrected) P/T limits, for the ANO-2 reactor vessel for 32 EFPY of operation. The PIT limits are calculated for the reactor vessel beltline, inlet and outlet nozzles, and closure head flange locations for normal cooldown. These limits are usually expressed in the form of a curve of allowable pressure versus temperature. The 1/4 t wall location is defined as a point within the vessel wall that is located at a distance of one-quarter of the vessel section thickness from the cladding-base metal interface.

Pressure/temperature limits were developed using Framatome ANP computer code PTPC.

The analytical approach is in accordance with the requirements of the ASME Code, Section Xl, Appendix G. The analytical procedures used to calculate P/T limits are based on linear 1 As noted in the Record of Revision for BAW-2405, Revision 2, Revision 1 of this report only modified a correction in a table value.

to 2CAN040405 Page 3 of 6 elastic fracture mechanics methods for calculating stress intensity factors at the maximum depths of postulated semi-elliptical surface flaws as discussed in Section 6 of BAW-2405, Revision 2.

This analysis was performed in accordance with the requirements of 1 OCFR50, Appendix G, ASME Code Section Xl, Appendix G, ASME Code Case N-640, Alternative Reference Fracture Toughness for Development of P-T Limit Curves, Section Xl, Division 1, ASME Code Case N-588, Attenuation to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels, Section Xl, Division 1.

ASME Code Case N-640 allows the use KIc reference fracture toughness curve from Section XI, Appendix A. When this Code Case is employed, LTOP systems are required to limit the maximum pressure in the vessel to 100% of the pressure allowed by the P/T limit curves.

When a given material is indexed to the KIc curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Plant operating limits can then be determined using these allowable stress intensity factors. ASME Code Case N-588 contains explicit methodology for calculating membrane tension (Mm) and contains an influence coefficient solution for KIT (thermal stress intensity factor). ASME Code Case N-588 also provides relief from the Appendix G requirement that all postulated flaws must be oriented normal to the direction of maximum stress by permitting flaws in circumferential welds to be oriented in the circumferential direction, such that the normal direction is aligned with the axis of the vessel. This is beneficial to plants whose limiting material is a circumferential weld since axial pressure stress is roughly one-half of the hoop stress. Since the limiting beltline material for ANO-2 is a plate material, the more limiting axial flaw is postulated.

For each analyzed transient and steady state condition, the allowable pressure is determined as a function of reactor coolant temperature considering postulated flaws in the reactor vessel beltline, closure head flange, and inlet nozzle and outlet nozzle. In the beltline region, flaws are postulated to be present at the 1/4 t and 3/4 t locations of the controlling material (shell plate, longitudinal weld, or circumferential weld), as defined by the fluence adjusted RTNDT. The reactor vessel nozzle flaws are located at the inside juncture of the nozzle.

These location specific P/T limits are obtained from the PTPC output files.

The maximum allowable pressure at a particular fluid temperature is taken as the minimum value of allowable pressure calculated for each flaw location and operating condition, including steady state. A P/T cooldown limit curve is then constructed as the collection of points that define the maximum allowable pressures as a function of fluid temperature for a particular mode of reactor operation. The P/T curves provided in BAW-2405, Revision 2 are adjusted for sensor location, but do not account for any instrument uncertainty. The P/T limit instrument uncertainty will be added to the cooldown curves contained in the ANO-2 operating procedures.

A summary of the new corrected cooldown P/T limits including composite cooldown P/T limits (bounding cooldown limits) are reported in Table 7-2 and illustrated in Figure 7-2 of BAW-2405, Revision 2. These composite cooldown P/T limits were established based on considering a linear ramp rate of 100 IF /hr as well as series of 50 OF step changes followed by half-an-hour hold periods from 560 IF to 50 IF.

to 2CAN040405 Page 4 of 6

5.0 REGULATORY ANALYSIS

5.1 Applicable Regulatory Requirements/Criteria ANO-2 Compliance with 10CFR50. Appendix G - 10CFR50, Appendix G, specifies minimum fracture toughness requirements for the ferritic materials of the pressure-retaining components of the RCPB of light water-cooled power reactors and provides specific guidelines for determining the pressure-temperature limitations. The revised P/T cooldown limits for the ANO-2 reactor vessel are developed in accordance with the requirements of 1 OCFR Part 50, Appendix G, and ASME Code Section Xl, Appendix G, and ASME Code Cases N-640 and N-588. Code Cases N-588 and N-640 have been approved for general use per Regulatory Guide 1.147, Revision 13 (June 2003).

Evaluation of ANO-2 Safety Analysis Report - Section 5.2.4 of the ANO-2 SAR provides ANO-2's assurance that the reactor coolant pressure boundary is within the fracture toughness requirements of 10CFR50, Appendix G. ANO-2 SAR Section 5.2.4.3.1 specifically discusses the approach and methodology for calculating the P/T limits for 32 EFPY. The P/T cooldown curve is not contained in the SAR, but is referenced to the ANO-2 TSs. Therefore, the discussions within this section are unchanged by this proposed amendment to the TSs other than to revise the reference to BAW-2405 which will now be Revision 2. The ANO-2 SAR will be modified to reflect the reference to Revision 2 of BAW-2405 in the next SAR revision cycle.

SAR Section 3.1.2 discusses ANO-2's compliance with the RCS pressure boundary General Design Criteria (GDC). Criterion 14, Reactor Coolant Pressure Boundary states that the reactor coolant pressure boundary (RCPB) shall be designed, fabricated, erected and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure and of gross rupture. To establish operating pressure and temperature limitations during startup and shutdown of the RCS, the fracture toughness rules defined in Appendix G of the ASME Code, Section Xl were followed. Quality control, inspection, and testing as required by this standard and allowable reactor P/T limits assure the integrity of the RCPB. Compliance with this section of the ANO-2 SAR is unaffected by the proposed change.

The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met. Entergy has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, other than the TS, and do not affect conformance with any GDC differently than described in the ANO-2 SAR.

5.2 No Sicqnificant Hazards Consideration The proposed amendment modifies the Operating License No. NPF-6 for Arkansas Nuclear One, Unit 2 (ANO-2) for the Reactor Coolant System (RCS) Pressure/Temperature (P/T) cooldown limits contained in Technical Specification (TS) 3/4.4.9. The revision to Limiting Condition for Operation (LCO) 3.4.9.1.b is proposed to be a maximum cooldown rate of 100°F per hour (constant) or 50° F in any half hour period (step) for RCS cold leg temperatures between 50° F and 5600 F. The current ANO-2 P/T cooldown limits contained in LCO 3.4.9.1.b provides a variable cooldown rate based on RCS cold leg temperature conditions.

The proposed amendment for the new P/T cooldown limits also includes a change to TS Figure 3.4-2B, Cooldown Curve - 32 EFPY (Effective Full Power Years).

to 2CAN040405 Page 5 of 6 Entergy Operations, Inc. has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three criteria set forth in 1 OCFR50.92, 'Issuance of amendment," as discussed below:

1.

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

Response: No.

The probability of occurrence of an accident previously evaluated for ANO-2 is not altered by the proposed amendment to the TSs. The accidents remain the same as currently analyzed in the ANO-2 Safety Analysis Report (SAR) as a result of the change to the cooldown P/T limits. The new P/T cooldown limits were based on NRC accepted methodologies along with ASME Code alternatives. The proposed change does not impact the integrity of the reactor coolant pressure boundary (RCPB) (i.e.,

there is no change to the operating pressure, materials, loadings, etc.) as a result of this change. In addition, there is no increase in the potential for the occurrence of a loss of coolant accident. The proposed P/T cooldown limit curve is not considered to be an initiator or contributor to any accident currently evaluated in the ANO-2 SAR.

The revised P/T cooldown limits ensure the long term integrity of the RCPB. For each analyzed transient and steady state condition, the allowable pressure was determined as a function of reactor coolant temperature considering postulated flaws in the reactor vessel beltline, inlet nozzle, outlet nozzle, and closure head flange.

Therefore, the proposed change 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?

Response: No.

The proposed changes to the P/T limits will not create a new accident scenario. The requirements to have PIT protection are part of the ANO-2 licensing basis. The proposed change in the P/T cooldown limits is based on NRC approved methodologies performed by Framatome ANP. This methodology complies with NRC and ASME requirements for protecting the RCS. Therefore, the revised P/T cooldown limits provide protection of the RCS from limiting transients during normal cooldown.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3.

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

Response: No.

The revision of the PIT limits and curves will ensure that ANO-2 continues to operate within the operating margins of the ASME Code.

The application of ASME Code Cases N-640 and N-588 presents alternative procedures for calculating P/T to 2CAN040405 Page 6 of 6 temperatures and pressures. These Code Cases allow certain assumptions to be conservatively reduced. However, the procedures allowed by these Code Cases still provide sufficient conservatism and ensure an adequate margin of safety in the development of P/T operating and pressure test limits to prevent non-ductile fractures.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Entergy concludes that the proposed amendment(s) present no significant hazards consideration under the standards set forth in 10CFR50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.3 Environmental Considerations The proposed amendment is confined to (i) changes to surety, insurance, and/or indemnity requirements, or (ii) changes to record keeping, reporting, or administrative procedures or requirements. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 1 OCFR51.22(c)(1 0). Therefore, pursuant to 1 OCFR51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 PRECEDENCE As discussed in Reference 1, Entergy had proposed a similar change that established new P/T limits for 32 EFPY in October of 2001. The NRC approved this amendment request on April 15, 2002, as discussed in the Safety Evaluation Report contained in Reference 2.

2CAN040405 Proposed Technical Specification Changes (mark-up) to 2CAN040405 Page 1 of 3 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2A, 3.4-2B and 3.4-2C during heatup/criticality, cooldown, and inservice leak and hydrostatic testing operations with:

a. A maximum heatup of 50'F, 60'F, 70'F or 80'F in any one hour period in accordance with Figure 3.4-2A.

b.

A maximum cooldown rate of 100OF per hour (constant) or 50° F in any half hour period (step) for RCS cold leg temperatures between 50° F and 5600 F.

AxiumG ooldewn Fate-based on:

RG:Fernperatufe4T-G)

Maximur-Gooldewn-Rate T0 G-_IOO2F 2F!5 T, 2rm T-T.42O2F n02F-pler-heur-(rnat)aor-50°F-i4 any-14alf-uFperied (step) arny-alf-houF-period (step) 25aF-per-hour-(ionstant)-aFr2.50F-in any-1alf-houF-period(step)

c.

