Text

Request to Revise Technical Specifications Associated with the Low Temperature Overpressure Protection System and the Pressure and Temperature Limits Report
	ML12229A521
Person / Time
Site:	Farley
Issue date:	08/15/2012
From:	Ajluni M; Southern Co, Southern Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	NL-12-0868
	Download: ML12229A521 (216)
	v • d • e

Mark J. Ajluni, P.E. Southern Nuclear Nuclear Licensing Director Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham. Alabama 35201 Tel 205 .992.7673 August 15, 2012 Fax 205.992.7885 SOUTHERN'\'

Docket Nos.: 50-348 50-364 NL-12-0868 COMPANY U. S. Nuclear Regulatory Commission AnN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Units 1 & 2 Request to Revise Technical Specifications Associated with the Low Temperature Overpressure Protection System and the Pressure and Temperature Limits Report Ladies and Gentlemen:

In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), Southern Nuclear Operating Company (SNC) is submitting a request for an amendment to the Technical Specifications (TS) associated with the Low Temperature Overpressure Protection (LTOP) System and the Pressure and Temperature Limits Report (PTLR) for Joseph M. Farley Nuclear Plant (FNP) .

Additionally, revised PTLRs for FNP Unit 1 and Unit 2 are provided in accordance with the reporting requirements of FNP TS 5.6.6, "Reactor Coolant System (RCS)

Pressure and Temperature Limits Report."

The PTLRs for FNP Unit 1 and Unit 2 were revised to implement new 54 Effective Full Power Years (EFPY) pressure and temperature limit curves. The revised pressure and temperature limits were determined in accordance with the NRC approved methodology in WCAP-14040-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 4, May 2004.

Based on the implementation of new 54 EFPY pressure and temperature limit curves discussed above, the following changes to the FNP TS are proposed:

TS 3.4.12, "Low Temperature Overpressure Protection (LTOP) System,"

Mode 4 applicability temperature (:::; 325°F) below which the LTOP System must be operable would be revised to:::; 275°F (this revised value will be located in the PTLR, as discussed below).

The methodology used to determine the RCS pressure and temperature limits identified in Specification 5.6.6, "Reactor Coolant System (RCS)

Pressure and Temperature Limits Report (PTLR): would be revised to reference WCAP-14D40-A, "Methodology Used to Develop Cold

U. S. Nuclear Regulatory Commission Nl-12-0868 2

WCAP-14040-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and Heatup and Cooldown limit Curves," Revision May 2004.

In addition, following PTlR and lTOP System changes to the FNP TS are proposed:

It The l TOP applicability temperature would be relocated from FNP to the PTlR. relocation of l TOP System applicability temperature is based on NRC approved changes 431, "Standard Specifications Westinghouse Plants." These changes were approved by NRC TS Force (TSTF) Traveler number TSTF-233-A, l TOP Arming Temperature to PTlR,"

Revision 0, by July 1 1998. The changes proposed in TSTF-233 were subsequently incorporated into Revision 2 1431. Consistent with TSTF-233, the relocation of the System arming temperature (Le., the FNP l TOP applicability temperature) to the PTlR would the following TS sections:

>> definition of the in Section 1.1, "Definitions,"

>> 3.4.6, "RCS loops - MODE 4,"

>> TS 3.4.7, loops - MODE 5, loops "

>> TS 3.4.10, Safety Valves,"

>> 3.4.1 "low Temperature Overpressure Protection (lTOP)

System,"

>> Specification 5.6.6, "Reactor Coolant (RCS) Pressure and Temperature limits Report (PTlR)."

It The definition of the PTlR in Section 1.1, "Definitions," and Specification "Reactor Coolant System (RCS) Pressure and limits Report (PTlR)," would be revised in accordance with TSTF-419-A, "Revise PTlR Definition and References in ISTS RCS

" Revision O. Traveler 9, Revision 0 was approved by the NRC by dated March ,2002. The changes proposed in TSTF 9 have subsequently been incorporated into Revision 3 of NUREG 1431.

.. lCO, Actions, and Surveillance Requirements of 12, "low Temperature Overpressure Protection (lTOP) System," would be revised to address a maximum of two charging pumps capable of injecting into the RCS clarify the TS requirements to preserve the applicable safety analysis.

Appropriate changes would also be made consistent with the discussed above.

Enclosure 1 provides the for the proposed change to FNP TS.

Enclosure 2 provides FNP and markup pages showing proposed changes. Enclosure 3 provides the FNP TS clean typed showing the proposed changes. Enclosure 4 provides the revised Unit 1 and Unit

U. S. Nuclear Regulatory Commission NL-12-0868 Page 3 2 PTLRs. Enclosure 5 provides WCAP-17122-NP, "J. M. Farley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," which contains the methodology and results of the development of the new 54 EFPY heatup and cooldown pressure and temperature limit curves for FNP Unit 1. Enclosure 6 provides WCAP-17123-NP, "J. M. Farley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," which contains the methodology and results of the development of the new 54 EFPY heatup and cooldown pressure and temperature limit curves for FNP Unit 2.

SNC requests approval of the proposed license amendment by August 6, 2013.

The proposed changes would be implemented within 60 days of issuance of the amendments.

In accordance with 10 CFR 50.91 (b)(1), "State Consultation," a copy of this application and its reasoned analysis about no significant hazards considerations is being provided to the designated Alabama officials.

This letter contains no NRC commitments. If you have any questions, please contact Jack Stringfellow at (205) 992-7037.

Mr. M. J. Ajluni states he is Nuclear Licensing Director of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.

Respectfully submitted, M. J. Ajluni Nuclear Licensing Director

""""'-~=.:.--_ _, 2012.

My commission expires: J l-- 07.-10t3 MJA/RMJ/lac

S. Nuclear Regulatory Commission NL-12-0868 4

Enclosures:

1. Basis for Proposed Change Technical Specifications and Bases Markup Pages Technical Specifications Clean Typed Joseph M. Farley Nuclear Plant Unit 1 and Unit 2 Pressure Temperature Limits Report

5. WCAP-17122-NP, Revision 0, "J. M. Farley Unit 1 Heatup Cooldown Limit Curves for Normal Operation," October 2009

6. WCAP-1 23-NP, Revision 1, "J. M. Farley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," July 2011 cc:

Mr. Kuczynski, Chairman, President &

Mr. D. G. Bost, President & Chief Nuclear Officer Mr. T. A. Lynch, Vice President - Farley Mr. B. L. Ivey, Vice President - Regulatory Affairs Mr. J. Adams, Vice President - Fleet Operations RTYPE: CFA04.054 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. R. Martin, NRR Project Manager - Farley Mr. L. Crowe, Senior Resident Inspector Farley

Joseph M. Farley Nuclear Plant Request to Revise Technical Specifications Associated with the Low Temperature Overpressure Protection System and the Pressure and Temperature Limits Report Enclosure 1 Basis for Proposed Change to NL-12-0868 Basis for Proposed Change Table of Contents

1. Summary Description

2. Detailed Description

3. Technical Evaluation

4. Regulatory Evaluation 4.1 Significant Hazards Consideration 4.2 Applicable Regulatory Requirements/Criteria 4.3 Conclusions

5. Environmental Consideration

6. References E1-2 NL-12-0868 Basis for Proposed Change

1. Summary Description This evaluation supports a request to amend Appendix A of Operating Licenses NPF-2 and NPF-8 for the Joseph M. Farley Nuclear Plant (FNP) Unit 1 and Unit 2, respectively.

The Pressure and Temperature Limits Reports (PTLRs) for FNP Unit 1 and Unit 2 were revised to implement new 54 Effective Full Power Years pressure and temperature limit curves. The revised pressure and temperature limits were determined in accordance with the NRC approved methodology in WCAP-14040-A, "Methodology Used to Develop Cold Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision (Reference 1). revised are provided in Enclosure 4.

WCAP-17122-NP, "J. M. Farley Unit 1 Heatup and Cooldown Limit for Normal Operation," (Reference 2) contains the methodology and results of the development of new 54 heatup and cooldown pressure and temperature limit curves for FNP Unit 1. WCAP-171 is provided in Enclosure WCAP 17123-NP, M. Farley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," (Reference 3) contains the methodology and results of the development of the new 54 heatup and cooldown pressure and temperature limit curves for FNP Unit 2. WCAP-17123-NP is provided in Enclosure 6.

Based on the implementation of new 54 pressure and temperature limit curves the following changes to FNP Technical Specifications (TS) are proposed:

III TS 3.4.1 "Low Temperature Overpressure Protection (LTOP) System," Mode 4 applicability temperature 325°F), below which the LTOP must operable, would be revised::::; (this revised value will be located in the PTLR, as discussed below).

III The methodology used to determine the pressure and temperature limits identified in Specification 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," would be revised to reference WCAP 14040-A, "Methodology to Develop Cold Mitigating System Setpoints and RCS Heatup and Cool down Limit Curves," Revision 4.

In addition, the following PTLR and LTOP System related changes to the FNP TS are also proposed:

III LTOP System applicability temperature would be relocated from the FNP to the PTLR. The relocation of the LTOP System applicability temperature is based on NRC approved changes to NUREG-1431, "Standard Technical Specifications - Westinghouse Plants" (Reference 4). These changes were approved by NRC via TS (TSTF) Traveler number TSTF-233-A, "Relocate LTOP Arming to PTLR," Revision 0 (Reference 5) by letter dated July 16, 1998 (Reference 6). The proposed in TSTF-233 have subsequently been incorporated into NUREG-1431. Consistent with TSTF relocation of the LTOP System arming temperature (I.e., the FNP LTOP applicability temperature) to the PTLR would affect the following sections:

E1 to NL-12-0868 Basis for Proposed Change

>> The definition of in Section 1.1 "Definitions,"

>> 3.4.6, "RCS Loops - MODE "

>> 3.4.7, Loops MODE Loops "

>> TS 0, "Pressurizer Safety Valves,"

>> TS 1 "Low Temperature Overpressure Protection (LTOP) System,"

>> Specification 5.6.6, "Reactor Coolant (RCS) Pressure and Temperature Limits (PTLR)"

The definition of the in TS Section 1.1 and Specification 5.6.6 would be revised in accordance with TSTF-419-A, "Revise PTLR Definition and References in ISTS PTLR," (Reference 7) to eliminate redundant references to the applicable TS. Traveler 9, Revision 0 was approved by NRC by letter dated March , 2002 (Reference 8). The proposed in 9 have subsequently been incorporated into NUREG-1

The LCO, Actions, and Surveillance Requirements (SRs) of 3.4.12 would be revised to address a maximum two charging pumps capable of injecting into RCS to clarify the requirements to preserve the applicable safety analysis.

