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.

Entergy - 1448SR 333 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 100 cF per hour (constant) or 50 0F in any half hour period (step) 120 0 F *TY *200OF 60 OF per hour (constant) or 30 IF in any half hour period (step)

Tc < 120OF 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 1000 Fperhour(constant) or 500 F in anyhalf hourperiod (step) for RCS cold leg temperatures between 500 F and 560 0F.

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 100 0F 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-2405 1 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

10CFR50, 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 10CFR 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 10CFR50.92, '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 10CFR51 .22(c)(1 0). Therefore, pursuant to 10CFR51 .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.

Attachment 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 2 n0F-pler-heur-(rnat)aor-50°F-i4 any-14alf- uFperied (step) 2F!5 T, 2rm arny-alf-houF-period (step)

T-T.42O2F 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 2000 F 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 Accetable 1500 am 3 Lowest Service N

1000 Un U,

a)

L-500

- -nimum Boltup

$Temperature, 50 F-0 0 100 200 300 400 500 Reactor Coolant Temperature T 0, OF (Curves do not include margins for instrument uncertainties)

THIS CURVE IS BEING DELETED ARKANSAS - UNIT 2 3/4 4-23a Amendmr 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 II I I I I . I .I !

.I I 111..

. . . . . . . . . . . . . . . . . I. . . I . . . .

2000 Cooldwn - Region

.rco CL a 1500 a-(n Lowest Service Temperature, II-_

I_ I-N 6-150 F- -_

0)u, 1000 U,

L 1 . i I I 4 1I I I I. I I I I. . . . . . . 1 I I 1 I i i I I i i i .- I

.,' . . . . 1-- 1

,IIIII IIIIIIIIIIII 500 I -I4--4-I-- I -I I 1 - + I II I . I I 1 -+- ..- i i I 4I i

~2Minimum Boltup 0

I0 100 200 300 400 5C 00 Reactor Coolant Temperature Tc, OF (Curves do not include margins for instrument uncertainties)

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

Attachment 3 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.

Attachment 4 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 Revision Description of Revision Date BAW-2405 Original Release 9/2001 BAW-2405, Rev. 1 Minor error on Table 7-2 & 7-6 at fluid temp. value of 195 F. 4/2004 No impact on P-T limit curves since pressure is above 2500 psig Affected Pages: i through v, 7-5, 7-15, 9-1 BAW-2405, Rev. 2 Complete Re-release with revised normal cooldown 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 LIST OF FIGURES (Cont'd.)

Figure Page 7-2 ANO-2 Composite Cooldown P/T Limit Curve (psig) .......................................... 7-23 7-3 ANO-2 ISLH P/T Limit Curve (psig) .......................................... 7-24 7-4 ANO-2 Ramp Heatup & Critical Core P/T Limit Curves (psia) ................................... 7-26 7-5 ANO-2 Composite Cooldown P/T Limit Curve (psia) .......................................... 7-28 7-6 ANO-2 ISLH P/T Limit Curve (psia) .......................................... 7-30 8-1 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," 1 3' 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."1 41 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 Composition 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 -30 3-2 Ak AR EVA

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% 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 270 0F.

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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 = 24.675 in.

Nozzle radius of inlet nozzle per WRCB 175 = 18.248 in.

Postulated flaws: 1/4T x 3/2T semi-elliptical longitudinal surface flaw in vessel beltline one-inch inside corner flaw in nozzle Closure head limits: When core not critical, 622 psig (uncorrected) up to 150 0 F (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-ft 2 -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-0 F) (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 (0 F)

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; Lm300 I-200 100 0

0 100 200 300 400 500 600 700 Time, minutes 5-6 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 50 100 150 200 250 300 350 Time, minutes 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 AR EVA

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 50 100 150 200 250 300 350 Time, minutes 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-C) 300 ca.

E is I--

200 100 0

0 50 100 150 200 250 300 350 Time, minutes 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.