A maximum temperature change of

100F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY:

At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the acceptable region of the applicable curve within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS Tc and pressure to less than 2000F and less than 500 psia, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ARKANSAS - UNIT 2 3/4 4-22 Amendment No. 424,242 to 2CAN040405 Page 2 of 3 Figure 3.4-2B COOLDOWN CURVE - 32 EFPY REACTOR COOLANT SYSTEM PRESSURE/TEMPERATURE LIMITS 2500 2000 am N

3 Un U,

a)

L-1500 1000 Accetable Lowest Service

-nimum Boltup

$Temperature, 50 F-500 0

0 100 200 300 400 500 Reactor Coolant Temperature T0, OF (Curves do not include margins for instrument uncertainties)

THIS CURVE IS BEING DELETED 3/4 4-23a Amendmr ARKANSAS - UNIT 2 mt No. 4-24,24 to 2CAN040405 Page 3 of 3 Figure 3.4-2B COOLDOWN CURVE - 32 EFPY REACTOR COOLANT SYSTEM PRESSURE/TEMPERATURE LIMITS 2500 I I I I

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00 ARKANSAS - UNIT 2 3/4 4-23a Amendment No. 424,242

2CAN040405 List of Regulatory Commitments to 2CAN040405 Page 1 of 1 List of Regulatory Commitments The following table identifies those actions committed to by Entergy in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

TYPE SCHEDULED (Check one)

COMPLETION COMMITMENT DATE (If ONE-CONTINUING Required)

TIME COMPLIANCE ACTION The P/T curves provided in BAW-2405, Revision X

Prior to 2 are adjusted for sensor location but do not implementation account for any instrument uncertainty. The P/T of the TS limit instrument uncertainty will be added to the Change cooldown curves contained in the ANO-2 operating procedures.

The ANO-2 SAR will be modified to reflect the X

Within the next reference to Revision 2 of BAW-2405.

ANO-2 SAR update cycle after NRC approval.

2CAN040405 BAW-2405, Revision 2, Appendix G Pressure-Temperature Limits for 32 EFPY, Using ASME Code Cases

BAW-2405, Rev. 2 April 2004 APPENDIX G PRESSURE-TEMPERATURE LIMITS FOR 32 EFPY, USING ASME CODE CASES, FOR Arkansas Nuclear One Unit 2 Power Plant At AR EVA

BAW-2405, Rev. 2 APPENDIX G PRESSURE-TEMPERATURE LIMITS FOR 32 EFPY, USING ASME CODE CASES, FOR Arkansas Nuclear One Unit 2 Power Plant By A. D. Nana AREVA Document No. 77-2405-02 Prepared for Entergy Operations, Inc.

Prepared by Framatome ANP, Inc.

An AREVA and Siemens Company 3315 Old Forest Road P. O. Box 10935 Lynchburg, Virginia 24506-0935 A,

AR EVA

BAW-2405, Rev. 2 RECORD OF REVISIONS Description of Revision Revision Date BAW-2405 BAW-2405, Rev. 1 BAW-2405, Rev. 2 Original Release Minor error on Table 7-2 & 7-6 at fluid temp. value of 195 F.

No impact on P-T limit curves since pressure is above 2500 psig Affected Pages: i through v, 7-5, 7-15, 9-1 Complete Re-release with revised normal cooldown 9/2001 4/2004 4/2004 ii Ai AR EVA

BAW-2405, Rev. 2 CONTENTS Section Page

1.

INTRODUCTION........................................................

1-1

2.

BACKGROUND........................................................

2-1

3.

ADJUSTED NIL-DUCTILITY TRANSITION REFERENCE TEMPERATURES............ 3-1

4.

PRESSURIZED THERMAL SHOCK REFERENCE TEMPERATURES....................... 4-1

5.

DESIGN BASIS FOR PRESSURE/TEMPERATURE LIMITS........................................ 5-1

6.

TECHNICAL BASIS FOR PRESSURE/TEMPERATURE LIMITS................

................. 6-1

7.

OPERATIONAL PRESSURE/TEMPERATURE LIMITS................................................. 7-1

8.

LTOP PRESSURE/TEMPERATURE LIMITS........................................................

8-1

9.

CERTIFICATION........................................................

9-1

10.

REFERENCES........................................................

10-1 iii Ai AR EVA

BAW-2405, Rev. 2 LIST OF TABLES Table Page 3-1 Description of ANO-2 Reactor Vessel Beltline Region Materials............................... 3-2 3-2 Adjusted Reference Temperatures for ANO-2 Applicable to 32 EFPY with Power Uprate.3-3 4-1 Pressurized Thermal Shock Reference Temperatures for ANO-2 Applicable to 32 EFPY with Power Uprate.....................................................

4-2 5-1 ANO-2 Design Data.....................................................

5-3 5-2 ANO-2 Material Properties.....................................................

5-4 5-3 Limiting RTNDT'S for ANO-2 Reactor Vessel Materials at 32 EFPY............................ 5-5 7-1 ANO-2 Summary of Ramp Heatup P/T Limits & Critical Core Limits (psig).............. 7-2 7-2 ANO-2 Cooldown Composite P/T Limits (psig).....................................................

7-5 7-3 ANO-2 ISLH P/T Limits (psig).....................................................

7-6 7-4 ANO-2 Steady-State "Isothermal Condition" P/T Limits (psig)................................... 7-9 7-5 ANO-2 Summary of Ramp Heatup P/T Limits & Critical Core Limits (psia).............. 7-12 7-6 ANO-2 Cooldown Composite P/T Limits (psia).....................................................

7-15 7-7 ANO-2 ISLH P/T Limits (psia).....................................................

7-16 7-8 ANO-2 Steady-State "Isothermal Condition" P/T Limits (psia)................................... 7-19 8-1 ANO-2 P/T Limits for LTOP (psig).....................................................

8-2 LIST OF FIGURES Figure Page 3-1 Location and Identification of ANO-2 Reactor Vessel Materials................................ 3-4 3-2 Location of ANO-2 Longitudinal Welds in the Reactor Vessel Upper and Lower Shell Courses......................................................

3-5 5-1 ANO-2 Ramp Heatup Temperature Transients......................................................

5-6 5-2 ANO-2 Ramp Cooldown Temperature Transient - Set 1 (normal shutdown)........... 5-7 5-3 ANO-2 Step Cooldown Temperature Transient - Set 1 (normal shutdown)............. 5-8 5-4 ANO-2 Ramp Cooldown Temperature Transient - Set 2 (acid reduction shutdown).5-9 5-5 ANO-2 Step Cooldown Temperature Transient - Set 2 (acid reduction shutdown)...........................................

5-10 7-1 ANO-2 Ramp Heatup & Critical Core P/T Limit Curves (psig)................................... 7-22 iv DA AR EVA

BAW-2405, Rev. 2 Figure 7-2 7-3 7-4 7-5 7-6 8-1 LIST OF FIGURES (Cont'd.)

Page ANO-2 Composite Cooldown P/T Limit Curve (psig)..........................................

7-23 ANO-2 ISLH P/T Limit Curve (psig)..........................................

7-24 ANO-2 Ramp Heatup & Critical Core P/T Limit Curves (psia)................................... 7-26 ANO-2 Composite Cooldown P/T Limit Curve (psia)..........................................

7-28 ANO-2 ISLH P/T Limit Curve (psia)..........................................

7-30 ANO-2 P/T Limits for LTOP (psig)..........................................

8-3 v

At AR EVA

BAW-2405, Rev. 2

1. INTRODUCTION This report presents pressure/temperature (PIT) limits for the Arkansas Nuclear One Unit 2 (ANO-2) reactor vessel at 32 effective full power years (EFPY) of operation including an estimated increase in fluence due to a proposed power uprate. The data used to develop these operational limits are based on the evaluation of the ANO-2 reactor vessel surveillance capsule.'1 Pressure-temperature limits are developed for normal heatup and cooldown operating conditions and inservice leak and hydrostatic (ISLH) test conditions.

AR EVA

BAW-2405, Rev. 2

2. BACKGROUND The ability of the reactor pressure vessel to resist fracture is the primary factor in ensuring the safety of the primary system in light water-cooled reactors. The beltline region of the reactor vessel is the most critical region of the vessel because it is exposed to neutron irradiation. The general effects of fast neutron irradiation on the mechanical properties of low-alloy ferritic steels such as SA533, Grade B plate material used in the fabrication of the ANO-2 reactor vessel, are well characterized and documented in the literature. The effects of irradiation on these steels include an increase in the yield and ultimate strengths and a decrease in ductility. The most significant effect, however, is an increase in the temperature associated with the transition from brittle to ductile fracture and a reduction in the Charpy upper-shelf energy value.

Appendix G to 10 CFR 50, "Fracture Toughness Requirements,"[21 specifies minimum fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of water-cooled power reactors, and provides specific guidelines for determining the pressure-temperature limitations on operation of the RCPB. The toughness and operational requirements are specified to provide adequate safety margins during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Although the requirements of Appendix G to 10 CFR 50 originally became effective on August 16, 1973, the requirements are applicable to all boiling and pressurized water-cooled nuclear power reactors, including those under construction or in operation on the effective date.

Appendix H to 10 CFR 50, "Reactor Vessel Materials Surveillance Program Requirements," 13' defines the material surveillance program required to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of water-cooled reactors resulting from exposure to neutron irradiation and the thermal environment. Fracture toughness test data are obtained from material specimens periodically withdrawn from the reactor vessel.

These data will permit determination of the conditions under which the vessel can be operated with adequate safety margin against fracture throughout its service life.

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BAW-2405, Rev. 2 A method for guarding against brittle fracture in reactor pressure vessels is described in Appendix G to the ASME Boiler and Pressure Vessel Code, Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components."141 This method utilizes fracture mechanics concepts and the reference temperature for nil-ductility transition, RTNDT, which is defined as the greater of the drop weight nil-ductility transition temperature (per ASTM E20815]) or the temperature that is 60 OF below that at which the material exhibits 50 ft-lbs and 35 mils lateral expansion. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KIa or Kc curve as applicable). The Kia curve appears in Appendix G of ASME Code Section XI.

ASME Code Case N-640t1" permits the use of the K1c curve as given in Appendix A of ASME Code Section Xl. When a given material is indexed to the K1, curve, allowable stress intensity factors can be obtained for this material as a function of temperature.