Appropriate Bases changes would also made consistent with the changes discussed above.

Detailed Description Due to number of changes proposed in this license amendment request, changes are categorized by type. Each type of proposed change has a different for the change and is discussed separately from the other types of changes.

Combining discussion of different types of changes is minimized in order avoid confusion. overall of the combination of changes can viewed on the TS markups included in Enclosure Revised PTlR and LTOP Applicability Temperature.

Consistent with the revised and temperature limits, a new LTOP applicability temperature (Le., temperature below which LTOP is required operable) was calculated. As a result, the LTOP applicability temperature will be revised from:::; to :::; revised value will located in the as discussed in Section 2.2.

use WCAP-14040-A, Revision 4, in determining the new pressure and temperature limits and LTOP System Applicability temperature results in a I"'n'~nr.c to the methodology identified in Specification "Reactor Coolant System (RCS) Temperature Report (PTLR)." FNP Specification contains the reporting requirements applicable to the Specification 5.6.6.b currently states:

'The analytical methods used to determine the and temperature limits be those previously reviewed and approved by specifically described in the NRC letters dated March 31, 1998 and April 3, 1 "

-4 to I\IL-12-0868 for Change NRC letters, dated March ,1998 and April 1998, are References 9 and 10, respectively.

The proposed change would revise Specification to "The analytical methods used to determine the RCS pressure and temperature limits shall be previously reviewed and approved by the NRC, specifically those described in WCAP-14040-A, 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004."

2.2. Relocation of L TOP Applicability Temperature to the PTLR (TSTF-233-A)

The System applicability temperature would relocated from the FNP to the relocation the LTOP System applicability temperature is based on NRC approved changes . These changes were approved by the NRC via TSTF Traveler number TSTF-233-A by letter dated July 16, 1 The changes proposed in were subsequently incorporated into Revision 2 of NUREG-1431.

FNP LTOP System design does not include an "arming" feature, and LTOP TS only specifies an applicability temperature (Le., the temperature below which the LTOP System is required operable). As in TSTF-233, change provides an option for replacing the explicit temperature below which the LTOP system must be operable with a to the temperature as specified in the PTLR." on to "The temperature defines the LCO 3.4.1 LTOP, Applicability in MODE 4." As such, the arming temperature referred to in TSTF-233 is same as the applicability re for the LTOP System. Therefore, the changes contained in apply to the LTOP System applicability temperature.

The changes to the LTOP System applicability temperature included in 233 affect numerous sections. Consistent with TSTF-233, the LTOP System applicability temperature is used in the TS to identify the applicability of the LTOP System TS and Safety Valve TS in Mode 4. In addition, the LTOP System applicability is used in the TS as a low temperature limit for starting an pump. In some cases, where the LTOP System applicability temperature is high enough to affect the Mode 4 to Mode 3 transition temperature (Le., 350°F), the LTOP System applicability temperature is used in a note to modify the requirements of LCO "ECCS - Operating." The FNP LTOP System applicability temperature is below Mode 4 to Mode 3 transition temperature and does not impact the requirements of LCO 3.5.2. Therefore, a change to the FNP LTOP System applicability temperature only affects the FNP TS for RCS loops, the LTOP System, the Safety Valves, the PTLR definition and PTLR reporting requirements. As such, implementation of to relocate the LTOP System applicability temperature would revise the FNP as follows:

-5 to NL-12-0868 Basis for Proposed Change

TS Section 1.1 contains the definition of the PTLR. The first sentence of the current FNP PTLR definition "The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including and cooldown rates, for the current reactor vessel fluence period."

The proposed change would the first sentence of the FNP PTLR definition as follows:

"The is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown and the Low Temperature Overpressure Protection System applicability temperature, for the current reactor vessel fluence period."

TS 3.4.6 specifies the loop operability requirements in Mode 4, including a note pertaining to the start of an RCS pump. LCO 3.4.6 Note 2 in part:

"No RCP shall started with any RCS cold leg temperature ~

The proposed change would revise Note 2 as follows:

"No RCP shall be started with any RCS cold leg temperature ~ the Low Temperature Overpressure Protection (LTOP) System applicability temperature in the PTLR "

TS 3.4.7 the RCS loop operability requirements in Mode 5 with RCS loops filled, including a note pertaining to start of an ReS pump. LCO 3.4.7 Note 3 in part:

"No Reactor coolant pump shall be started with one or more RCS cold leg temperatures ~ 325°F unless: ... "

proposed change would Note 3 as follows:

"No Reactor coolant pump shall be started with one or more RCS cold leg temperatures ~ the Low Temperature Overpressure Protection (LTOP)

System applicability temperature in the PTLR unless: .. ,"

TS 3.4.10 the safety valve operability requirements including the Mode 4 applicability. The Mode 4 applicability for LCO 3.4.10 states:

"MODE 4 with all cold leg temperatures> 325°F."

E1-6

1 to NL-12-0868 Basis for Proposed Change The proposed change would revise the Mode 4 applicability as follows:

"MODE 4 with all RCS cold temperatures> the Low Temperature Protection (LTOP) System applicability temperature specified in the PTLR."

'I 3.4.10 specifies the Required Actions for the pressurizer safety valves which refer to the Mode 4 applicability of LCO 3.4.10 discussed Specifically, Required Action which in MODE 4 with any RCS cold leg temperatures s 325°F."

The proposed change would revise Required Action as follows:

in MODE 4 with any RCS cold temperatures s the LTOP System applicability temperature specified in the PTLR."

'I TS 12 specifies the operability requirements for the LTOP System including the Mode 4 applicability. The Mode 4 applicability LCO 3.4.12 "MODE 4 when the temperature of one or more RCS cold legs is s 325°F."

The proposed change would revise the Mode 4 applicability as follows:

"MODE 4 when the temperature of one or more RCS cold legs is s the L TOP System applicability temperature specified in the PTLR."

'I TS 12 contains the Required Actions for the LTOP System which refer to the Mode 4 applicability temperature requirement discussed above. Specifically, Required Action C.1 which ",t~i'o",'

"Increase cold leg temperatures to > 325°F."

The proposed change would revise Required Action C.1 as follows:

"Increase RCS cold leg temperatures to > the LTOP System applicability temperature specified in the PTLR."

'I Specification contains the requirements for the PTLR report. Specification states in "The reactor coolant system pressure and temperature limits, including heatup and cooldown rates, shall be established and documented in the

-7

1 to Nl-12-0868 Basis for Proposed Change The proposed change would Specification 5.6.6.a as follows:

"The reactor coolant system and temperature limits, including heatup and cooldown rates and the l TOP System applicability temperature, shall be established and documented in the PTlR ... "

Revision of PTLR Definition and References in the PTLR Report (TSTF-41 A) 9-A, Revision 0, was approved by the NRC by letter dated March 2002. The changes proposed in 9 have subsequently been incorporated into I\IUREG-1431. Due to other changes made to NUREG-1 ,

the PTlR requirements in NUREG-1 are in Section 5.6.4. In 9, as well as the FNP TS, the PTlR requirements are in Section 5.6.6 of the TS. This difference is noted only to avoid any potential confusion regarding the different numbering system used in later versions of NUREG-1431 and does not represent a technical change from the NRC approved 9 and NUREG 1431.

TSTF-419 contains two changes. One change simply eliminates the duplication of lCOs identified in the PTlR definition and in Section of the The lCOs identified in these two locations would commonly include lCO 3.4.3, "RCS Pressure and Temperature (PfT) limits," to address the heatup and cooldown curves moved to the and lCO 3.4.12 to address the lTOP System applicability temperature moved to the PTlR. This deletes the lCO referenced in the PTlR definition while leaving the referenced in Section 5.6.6 of TS. details of this change to the FNP TS are . . . o"',...ron below.

The second change in 9 added a bracketed description to part b of

elcncm 5.6.6 in the TS. The bracketed text describes how methodology used to determine the pressure and temperature limits should referenced in Section 5.6.6 of TS. change would allow revision number and approval date of the referenced methodology to be contained in PTlR instead of in Section 5.6.6.b of the However, in with the NRC letter to the TSTF, dated August 2011 (Reference 11), in order to approve amendment requests implementing full topical report or methodology citation (Le., WCAP-14040-A, "Methodology Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004) must included in 5.6.6.b of the TS and not in the PTLR. Therefore, as described in Section 2.1 above, the proposed the methodology in Section of FNP includes the complete citation of WCAP-14040-A, including the revision number and such, implementation of TSTF-419 results in following proposed changes to the FNP

TS Section 1.1 contains the definition of the PTLR. last sentence of the current FNP PTlR definition sta1tes:

-8 to NL-12-0868 Basis for Proposed Change "Plant operation within these operating limits is addressed in LCO 3.4.3, "RCS and Temperature (PfT) Limits."

The proposed change would delete this last from the FNP PTLR definition.

Specification 5.6.6 contains the reporting requirements applicable to the FNP Specification 5.6.6.a states in part:

" ... shall established and documented in the PTLR for LCO 3.4.3."

The proposed change would this part of the FNP Specification as follows:

" ... shall be established and documented in the PTLR for the following:

LCO 3.4.3, "RCS and Temperature (PfT) Limits," and LCO 3.4.1 "Low Temperature Over Protection (LTOP)

System.""

The addition of LCO 3.4.12 to Specification 5.6.6.a is consistent with relocation of the LTOP System applicability temperature to the PTLR, discussed in Section 2.2 above. relocated applicability temperature is associated with the LTOP System, and LCO 2 contains the requirements for the LTOP System.

2.4. Revision of 3.4.12 to Incorporate the Requirement for Two Charging Pumps.

The current 3.4.12 LCO, Actions and SRs only address a maximum of one charging pump capable of injecting into the TS 12 is currently applicable in Mode 4 when one or more RCS cold leg temperatures are s However, TS 12 Note 1, which modifies the LCO requirement, in part that:

"The requirement to have only one charging pump capable of injecting into RCS is only applicable when one or more of the RCS cold legs is s 180°F; ...".