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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) r 2 +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) 50 593 593 593 593 55 593 593 593 593 60 593 593 593 593 65 593 593 593 593 70 593 593 593 593 75 593 593 593 593 80 593 593 593 593 85 593 593 593 593 90 593 593 593 593 95 593 593 593 593 100 593 593 593 593 105 593 593 593 593 110 593 593 593 593 115 593 593 593 593 120 593 593 593 593 125 593 593 593 593 130 593 593 593 593 135 593 593 593 593 140 593 593 593 593 145 593 593 593 593 150 593 593 593 593 150 1148 1066 995 934 155 1215 1125 1047 980 160 1289 1191 1105 1031 165 1371 1263 1170 1088 170 1462 1344 1241 1152 175 1562 1433 1320 1222 180 1673 1532 1408 1300 185 1795 1641 1505 1386 190 1931 1762 1613 1482 0 190 1931 1762 1613 1482 934 195 2081 1895 1732 1588 980 200 2204 2005 1825 1667 1031 205 2378 2168 1970 1797 1088 210 2569 2348 2131 1940 1152 215 2781 2546 2308 2098 1222 220 3015 2766 2504 2273 1300 225 3263 3008 2720 2466 1386 230 3263 3263 2959 2679 1482 235 3263 3263 3223 2915 1588 240 3263 3263 3263 3175 1667 245 3263 3263 3263 3263 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 3263 3263 3263 3263 1940 255 3263 3263 3263 3263 2098 260 3263 3263 3263 3263 2273 265 3263 3263 3263 3263 2466 270 3263 3263 3263 3263 2679 275 3263 3263 3263 3263 2915 280 3263 3263 3263 3263 3175 285 3263 3263 3263 3263 3263 290 3263 3263 3263 3263 3263 295 3263 3263 3263 3263 3263 300 3263 3263 3263 3263 3263 305 3263 3263 3263 3263 3263 310 3263 3263 3263 3263 3263 315 3263 3263 3263 3263 3263 320 3263 3263 3263 3263 3263 325 3263 3263 3263 3263 3263 330 3263 3263 3263 3263 3263 335 3263 3263 3263 3263 3263 340 3263 3263 3263 3263 3263 345 3263 3263 3263 3263 3263 350 3263 3263 3263 3263 3263 355 3263 3263 3263 3263 3263 360 3263 3263 3263 3263 3263 365 3263 3263 3263 3263 3263 370 3263 3263 3263 3263 3263 375 3263 3263 3263 3263 3263 380 3263 3263 3263 3263 3263 385 3263 3263 3263 3263 3263 390 3263 3263 3263 3263 3263 395 3263 3263 3263 3263 3263 400 3263 3263 3263 3263 3263 405 3263 3263 3263 3263 3263 410 3263 3263 3263 3263 3263 415 3263 3263 3263 3263 3263 420 3263 3263 3263 3263 3263 425 3263 3263 3263 3263 3263 430 3263 3263 3263 3263 3263 435 3263 3263 3263 3263 3263 440 3263 3263 3263 3263 3263 445 3263 3263 3263 3263 3263 450 3263 3263 3263 3263 3263 455 3263 3263 3263 3263 3263 460 3263 3263 3263 3263 3263 465 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) (psig) (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. Temp.