Plant operating limits can then be determined using these allowable stress intensity factors.

The RTNDT of the reactor vessel materials, and in turn, the pressure/temperature limits of a reactor vessel, must be adjusted to account for the effects of irradiation. Neutron embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel are monitored by a surveillance program consisting of periodic removal of surveillance capsules from an operating reactor and testing of reactor vessel material specimens obtained from the capsules. The increase in the Charpy V-notch 30 ft-lb temperature is added to the unirradiated RTNDT to adjust it for neutron embrittlement. This adjusted RTNDT is used to index the material to the K1, curve, which in turn, is used to set new operating limits for the nuclear power plant.

These new limits take into account the effects of irradiation on the reactor vessel materials.

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BAW-2405, Rev. 2

3. ADJUSTED NIL-DUCTILITY TRANSITION REFERENCE TEMPERATURES The reactor pressure vessel beltline region consists of two shells, containing six heats of base metal plate, six longitudinal weld seams, and one circumferential weld seam. Table 3-1 presents a description of the reactor vessel beltline materials including their copper and nickel chemical contents and their unirradiated mechanical properties. The locations of the materials within the reactor vessel beltline region are shown in Figures 3-1 and 3-2.

The adjusted reference temperatures for the reactor vessel beltline region materials are calculated in accordance with Regulatory Guide 1.99, Revision 2161. The adjusted reference temperatures are calculated by adding the initial RTNDT, the predicted radiation-induced ARTNDT, and a margin term to cover the uncertainties in the values of initial RTNDT, copper and nickel contents, fluence, and the calculational procedures. The predicted radiation induced ARTNDT is calculated using the respective reactor vessel beltline materials copper and nickel contents and the neutron fluence applicable to 32 EFPY including an estimated increase in flux due to a proposed power uprate. The 1/4-thickness (%T) and 3/4-thickness (3 4T) wall locations for each beltline material are determined by adding the thickness of the cladding to the distance into the base metal at the /4T and 3/4T locations (i.e., 1/4T = (7.875

4] + 0.125 = 2.094 inches and YAT =

[7.875 - 3/4] + 0.125 = 6.031 inches).

The 'AT and %4T adjusted reference temperature resultsill for the ANO-2 reactor vessel beltline region materials applicable to 32 EFPY are presented in Table 3-2. Based on these results, the controlling beltline material for the ANO-2 reactor vessel is the lower shell plate C-801 0-1.

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BAW-2405. Rev. 2 Table 3-1. Description of Reactor Vessel Beitline Region Materials Chemical Comp osition Unirradiated Beltline Material Material Material Cu, Ni,

RTNDT, Region Location Identification Type Heat No.

wl%

Wt%

F Inter. Shell Long. Welds 2-203 Subarc Weld 10120 0.046 0.082

-56 Lower Shell Long. Welds 3-203 Subarc Weld 10120 0.046 0.082

-56 Lower/Inter. Shell Girth Weld 9-203 Subarc Weld 83650 0.045 0.087

-10 Intermediate Shell C-8009-1 SA-533B C1.1 C8161-3 0.098 0.605

-26 Intermediate Shell C-8009-2 SA-533B Cl.1 C8161-1 0.085 0.600 0

Intermediate Shell C-8009-3 SA-533B CI.1 C8182-2 0.096 0.580 0

Lower Shell C-8010-1 SA-533B C1.1 C8161-2 0.085 0.585 12 Lower Shell C-8010-2 SA-533B C1.1 B2545-1 0.083 0.668

-28 Lower Shell C-8010-3 SA-533B C0.1 B2545-2 0.080 0.653

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BAW-2405, Rev. 2 Table 3-2. Adjusted Reference Temperatures Applicable to 32 EFPY with Power Uprate Chemical 2

ARTNDT, F ARTNDT, F Material Description Composition 32 EFPY Fluence, nlcm2 at 32 EFPY at 32 EFPY Reactor Vessel Matl.

Heat Cu Ni Initial Chemistry Inside T14 3/4T T14 314T T/4 314T Beltline Location Ident.

Number Base Metal /

wt%

RTNOT Factor Wetted Location Location Location Location Location Location Flux Type Surface Regulatory Guide 1.99, Revision 2, Position 1.1 Inter. Shell Long. Weld 2-203 A 10120 Linde 0091 0.046 0.082

-56 34.0 3.568E+19 2.159E+19 8.391 E+18 41.1 32.3 38.4 23.2 Inter. Shell Long. Weld 2-203 B 10120 Linde 0091 0.046 0.082

-56 34.0 2.754E+19 1.666E+19 6.477E+18 38.8 29.9 34.4 19.2 Inter. Shell Long. Weld 2-203 C 10120 Linde 0091 0.046 0.082

-56 34.0 2.754E+19 1.66613+19 6.477E+18 38.8 29.9 34.4 19.2 Lower Shell Long. Weld 3-203 A 10120 Linde 0091 0.046 0.082

-56 34.0 3.582E+19 2.167E+19 8.4251E+18 41.1 32.4 38.4 23.4 Lower Shell Long. Weld 3-203 B 10120 Linde 0091 0.046 0.082

-56 34.0 2.765E+19 1.673E+19 6.5021E+18 38.8 29.9 34.4 19.2 Lower Shell Long. Weld 3-203 C 10120 Linde 0091 0.046 0.082

-56 34.0 2.765E+19 1.673E+19 6.502E+18 38.8 29.9 34.4 19.2 Inter./Lower Shell Girth 9-203 83650 Linde 0091 0.045 0.087

-10 34.1 3.776E+19 2.284E+19 8.88013+18 41.7 33.0 73.4 56.0 Weld Intermediate Shell Plate C-8009-1 C8161-3 SA-533B CI.1 0.098 0.605

-26 63.6 3.776E+19 2.28413+19 8.880E+18 77.8 61.5 85.8 69.5 Intermediate Shell Plate C-8009-2 C8161-1 SA-533B CI.1 0.085 0.600 0

54.5 3.776E+19 2.284E+19 8.880E+18 66.7 52.7 100.7 86.7 Intermediate Shell Plate C-8009-3 C8182-2 SA-533B CI.1 0.096 0.580 0

62.2 3.776E+19 2.28413+19 8.880E+18 76.1 60.1 110.1 94.1 Lower Shell Plate C-8010-1 C8161-2 SA-533B CI.1 0.085 0.585 12 54.5 3.791E+19 2.29313+19 8.915E+18 66.7 52.8

[112.7]

[98.8]

Lower Shell Plate C-8010-2 B2545-1 SA-533B CI.1 0.083 0.668

-28 53.1 3.791E+19 2.293 E+19 8.915E+18 65.0 51.4 71.0 57.4 Lower Shell Plate C-8010-3 B2545-2 SA-533B CI.1 0.080 0.653

-30 51.0 3.791E+19 2.29313+19 8.915E+18 62.4 49.4 66.4 53.4 Regulatory Guide 1.99, Revision 2, Position 2.1 Inter. Shell Long. Weld 2-203 A 10120 Linde 0091 0.046 0.082

-56 14.9 3.568E+19 2.159E+19 8.391E+18 18.0 14.2 0.5

-5.0 Inter. Shell Long. Weld 2-203 B 10120 Linde 0091 0.046 0.082

-56 14.9 2.754E+19 1.666E+19 6.477E+18 17.0 13.1

-1.0

-6.5 Inter. Shell Long. Weld 2-203 C 10120 Linde 0091 0.046 0.082

-56 14.9 2.7541E+19 1.66613+19 6.477E+18 17.0 13.1

-1.0

-6.5 Lower Shell Long. Weld 3-203 A 10120 Linde 0091 0.046 0.082

-56 14.9 3.582E+19 2.167E+19 8.425E+18 18.0 14.2 0.5

-5.0 Lower Shell Long. Weld 3-203 B 10120 Linde 0091 0.046 0.082

-56 14.9 2.765E+19 1.673E+19 6.502E+18 17.0 13.1

-1.0

-6.5 Lower Shell Long. Weld 3-203 C 10120 Linde 0091 0.046 0.082

-56 14.9 2.765E+19 1.673E+19 6.5021E+18 17.0 13.1

-1.0

-6.5 Intermediate Shell Plate C-8009-3 C8182-2 SA-533B CI.1 0.096 0.580 0

40.7 3.776E+19 2.284E+19 8.880E+18 49.8 39.4 66.8 56.4

[ ] - Controlling values of the adjusted reference temperatures.

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BAW-2405, Rev. 2 Figure 3-1. Location and Identification of Reactor Vessel Materials Girth Weld 8-203 Intermediate Shell Plates C8009-1, -2, -3 Girth Weld 9-203 Lower Shell Plates C8010-1, -2, -3 up_ Girth Weld 10-203 3-4 EA AR EVA

BAW-2405, Rev. 2 Figure 3-2. Location of Longitudinal Welds in the Reactor Vessel Upper and Lower Shell Courses 1800 900 Vessel Vertical Weld Seam Locations O

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BAW-2405, Rev. 2

4. PRESSURIZED THERMAL SHOCK REFERENCE TEMPERATURES A pressurized thermal shock (PTS) evaluation for the ANO-2 reactor vessel beltline materials was performed in accordance with Code of Federal Regulation, Title 10, Part 50.61 (10 CFR 50.61).171 The results of the PTS evaluation are shown in Table 4-1. These results demonstrate that the ANO-2 reactor vessel beltline materials will not exceed the PTS screening criteria before 32 EFPY. The controlling beltline material for the ANO-2 reactor vessel with respect to PTS is the lower shell plate C-8010-1, with a RTPTS value of 118.80F that is well below the PTS screening criterion of 2700F.

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BAW-2405, Rev. 2 Table 4-1. Pressurized Thermal Shock Reference Temperatures Applicable to 32 EFPY with Power Uprate Chemical Vessel/Clad Material Description Composition Interface Fluence at Reactor Vessel Matd.

Ieat Base Metal /

Cu Ni Initial Chemistry 32 EFPY; Fluence ARTp.s, Margin

RTprs, Screening Beltline Location Ident.