The current TS 12 not clearly address the requirement for a maximum of two RCS charging pumps capable of injecting into the RCS when all RCS cold leg temperatures are> 180°F and one or more cold leg temperatures are s In order to incorporate requirement for a maximum of two charging pumps capable of injecting into the when all cold leg are> 180°F, the proposed change would the following elements of TS 3.4.1

The current LCO 3.4.12 states in part:

-9 to NL-12-0868 for Proposed Change "An LTOP System shall be with a maximum one charging pump capable of injecting into the RCS the accumulators isolated and either a or b below."

proposed change would revise the statement as follows:

"An LTOP System shall be OPERABLE with a maximum of one charging pump of injecting into the RCS when one or more of the RCS cold is :::; 180°F a maximum of two charging pumps capable of injecting into the RCS when all the cold legs are> 180°F the accumulators isolated and either a or b below."

III The two notes modifying LCD requirements would moved from the Applicability statement to below LCD statement, consistent with the position of the similar Notes in the standard 3.4.12 in NUREG-1431, Revision 4. The Notes modify the LCD, not the applicability. The change in the location the Notes addresses a format and is not technical In addition, Note 1 would be revised as follows:

Note 1 currently in "The requirement to have only one charging pump capable of injecting into the is only applicable when one or more of the cold legs is

~ 180°F; however, in this condition, two charging pumps may capable of injecting into the RCS during pump swap operations for a period of no more than 15 minutes .... "

the proposed change to the LCD (discussed above), which clarifies the cold temperatures and associated maximum required charging pumps, the proposed change to Note 1 would clarify Note to state:

'With one or more RCS cold legs ~ 180°F, two charging pumps may be capable of injecting into the RCS during pump swap operations for a period of no more than 15 minutes ...."

III The current Action Condition A "Two or more charging pumps capable of injecting into the "

proposed change would revise Condition A be applicable for either charging pump condition (a maximum of one ~ 1 or a maximum of two>

180°F), as in the revised LCD discussion above. The proposed Condition A would read as follows:

"More than the maximum required charging pump(s) capable of injecting into RCS."

III The current modifying Required Action A.1 "Two charging pumps may be capable of injecting the during pump swap operation for:::; 15 minutes."

-10 to NL*12*0868 Basis for Proposed Change the LCO is already modified by a similar, but more detailed Note (including the applicable temperature of ::; 180°F), Required Action A.1 would not applicable while complying with the Note, as the LCO (as modified by the Note) is still met. Therefore, the Note modifying Required Action Ai is not needed and would be deleted.

The current Required Action A.1 "Initiate action to verify a maximum of one charging pump is capable of injecting into the RCS."

Similar to the changes proposed for Actions Condition A, Required Action Ai would revised to provide the appropriate Required Action for either charging pump condition (a maximum of one::; 180°F or a maximum of two

> 180°F) as stated in the revised LCO discussion above. The proposed Required Action Ai would state:

"Initiate action to verify s maximum required charging pump(s) capable of injecting into the RCS."

current SR 1 1 states:

"Verify a maximum one charging pump is capable of injecting into the The proposed change would SR 3.4.1 1 to be applicable when a maximum of one charging pump is capable of injection into the RCS

180°F). such, the proposed SR 1 would "Verify a maximum of one charging pump is capable of injecting into the when one or more RCS cold legs is: :; 180°F."

In order to address the proposed LCO condition of a maximum of two charging pumps capable of injecting into RCS (when all RCS cold leg temperatures are> 180"F), a new SR is proposed. The new SR would state:

"Verify a maximum of two charging pumps are ca[)aOile of injecting into the RCS when all cold legs are> 180"F."

The Frequency of proposed 3.4.1 would be the same as existing SR 3.4.12.1 (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ).

As a result of introducing a new SR, all the subsequent (existing SR 3.4.12.2 through SR 3.4.1 would renumbered accordingly.

E1-11 to NL-12-0868 for Proposed Change

3. Technical Analysis 3.1. Revised PTLR and LTOP Applicability Temperature.

The PTLR contains pressure and temperature related limits that are plant specific, vary with fluence, and can be revised without NRC approval provided that they are calculated using an NRC approved methodology.

FNP Specification 5.6.6 contains the requirements applicable to the FNP Unit 1 and Unit 2 including the methodology to determine the pressure and temperature limits. Specifically, FNP Specification 5.6.6.b currently states:

"The analytical methods used determine the RCS and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the NRC letters dated March 31, 1998 and April 3, 1998."

This TS license amendment request would change the NRC approved methodology currently cited in FNP Specification 5.6.6.b to the latest NRC approved methodology (Le., WCAP-14040-A, Rev. 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cool down Limit Curves," May 2004). WCAP-14040-A been approved by the NRC.

The proposed change in methodology is acceptable as it meets the requirement of FNP Specification 5.6.6 part b. Part b of Specification requires the NRC approved methodology used for determining the RCS pressure and temperature limits to in the TS. Therefore, consistent with requirements of Specification 5.6.6, any future changes to the FNP pressure and temperature limits contained in the PTLR will determined in accordance with this NRC approved methodology.

In addition, based on the revised FNP pressure and temperature limits (for EFPY), a new LTOP System applicability temperature was calculated. As a result, the applicability temperature would be revised from s; to s 275°F.

LCO 3.4.12 contains the TS requirements for the LTOP System. The LTOP controls pressure at low temperatures, so the integrity of the reactor coolant pressure boundary (RCPS) is not compromised by violating the pressure and temperature limits of 10 CFR 50, Appendix "Fracture Toughness Requirements." The reactor vessel is the limiting component for demonstrating such protection. The reactor vessel material is less tough at low temperatures than normal operating temperature. Therefore, RCS pressure is maintained low at low temperatures and is increased only as temperature is increased. The LTOP applicability temperature is limiting RCS cold temperature below which the reactor vessel may damage from a cold overpressure event. Fracture mechanics analyses establish the applicability temperature. such, the LTOP applicability temperature is the temperature below which the LTOP System is required to operable.

E1-12 to NL-12-0868 Basis for Proposed Change In addition to identifying the applicability of the LTOP System in Mode 4, the LTOP System applicability temperature is used in the FNP TS identify the applicability of LCO 3.4.10 in Mode 4. Above the LTOP System applicability temperature, the Pressurizer safety are required operable to provide the required overpressure protection. Additionally, the LTOP applicability temperature is used in the FNP TS for a low temperature limit when starting an RCS pump in LCO 3.4.6 LCO 3.4.7. The temperature limit for starting an RCS pump prevents a heat input transient due temperature asymmetry within the RCS or between the RCS and steam generators. The heat input transient that may result from start of an RCS pump could cause a cold overpressure event. As such, the LTOP applicability temperature is included in several FNP TS.

revised applicability temperatures for Units 1 and 2 were calculated in accordance with ASME N-641 , "Alternative Pressure-Temperature Relationship and Low Overpressure Protection System RequirementsSection XI, Division 1," January 17, 2000 (Reference 12).

ASME Code Case N-641 presents alternative procedures for calculating pressure temperature relationships and LTOP System effective temperatures and allowable pressures. procedures provided in Code Case N-641 take into account alternative fracture toughness properties, circumferential and axial reference flaws, and plant-specific LTOP applicability temperature calculations.

ASME Code Case N-641 was first accepted by the NRC without conditions in Regulatory 1.147, "Inservice Inspection Code Acceptability, ASME Section XI, Division 1, Revision 1 June 2003," (Reference 13). ASME Code Case N-641 continues to be identified as accepted without conditions in the current version of Regulatory Guide 1.147, Revision 16, October 2010, (Reference 14). Regulatory Guide 1.147, Revision 1 is identified in 1 50.55a(b), "Standards approved for Incorporation by Reference," (Reference 15).

WCAP-14040-A, Revision 4, also utilizes the use of ASME Code Case N-641 for the calculation of the LTOP applicability temperature. such, proposed FNP Unit 1 and Unit 2 LTOP applicability temperatures were calculated in accordance with ASME Code N-641 , consistent with the Simplified equation provided in Section WCAP-14040-A for a three loop plant. The calculated LTOP applicability temperatures for Units 1 and 2 for the new pressure and temperature limit curves are (Unit 1) and 251°F (Unit revised LTOP System applicability temperature value that would be specified in the PTLR includes an allowance for temperature measurement uncertainty, which is OF. Accounting for instrument uncertainty ensures that the LTOP System is operable at the temperatures where it is required to be operable. The uncertainty used for the LTOP System applicability temperature is the same measurement uncertainty previously used for this purpose and approved in NRC Letter, "Joseph M. Farley Nuclear Plant, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report" dated March 31, 1998 (Safety Evaluation Section 3.3.1 in Reference In addition, consistent with the current FNP and in order to minimize unit differences, the highest (most limiting) calculated temperature for FNP Unit 2 was chosen for both units. Thus, proposed LTOP applicability temperature for

-13 to NL-12-0868 Basis for Proposed Change inclusion in the Unit 1 and Unit 2 PTLRs is ::; 275°F (i.e., 251°F calculated Unit 2 temperature + 21°F measurement uncertainty = 272°F conservatively rounded up to 275°F).

The relocation of the revised LTOP applicability temperature from the TS to the PTLR is discussed below in Section 3.2.

3.2. Relocation of LTOP Applicability Temperature to the PTLR (TSTF-233-A)

In many LTOP System designs, a plant specific arming feature exists that places the LTOP system in operation at the applicability temperature. The arming design feature is typically associated with plants that use power operated relief valves in the LTOP System. In plants with the arming design feature, the LTOP applicability temperature is often referred to as the "LTOP arming temperature."

The FNP LTOP System design does not include power operated relief valves and the associated "arming feature." The FNP LTOP System pressure relief capability consists of two redundant RHR relief valves or a depressurized RCS with an RCS vent of sufficient size. Therefore, the changes contained in TSTF 233 are applicable to the FNP LTOP applicability temperature and not the plant specific term of "LTOP arming temperature."