(F) (psig) (F) (psig) (F) (psig) 50 502 265 2699 485 2952 55 524 270 2700 490 2969 60 547 275 2700 495 2986 65 573 280 2701 500 3004 70 593 285 2702 505 3023 75 593 290 2703 510 3043 80 593 295 2704 515 3064 85 593 300 2706 520 3086 90 593 305 2707 525 3108 95 593 310 2709 530 3132 100 593 315 2710 535 3156 105 593 320 2712 540 3181 110 593 325 2716 545 3206 115 593 330 2718 550 3231 120 593 335 2720 555 3252 125 593 340 2723 560 3263 130 593 345 2726 135 593 350 2729 140 593 355 2732 145 593 360 2736 150 593 365 2739 150 1410 370 2743 155 1497 375 2747 160 1593 380 2750 165 1699 385 2755 170 1816 390 2761 175 1946 395 2767 180 2089 400 2773 185 2247 405 2779 190 2422 410 2786 195 2616 415 2794 200 2700 420 2801 205 2700 425 2810 210 2700 430 2818 215 2700 435 2828 220 2699 440 2837 225 2700 445 2847 230 2699 450 2858 235 2699 455 2870 240 2699 460 2882 245 2699 465 2894 250 2699 470 2908 255 2699 475 2922 260 2699 480 2937 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) (psig) (psig) I 50 943 1935 1408 593 593 55 953 202E 1461 593 593 60 967 2152 1531 593 593 65 984 2295 1612 593 593 70 1004 2457 1704 593 593 75 1026 2636 1807 593 593 80 1050 284C 1921 593 593 85 1077 3063 2048 593 593 90 1107 3311 2188 593 593 95 114C 3586 2344 593 593 100 1176 388E 2515 593 593 105 1216 4224 2705 593 593 110 1261 4362 2916 593 593 115 1310 4362 3148 593 593 120 1364 4362 3405 593 593 125 1424 4362 3688 593 593 130 1490 4362 4002 593 593 135 1563 4362 4348 593 593 140 1643 4362 4491 593 593 145 1733 4362 4491 593 593 150 1831 4362 4491 593 593 150 1831 4362 4491 2500 1831 155 1940 4362 4491 194C 160 2061 4362 4491 2061 165 2194 4362 4491 2194 170 2341 4362 4491 2341 175 2503 4362 4491 250 180 2683 4362 4491 2683 185 2881 4362 4491 2881 190 3100 4362 4491 310O 195 3342 4362 4491 3342 200 3572 4362 4453 3572 205 386E 4362 4453 386E 21C 4196 4362 4453 419E 215 4557 4362 4453 4362 220 4957 4362 4453 4362 225 4984 4362 4453 4362 23C 4984 4362 4453 4362 235 4984 4362 4453 4362 240 4984 4362 4453 4362 245 4984 4362 4453 4362 25C 4984 4362 4453 4362 255 4984 4362 4453 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) (psig) (psig) 26C 4984 4362 4453 4362 265 4984 4362 4453 4362 27C 4984 4362 4453 4362 275 4984 4362 4453 4362 28C 4984 4362 4453 4362 285 4984 4362 4453 4362 29C 4984 4362 4453 4362 295 4984 4362 4453 4362 30C 4984 4362 4453 4362 305 4984 4362 4453 4362 31C 4984 4362 4453 4362 315 4984 4362 4453 4362 32C 4984 4362 4453 4362 325 4984 4362 4453 4362 33C 4984 4362 4453 4362 335 4984 4362 4453 4362 34C 4984 4362 4453 4362 345 4984 4362 4453 4362 35C 4984 4362 4453 4362 355 4984 4362 4453 4362 36C 4984 4362 4453 4362 365 4984 4362 4453 4362 37C 4984 4362 4453 4362 37! 4984 4362 4453 4362 38C 4984 4362 4453 4362 385 4984 4362 4453 4362 39C 4984 4362 4453 4362 395 4984 4362 4453 4362 40C 4984 4362 4453 4362 405 4984 4362 4453 4362 41C 4984 4362 4453 4362 415 4984 4362 4453 4362 42C 4984 4362 4453 4362 425 4984 4362 4453 4362 43C 4984 4362 4453 4362 435 4984 4362 4453 4362 44C 4984 4362 4453 4362 44E 4984 4362 4453 4362 45C 4984 4362 4453 4362 455 4984 4362 4453 4362 46C 4984 436 4453 4362 465 4984 436 4453 4362 47C 4984 4362 4453 4362 475 4984 4362 4453 4362 48C 4984 4362 4453 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) (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)(