Number Flux Type wt%

wt%

RTNDT Factor n/cm2 ta)

Factor F

F F

Criteria RTpTs Calculation Per 10 CFR 50.61 Using Tables Inter. Shell Long. Weld 2-203 A 10120 Linde 0091 0.046 0.082

-56 34.0 3.432E+19 1.322 44.9 56.4 45.3 270 Inter. Shell Long. Weld 2-203 B 10120 Linde 0091 0.046 0.082

-56 34.0 2.646E+19 1.260 42.8 54.7 41.5 270 Inter. Shell Long. Weld 2-203 C 10120 Linde 0091 0.046 0.082

-56 34.0 2.646E+19 1.260 42.8 54.7 41.5 270 Lower Shell Long. Weld 3-203 A 10120 Linde 0091 0.046 0.082

-56 34.0 3.446E+19 1.323 45.0 56.4 45.4 270 Lower Shell Long. Weld 3-203 B 10120 Linde 0091 0.046 0.082

-56 34.0 2.657E+19 1.261 42.9 54.8 41.7 270 Lower Shell Long. Weld 3-203 C 10120 Linde 0091 0.046 0.082

-56 34.0 2.657E4+19 1.261 42.9 54.8 41.7 270 Inter./LowerShellGirth 9-203 83650 Linde0091 0.045 0.087

-10 34.1 3.613E+19 1.334 45.5 45.6 81.1 300 Weld Intermediate Shell Plate C-8009-1 C8161-3 SA-533B CI.l 0.098 0.605

-26 63.6 3.613E+19 1.334 84.8 34.0 92.8 270 Intermediate Shell Plate C-8009-2 C8161-1 SA-533B CI.l 0.085 0.600 0

54.5 3.613E+19 1.334 72.7 34.0 106.7 270 Intermediate Shell Plate C-8009-3 C8182-2 SA-533B CH.l 0.096 0.580 0

62.2 3.613E+19 1.334 83.0 34.0 117.0 270 Lower Shell Plate C-8010-1 C8161-2 SA-533B C.I 0.085 0.585 12 54.5 3.627E+19 1.335 72.8 34.0

[118.8]

270 Lower Shell Plate C-8010-2 B2545-1 SA-533B Cl.l 0.083 0.668

-28 53.1 3.627E+19 1.335 70.9 34.0 76.9 270 Lower Shell Plate C-8010-3 B2545-2 SA-533B CI.1 0.080 0.653

-30 51.0 3.627E+19 1.335 68.1 34.0 72.1 270 RTm Calculation Per 10 CFR 50.61 Using Surveillance Data Inter. Shell Long. Weld 2-203 A 10120 Linde 0091 0.046 0.082

-56 14.9 3.432E+19 1.322 19.7 39.3 3.0 270 Inter. Shell Long. Weld 2-203 B 10120 Linde 0091 0.046 0.082

-56 14.9 2.646E+19 1.260 18.8 38.9 1.7 270 Inter. Shell Long. Weld 2-203 C 10120 Linde 0091 0.046 0.082

-56 14.9 2.646E+19 1.260 18.8 38.9 1.7 270 Lower Shell Long. Weld 3-203 A 10120 Linde 0091 0.046 0.082

-56 14.9 3.446E+19 1.323 19.7 39.3 3.0 270 Lower Shell Long. Weld 3-203 B 10120 Linde 0091 0.046 0.082

-56 14.9 2.657E+19 1.261 18.8 38.9 1.7 270 Lower Shell Long. Weld 3-203 C 10120 Linde 0091 0.046 0.082

-56 14.9 2.657E+19 1.261 18.8 38.9 1.7 270 Intermediate Shell Plate C-8009-3 C8182-2 SA-533B CI.A 0.096 0.580 0

40.7 3.613E+19 1.334 54.3 17.0 71.3 270 (a) The inside surface fluence is the calculated value at the clad - base metal interface of the reactor vessel; attenuation through the cladding is based on deterministic methods.

l - Limiting reactor vessel beltline material in accordance with 10 CFR 50.61.

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BAW-2405, Rev. 2

5.

DESIGN BASIS FOR PRESSUREITEMPERATURE LIMITS Essential geometric data and analytical parameters used in the preparation of ANO-2 P/T limits are summarized in Table 5-1. Table 5-2 presents material properties utilized in analyzing the intermediate and lower beltline shells (SA533, Grade B, Class 1 plate material) and the reactor vessel inlet nozzles (SA508, Class 2 forging material). Limiting values of the adjusted reference temperature, RTNDT, are listed in Table 5-3.

For normal heatup operation, four ramped heatup transient conditions are considered in the evaluation. These transient conditions are simulated by increasing the reactor coolant system (RCS) cold leg temperature from 50 OF to 560 OF at constant rates of 50, 60, 70 and 80 OF/hr. The normal heatup transients are illustrated in Figure 5-1. The inservice leak and hydrostatic (ISLH) heatup test condition is also evaluated using the above RCS cold leg temperature ranges, at a ramp rate of 10 OF/hr.

For normal cooldown operation, the following temperature dependant rates for ramped and stepped cooldown transients are considered in the evaluation.

Actual RCS Cold Leg Temperature Maximum Cooldown Rate 50 0F < Tc *560 0F 100 OF/hr (ramp) or 50 0F in any half hour period (step)

A step change is also included in the ramped and step cooldown transients to simulate the temperature change that occurs at the initiation of shutdown cooling when the last reactor coolant pump is secured.

Cooldown Transient Set 1 For the case when acid reducing is not performed during shutdown cooling, a 50 OF step is modeled at 200 OF, followed by a 30 minute hold period. The ramped and stepped cooldown transients, for this normal shutdown condition, are depicted in Figures 5-2 and 5-3, respectively.

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BAW-2405, Rev. 2 Cooldown Transient Set 2 During an acid reducing shutdown, a 30 0F step is modeled at 120 0F, followed by a 30 minute hold period, as illustrated in Figures 5-4 and 5-5 for ramped and stepped cooldown transients, respectively.

The ISLH cooldown test condition is assessed for the RCS temperature decreasing from 560 OF to 50 OF at a ramp rate of 10 OF/hr. Steady state limits are also calculated at 5 OF intervals during heatup and cooldown, thereby providing results for "soak periods" where there is no change in the temperature of the reactor coolant (0 OF/hr rate).

Since overpressure events most likely occur during isothermal conditions in the RCS, the steady state Appendix G limit was used in developing the LTOP P/T limits. This is consistent with the Westinghouse standard methodology (WCAP-14040-NP-A) that endorses the use of steady-state Appendix G limit as the LTOP design limit. This methodology has been previously reviewed and approved by the USNRC staff. Per Code Case N-640 1"j, the maximum allowable pressure in the RV is limited to 100% of the Appendix G P-T limit, which in this case is the steady state Appendix G limit.

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BAW-2405. Rev. 2 Table 5-1 Design Data Vessel geometry:

Beltline inner radius

= 79.719 in.

Beltline outer radius

= 87.594 in.

Beltline wall thickness

= 7.875 in.

Nozzle belt inner radius

= 79.125 in.

Nozzle belt outer radius

= 89.625 in.

Nozzle belt wall thickness

= 10.500 in.

Cladding thickness

= 0.2188 in.

Core barrel outer radius

= 70.500 in.

Nozzle radius of outlet nozzle per WRCB 175 Nozzle radius of inlet nozzle per WRCB 175

= 24.675 in.

= 18.248 in.

Postulated flaws:

Closure head limits:

1/4T x 3/2T semi-elliptical longitudinal surface flaw in vessel beltline one-inch inside corner flaw in nozzle When core not critical, 622 psig (uncorrected) up to 150 0F (RTNDT + 120 0F)

Upper shelf fracture toughness:

200 ksidin for vessel plate material and nozzle forging material Safety factors on pressure:

2.0 for normal heatup and cooldown conditions 1.5 for ISLH heatup and cooldown conditions Adjustments for sensor location and instrument error:

Beltline & Inlet nozzle: Pressure = 47.0 psig (< 200 OF), 85.1 psig (2 200 OF)

Outlet nozzle:

Pressure = 32.7 psig Closure head:

Pressure = 29.1 psig Temperature

= 0 OF Convection film coefficients:

520 Btu/hr-ft2-PF at clad-base metal interface 0 Btu/hr-ft2?-F at the outside surface (perfectly insulated) 5-3 At AR EVA

BAW-2405, Rev. 2 Table 5-2 Material Properties(1)

Temp.

Elastic Thermal Thermal Specific Density Poisson's Modulus Expansion Conductivity Heat Ratio (0F)

(106 psi)

(104 inlinf F)

Btu/hr-ft-0F)

(Btu/lb-9F)

(Ib/ft3) 50 29.28 6.99 22.1 0.105 491.2 0.3 70 29.20 7.02 22.3 0.106 490.9 0.3 100 29.04 7.06 22.6 0.108 490.5 0.3 150 28.77 7.16 23.1 0.111 489.9 0.3 200 28.50 7.25 23.4 0.114 489.2 0.3 250 28.25 7.34 23.7 0.117 488.6 0.3 300 28.00 7.43 23.8 0.120 487.9 0.3 350 27.70 7.50 23.8 0.122 487.3 0.3 400 27.40 7.58 23.8 0.126 486.7 0.3 450 27.20 7.63 23.7 0.129 486.0 0.3 500 27.00 7.70 23.5 0.132 485.4 0.3 550 26.70 7.77 23.2 0.135 484.7 0.3 600 26.40 7.83 23.0 0.139 484.1 0.3 650 25.85 7.90 22.7 0.142 483.4 0.3 700 25.30 7.94 22.3 0.145 482.8 0.3 (1) Based on the 1995 Edition with Addenda through 1996 of the ASME Boiler and Pressure Vessel Code, Section 1II, Division 1, Appendices using the limiting beltline shell material, SA533, Grade B, Class 1.

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BAW-2405. Rev. 2 Table 5-3 Limiting RTNDT'S for Reactor Vessel Materials at 32 EFPY Adjusted RTNDT (0F)

Component at 1/4T at 3/4T Beltline Region, 112.7(')

98.8(')

Lower Shell Plate (C-801 0-1)

Inlet Nozzle 30.0(2)

Outlet Nozzle 0.0(2)

Closure Flange Region 30.0 30.0 (1) the P/T limit analysis conservatively used the 1/4T and 3/4T RTNDT values of 113 OF and 99 OF, respectively.