The specific value for the LTOP applicability temperature is reactor vessel plant specific and varies with vessel fluence. The use of a plant specific value in the TS, which will require periodic amendments, is not consistent with the PTLR philosophy. Reference to the PTLR for other plant specific values (e.g., LCO 3.4.3, "RCS Pressure and Temperature (PfT) limits") has been found to be acceptable, and results in simplifying the revision process when the values change with reactor fluence. Similar to the PfT limits, periodic updates to the LTOP applicability temperature can also be made without going through the license amendment process, as the methodology used to determine the limiting temperature is controlled by TS and requires NRC approval for changes. WCAP 14040-A, Revision 4, is the NRC approved methodology being proposed in the FNP TS Section 5.6.6.b, for the calculation of the LTOP applicability temperature.

As such, future changes to the LTOP applicability temperature will be made in accordance with the NRC approved methodology specified in the TS.

Based on the discussion above, the implementation of TSTF-233 in the TS for FNP Units 1 and 2 is acceptable.

3.3. Revision of PTLR Definition and References in the PTLR Report (TSTF-419 A)

TSTF-419-A, "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR," Revision 0 was approved by the NRC by letter dated March 21, 2002.

The changes proposed in TSTF-419 have subsequently been incorporated into NUREG-1431.

TSTF-419 contains two changes, as discussed in Section 2.3 above. The only change being proposed for the FNP TS from TSTF-419 is the change that eliminates the duplication of TS LCOs identified in the PTLR definition and in Section 5.6.6 of the TS. The TS LCOs identified in these two TS locations E1-14 to I\IL-12-0868 Basis for Proposed Change include LCO 3.4.3 to the heatup and cooldown curves moved to the PTLR and LCO 3.4.12 to address the LTOP System temperature moved to the PTLR. This change deletes the TS referenced in the PTLR definition while leaving the TS LCOs referenced in Section 5.6.6 of the TS. The proposed change is administrative in nature and serves to eliminate duplication of information in the In addition, the proposed change was approved in 419. Therefore, the proposed change the FI\IP Unit 1 and 2 is acceptable.

3.4. Revision of TS 3.4.12 to Incorporate Requirement for Two Charging Pumps.

The FNP LTOP System pressure relief capability consists of two redundant RHR relief valves or a with a RCS vent of sufficient One RHR relief valve or the open RCS vent is sufficient to terminate an increasing pressure event. During cooldown operation, the LTOP System must operable to mitigate overpressure transients starting at an RCS cold leg temperature equal to 3.4.12 applicabiiity temperature in Mode 4 until the reactor vessel head is in Mode 6. Once reactor vessel head is raised, a sufficient RCS vent is created to mitigate any overpressure event transient. During heatup operation, the LTOP System is required operable the reactor vessel head is seated on the vessel up to when all RCS cold leg temperatures exceed the LTOP system applicability temperature. The current RHR suction relief valve setpolnt of .:s 450 psig and RCS vent of 2.85 square inches remain valid for new 54 pressure and temperature limit curves and the LTOP System applicability temperature.

current TS 3.4.12 LCO only a single charging pump capable of injecting into the RCS when one or more RCS cold leg temperatures is:S: 180°F. it is silent regarding the condition when all RCS cold leg temperatures are> 180°F but still below the TS 3.4.12 applicability temperature.

In addition, current LCO is modified by Note 1 that states:

requirement to have only one charging pump capable of injecting into the RCS is only applicable when one or more of the RCS cold legs is

180°F;"

The current 3.4.12 Bases explains Note 1 as:

"Note 1 that the to have only one charging pump capable of injecting into the RCS is only applicable when one or more of the RCS cold legs is s 180°F. This Note permits more than one charging pump to capable o'f injecting into the RCS in MODE 4 at temperatures> 180°F and specifies that the charging pump surveillance requirement need only be performed at temperatures :s: 180"F."

Instead relying on the explanation of Note 1 in Bases, proposed change would modify the TS 3.4.12 LCO and to clarify the requirements as follows:

" Adding an LCO requirement for a maximum of two charging pumps capable of injecting into the RCS when all cold temperatures are> 180°F to E1-15 to NL-12-0868 Basis for Proposed Change complement the existing requirement for a single charging pump capable of injecting into the RCS when any cold leg temperature is s 180°F,

.. Modifying existing 12.1, which verifies a maximum of one charging pump is capable of injecting into the RCS, to clarify that this is required when one or more RCS cold leg temperatures is "S 1

.. Adding new SR 12.2, which would verify a maximum of two charging pumps are capable of injecting into RCS when all the cold leg temperatures are > 180°F with a Frequency in accordance with the Surveillance Frequency Control Program (the same as SR 12.1).

The proposed change also includes changes to the 3.4.12 Actions, and to accommodate clarifications described above. All the related changes are described in detail in Section above.

The applicable aspect of the FNP LTOP System design (I.e., the RHR relief capacity) is discussed in the for TS 3.4.12 as follows:

"During LTOP MODES, the RHR System is operated for decay heat removal and low pressure letdown control. Therefore, the RHR suction isolation valves are open in piping from the hot legs to the inlets of the RHR pumps. While are open and the RHR suction valves are open, the RHR suction relief valves are exposed to the RCS and are able to relieve pressure transients in the RCS.

The RHR suction isolation valves and the RHR suction valves must open to make RHR suction valves OPERABLE for RCS overpressure mitigation. The RHR suction relief valves are spring loaded, bellows type water relief valves with pressure tolerances and accumulation limits established by Section III of the American Society of Mechanical Engineers (ASME) Code (Ref. 3) for Class 2 relief valves. Each valve capacity to mitigate over pressurization in worst case of inadvertent startup of three charging pumps injecting into a solid RCS."

The applicable transient analysis for the LTOP System is discussed in FI\lP UFSAR "Pressure Transient Analyses" (Reference 16).

This UFSAR section the high-head safety injection pumps whiCh, for FNP design, are the same as the charging pumps. Regarding the capability the LTOP System to mitigate an overpressure tranSient due to the inadvertent operation of charging pumps, FNP UFSAR Section 5.2.2.4.3 Section III, Appendix establishes guidelines for RCS pressure during low temperature operation (~350°F). The relief system discussed in paragraph 1 serves to mitigate overpressure excursions to within these allowable limits. The worst-case mass input event was assumed to be the inadvertent operation of three high-head safety injection pumps with a maximum total flowrate of 1000 gal/min at 0 psig backpressure at RCS temperatures ~ 180°F. Due to Technical Specification restrictions that allow only one operable charging pump at temperatures < 180°F, worst

-16 to NL-12-0868 Basis for Proposed Change case mass injection is limited to the start of a single charging pump at temperatures < 180°F."

Although the analysis described in Bases and UFSAR (quoted above) conservatively assumes the capacity of three charging pumps, the plant design limits the number of charging pumps that can be in service one time to two.

Based on the discussions above, the proposed change provides a clarification of the TS 3.4.12 requirements that is consistent with the intent of existing Bases, and the FNP design. It preserves the applicable FNP UFSAR pressure transient analysis for the LTOP System. Therefore, the proposed change will not adversely affect the FNP LTOP System capability to perform required safety function.

-17 to 2~0868 for Proposed Change

4. Regulatory Safety Analysis 4.1. Significant Hazards Consideration proposed Amendment would the J. M. Nuclear Plant (FNP)

Technical Specifications (TS) based on the implementation of new 54 Effective Full Power Years pressure and temperature limit curves. The proposed changes would also revise the TS 3.4.1 "Low Temperature Overpressure Protection (LTOP) System," Mode 4 applicability temperature and the methodology used to determine the RCS pressure and temperature limits identified in Specification "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)."

In addition, the proposed amendment would relocate Low Temperature Overpressure Protection (LTOP) applicability temperature from FNP to the The relocation of the LTOP System applicability temperature is based on NRC approved changes to 431, "Standard Technical Specifications

- Westinghouse Plants," (i.e., Task (TSTF) Traveler number 233~A, "Relocate LTOP Arming Temperature PTLR," Revision 0).

relocation of the FNP LTOP System applicability temperature to the PTLR would the following

definition of PTLR in Section 1.1, "Definitions,"

TS "RCS Loops - MODE 4,"

3.4.7, "RCS Loops MODE 5, Loops Filled,"

TS 10, Safety Valves,"

.. 3.4.12, "Low Temperature Overpressure Protection (LTOP)

.. Specification 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature (PTLR)"

The proposed amendment would also the definition of the PTLR in TS Section 1.1, "Definitions," and Specification 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," to eliminate redundant references to the applicable This change is consistent with NRC approved changes to NUREG~1431 (I.e., "Revise Definition and References in 5.6.6, RCS PTLR," Revision 0).

Additionally, the proposed amendment would revise 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," to more clearly state the requirement for a maximum of two charging pumps being capable of injecting into the to preserve the applicable analysis.

As required by 10 CFR 50.91 (a), Southern Nuclear Operating Company (SNC) has evaluated the proposed changes to the FNP TS using the criteria in 10 50.92 and has determined that the proposed changes do not involve a significant consideration. An analysis of the of no significant hazards consideration is below:

E1-18 NL-12-0868 Basis for Proposed Change

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

Response: No proposed amendment involves changes to the TS requirements to incorporate new and limit curves were determined with an NRC approved methodology for the LTOP system, as well as incorporating that methodology into the pressure and temperature limit curves preserve the integrity of the reactor The LTOP System provides overpressure protection during operation low RCS temperatures. In addition, amendment proposes to adopt the NRC approved and TSTF-419-A Adoption of these will relocate the LTOP applicability temperature from TS to the PTLR and will eliminate redundant references in Sections 1.1 and 5.6.6 of the TS. Lastly, the proposed change includes clarifications to the LTOP System TS requirements that are consistent with the FNP design and n~""(>"\1 the applicable safety analyses. The proposed changes are based on NRC approved methods, and NRC approved changes to the Standard for Westinghouse Plants.

The proposed change to the does not affect the initiators of any analyzed accident. In addition, operation in accordance with the proposed TS change ensures that previously evaluated accidents will continue to be mitigated as analyzed. Thus, the proposed change does not adversely affect the design function or operation of any structures, systems, and components important to safety.