50 696 1444 1045 593 593 55 708 1541 110 593 59 60 721 1649 1161 593 59 65 735 1768 122 593 59 70 751 1900 130 593 59 75 768 2045 1386 593 59 81 788 2206 1477 593 59 85 809 2384 1578 593 59 90 833 2581 1689 593 59 95 859 2799 1813 593 59 100 888 3039 1949 593 59 105 920 3263 2099 593 59 110 955 3262 2266 593 59 115 994 3263 2449 593 59 120 1037 3263 2652 593 59 125 1085 3263 2877 593 59 130 1138 3263 3125 593 593 135 1196 3263 3356 593 593 140 1260 3263 3356 593 593 145 1331 3263 3356 593 593 150 1410 3263 3356 593 593 150 1410 3263 3356 2500 141 155 1497 3263 3356 149 160 1593 3263 3356 159 165 1699 3263 3356 169 170 181E 3263 3356 181 175 194E 3263 3356 194 180 2089 3263 3356 208 185 2247 3263 3356 224 190 2422 3263 3356 242 195 2616 3263 3356 261 200 2791 3263 3318 2791 205 3027 3263 3318 3027 21C 328E 3268 3318 326 215 3577 3263 3318 326 22C 3716 3263 3318 3263 225 3716 3263 3318 3263 230 3716 3263 3318 3263 235 3716 3263 3318 3263 240 3716 3263 3318 3263 245r 3716 3263 3318 3263 25C 3716 3263 3318 3263 255 3716 3263 3318 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) L (psig)p pig) 260 3716 3262 3318 326 265 3716 3262 3318 326 27C 3716 3262 3318 326 275 3716 3262 3318 326 280 3716 3262 3318 326 285 3716 3262 3318 326 290 3716 3262 3318 326 295 3716 3262 3318 3263 30C 3716 3262 3318 3263 305 3716 3262 3318 3263l 31C 3716 3262 3318 3263 315 3716 3262 3318 3263 32C 3716 3262 3318 3263 325 3716 3262 3318 3263 330 3716 3262 3318 3263l 335 3716 3262 3318 326 340 3716 3262 3318 326 345 3716 3262 3318 326 350 3716 3262 3318 3263 355 3716 3262 3318 3263 360 3716 3262 3318 3263 365 3716 3262 3318 3263 370 3716 3262 3318 326 375 3716 3262 3318 326 380 3716 3262 3318 326 385 3716 3262 3318 3263 390 3716 3262 3318 3263 395 3716 3262 3318 3263 40C 3716 3262 3318 3263 405 3716 326 3318 3263 41C 3716 3262 3318 3263 415 3716 3262 3318 3263 420 3716 3262 3318 3263 425 3716 3262 3318 3263 430 3716 3262 3318 3263 435 3716 3262 3318 3263 440 3716 3262 3318 3263 44E 3716 3262 3318 3263 450 3716 3262 3318 3263 455 3716 3262 3318 3263 460 3716 3262 3318 3263 465 3716 3262 3318 3263 470 3716 3262 3318 3263 475 3716 3262 3318 3263 48C 371E 3262 3318 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) (psig) (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) (psia) (psia) (psia) 50 608 608 608 608 55 608 608 608 608 60 608 608 608 608 65 608 608 608 608 70 608 608 608 608 75 608 608 608 608 80 608 608 608 608 85 608 608 608 608 90 608 608 608 608 95 608 608 608 608 100 608 608 608 608 105 608 608 608 608 110 608 608 608 608 115 608 608 608 608 120 608 608 608 608 125 608 608 608 608 130 608 608 608 608 135 608 608 608 608 140 608 608 608 608 145 608 608 608 608 150 608 608 608 608 150 1163 1081 1010 949 155 1230 1140 1062 995 160 1304 1206 1120 1046 165 1386 1278 1185 1103 170 1477 1359 1256 1167 175 1577 1448 1335 1237 180 1688 1547 1423 1315 185 1810 1656 1520 1401 190 1946 1777 1628 1497 15 190 1946 1777 1628 1497 949 195 2096 1910 1747 1603 995 200 2219 2020 1840 1682 1046 205 2393 2183 1985 1812 1103 210 2584 2363 2146 1955 1167 215 2796 2561 2323 2113 1237 220 3030 2781 2519 2288 1315 225 3278 3023 2735 2481 1401 230 3278 3278 2974 2694 1497 235 3278 3278 3238 2930 1603 240 3278 3278 3278 3190 1682 245 3278 3278 3278 3278 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 3278 3278 3278 3278 1955 255 3278 3278 3278 3278 2113 260 3278 3278 3278 3278 2288 265 3278 3278 3278 3278 2481 270 3278 3278 3278 3278 2694 275 3278 3278 3278 3278 2930 280 3278 3278 3278 3278 3190 285 3278 3278 3278 3278 3278 290 3278 3278 3278 3278 3278 295 3278 3278 3278 3278 3278 300 3278 3278 3278 3278 3278 305 3278 3278 3278 3278 3278 310 3278 3278 3278 3278 3278 315 3278 3278 3278 3278 3278 320 3278 3278 3278 3278 3278 325 3278 3278 3278 3278 3278 330 3278 3278 3278 3278 3278 335 3278 3278 3278 3278 3278 340 3278 3278 3278 3278 3278 345 3278 3278 3278 3278 3278 350 3278 3278 3278 3278 3278 355 3278 3278 3278 3278 3278 360 3278 3278 3278 3278 3278 365 3278 3278 3278 3278 3278 370 3278 3278 3278 3278 3278 375 3278 3278 3278 3278 3278 380 3278 3278 3278 3278 3278 385 3278 3278 3278 3278 3278 390 3278 3278 3278 3278 3278 395 3278 3278 3278 3278 3278 400 3278 3278 3278 3278 3278 405 3278 3278 3278 3278 3278 410 3278 3278 3278 3278 3278 415 3278 3278 3278 3278 3278 420 3278 3278 3278 3278 3278 425 3278 3278 3278 3278 3278 430 3278 3278 3278 3278 3278 435 3278 3278 3278 3278 3278 440 3278 3278 3278 3278 3278 445 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) (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. Temp.