(2) applicable at the one-inch flaw depth 5-5 A'A AR EVA

BAW-2405, Rev. 2 Figure 5-1. ANO-2 Ramp Heatup Temperature Transients 600 500 400 L-a; I-Lm 300 200 100 0

0 100 200 300 400 500 600 Time, minutes 5-6 700 AR EVA

BAW-2405. Rev. 2 Figure 5-2. ANO-2 Ramp Cooldown Temperature Transient - Set I (normal shutdown) 600 500 400 U-Ca 1 300 C}

00 200 100 0

0 50 100 150 200 250 300 Time, minutes 350 5-7 AREVA

BAW-2405, Rev. 2 Figure 5-3. ANO-2 Step Cooldown Temperaure Transient - Set 1 (normal shutdown) 600 560 F 500 50 F Steps with 30 min. hold periods 400 U-2 300-E 200 F 200 -

50 F step due to last RCP trip 150 F 100_

50 F 0

0 50 100 150 200 250 300 350 Time, minutes 5-8 A

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BAW-2405, Rev. 2 Figure 5-4. ANO-2 Ramp Cooldown Temperature Transient - Set 2 (acid reduction shutdown) 600 500 400 L-0; a.

E I--

200 100 0

0 50 100 150 200 250 300 Time, minutes 350 5-9 AR EVA

BAW-2405, Rev. 2 Figure 5-5. ANO-2 Step Cooldown Temperature Transient - Set 2 (acid reduction shutdown) 600 500 400 U-

' 300 C)ca.

E isI--

200 100 0

0 50 100 150 200 250 300 Time, minutes 350 5-10 AR EVA

BAW-2405, Rev. 2

6. TECHNICAL BASIS FOR PRESSURE/TEMPERATURE LIMITS Pressure/temperature limits for the ANO-2 reactor vessel are calculated to satisfy the requirements of 10 CFR Part 50, Appendix G using analytical methods and acceptance criteria of the ASME Boiler and Pressure Vessel Code, Section Xl, Appendix G [4] and ASME Code Case N-640[111 for use of the K,C fracture toughness curve and ASME Code Case N-588[121 for influence coefficient solution for Kit and explicit method for calculating membrane correction factor (Mm).

The methods and criteria employed to establish operating pressure and temperature limits are described below. The objective of these limits is to prevent nonductile failure during normal operating conditions, including anticipated operational occurrences and system hydrostatic pressure and leak tests.

Of all the components of the RCPB that are subject to the requirements of 10 CFR 50, Appendix G, the only regions that regulate the pressure/temperature limits are the closure head flange, inlet and outlet nozzle, and beltline regions of the reactor vessel. The closure head region can be significantly stressed at relatively low temperatures due to mechanical loads resulting from bolt preload and pressure. High stresses, of the order of two to three times the shell membrane stress, can also occur at the inside corners of the reactor vessel nozzles due to local stress concentrations.

Typically, the closure head and nozzle regions influence the pressure-temperature limits only during the first several service periods, prior to significant neutron embrittlement of the reactor vessel beltline materials.

After several years of exposure to neutron irradiation, the increase in the RTNDT of the beltline region materials is such that the RCPB pressure/temperature limits are usually controlled by the beltline region of the reactor vessel. The pressure/temperature limits contained in this report are established by determining the minimum allowable pressure, as a function of fluid temperature, considering the closure head, the inlet and outlet nozzles, and the beltline regions of the reactor vessel.

The analytical procedures used to calculate P/T limits are based on linear elastic fracture mechanics methods for calculating stress intensity factors at the maximum depths of postulated semi-elliptical surface flaws.

6-1 Alk AR EVA

BAW-2405, Rev. 2 The basic equation for allowable pressure is:

alloW = SFK

-IT

where, Pallow = allowable pressure KIR

= reference stress intensity factor (K1a or Kic)

KIT

= thermal stress intensity factor Klp

= unit pressure stress intensity factor (due to 1 psig)

SF

= safety factor For each analyzed transient and steady state condition, the allowable pressure is determined as a function of reactor coolant temperature considering postulated flaws in the reactor vessel beltline, inlet nozzle, outlet nozzle, and closure head. In the beltline region, flaws are presumed to be present at the 1/4t and 3/4t locations of the controlling material (shell plate or weld), as defined by the fluence adjusted RTNDT. The nozzle flaw is located at the inside juncture (corner) of the nozzle.

The closure head flange limit is not explicitly calculated. However, for the condition when the core is not critical, the uncorrected closure flange allowable pressure of 622 psig (20% of preservice hydrostatic test pressure of 3110 psig) is maintained as the limit for temperatures up to 150 OF (30 OF RTNDT + 120 OF margin) per Table 1, item 2.b of 10 CFR50, Appendix G (21. Above, 150 OF, the closure flange allowable pressure is 2500 psig. P-T limits for the beltline and nozzle regions are calculated using a factor of safety of 2 for normal operation and 1.5 for ISLH operation. These location specific P-T limits are calculated using the FRA-ANP proprietary computer code PTPC181.

The maximum allowable pressure at a particular fluid temperature is taken as the minimum value of allowable pressure calculated for each flaw location and operating condition, including steady state. A P/T limit curve is then constructed as the collection of points that define the maximum allowable pressures as a function of fluid temperature for a particular mode of reactor operation.

The P-T curves provided in this report are adjusted for sensor location but does not include instrument error. They are, "refined" as necessary to eliminate regions of negative slope by lowering the allowable pressure for temperatures less than that corresponding to the minimum pressure.

The criticality limit temperature is obtained by satisfying the requirement of Item 2.d in Table 1 of 10 CFR 50, Appendix G I2J. It requires the minimum temperature to be the larger of minimum permissible temperature for inservice system hydrostatic pressure test (taken as the leak test 6-2 A

AR EVA

BAW-2405, Rev. 2 temperature corresponding to the ISLH limit pressure of 2500 psig with heatup and cooldown rates up to 10 0F/hr) or the RTNDT of the closure flange material + 160 OF.

Various aspects of the calculational procedures utilized in the development of P/T limits are discussed below.

6.1 Fracture Toughness The fracture toughness of reactor vessel steels is expressed as a function of crack-tip temperature, T, indexed to the adjusted reference temperature of the material, RTNDT.

Pressure/temperature limits developed in accordance to ASME Code, Section Xi, Appendix G utilize the expression for crack arrest fracture toughness, Kla = 26.8 + 1.233 exp [0.0145 ( T - RTNDT + 160 F)]

Exemptions to 10 CFR 50, Appendix G, that cite ASME Code Case N-640 (utilized in the generation of the P-T limits contained in this report), utilize the crack initiation fracture toughness, KIc = 33.2 + 2.806 exp [ 0.02 ( T - RTNDT + 100 F)]

The upper shelf fracture toughness is limited to an upper bound value of 200 ksi 'in. The crack-tip temperature needed for these fracture toughness equations is obtained from the results of a transient thermal analysis, described below.

6.2 Thermal Analysis and Thermal Stress Intensity Factor Through-wall temperature distributions are determined by solving the one-dimensional transient axisymmetric heat conduction equation, CP a8T = k( a 2 T + 1 aT a

r2 r ar subject to the following boundary conditions:

at the inside surface, where r = Ri,

- kaT = h(Tw - Tb) ar 6-3 A

AR EVA

BAW-2405, Rev. 2 at the outside surface, where r = R0, aT =0 ar

where, p = density Cp = specific heat k = thermal conductivity T = temperature r = radial coordinate t = time h = convection heat transfer coefficient T, = wall temperature Tb = bulk coolant temperature Ri = inside radius of vessel R0 = outside radius of vessel The above equation is solved numerically using a finite difference technique to determine the temperature at 17 points through the wall as a function of time for prescribed changes in the bulk fluid temperature, such as multi-rate ramp and step changes for heatup and cooldown transients.

An equivalent linear thermal bending stresses (based on AT through the wall) is derived from the through-wall temperature distribution at each solution time point.

Through-wall thermal stress distributions are determined by trapezoidal integration of the following expression:

Thermal hoop stresses:

a(r)

=

r2

+R2

2 Trdr+

Trdr-Tr2J

[9, Eqn (255)]

Expressing the thermal stress distributions by a(x) = C0 + C1 (x/a) + C2 (x/a)2 + C3 (x/a)3, 6-4 AR EVA

BAW-2405, Rev. 2

where, x = is a dummy variable that represents the radial distance from the appropriate (i.e., inside or outside) surface, in.

a = the flaw depth, in.,

the thermal stress intensity factors are defined by the following relationships:

For a 1/4-thickness inside surface flaw during cooldown, Kit = (1.0359 C0 + 0.6322 C1 + 0.4753 C2 + 0.3855 C 3) 4T For a 1/4-thickness outside surface flaw during heatup, Kit = (1.043 C0 + 0.630 C, + 0.481 C2 + 0.401 C3) 4-H 6.3 Unit Pressure Stress Intensity Factor for Reactor Vessel Beltline Region The membrane stress intensity factor in the reactor vessel shell due to a unit pressure load is Kjm = Mm x Ri/t where R1 = vessel inner radius, in.

t = vessel wall thickness, in.

For a longitudinal 1/4-thickness x 3/2-thickness semi-elliptical surface flaw:

at the inside surface, Mm= 1.85 for qt<<2

= 0.926 't for 2 * 't

3.464

= 3.21 for 't > 3.464 at the outside surface, Mm= 1.77 for 4t<2

= 0.893 'It for 2 * 't

3.464

= 3.09 for At > 3.464 6.4 Unit Pressure Stress Intensity Factor for Reactor Vessel Nozzles Considering a nozzle as a hole in a shell, WRC Bulletin 175[101 presents the following method for estimating stress intensity factors for a nozzle corner flaw:

6-5 A

AR EVA

BAW-2405, Rev. 2 Kim = crfib F(a/rr) where a= Ri/t Ri = nozzle belt shell inner radius, in.

t = nozzle belt shell wall thickness, in.

a = flaw depth, in.

r,= apparent radius of nozzle, in.

= r + 0.29r, r = inner radius of nozzle, in.

rc= nozzle corner radius, in.

and F(aIrn) = 2.5 - 6.108(a/rn) + 12(a/rn)2 - 9.1664(a/r,) 3 6-6 At AR EVA

BAW-2405, Rev. 2

7. OPERATIONAL PRESSURE/TEMPERATURE LIMITS Results of the thermal and fracture mechanics analyses '31 performed for the ANO-2 reactor vessel are presented in the form of P/T curves for (three) operating conditions; normal heatup, normal cooldown, and ISLH operations.

These P/T curves are location adjusted to account for the differences between the controlling pressure location and the point of system pressure measurement in the pressurizer. They do not account for instrument error.