Therefore, it is concluded that the proposed not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No The proposed amendment involves changes to the TS requirements to incorporate new pressure and limit curves that were determined with an NRC approved methodology for the LTOP system, as well as incorporating that methodology into TS. The and temperature limit curves the integrity of the reactor vessel. The LTOP System provides overpressure protection during operation at low temperatures. In addition, this amendment proposes to adopt the NRC approved TSTF-233-A and TSTF-419-A Adoption of these TSTFs will relocate the LTOP applicability temperature from the to the PTLR and will eliminate redundant references in Sections 1.1 and of the TS. Lastly, the proposed change includes clarifications to the System TS requirements that are consistent with the FNP design and the applicable safety analyses. The proposed changes are E1-19 to NL-12-0868 Basis for Proposed Change based on NRC approved methods and NRC approved changes to the Standard TS for Westinghouse Plants.

proposed change does not involve a physical alteration of the plant (no new or different type of equipment will installed). The proposed change not create any new failure modes for existing equipment or any new limiting single failures. Additionally the proposed change does not involve a change in the methods governing normal plant operation and all safety functions will continue to perform as previously assumed in accident analyses. The pressure and temperature limit curves will continue to the integrity of the reactor vessel. The System will continue to ensure that the appropriate fracture toughness margins are maintained to protect against reactor vessel failure during low temperature operation. Thus, the proposed change does not adversely affect the design function or operation of any structures, systems, and components important to safety.

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

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

Response: No proposed amendment involves to the requirements to incorporate new pressure and temperature limit curves that were determined with an NRC approved methodology for the LTOP system, as well as incorporating that methodology into the The pressure and temperature limit curves preserve the integrity of reactor \J"'C;;~<:!"'I LTOP System provides overpressure protection during operation at low RCS temperatures. In addition, this amendment proposes to adopt the NRC approved and TSTF-419-A. Adoption of these will relocate the LTOP applicabiiity temperature from TS to the and will eliminate redundant references in Sections 1.1 and 5.6.6 of the TS. Lastly, the proposed change includes clarifications to the LTOP System TS requirements that are consistent with the FNP design and preserve the applicable safety analyses. The proposed changes are based on NRC approved methods and NRC approved changes to Standard for Westinghouse Plants.

The proposed change will not adversely affect the operation of plant equipment or function equipment in the accident analysis.

pressure-temperature limit curves and System applicability temperature have determined in accordance with NRC approved methodologies. The proposed changes to the LTOP System requirements remain consistent with applicable LTOP System design, and preserve applicable safety analYSis assumptions. Additionally, no are made to the System function as assumed in the applicable safety analysis.

-20 to NL-12-0868 for Proposed Change Therefore, it is concluded that proposed change does not involve a significant reduction in a margin of safety.

Based upon the above analysis, SNC concludes that the proposed amendment does not involve a significant hazards consideration, under standards forth in 10 50.92(c), "Issuance of Amendment," and accordingly, a finding of "no significant hazards consideration" is justified.

4.2. Applicable Regulatory Requirements/Criteria The proposed amendment involves changes to requirements for the pressure temperature limits, and LTOP System. LCO 3.4.3, "RCS and Temperature (PfT) " contains TS requirements for the pressure limits. pressure and temperature limit curves preserve the vessel. LCO 3.4.1 "Low Temperature Overpressure Protection (LTOP) " contains the requirements for the LTOP System.

The LTOP System controls RCS at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is not compromised by violating pressure temperature limits of 10 CFR 50, Appendix G, "Fracture Toughness Requirements." reactor is the limiting RCPB component for demonstrating protection. The reactor vessel material is tough at low temperatures than at normal operating temperature. pressure, therefore, is maintained low at low temperatures and is only as temperature is increased. The following regulatory criteria are applicable to proposed amendment to PfT limits and LTOP requirements:

.. 10 CFR 50.55a(b), "Standards approved for incorporation by reference,"

which states in part:

"(b) Standards approved for incorporation by reference. and components of boiling and water cooled nuclear power reactors meet the requirements of the following standards referenced in paragraphs (b}(1), (b)(2), (b)(3), (b)(4), (b)(5), and (b)(6) of section: The ASME Boiler and Pressure Vessel Section III, Division 1 (excluding Non-mandatory Appendices), and XI, Division 1; the ASME Code for Operation and Maintenance of Nuclear Power NRC Regulatory Guide (RG) 1 Revision "DeSign, Fabrication, and Code Acceptability, ASME Section III" (July 2010), RG 1.1 Revision 1 "Inservice Inspection Code Acceptability, ASME Section XI, Division 1" (July 2010), and ... "

.. 10 50.61, "Fracture toughness requirements for protection against pressurized thermal events," provides protection an event or transient in pressurized water (PWRs) that could cause severe overcooling (thermal shock) concurrent with or followed by significant in the vessel.

.. 10 Appendix A, "General Design for Nuclear Power Plants," Criterion 31, "Fracture prevention of coolant pressure boundary," which states:

-21 to NL-12-0868 Basis for Proposed Change "The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a non brittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws."

10 CFR 50, Appendix G, "Fracture Toughness Requirements," which states in part:

"This appendix specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety 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 ASME Code forms the basis for the requirements of this appendix.

"ASME Code" means the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. If no section is specified, the reference is to Section III, Division 1, "Rules for Construction of Nuclear Power Plant Components." "Section XI" meansSection XI, Division 1, "Rules for Inservice Inspection of Nuclear Power Plant Components." If no edition or addenda are specified, the ASME Code edition and addenda and any limitations and modifications thereof, which are specified in § 50.55a, are applicable.

The sections, editions and addenda of the ASME Boiler and Pressure Vessel Code specified in § 50.55a have been approved for incorporation by reference by the Director of the Federal Register. A notice of any changes made to the material incorporated by reference will be published in the Federal Register."

Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 16, October 2010, which states in part:

"General Design Criterion (GDC) 1, "Quality Standards and Records," of Appendix A, "General Design Criteria for Nuclear Power Plants," to Title 10, Part 50, of the Code of Federal Regulations (10 CFR Part 50),

"Domestic licensing of Production and Utilization Facilities" (Ref. 1),

requires, in part, that structures, systems, and components important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, E1-22 NL-12-0868 for Proposed Change Criterion 1 requires that they be identified and evaluated determine their applicability, adequacy, and sufficiency and be supplemented or modified as necessary to ensure a quality product in keeping with the required function."

The proposed amendment is acceptable, since the PfT limits were determined using NRC approved methodologies, and the design and function of LTOP system and associated applicability are maintained consistent with the assumptions the applicable safety analyses, the design basis of unit, and the FI\lP compliance with regulatory criteria cited above.

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

5. Environmental Consideration A review determined that the proposed amendment would change a requirement with respect to installation or use a facility component located within restricted area, as defined in 10 20, and would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant consideration, (ii) a significant change in types or significant increase in the amounts of any effluents may be offsite, or (iii) a significant in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR

.22(c)(9). Therefore, pursuant to 10 CFR 51 no environmental impact statement or environmental assessment need be prepared in connection with proposed amendment.

6. References

1. WCAP-14040-A, "Methodology Used to Develop Cold Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,"

Revision May WCAP-1 22-NP, "J. M. Farley Unit 1 Heatup and Cooldown Limit for Normal Operation," Revision 0, October 2009.

3. WCAP-17123-NP, "J. M. Farley Unit 2 Heatup and Cooldown Limit for Normal Operation," Revision 1, July 2004.

4. NUREG-1431, Standard Technical Specifications Westinghouse Plants, Revision 4.0.

5. TSTF-233-A, "Relocate LTOP Arming Temperature to PTLR," Revision 0, July 16, 1998.

-23 to NL-12-0868 for Proposed Change

6. NRC to James (NEI) from William D. Beckner, dated July 16, 1998.

9-A, PTLR Definition and 5.6.6, PTLR," March 21, 2002.

8. NRC to Anthony R. Pietrangelo (NEI) from William D. Beckner, dated

,2002 (ML020800488).

9. NRC letter, to N. Morey (SNC) from Herbert N. Berkow, "Joseph M. Farley Nuclear Plant, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Report," dated March 31,1998 (TAC Nos. M99338 and M99339).

10. NRC Letter, to D. N. Morey (SNC) from Jacob I. Zimmerman, "Joseph M.

Farley Nuclear Plant, Units 1 and 2, Correction to Acceptance Letter for Referencing of Pressure Temperature Limits Report," dated April 1998 (TAC Nos. M99338 and M99339).

11. NRC letter to TSTF from John R. Jolicoeur, "Implementation of Travelers TSTF-363, Revision 0, "Revise Topical Report References in ITS COLR [Core Operating Limits Report]," Revision 1, "Relocation Of LTOP [Low Temperature Overpressure Protection] Enable Temperature and PORV [Power-Operated Relief Valve] Lift to the [Pressure-Temperature Limits Report]," and TSTF-419, Revision 0, "Revise PTLR Definition and References in [Improved Standard Technical Specification] 5.6.6, RCS [Reactor Coolant System] PTLR," dated August 4, 1.

1 ASME Code Case N-641 , "Alternative Pressure-Temperature Relationship Low Temperature Overpressure Protection System RequirementsSection XI, Division 1 ," January 1 2000.

1 Regulatory Guide 1.147, "Inservice Inspection Code Acceptability, ASME Section XI, Division 1," Revision 13, June 2003.

14. Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 16, October 2010.

15. 10 CFR 50.55a(b), "Standards approved for Incorporation by ""OTorOnl"'O 1 FNP Unit 1 and Unit 2 UFSAR, Section "Pressure Transient Analyses," Revision 24, March 2012.

E1

Joseph M. Farley Nuclear Plant Request to Revise Technical Specifications Associated with Low Temperature Overpressure Protection System and the Pressure and Temperature Limits Report Enclosure 2 Technical Specifications and Bases Markup Pages

and the Low Temperature Definitions Overpressure Protection 1.1 System applicability temperature 1.1 I PRESSURE AND The PTLR is the unit speci c document that provides the TEMPERATURE LIMITS reactor vessel pressure an limits, including (PTLR) heatup and cooldown for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification Pial it operation \tw'ithin these cperatiflg limits is addressed 11'1 LCO 3.4.3, "RCS Pressure and Telilpelature (PIT) limits."