(F) (psia) . (F) (psia) . (F) (psia) 50 517 265 2714 485 2967 55 539 270 2715 490 2984 60 562 275 2715 495 3001 65 588 280 2716 500 3019 70 608 285 2717 505 3038 75 608 290 2718 510 3058 80 608 295 2719 515 3079 85 608 300 2721 520 3101 90 608 305 2722 525 3123 95 608 310 2724 530 3147 100 608 315 2725 535 3171 105 608 320 2727 540 3196 110 608 325 2731 545 3221 115 608 330 2733 550 3246 120 608 335 2735 555 3267 125 608 340 2738 560 3278 130 608 345 2741 135 608 350 2744 140 608 355 2747 145 608 360 2751 150 608 365 2754 150 1425 370 2758 155 1512 375 2762 160 1608 380 2765 165 1714 385 2770 170 1831 390 2776 175 1961 395 2782 180 2104 400 2788 185 2262 405 2794 190 2437 410 2801 195 2631 415 2809 200 2715 420 2816 205 2715 425 2825 210 2715 430 2833 215 2715 435 2843 220 2714 440 2852 225 2715 445 2862 230 2714 450 2873 235 2714 455 2885 240 2714 460 2897 245 2714 465 2909 250 2714 470 2923 255 2714 475 2937 260 2714 480 2952 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) ,(psia) 50 958 1950 1423 608 608 55 968 2044 1476 608 608 60 982 2167 1546 608 608 65 999 2310 1627 608 608 70 1019 2472 1719 608 608 75 1041 2653 1822 608 608 80 1065 2855 1936 608 608 85 1092 3078 2063 608 608 90 1122 3326 2203 608 608 95 1155 3601 2359 608 608 100 1191 3904 2530 608 608 105 1231 4239 2720 608 608 110 1276 4377 2931 608 608 115 1325 4377 3163 608 608 120 1379 4377 3420 608 608 125 1439 4377 3703 608 608 130 1505 4377 4017 608 608 135 1578 4377 4363 608 608 140 1658 4377 4506 608 608 145 1748 4377 4506 608 608 150 1846 4377 4506 608 608 150 1846 4377 4506 2515 1846 155 1955 4377 4506 1955 160 2076 4377 4506 2076 165 2209 4377 4506 2209 170 2356 4377 4506 2356 175 2518 4377 4506 2518 180 2698 4377 4506 2698 185 2896 4377 4506 2896 190 3115 4377 4506 3115 195 3357 4377 4506 3357 200 3587 4377 4468 3587 205 3883 4377 4468 3883 210 4211 4377 4468 4211 215 4572 4377 4468 4377 220 4972 4377 4468 4377 225 4999 4377 4468 4377 230 4999 4377 4468 4377 235 4999 4377 4468 4377 240 4999 4377 4468 4377 245 4999 4377 4468 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) (psia) (psia) (psia) 250 4999 4377 4468 4377 255 4999 4377 4468 4377 260 4999 4377 4468 4377 265 4999 4377 4468 4377 270 4999 4377 4468 4377 275 4999 4377 4468 4377 280 4999 4377 4468 4377 285 4999 4377 4468 4377 290 4999 4377 4468 4377 295 4999 4377 4468 4377 300 4999 4377 4468 4377 305 4999 4377 4468 4377 310 4999 4377 4468 4377 315 4999 4377 4468 4377 320 4999 4377 4468 4377 325 4999 4377 4468 4377 330 4999 4377 4468 4377 335 4999 4377 4468 4377 340 4999 4377 4468 4377 345 4999 4377 4468 4377 350 4999 4377 4468 4377 355 4999 4377 4468 4377 360 4999 4377 4468 4377 365 4999 4377 4468 4377 370 4999 4377 4468 4377 375 4999 4377 4468 4377 380 4999 4377 4468 4377 385 4999 4377 4468 4377 390 4999 4377 4468 4377 395 4999 4377 4468 4377 400 4999 4377 4468 4377 405 4999 4377 4468 4377 410 4999 4377 4468 4377 415 4999 4377 4468 4377 420 4999 4377 4468 4377 425 4999 4377 4468 4377 430 4999 4377 4468 4377 435 4999 4377 4468 4377 440 4999 4377 4468 4377 445 4999 4377 4468 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) (psia) (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) (psia) (psia) 50 711 1459 1060 608 608 55 723 1556 1115 608 608 60 736 1664 1176 608 608 65 750 1783 1243 608 608 70 766 1915 1318 608 608 75 783 2060 1401 608 608 80 803 2221 1492 608 608 85 824 2399 1593 608 608 90 848 2596 1704 608 608 95 874 2814 1828 608 608 100 903 3054 1964 608 608 105 935 3278 2114 608 608 110 970 3278 2281 608 608 115 1009 3278 2464 608 608 120 1052 3278 2667 608 608 125 1100 3278 2892 608 608 130 1153 3278 3140 608 608 135 1211 3278 3371 608 608 140 1275 3278 3371 608 608 145 1346 3278 3371 608 608 150 1425 3278 3371 608 608 150 1425 3278 3371 2515 1425 155 1512 3278 3371 1512 160 1608 3278 3371 1608 165 1714 3278 3371 1714 170 1831 3278 3371 1831 175 1961 3278 3371 1961 180 2104 3278 3371 2104 185 2262 3278 3371 2262 190 2437 3278 3371 2437 195 2631 3278 3371 2631 200 2806 3278 3333 2806 205 3042 3278 3333 3042 210 3303 3278 3333 3278 215 3592 3278 3333 3278 220 3731 3278 3333 3278 225 3731 3278 3333 3278 230 3731 3278 3333 3278 235 3731 3278 3333 3278 240 3731 3278 3333 3278 245 3731 3278 3333 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) (psia) .