Pressure-temperature limits (in units of psig) for normal heatup (including criticality core limits) at 32 EFPY are presented in Table 7-1 and illustrated in Figure 7-1.

The criticality limit temperature is 190 "F. It is based on the RTNDT of the closure flange material (30 OF) plus 160 OF which is larger than the 175 "F value that corresponds to the ISLH limit pressure of 2500 psig per Table 7-3.

Considering all the ramp and step cooldown transient scenarios shown in Figures 5-2 through 5-5, composite cooldown P/T limits are determined as reported in Table 7-2 and depicted by Figure 7-2. The ISLH P/T limits are given in Table 7-3 and illustrated in Figure 7-3. The steady state P/T limits are provided in Table 7-4. Protection against nonductile failure is ensured by using these curves to limit the reactor coolant pressure. Acceptable pressure and temperature combinations for reactor operation are below and to the right of the pressure-temperature limit curves.

Additionally, to better facilitate use by plant operations, the P/T limits are provided in units of psia. The P/T limits for normal heatup, normal cooldown, ISLH and steady state conditions are provided in Tables 7-5 though 7-8, respectively. The P/T limits for normal heatup, normal cooldown and ISLH conditions are also depicted in Figures 7-4 through 7-6, respectively.

7-1 At ARE VA

BAW-2405, Rev. 2 Table 7-1. ANO-2 Summary of Ramp Heatup P/T Limits & Critical Core Limits (psig)

Ramp Heatup Fluid Critical Temp.

50 F/hr 60 F/hr 70 F/hr 80 F/hr Core I

(F)

I (psig) I (psig) I (psig) I (psig)

I (psig) 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 150 155 160 165 170 175 180 185 190 190 195 200 205 210 215 220 225 230 235 240 245 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 1148 1215 1289 1371 1462 1562 1673 1795 1931 1931 2081 2204 2378 2569 2781 3015 3263 3263 3263 3263 3263 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 1066 1125 1191 1263 1344 1433 1532 1641 1762 1762 1895 2005 2168 2348 2546 2766 3008 3263 3263 3263 3263 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 995 1047 1105 1170 1241 1320 1408 1505 1613 1613 1732 1825 1970 2131 2308 2504 2720 2959 3223 3263 3263 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 934 980 1031 1088 1152 1222 1300 1386 1482 1482 1588 1667 1797 1940 2098 2273 2466 2679 2915 3175 3263 0

934 980 1031 1088 1152 1222 1300 1386 1482 1588 1667 1797 7-2 At AR EVA

BAW-2405, Rev. 2 Table 7-1. ANO-2 Summary of Ramp Heatup PIT Limits & Critical Core Limits (Cont'd)

Ramp Heatup Fluid Critical Temp.

50 F/hr 60 F/hr 70 F/hr 80 F/hr Core iSig) l(psig)

I(psig)

(psig)

(psig) 250 255 260 265 270 275 280 285 290 295 300 305 310 315 320 325 330 335 340 345 350 355 360 365 370 375 380 385 390 395 400 405 410 415 420 425 430 435 440 445 450 455 460 465 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 1940 2098 2273 2466 2679 2915 3175 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 7-3

,A AR EVA

BAW-2405, Rev. 2 Table 7-1. ANO-2 Summary of Ramp Heatup P/T Limits &Critical Core Limits (Cont'd)

Ramp Heatup Fluid Critical Temp.

50 F/hr 60 F/hr 70 F/hr 80 F/hr Core (F)

(psig)

(psig) 470 3263 3263 3263 3263 3263 475 3263 3263 3263 3263 3263 480 3263 3263 3263 3263 3263 485 3263 3263 3263 3263 3263 490 3263 3263 3263 3263 3263 495 3263 3263 3263 3263 3263 500 3263 3263 3263 3263 3263 505 3263 3263 3263 3263 3263 510 3263 3263 3263 3263 3263 515 3263 3263 3263 3263 3263 520 3263 3263 3263 3263 3263 525 3263 3263 3263 3263 3263 530 3263 3263 3263 3263 3263 535 3263 3263 3263 3263 3263 540 3263 3263 3263 3263 3263 545 3263 3263 3263 3263 3263 550 3263 3263 3263 3263 3263 555 3263 3263 3263 3263 3263 560 3263 3263 3263 3263 3263 7-4 DA AR EVA

BAW-2405. Rev. 2 Table 7-2. ANO-2 Cooldown Composite P/T Limits (psig)

Composite Composite Composite Cooldown Coo ldown Cooldown Fluid Limit Fluid Limit Fluid Limit Temp.

Temp.

(F)

(psig)

(F)

(psig)

(F)

(psig) 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 502 524 547 573 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 1410 1497 1593 1699 1816 1946 2089 2247 2422 2616 2700 2700 2700 2700 2699 2700 2699 2699 2699 2699 2699 2699 2699 265 270 275 280 285 290 295 300 305 310 315 320 325 330 335 340 345 350 355 360 365 370 375 380 385 390 395 400 405 410 415 420 425 430 435 440 445 450 455 460 465 470 475 480 2699 2700 2700 2701 2702 2703 2704 2706 2707 2709 2710 2712 2716 2718 2720 2723 2726 2729 2732 2736 2739 2743 2747 2750 2755 2761 2767 2773 2779 2786 2794 2801 2810 2818 2828 2837 2847 2858 2870 2882 2894 2908 2922 2937 485 490 495 500 505 510 515 520 525 530 535 540 545 550 555 560 2952 2969 2986 3004 3023 3043 3064 3086 3108 3132 3156 3181 3206 3231 3252 3263 7-5

'AE AR EVA

BAW-2405. Rev. 2 Table 7-3. ANO-2 ISLH P/T Limits (psig)

Allowable Pressures l

Fluid Limiting Outlet Inlet Closure Minimum Temp.

Beltline Nozzle Nozzle Head (F) (psi (psig)

(psig)

I 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 150 155 160 165 170 175 180 185 190 195 200 205 21C 215 220 225 23C 235 240 245 25C 255 943 953 967 984 1004 1026 1050 1077 1107 114C 1176 1216 1261 1310 1364 1424 1490 1563 1643 1733 1831 1831 1940 2061 2194 2341 2503 2683 2881 3100 3342 3572 386E 4196 4557 4957 4984 4984 4984 4984 4984 4984 4984 1935 202E 2152 2295 2457 2636 284C 3063 3311 3586 388E 4224 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 1408 1461 1531 1612 1704 1807 1921 2048 2188 2344 2515 2705 2916 3148 3405 3688 4002 4348 4491 4491 4491 4491 4491 4491 4491 4491 4491 4491 4491 4491 4491 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 2500 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 1831 194C 2061 2194 2341 250 2683 2881 310O 3342 3572 386E 419E 4362 4362 4362 4362 4362 4362 4362 4362 4362 7-6 DA AR EVA

BAW-2405, Rev. 2 Table 7-3. ANO-2 ISLH P/T Limits (Cont'd.)

Allowable Pressures l

Fluid Limiting Outlet I Inlet Closure Minimum Temp.

Beltline Nozzle Nozzle Head (F)

, (ps)

(psig)

(psig) 26C 265 27C 275 28C 285 29C 295 30C 305 31C 315 32C 325 33C 335 34C 345 35C 355 36C 365 37C 37!

38C 385 39C 395 40C 405 41C 415 42C 425 43C 435 44C 44E 45C 455 46C 465 47C 475 48C 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4984 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 436 436 4362 4362 4362 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4453 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 4362 7-7 At AR EVA

BAW-2405, Rev. 2 Table 7-3. ANO-2 ISLH P/T Limits (Cont'd.)

Allowable Pressures Fluid Limiting Outlet Inlet Closure Minimum Temp.

Beltline Nozzle Nozzle Head (F)

(psig)

.(psig)

(psig)

(psig) 485 4984 436 4453 4362 490 4984 436 4453 4362 495 4984 436 4453 4362 500 4984 436 4453 4362 505 4984 436 4453 4362 510 4984 436 4453 4362 515 4984 436 4453 4362 520 4984 436 4453 4362 525 4984 436 4453 4362 530 4984 436 4453 4362 535 4984 436 4453 4362 540 4984 436 4453 4362 545 4984 436 4453 4362 550 4984 436 4453 4362 555 4984 4362 4453 4362 560 4984 4362 4453 4362 7-8 AD AR EVA

BAW-2405, Rev. 2 Table 7-4. ANO-2 Steady-State "Isothermal Condition" P/T Limits (psig)

Allowable Pressures l

Limiting Outlet Inlet Closure Minimum Fluid Beltline Nozzle Nozzle Head (F) l(psig)

(p (psig)

(psig)I (psig)(

50 55 60 65 70 75 81 85 90 95 100 105 110 115 120 125 130 135 140 145 150 150 155 160 165 170 175 180 185 190 195 200 205 21C 215 22C 225 230 235 240 245r 25C 255 696 708 721 735 751 768 788 809 833 859 888 920 955 994 1037 1085 1138 1196 1260 1331 1410 1410 1497 1593 1699 181E 194E 2089 2247 2422 2616 2791 3027 328E 3577 3716 3716 3716 3716 3716 3716 3716 3716 1444 1541 1649 1768 1900 2045 2206 2384 2581 2799 3039 3263 3262 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3268 3263 3263 3263 3263 3263 3263 3263 3263 3263 1045 110 1161 122 130 1386 1477 1578 1689 1813 1949 2099 2266 2449 2652 2877 3125 3356 3356 3356 3356 3356 3356 3356 3356 3356 3356 3356 3356 3356 3356 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 2500 593 59 59 59 59 59 59 59 59 59 59 59 59 59 59 59 593 593 593 593 593 141 149 159 169 181 194 208 224 242 261 2791 3027 326 326 3263 3263 3263 3263 3263 3263 3263 3263 7-9 At AR EVA

BAW-2405, Rev. 2 Table 7-4. ANO-2 Steady-State "Isothermal Condition" P/T Limits (Cont'd)

Allowable Pressures Limiting Outlet Inlet Closure Minimum Fluid Beltline Nozzle Nozzle Head Temp.