QUADRANT POWER TILT QPTR shall be the ratio of the maximum excore RATIO (QPTR) detector calibrated output to the of the upper excore detector outputs, or the ratio of the maximum lower excore detector calibrated output to the average the lower excore detector calibrated outputs. whichever is greater.

RATED THERMAL POWER RTP be a total reactor core heat transfer rate to the (RTP) reactor coolant of 2175 MWt.

REACTOR TRIP The RTS RESPONSE TIME shall be that time interval from SYSTEM (RTS) RESPONSE when the monitored parameter exceeds its RTS trip setpoint TIME at the channel sensor until loss of stationary gripper coil response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be suberitical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted for the single RCCA of nll'1nA':::J reactivity worth. which is to be fully withdrawn.

With any RCCA not capable of being fully the worth of the RCCA must be accounted for in the determination of SDM; and Farley Units 1 and 2 1.1-5 Amendment No. t.ffl' (Unit 1)

Amendment No. t>tt (Unit 2)

RCS Loops- 4 3.4.6 3.4 REACTOR COOLANT SYSTEM (RCS) the Low Temperature Overpressure Protection 3.4.6 RCS Loops-MODE 4 (LTOP) System applicability temperature specified in the PTLR LCO 3.4.6 Two loops consisting of any combination of RCS loops "'..,,,...."" heat removal (RHR) loops shall be OPERABLE. and one loop shall operation.

1. All reactor coolant pumps (RCPs) and RHR pumps may not be in operation for::;; 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a. No operations are permitted that would cause reduction of the RCS boron concentration; and

b. Core outlet temperature is maintained at least 10°F below saturation temperature.

2. No RCP shall be started with any RCS leg temperature ::;;

a. The secondary side water temperature each steam generator (SG) is < 50°F above each of the RCS cold leg temperatures; or

b. The water volume is less than 770 cubic (24 %

of wide range, cold, pressurizer level indication).

APPLICABILITY: 4.

CONDITION REQUIRED ACTION COMPLETION TIME A. One RCS loop A.1 Initiate action to restore a Immediately inoperable. second loop to OPERABLE status.

TwoRHR Inoperable.

Units 1 and 2 3.4.6-1 Amendment No. f46' (Unit 1)

Amendment No. t9T (Unit

RCS Loops - MODE 5, Filled 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Loops MODE 5, Loops Filled 3.4.7 residual heat removal (RHR) loop shall be OPERABLE and in operation. and either:

a. One additional RHR shall be OPERABLE; or

b. The secondary side water level of at two steam generators (SGs) shall be 2: 75°10 (wide range).

1. The RHR pump of the loop in operation not be in operation for

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided

a. No operations are permitted that would cause reduction of the RCS boron concentration; and

b. Core outlet temperature is maintained at least 1Q°F below saturation temperature.

2. One required RHR loop may be inoperable for::; 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided that the other RHR is OPERABLE and!n

3. No reactor coolant pump shall be started with one or more cold leg temperatures S unless:

The secondary side water temperature of each SG is < 50°F above each of the leg or

b. The pressurizer water volume is less than 770 cubic feet (24%

of wide range, cold, level indication).

4. All RHR loops may be removed from operation during planned heatup to MODE 4 when at least one RCS loop is in operation.

5. The number of operating Reactor Coolant Pumps is limited to one at RCS temperatures < 110"F with the exception that a second pump may be started for the purpose of maintaining continuous flow while taking the operating pump out of service.

Units 1 and 2 3.4.7-1 Amendment No. t4T- (Unit 1)

Amendment No. 1"51t (Unit 2)

Pressurizer Safety Valves 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Pressurizer Safety Valves LCO 3.4.10 Three pressurizer safety valves shall be OPERABLE with lift settings

~ 2460 psig and ~ 2510 psig. the Low Temperature Overpressure Protection (L TOP) System applicability temperature specified in the PTLR APPLICABILITY: MODES 1, 2, and 3, MODE 4 with all RCS cold leg temperatures>

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

The lift settings are not required to be within the LCO limits during MODES 3 and 4 for the purpose of setting the pressurizer safety valves under ambient (hot) conditions. This exception is allowed for 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months following entry into MODE 3 provided a preliminary cold setting was made prior to heatup.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pressurizer safety A.1 Restore valve to 15 minutes valve inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND OR B.2 Be in MODE 4 with any 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Res cold leg Two or more pressurizer temperatures s;.~

safety valves inoperable.

/

I~he LTOP System applicability temperature speCified in I the PTLR Farley Units 1 and 2 3.4.10-1 Amendment No. 1-4& (U nit 1)

Amendment No. 43T- (Unit 2)

LTOP System 3.4.12 3.4 REACTOR COOLANT SYSTEM when one or more of the RCS cold legs is S 180°F and a maximum two charging pumps capable of 3.4.12 Low Temperature Overpressure Protection (lTOP) System injecting into the RCS when all of the RCS cold legs are:> 180°F LCO 3.4.12 An LTOP System shall be OPERABLE th a maximum of one charging pump capable of injecting into the RCS nd the accumulators isolated and either a or b below.

NOTES-------

1. With one or mora of the RCS cold a. Two residual heat removal (RHR) suction valves with setpoints legs S 180°F, two s450 2.

The RCS depressurized and an RCS vent of ? 2.85 square inches.

MODE 4 when the temperature of one or more RCS cold Ie s is s ':t.,,~~~

MODE 5, the LTOP System MODE 6 when the reactor vessel head is on. applicability temperature specified in the PTLR 1.

c argmg pumps may e capa eo Injectmg Into t e unng pump swap operations for a period of no more than 15 minutes provided that the is in a non-water solid condition and both RHR relief valves are or the RCS is vented via an opening of no less than 5.7 inches in area.

Accumulator isolation is only required when accumulator pressure is than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed by PIT limit curves provided in the PTlR.

Farley Units 1 and 2 3.4.12-1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

LTOP System 3.4.12 ACTIONS

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

LCO 3.0.4b is not applicable when entering MODE 4.

CONDITION REQUIRED ACTION COMPLETION TIME A. ::r'NO or more charging A.1 ----------- NOTE----

pump apable of T I injecti RCS.

into the (s)

IMore than the maximum requiredl B. An accumulator not B.1 Isolate affected 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months isolated when the accumulator.

accumulator pressure is the L TOP System greater than or equal to applicability temperature the maximum RCS pressure for existing cold specified in the PTLR leg temperature allowed in the PTLR.

C. Required Action and C.1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion temperature to >

Time of Condition B not OR met.

C.2 Depressurize affected 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months accumulator to less than the maximum RCS pressure for existing cold leg temperature allowed in the PTLR.

Farley Units 1 and 2 3.4.12-2 Amendment No. 1T&- (Unit 1)

Amendment No. 1 (Unit 2)

LTOP System 3.4.12 SURVEILLANCE FREQUENCY Verify a maximum of one charging pump is In accordance with the Surveillance Frequency Control Program 3.4.12x ~ Verify each accumulator is isolated. In accordance with the Frequency Control Program 2."5- Verify RHR suction isolation valves are open for each In accordance with RHR suction relief valve. the Surveillance Frequency Control Program SR 3.4.12:4-~

~ Only to LCO 3.4.12.b.

Verify RCS vent ~ 2.85 square open. In accordance with the Surveillance

. Frequency Control

. Program SR 3.4.12:5"' Verify each required RHR suction relief valve In accordance with

~setPoint. the Inservice Testing Program Program Farley Units 1 and 2 3.4.124 Amendment No. 485- (Unit 1)

Amendment No. -t6tt (Unit 2)

Reporting Requirements 5.6 5.6 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

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

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

5.6.6 LCO 3.4.3,

b. The analytical methods used to determine the RCS and temperature limits shall be those fHVIHVlIHn and approved by the NRC. specifically those aAs pril a, 1998.

c. The shall be provided to the NRC upon issuance for each reactor fluence period and for any revision or supplement thereto.

WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and 5.6.7 Heatup and Cooldown limit .. May 2004 If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures shall be reported within 30 days.

Reports on EDG failures shall include a description of the failures, underlying causes, corrective actions taken per the Emergency Diesel Generator Reliability Monitoring Program.

When a report is required by Condition B or F of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for ro.,cYlowonn the instrumentation channels of the Function to OPERABLE status.

Farley Units 1 and 2 5.6-5 Amendment No. -t&4- (Unit 1)

Amendment No. -t=f8- (Unit 2)

INSERT 1:

SR 3.4.12.2 Verify a maximum of two charging pumps are capable of In accordance with injecting into the RCS when all RCS cold legs are:> 180"F. the Surveillance Frequency Control I Program

RCS limits B 3.4.3 BASES SURVEILLANCE Verification that operation is within the PTLR limits is when RCS pressure and temperature conditions are undergoing planned changes. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Surveillance for heatup, cooldown. or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.

This is modified by a Note that only requires this SR to performed during system heatup. cooldown. and ISlH testing. No is given for criticality operations because lCO 3.4.2 contains a more restrictive requirement.

2. 10 CFR 50, Appendix G.

3. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix G.

4. ASTM E 1 July 1982.

5. 10 CFR 50, Appendix H.

6. Regulatory Guide 1.99, Revision 2, May 1988.

7. ASME, and Pressure Code,Section XI. Appendix E.

Farley Units 1 and 2 B 3.4.3-7 Revision

RCS Loops - MODE 4 B 3.4.6 LCO .... nn'<::.<::f of any combination of RCS loops and RHR loops. Anyone (continued) loop in operation provides enough flow to remove decay heat from the core with circulation. An additional is required to be OPERABLE to provide redundancy for removal.

Note 1 permits all RCPs or RHR pumps to not be in operation for

~ 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months The purpose of the Note is to permit tests that are deSigned to validate various accident analyses values.

One of the tests performed during the startup testing program is the validation of rod drop times during cold conditions, both with and without flow. The no flow test be performed in MODE 3, 4, or 5 and that the be stopped a short period of time.

Note permits the stopping of the pumps in order to perform this test and validate the assumed analysis values. If changes are made to the that would cause a change to the flow characteristics of the the input values must be by conducting test again. 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months time period is adequate to perform the test, and operating experience has shown that boron stratification is not a problem during this short period with no forced flow.