(psia) 250 3731 3278 3333 3278 255 3731 3278 3333 3278 260 3731 3278 3333 3278 265 3731 3278 3333 3278 270 3731 3278 3333 3278 275 3731 3278 3333 3278 280 3731 3278 3333 3278 285 3731 3278 3333 3278 290 3731 3278 3333 3278 295 3731 3278 3333 3278 300 3731 3278 3333 3278 305 3731 3278 3333 3278 310 3731 3278 3333 3278 315 3731 3278 3333 3278 320 3731 3278 3333 3278 325 3731 3278 3333 3278 330 3731 3278 3333 3278 335 3731 3278 3333 3278 340 3731 3278 3333 3278 345 3731 3278 3333 3278 350 3731 3278 3333 3278 355 3731 3278 3333 3278 360 3731 3278 3333 3278 365 3731 3278 3333 3278 370 3731 3278 3333 3278 375 3731 3278 3333 3278 380 3731 3278 3333 3278 385 3731 3278 3333 3278 390 3731 3278 3333 3278 395 3731 3278 3333 3278 400 3731 3278 3333 3278 405 3731 3278 3333 3278 410 3731 3278 3333 3278 415 3731 3278 3333 3278 420 3731 3278 3333 3278 425 3731 3278 3333 3278 430 3731 3278 3333 3278 435 3731 3278 3333 3278 440 3731 3278 3333 3278 445 3731 3278 3333 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) (psia) (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 50 100 150 200 250 300 350 400 450 500 Temperature, F 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 0