(F)

, ip (psig)

(psig)

(psig)p L

pig) 260 265 27C 275 280 285 290 295 30C 305 31C 315 32C 325 330 335 340 345 350 355 360 365 370 375 380 385 390 395 40C 405 41C 415 420 425 430 435 440 44E 450 455 460 465 470 475 48C 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 3716 371E 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 326 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3262 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 3318 326 326 326 326 326 326 326 3263 3263 3263l 3263 3263 3263 3263 3263l 326 326 326 3263 3263 3263 3263 326 326 326 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 7-10 EA AR EVA

BAW-2405, Rev. 2 Table 7-4. ANO-2 Steady-State 'Isothermal Condition" P/T Limits (Cont'd)

Allowable Pressures l

Limiting Outlet Inlet Closure Minimum Fluid Beltline Nozzle Nozzle Head Temp.

(F)

(psig)

(psig) 485 3716 3263 3318 3263 490 3716 3263 3318 326 495 3716 3263 3318 3263 50C 3716 3263 3318 3263 505 3716 3263 3318 3263 51C 3716 3263 3318 3263 51' 3716 3263 3318 3263 520 3716 3263 3318 3263 525 3716 3263 3318 3263 530 3716 3263 3318 3263 535 3716 3263 3318 3263 540 3716 3263 3318 3263 545 3716 3263 3318 3263 55C 3716 3263 3318 3263 55C 3716 3263 3318 3263 560 371 £ 3263 3318 3263 7-11 DA AR EVA

BAW-2405. Rev. 2 Table 7-5. ANO-2 Summary of Ramp Heatup PIT Limits & Critical Core Limits (psia)

Ramp Heatup l Critical 50 F/hr60 F/hr70 F/hr 80 F/hr Core Fluid Temp.

l (F)

(psia)

(psia) 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 150 155 160 165 170 175 180 185 190 190 195 200 205 210 215 220 225 230 235 240 245 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 1163 1230 1304 1386 1477 1577 1688 1810 1946 1946 2096 2219 2393 2584 2796 3030 3278 3278 3278 3278 3278 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 1081 1140 1206 1278 1359 1448 1547 1656 1777 1777 1910 2020 2183 2363 2561 2781 3023 3278 3278 3278 3278 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 1010 1062 1120 1185 1256 1335 1423 1520 1628 1628 1747 1840 1985 2146 2323 2519 2735 2974 3238 3278 3278 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 949 995 1046 1103 1167 1237 1315 1401 1497 1497 1603 1682 1812 1955 2113 2288 2481 2694 2930 3190 3278 15 949 995 1046 1103 1167 1237 1315 1401 1497 1603 1682 1812 7-12 AN AREVVA

BAW-2405, Rev. 2 Table 7-5. ANO-2 Summary of Ramp Heatup P/T Limits & Critical Core Limits (Cont'd.)

Ramp Heatup Critical 50 F/hr 60 F/hr 70 F/hr 80 F/hr Core Fluid I

_ I__

T em p.

(jF) I(psia),(psia)

_(psia) I(psia) ]E(psia) 250 255 260 265 270 275 280 285 290 295 300 305 310 315 320 325 330 335 340 345 350 355 360 365 370 375 380 385 390 395 400 405 410 415 420 425 430 435 440 445 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 1955 2113 2288 2481 2694 2930 3190 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 7-13 At AR EVA

BAW-2405, Rev. 2 Table 7-5. ANO-2 Summary of Ramp Heatup P/T Limits & Critical Core Limits (Cont'd.)

Ramp Heatup Critical 50 F/hr60 F/hr70 F/hr80 F/hr Core Fluid Temp.__

(F)

(psia)

(psla)

(psia)

(Psia) 450 3278 3278 3278 3278 3278 455 3278 3278 3278 3278 3278 460 3278 3278 3278 3278 3278 465 3278 3278 3278 3278 3278 470 3278 3278 3278 3278 3278 475 3278 3278 3278 3278 3278 480 3278 3278 3278 3278 3278 485 3278 3278 3278 3278 3278 490 3278 3278 3278 3278 3278 495 3278 3278 3278 3278 3278 500 3278 3278 3278 3278 3278 505 3278 3278 3278 3278 3278 510 3278 3278 3278 3278 3278 515 3278 3278 3278 3278 3278 520 3278 3278 3278 3278 3278 525 3278 3278 3278 3278 3278 530 3278 3278 3278 3278 3278 535 3278 3278 3278 3278 3278 540 3278 3278 3278 3278 3278 545 3278 3278 3278 3278 3278 550 3278 3278 3278 3278 3278 555 3278 3278 3278 3278 3278 560 3278 3278 3278 3278 3278 7-14 EA

.AR EVA

BAW-2405. Rev. 2 Table 7-6. ANO-2 Cooldown Composite P/T Limits (psia)

Composite Composite Composite Cooldown Cooldown Cooldown Fluid Limit Fluid Limit Fluid Limit Temp.

Temp.

(F)

(psia)

(F)

(psia)

(F)

(psia) 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 517 539 562 588 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 1425 1512 1608 1714 1831 1961 2104 2262 2437 2631 2715 2715 2715 2715 2714 2715 2714 2714 2714 2714 2714 2714 2714 265 270 275 280 285 290 295 300 305 310 315 320 325 330 335 340 345 350 355 360 365 370 375 380 385 390 395 400 405 410 415 420 425 430 435 440 445 450 455 460 465 470 475 480 2714 2715 2715 2716 2717 2718 2719 2721 2722 2724 2725 2727 2731 2733 2735 2738 2741 2744 2747 2751 2754 2758 2762 2765 2770 2776 2782 2788 2794 2801 2809 2816 2825 2833 2843 2852 2862 2873 2885 2897 2909 2923 2937 2952 485 490 495 500 505 510 515 520 525 530 535 540 545 550 555 560 2967 2984 3001 3019 3038 3058 3079 3101 3123 3147 3171 3196 3221 3246 3267 3278 7-15 Ae AR EVA

BAW-2405, Rev. 2 Table 7-7. ANO-2 ISLH P/T Limits (psia)

Allowable Pressures Limiting Outlet Inlet Closure Minimum Fluid Beltline Nozzle Nozzle Head Temp.

(F)

,(psia)

(psia)

(psia),(psia) 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 958 968 982 999 1019 1041 1065 1092 1122 1155 1191 1231 1276 1325 1379 1439 1505 1578 1658 1748 1846 1846 1955 2076 2209 2356 2518 2698 2896 3115 3357 3587 3883 4211 4572 4972 4999 4999 4999 4999 4999 1950 2044 2167 2310 2472 2653 2855 3078 3326 3601 3904 4239 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 1423 1476 1546 1627 1719 1822 1936 2063 2203 2359 2530 2720 2931 3163 3420 3703 4017 4363 4506 4506 4506 4506 4506 4506 4506 4506 4506 4506 4506 4506 4506 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 2515 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 1846 1955 2076 2209 2356 2518 2698 2896 3115 3357 3587 3883 4211 4377 4377 4377 4377 4377 4377 4377 7-16 JA AR EVA

BAW-2405, Rev. 2 Table 7-7. ANO-2 ISLH P/T Limits (Cont'd.)

Allowable Pressures Limiting Outlet Inlet Closure Minimum Fluid Beitline Nozzle Nozzle Head (F)

(psia)

(psia) 250 255 260 265 270 275 280 285 290 295 300 305 310 315 320 325 330 335 340 345 350 355 360 365 370 375 380 385 390 395 400 405 410 415 420 425 430 435 440 445 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4999 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4468 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 4377 7-17 At AR EVA

BAW-2405, Rev. 2 Table 7-7. ANO-2 ISLH P/T Limits (Cont'd.)

Allowable Pressures Limiting Outlet Inlet Closure Minimum Fluid Beltline Nozzle Nozzle Head Temp.

(F)

(psia)

(psia) 450 4999 4377 4468 4377 455 4999 4377 4468 4377 460 4999 4377 4468 4377 465 4999 4377 4468 4377 470 4999 4377 4468 4377 475 4999 4377 4468 4377 480 4999 4377 4468 4377 485 4999 4377 4468 4377 490 4999 4377 4468 4377 495 4999 4377 4468 4377 500 4999 4377 4468 4377 505 4999 4377 4468 4377 510 4999 4377 4468 4377 515 4999 4377 4468 4377 520 4999 4377 4468 4377 525 4999 4377 4468 4377 530 4999 4377 4468 4377 535 4999 4377 4468 4377 540 4999 4377 4468 4377 545 4999 4377 4468 4377 550 4999 4377 4468 4377 555 4999 4377 4468 4377 560 4999 4377 4468 4377 7-18 A

AR EVA

BAW-2405, Rev. 2 Table 7-8. ANO-2 Steady-State "Isothermal Condition" P/T Limits (psia)

Allowable Pressures Limiting Outlet Inlet Closure Minimum Fluid Beltline Nozzle Nozzle Head Temp.

(F)AE (psia)

(psia)

(psia) 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 711 723 736 750 766 783 803 824 848 874 903 935 970 1009 1052 1100 1153 1211 1275 1346 1425 1425 1512 1608 1714 1831 1961 2104 2262 2437 2631 2806 3042 3303 3592 3731 3731 3731 3731 3731 3731 1459 1556 1664 1783 1915 2060 2221 2399 2596 2814 3054 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 1060 1115 1176 1243 1318 1401 1492 1593 1704 1828 1964 2114 2281 2464 2667 2892 3140 3371 3371 3371 3371 3371 3371 3371 3371 3371 3371 3371 3371 3371 3371 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 2515 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 608 1425 1512 1608 1714 1831 1961 2104 2262 2437 2631 2806 3042 3278 3278 3278 3278 3278 3278 3278 3278 7-19 At AR EVA

BAW-2405, Rev. 2 Table 7-8. ANO-2 Steady-State "Isothermal Condition" P/T Limits (Cont'd)

Allowable Pressures Limiting Outlet Inlet Closure Minimum Fluid Beltline Nozzle Nozzle Head Temp.

(F)

I (psia)

(psia)

(psia) 250 255 260 265 270 275 280 285 290 295 300 305 310 315 320 325 330 335 340 345 350 355 360 365 370 375 380 385 390 395 400 405 410 415 420 425 430 435 440 445 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3731 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3333 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 3278 7-20 At AR EVA

BAW-2405, Rev. 2 Table 7-8. ANO-2 Steady-State "Isothermal Condition" P/T Limits (Cont'd)

Allowable Pressures Limiting Outlet Inlet Closure Minimum Fluid Beltline Nozzle Nozzle Head Temp.