Utilization of Note 1 is permitted provided the following conditions are met along with any other conditions imposed by initial startup test procedures:

a. No operations are permitted that would dilute the boron concentration, therefore maintaining the to criticality.

Boron reduction is prohibited because a uniform concentration distribution throughout the RCS cannot be ensured when in natural circulation; and

b. Core outlet temperature is maintained at least 10 F below Q saturation temperature, so that no vapor bubble may and possibly cause a natural circulation flow obstruction.

side water temperature of each above each of the cold temperatures or that water volume is less than cub1c feet of wide range, cold, level indication) before the start of an RCP with RCS cold leg temperature :s o . This restraint is to prevent a low temperature overpressure vent due to a thermal transient when an RCP is started.

the LoW Temperature Overpressure Protection (LTOP) applicability temperature specified in PTLR Farley Units 1 2 B Revision t1"

RCS Loops MODE 5, Loops Filled B 3.4.7 BASES LCO distribution throughout the RCS cannot be ensured when in (continued) natural circulation; and

b. Core outlet temperature is maintained at least 10°F below saturation temperature, so that no vapor bubble may form and possibly cause a natural circulation flow obstruction.

Note 2 allows one RHR loop to be inoperable for a period of up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , provided that the other RHR loop is and in operation. This permits periodic surveillance tests to be performed on the inoperable loop during the only time when such testing is and Note 3 that the secondary water temperature of each be < 50°F above each of RCS cold leg temperatures or that the pressurizer water volume is than 770 cubic feet (24% of wide range, cold, pressurizer level indication) before the start of a reactor coolant pump (Rep) with an RCS cold leg temperature ~ 0 This restriction is to prevent a low temperature overpressure event to a thermal transient when an Rep is started.

Note 4 provides for an orderly transition from MODE 5 to MODE 4 during a planned heatup by permitting removal of RHR loops from operation when at least one loop is in operation. This Note provides for the transition to MODE 4 where an RCS loop is permitt to be in operation and replaces the ReS circulation function provided by the RHR loops.

Note 5 restricts the number of operating reactor coolant pumps at ReS temperatures less than 110°F. Only one reactor coolant pump is allowed to be in operation below 110°F (except during pump swap operations) consistent with the assumptions the prr Limits Curve.

RHR pumps are if they are capable of being powered and are able to provide flow if required. A SG can perform as a heat sink via natural circulation when it has an adequate water level and is OPERABLE.

the Low Temperature Overpressure Protection (LTOP) System applicability temperature speCified in the PTLR Units 1 and 2 B 3.4.7-3 Revision -24'"

Pressurizer Safety Valves B 3.4.10 B COOLANT (RCS) 83.4.10 Pressurizer Safety Valves BACKGROUND The pressurizer safety valves provide, in conjunction with the Reactor Protection System, overpressure protection for the RCS. The pressurizer safety valves are totally pop type, spring Ivau"" .. ,

valves with compensation. The safety valves are designed to prevent the system pressure from exceeding the system Safety Limit (SL), 2735 psig, which is 110% of the design pressure.

OI;:'....aLI;)1;: the safety valves are totally enclosed and self actuating. they are considered independent components. The relief capacity for each valve, 345,000 Ib/hr, is based on postulated overpressure transient conditions resulting from a complete loss of steam flow to the turbine.

This event results in the maximum rate into the which the minimum relief capacity for the valves. The discharge flow from the safety valves is directed to the pressurizer relief tank. This discharge flow is indicated by an increase in temperature downstream of the pressurizer safety valves or the Low Temperature increase in the pressurizer relief tank temperature or level.

Protection (LTOP) System applicability protection is required in MODES 1, 2, 3,4, and 5; temperature speCified in the wever, in MODE 4, with one or more cold leg temperatures o ,and MODE 5 and MODE 6 with the reactor vessel head on,

,PTLR overpressure protection is provided by operating procedures and by meeting the requirements of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System."

The upper lower limits are on the +/- 1% tolerance requirement 1) for lifting pressures above 1000 pSig. The lift setting is for ambient conditions associated with MODES 1, 2, and 3. This requires either that the valves be set hot or that a correlation between hot and cold settings be established.

The pressurizer safety valves are part of the primary success path and mitigate the effects of postulated accidents. OPERABILITY of the safety valves ensures that pressure will be limited to 110% of design pressure. The consequences of exceeding the Farley Units 1 and 2 B 3.4.10-1 Revision1t

Pressurizer Safety Valves B 3.4.10 BASES LCO is the reactor coolant pressure boundary (RCPB) SL of 110% of (continued) design pressure. Inoperability of one or more valves could result in exceeding the SL if a transient were to occur. The consequences of exceeding the ASME pressure limit could include damage to one or more RCS components, increased leakage. or additional stress analysis being required prior to resumption of reactor 0 eration.

when all RCS cold leg temperatures are>

APPLICABILITY In MODES 1 , and 3, and portions of MODE 4 aOOve the L TO temperature 0 , OPERABILITY of three valves is required because the combined capacity is required to keep reactor coolant pressure below 110% of its design value during certain accidents.

the LTOP System MODE 3 and portions of MODE 4 are conservatively included, applicability ough the listed accidents may not require the safety valves for temperature specified in prote . one or more the PTLR icable in MODE 4 when RCS cold leg temperatures are =s 0 or in MODE 5 because LTOP is provided.

Overpressure protection is not required in MODE 6 with reactor vessel head detensioned.

Normally demonstration of the safety valves' lift settings will occur during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.

The Note allows entry into MODES 3 and 4 with the lift settings outside the LCO limits. This permits testing and examination of the safety valves at high pressure and temperature near their normal operating range, but only after the valves have had a preliminary cold setting. The cold setting gives assurance that the valves are OPERABLE near their design condition. Only one valve at a time will be removed from service for testing. The 54 hour6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months exception is based on 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months outage time for each of the three valves. The 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months period is derived from operating experience that hot testing can be performed in this timeframe.

Farley Units 1 and 2 B 3.4 .10-3 Revision "6

Pressurizer Safety Valves B 3.4.10 BASES ACTIONS With one pressurizer safety valve inoperable, restoration must take place within minutes. The Completion Time of 15 minutes rOT""""""

the importance maintaining the Overpressure Protection System. An inoperable safety valve coincident with an RCS overpressure event could challenge the integrity of the pressure boundary.

applicability temperature specified in the PTlR If the Required Action of A.1 cannot b et within the required Completion Time or if two or more p s safety valves are inoperable, plant must brou t t a MODE in which the requirement does not apply. To hie e this status, the plant must be brought to at least MODE 3 wit* 6 h urs and to MODE 4 with any RCS cold leg temperatures .s ., w thin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, b ed on operating experience, to the plant conditions fr m full power conditions in an orderly manner and without challen . g plant systems. With any RCS cold leg temperatures at or below ." overpressure protection is provided by the l TOP System. The change from MODE 1, 2, or 3 to MODE 4 reduces the RCS energy (core power and pressure), lowers the potential for large pressurizer insurges, and thereby removes the need for overpressure protection by three pressurizer safety valves.

SURVEILLANCE SR 3.4.10.1 REQUIREMENTS Pressurizer safety valves are to be tested in accordance with the requirements of Section XI of the ASME Code (Ref. 4), which provides the activities and Frequencies necessary to satisfy the SRs.

No additional requirements are specified.

The valve setpoint is +/- 1% for OPERABILITY.

Farley Units 1 and 2 B 3.4.10-4 Revision"*

LTOP System B 3.4.12 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.12 Low Temperature Overpressure Protection (LTOP) System BASES BACKGROUND LTOP System controls RCS pressure at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is not by violating the pressure and temperature (PIT) limits 10 CFR 50, Appendix G (Ref. 1). The reactor vessel is the limiting RCPB component for demonstrating such protection. This Technical Specification provides the maximum allowable actuation setpoints for the RHR rellef valves and the the maximum RCS pressure for the existing RCS cold temperature during cooldown.

shutdown, and heatup to meet the Ref r nc 1 re ire ts durin the LTOP MODES. the LTOP System The reactor material is normal operating temperature. As the vessel neutron exposure accumulates, the material toughness decreases and becomes less resistant to pressure stress at low temperatures (Ref. 2). RCS pressure, therefore, is maintained low at low temperatures and is in""A:::"~AI1 only as is increased.

The potential for vessel overpressurization is most acute when is water solid, occurring only while shutdown; a pressure fluctuation can occur more quickly than an operator can react to relieve the condition. Exceeding the PIT limits by a significant amount could cause brittle of the reactor vessel. LeO "RCS and Temperature (PIT) limits," administrative control of RCS pressure and temperature during heatup and

a. A maximum of one cooldown to prevent exceeding the PTLR limits.

charging pump of injecting into the RCS when ThiS LCO provides RCS overpressure protection by aving a one or more RCS cold leg minimum coolant input capability and haVing adeq te pressure relief temperatures are s 180"F: capacity. Limiting coolant input capability

y & *

b. A maximum of two charging pumps capable of injecting aooumulators. The pressure relief capacity requires either two into the RCS when all the RCS redundant RHR relief valves or a depressurized RCS and an RCS cold leg temperatures are vent sufficient size. One RHR relief valve or the open RCS vent is

>180"F;and the overpressure protection device that acts to terminate an event.

c. Isolating the accumulators.

Farley Units 1 and 2 83.4.12-1 Revision

LTOP System B3.4.12 BASES BACKGROUND (continued)

LTOP mass or heat input transient, and maintaining pressure below the PIT limits. The required vent capacity may be provided by one or more vent paths. The vent path(s) must be above the level of reactor LTOP System applicability coolant, so as not to RCS when temperature in the PTLR APPLICABLE analyses (Ref. 4) demonstrate that th reactor vessel is SAFETY ANALYSES ely protected against exceeding t 1 PIT limits. In r;;;"]

MODES 2. and 3, in MODE 4 with cold leg temperature~L!J exceeding ", the safety valves will prevent RCS With one or more cold leg pressure from exceeding the Reference 1 limits. l\t abeut 326"F and temperatures s the LTOP System ""'""'tl~w. overpressure prevention falls to two OPERABLE RHR relief applicabifity temperature or to a and a sufficient sized vent.

specified in the PTLR Each of these means has a limited overpressure relief capability.