w XL 1000 750 500 70 250 0 50 100 150 200 250 300 350 400 450 500 Temperature, F 7-23 At A R EVA

BAW-2405, Rev. 2 Figure 7-3. ANO-2 ISLH PIT Limit Curve (psig) 2500 2250 2000 1750 1500 co (a

0 cl.

1250 0

a.

1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 Temperature, F 7-24 A 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 l70 F/hr tLt iF/r__1kl

.U NonCrtical Core 0 F/

- 1500 a) -- on-Cri ical Core (n

05 L.-

N L..

-Fh CriicX-a Core

2a' 1000 P AAcceptable 500 TIITI!T~llillIITITITITRegIion~lII 0

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 II I I .I I Ii . . . 1. 1. 1. . 1. 1. I . . . I I I I I 1I I 1 I I 2000 In 0.(

f 1500 n

In 0

IL.

I-a)

N

]Temperature,E------------------ I I. .5 Flllii u) 1000 0 II I I I. . . I . . . I.I.II.I.I

. . . . . . . . . . . . I .I. I I. I I . . . . I I I I I (L

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 II I I I I. I I II I I I. . . . I X ~~MiiuBoltup t\i\ttt 0

I0 100 200 300 400 50i0 Reactor Coolant Temperature T,, °F (Curves do not Include margins for Instrument uncertainties) 7-28 ARE AR 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

~tatic I___--I H1tgysdero :

2000

. t]1 j ~~Acceptable __ _

1500 n Lowest Service U) Temperature, 150 N

in1000 111111 11 ll ll liilit Tit I Miiu Blu 500 4 0\iit-1 ___1 00 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. Temp.

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

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 50 100 150 200 250 300 350 400 450 500 Temperature, F 8-3 A

AR EVA

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 Date Program Manager 9-1 AREVA

BAW-2405, Rev. 2

10. REFERENCES

1. J. B. Hall and J. W. Newman, Jr., "Analysisof 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

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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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