(F)

(psia)

(psia) 450 3731 3278 3333 3278 455 3731 3278 3333 3278 460 3731 3278 3333 3278 465 3731 3278 3333 3278 470 3731 3278 3333 3278 475 3731 3278 3333 3278 480 3731 3278 3333 3278 485 3731 3278 3333 3278 490 3731 3278 3333 3278 495 3731 3278 3333 3278 500 3731 3278 3333 3278 505 3731 3278 3333 3278 510 3731 3278 3333 3278 515 3731 3278 3333 3278 520 3731 3278 3333 3278 525 3731 3278 3333 3278 530 3731 3278 3333 3278 535 3731 3278 3333 3278 540 3731 3278 3333 3278 545 3731 3278 3333 3278 550 3731 3278 3333 3278 555 3731 3278 3333 3278 560 3731 3278 3333 3278 7-21 JA AR EVA

BAW-2405, Rev. 2 Figure 7-1. ANO-2 Ramp Heatup & Critical Core PIT Limit Curves (psig) 2500 2250 2000 1750 m 1500 a.

0.

2 1250 al I-X 1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 Temperature, F 500 7-22 AR EVA

BAW-2405, Rev. 2 Figure 7-2. ANO-2 Composite Cooldown PIT Limit Curve at 32 EFPY (psig) 2500 2250 2000 1750 0 1500

'V; U)

,- 1250 0w XL 1000 750 500 250 70 0

50 100 150 200 250 300 350 400 450 Temperature, F 7-23 A

500 At R EVA

BAW-2405, Rev. 2 Figure 7-3. ANO-2 ISLH PIT Limit Curve (psig) co 0(a cl.

0 a.

2500 2250 2000 1750 1500 1250 1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 Temperature, F 7-24 A

500 AR EVA

BAW-2405, Rev. 2 Figure 7-4 The following page contains Figure 7-4, which depicts ANO-2 Ramp Heatup and Critical Core P/T Limit Curves for 32 EFPY (psia).

7-25 AR EVA

BAW-2405, Rev. 2 2500 2000

.U 1500 a)

(n 05 P

L.-

N L..

2 1000 a'

500 0

l70 F/hr tLt iF/r__1kl NonCrtical Core 0 F/

on-Cri ical Core

-Fh Criica Core X-AAcceptable TIITI!T~llillIITITITITRegIion~lII 0

100 200 300 400 500 Reactor Coolant Temperature Tc, 0F (Curves do not Include margins for Instrument uncertainties) 7-26 Ak AR EVA

BAW-2405, Rev. 2 Figure 7-5 The following page contains Figure 7-5, which depicts ANO-2 Composite Cooldown PIT Limit Curve for 32 EFPY (psia).

7-27 DA ARE VA

BAW-2405, Rev. 2 2500 I

I I I I I I I I I I I I 1 1{

1 1 1 I 1 I 1 I I I I I I 1 I I I I 1 I.

I 1 1 I I 1 I I

I I I I I

.I i

I I 1 1 1 I

1 1

.. I I I I I I I I 1

I I

1 2000 In 0.(

f 1500 nIn 0

IL.

I-a)

N I.

u) 1000 0

(L

]Temperature,E------------------

I

.5 Flllii II I

I I..

I..

I I.I.

I.I.II.I.I I

I I I I I I I

I I

I. I I I I..

.I II

. II I I I I I

I.I l

I 500

.IIII I

I I I

I _ _ I _ _ I _ _ _ I I I I I I I I.

I I I I I I I.

I 0

I X

~~MiiuBoltup t\\i\\ttt 0

100 200 300 400 50 Reactor Coolant Temperature T,, °F (Curves do not Include margins for Instrument uncertainties) 7-28 i0 ARE A R E VA

BAW-2405, Rev. 2 Figure 7-6 The following page contains Figure 7-6, which depicts ANO-2 Inservice Leak and Hydrostatic (ISLH) P/T Limit Curve for 32 EFPY (psia).

7-29 A

ARE VA

BAW-2405, Rev. 2 25001 111111 11111111 11111 ll H1tgysdero

~tatic I___--I 2000 t]1 j ~~Acceptable 1500 n

Lowest Service U)

Temperature, 150 N

in1000 111111 11 ll ll liilit Tit I

Miiu Blu 00 500 4

0\\iit-1

___1 0

100 200 300 400 500 Reactor Coolant Temperature Tc, OF (Curves do not Include margins for Instrument uncertainties) 7-30 A

AR EVA

BAW-2405, Rev. 2

8. LTOP PRESSURE/TEMPERATURE LIMITS The pressure/temperature results developed for K1, measure of fracture toughness is used to develop LTOP P/T limits [¶3]

The ASME Code, Section Xl, Appendix G141 states that LTOP systems shall be effective at coolant temperatures less than 200 OF or at coolant temperatures corresponding to a reactor vessel metal temperature less than RTNDT + 50 'F, whichever is greater. Since the RTNDT of the controlling beltline material is 113 OF, the required metal temperature at the 1/4T depth from the inside surface of the beltline region is RTNDT + 50 OF or 163 'F. During normal plant heatup the metal temperature is lower and lags the coolant temperature. The maximum temperature difference occurs during the maximum plant heatup rate at 80 OF/hr when the corresponding coolant temperature is 186.4 'F. The minimum LTOP enable temperature for ANO-2 is therefore the greater of 186.4 'F, plus any adjustment for instrument error, or 200 OF.

LTOP systems must also limit the maximum pressure in the vessel to 100% of the pressure associated with the P/T limits reported in Section 7 when KC is used for fracture toughness".

LTOP P/T limits are presented in Table 8-1 and illustrated in Figure 8-1.

8-1 AR EVA

BAW-2405, Rev. 2 Table 8-1. ANO-2 PIT Limits for LTOP (psig)

Minimum Minimum Minimum Fluid Fluid Fluid Temp.

Temp.

)

(psig)

L(F)

Wpig)

(F)

WiPg) 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 125 13C 135 14C 145 15I 15C 155 16C 165 17C 175 18C 185 19C 195 20C 205 21C 215 22C 225 23C 235 24C 245 25C 25' 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 593 1410 1497 1593 1699 1816 1946 2089 2247 2422 2616 2791 3027 3263 3263 3263 3263 3263 3263 3263 3263 3263 326' 26 26 27 27 28 28 29 29 30 30 31 31 32 32 33 33 34 345 35 355 360 365 370 375 380 385 390 395 400 405 410 415 420 425 430 435 440 445 450 455 460 465 470 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 475 480 485 490 495 500 505 510 515 520 525 530 535 540 545 550 555 560 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 3263 8-2 At AR EVA

BAW-2405, Rev. 2 Figure 8-1. ANO-2 PIT Limits for LTOP (psig) 2500 2250 2000 1750 1500

,a.

, 1250 at 1000 750 500 250 0

0 50 100 150 200 250 300 350 400 Temperature, F 8-3 450 A

AR EVA 500

BAW-2405. Rev. 2

9.

CERTIFICATION Pressure/temperature limits for the ANO-2 reactor vessel have been calculated to satisfy the requirements of 10 CFR Part 50, Appendix G using analytical methods and acceptance criteria of the ASME Boiler and Pressure Vessel Code, Section Xl, Appendix G, 1995 Edition with Addenda through 1996 and ASME Code Cases N-588 & N640.

A.D. Nar

~-

Date Materials and Structural Analysis Unit This report has been reviewed for technical content and accuracy.

SX. Qsnm6,

&r-.^J/

.B.'Aiall (Material Analysis)

Date Materials and Structural Analysis Unit E Killan (Fracture Analysis)

Date Materials and Structural Analysis Unit Verification of independent review.

A. D. McKim Date Manager, Materials and Structural Analysis Unit This report is approved for release.

D.L. Howell Program 9-1 Date Manager AREVA

BAW-2405, Rev. 2

10.

REFERENCES

1.

J. B. Hall and J. W. Newman, Jr., "Analysis of Capsule W-104 Entergy Operations, Inc.,

Arkansas Nuclear One Unit 2 Power Plant, Reactor Vessel Material Surveillance Program," BAW-2399, Framatome ANP, Inc., Lynchburg, Virginia, September 2001.

2.

Code of Federal Regulation, Title 10, Part 50, "Domestic Licensing of Production and Utilization Facilities," ApDendix G. Fracture Toughness Requirements, Federal Register, December 19,1995.

3.

Code of Federal Regulation, Title 10, Part 50, "Domestic Licensing of Production and Utilization Facilities," Appendix H. Reactor Vessel Material Surveillance Program Requirements, Federal Register, December 19,1995.

4.

American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components," Appendix G. Fracture Toughness Criteria for Protection Against Failure, 1995 Edition with Addenda through 1996.

5.

ASTM Standard E 208-81, "Standard Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels," American Society for Testing and Materials, Philadelphia, Pennsylvania.

6.

U.S. Nuclear Regulatory Commission, "Radiation Embrittlement of Reactor Vessel Materials," Regulatory Guide 1.99. Revision 2, May 1988.

7.

Code of Federal Regulation, Title 10, Part 50, "Domestic Licensing of Production and Utilization Facilities," Section 50.61, "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," Federal Register, December 19,1995.

8.

FRA-ANP Document 32-1171775-08, 'Verification of PTPC & User's Manual," November 1999.

9.

Timoshenko, S.P., and Goodier, J.N., Theory of Elasticit, Third Edition, McGraw-Hill Book Company, 1970.

1A 10-1 AREVA

BAW-2405, Rev. 2

10.

PVRC Ad Hoc Group on Toughness Requirements, "PVRC Recommendations on Toughness Requirements for Ferritic Materials," Bulletin No. 175, Welding Research Council, August 1972.

11.

ASME Boiler and Pressure Vessel Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves," Section Xl, Division 1. Approval date:

February 26, 1999.

12.

ASME Boiler and Pressure Vessel Code Case N-588, 'Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels," Section Xl, Division 1. Approval date: December 12, 1997.

13.

FRA-ANP Document 32-5014182-00, "ANO-2 KIC Based Corrected P-T Limits at 32 EFPY," September, 2001.

10-2 AREVA

2CAN040405, License Amendment Request Technical Specification Change Request for Modification of ANO-2 RCS

Contents

Text

SUBJECT:

REFERENCES:

Dear Sir or Madam:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

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2CAN040405, License Amendment Request Technical Specification Change Request for Modification of ANO-2 RCS

Text

SUBJECT:

REFERENCES:

Dear Sir or Madam:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

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