The actual temperature at which the pressure in the PIT limit curve falls below the pressurizer safety valve setpoint increases as the reactor vessel material toughness due to neutron embrittlement. time the PTLR curves are the L TOP System must be to ensure its functional requirements can still be met using the RHR relief valve method or the depressurized and vented RCS condition.

The PTLR contains acceptance limits that define the LTOP requirements. Any to the must evaluated the Reference 4 analyses to determine the impact the LTOP acceptance limits.

Transients that are capable of overpressurizing the RCS are categorized as either mass or heat input transients. examples of which follow:

a. Inadvertent safety InU"'rTu,n or

b. Charginglletdown flow mismatch.

Farley Units 1 and 2 B Revision -e

LTOP System B 3.4.12 BASES APPLICABLE Heat Input Type Transients SAFETY ANALYSES (continued) a. Inadvertent actuation of

b. Loss of RHR cooling; or A maximum of one charging c. coolant pump (RCP) with temperature asymmetry p of injecting into within the RCS or between the RCS and steam generators.

e RCS when one or more RCS cold leg temperatures are The following are required during the LTOP MODES to ensure that s 180°F and a maximum of two mass and heat input transients do not occur. which of the LTOP charging pumps capable of protection means cannot handle:

injecting into the RCS when all the leg Rendering all but one oharging pump inoapable of injootion; are >180°F.

b. Deactivating the accumulator discharge isolation valves in their closed positions; and In the Reference 4 analyses, the worst case mass input c. Disallowing start of an RCP if secondary temperature is more than event was assumed to be 50"F above primary temperature in anyone loop except as inadvertent operation of three provided for in LCO 3.4.6, "RCS - MODE 4," and high-head safety injection LCO "RCS Loops-MODE 5, Loops Filled."

pumps charging with a maximum total flowrate of 1000 gal/min at 0 psig maintain RCS pressure bele'.... limits hen enly one oharging pump is at RCS actuated. Thl:ls. the LCO allows only one Gharging pump OPI!!RA8bl!!

~ 180°F. The Gti1FlfKHfle-t::-H:.H-'-IM\:;~~ Since one RHR relief valve has not been analysis conservatively demonstrated to be able to handle the need from assumes the operation of three accumulator injection, when RCS temperature is low, the LCO also charging pumps although the requires the accumulators isolated when accumulator pressure is plant design limits the total greater than or equal to the maximum RCS pressure for the eXisting number of operating charging RCS cold temperature allowed in the PTLR.

pumps to two pumps at a time.

Additionally. Reference 4 The isolated accumulators must have their discharge valves closed states that due to the and the valve power supply breakers fixed in their open ,..-;:----:-,---,

Technical Specification restrictions that allow only one Fracture mechanics analyses establishe4 the temperature charging pump capable of injecting into the RCS at RCS temperatures <180°F. the The of a small break loss of coolant accident (LOCA) worst case mass injection is in LTOP MODE 4 conform to 10 CFR 50.46 and 10 CFR 50, limited to the start of a single Appendix K (Refs. 5 and 6), requirements by having a maximum of charging pump. one charging pump Farley Units 1 and 2 B 3.4.12-4 Revisioni7

LTOP ....\/." ........

B 3.4.12 BASES LCO This that L TOP is OPERABLE. LTOP System is OPERABLE when the minimum coolant input and pressure

a. A maximum of one relief capabilities are OPERABLE. Violation of this LCO could to charging pump """~JaUlv loss of low temperature overpressure mitigation and violation of of injecting into the RCS the Reference 1 limits as a result of an operational T"<>in""J"~~'frill;:;;;;;;:;;;:---]

when one or more RCS cold temperatures To limit the coolant input capability, the LCO are s 180"F; ~9F*~*IFij:f-l*:IFfII:.-s;:.tI*II':}H:t-eHfltet;Q*~-lAiEHl!'-l&-~~af'H~ accumulator

b. A maximum of two discharge isolation valves closed and immobilized when accumulator charging pumps pressure is greater than or equal to the maximum RCS pressure for of injecting into the the existing RCS cold leg temperature allowed in the PTLR.

when all the cold leg temperatures are The the that provide low overpressure

>180"F;and mitigation through pressure relief are:

c. All

a. Two ............ "'"'"" .... RHR suction relief valves; or An RHR suction valve is OPERABLE for LTOP when its RHR suction isolation valve and its RHR suction valve are open, its setpoint is :s 450 and testing proven its ability to open at this setpoint.

b. A depressurized RCS an RCS vent.

An vent is OPERABLE when open with an area of ~ 2.85 square inches.

prevention is of vessel head is on fuUy seated on the reactor vessel flange, with or without studs). The pressurizer safety valves provide overpressure protection that meets the Reference 1 prr limits 0

When the reactor vessel head is raised, such that a total ven rea of ~ 2.85 square inches is created. seated on blocks providing an equivalent vent area, or off, overpressurization cannot occur.

when all the RCS cold leg temperatures are> the System applicability in the PTLR Farley Units 1 and 2 B 3.4.12-6 Revision -(t

when all the cold leg temperatures are>

System applicability temperature specified in the PTLR BASES

/

APPLICABILITY LCO provides the operational PfT limits all MODES.

(continued) LCO 3.4.10, *Pressurizer Safety Valves," req . es the OPERABILITY of the pressurizer safety valves that provide verpressure protection during MODES 1, 2. and 3, and MODE 4 0 Low temperature overpressure prevention is most critical during shutdown when the RCS is water solid, and a mass or heat input transient can cause a very rapid increase in RCS pressure with little or no time allowed for operator to mitigate the event.

soal. Particles in the RCS vlater May Gause '/loar on tho soal surfaces and IOS6 of seal injection pressure May Gause the seal Rot to fully reseat whon pressure is reapplied. Note 2 states that aOGuMtllatoF iselation is enly required *....he" tho aOGumulator pressure performed only under these pressure and tomperetuFO conditions.

Farley Units 1 and 2 B 3.4.1 Revision

LTOP System B 3.4.12 ACTIONS A Note prohibits the application of LCO 3.0.4b to an inoperable LTOP system when MODE 4. There is an risk associated with entering MODE 4 from MODE 5 with LTOP inoperable and the provisions of LCO 3.0.4b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment inoperable systems and components, should not be applied in this circumstance.

To immediately initiate action to restore restricted coolant input capability to the reflects the urgency of removing the RCS from this condition.

8.1. C.1. and C.2 unisolated accumulator requires isolation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This is only required when the accumulator pressure is at or more than the maximum RCS erature allowed b the PIT limit curves. the LTOP System aoolicabilitv temoerature soecified in the PTLR If isolation is needed and cannot e accomplished in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Required Action C.1 and Required Actio provide two options, either of which must be performed next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . By the RCS temperature an accumulator pressure of 600 650 psig cannot exceed the LTOP limits if the accumulators are fully injected. Depressurizing the accumulators below the LTOP limit from the PTLR also gives this protection.

The Completion Times are based on that these activities can be accomplished in these time and on engineering evaluations indicating that an event requiring is not likely in the allowed times.

Units 1 and 2 B 3.4.1 Revision~

LTOP System B 3.4.12 BASES the LTOP System applicability temperature specified in the ACTIONS D.1! D.2. and D.3 L.P....:..T=LRO-:--_

.:. . _ _ _ _ _ _.."..-_ _ _ _ _ _ _ _---.J (continued)

In MODE 4 when any RCS cold leg temperature is ~~, with one required RHR relief valve inoperable, the pressurizer level must be reduced to ~ 30% (cold calibrated) and a dedicated operator must be assigned for RCS pressure monitoring and control within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

These actions provide additional assurance that an RCS pressure transient will be rapidly identified and operator action taken to limit the transient. The RHR relief valve must be restored to OPERABLE status within a Completion Time of 7 days. Two RHR relief valves are required to provide low temperature overpressure mitigation while withstanding a single failure of an active component.

The 7 day Completion Time considers the facts that only one of the RHR relief valves is required to mitigate an overpressure transient, the actions taken to reduce pressurizer level and monitor RCS pressure, and that the likelihood of an active failure of the remaining valve path during this time period is very low.

The RCS must be depressurized and a vent must be established within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when:

a. Both required RHR relief valves are inoperable; or

b. A Required Action and associated Completion Time of Condition A, C, or D is not met; or

c. The L TOP System is inoperable for any reason other than Condition A, B, C, or D.

The vent must be sized ~ 2.85 square inches to ensure that the flow capacity is greater than that required for the worst case mass input transient reasonable during the applicable MODES. This action is needed to protect the RCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel.

The Completion Time considers the time required to place the plant in this Condition and the relatively low probability of an overpressure event during this time period due to increased operator awareness of administrative control reqU irements.

Farley Units 1 and 2 B 3.4.12-9 Revision~

LTOPSystem B 3.4.12 REQUIREMENTS To potential for a limiting the mass input capability, H-f~i(tffitHl'H*"ef1I&Elf'\efeHrte-Btif'fl&i9-verified capable of iMjeeting ifllo ttle Res end the eceuffil::Jlator

a. A maximum of one charging pump capable of The charging pumps are rendered incapable of injecting into the RCS injecting into the ReS through removing the power from the pumps by racking the breakers when one or more RCS out under administrative control. An alternate method of LTOP control cold are may be employed using at least two independent means to a s 180°F; pump start such that a failure or single action will not result in

b. A maximum of two an injection into the This may be accomplished through the Hot pumps capable Shutdown Panel Local/Remote and pump control switches being of injecting into the RCS placed in the Local and Stop positions, respectively. and at least one when all the cold leg valve in the discharge flow path being closed with the position of temperatures are >180°F; ..........,., ... ,......."'....t., controlled administratively.

and

c. The accumulator The Surveillance Frequency is controlled under the Surveillance discharge isolation valves Frequency Control Program.

are verified closed and locked out.

Each required RHR suction relief valve shall demonstrated OPERABLE by verifying its RHR suction isolation valves (8701A, 8701 B, 8702A and 8702B) are open. This Surveillance is only required to be performed if the RHR suction relief valve is being used to meet this LCO.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The RCS vent of 2: 2.85 square inches is proven OPERABLE by verifying its open condition.

The Surveillance Frequency is controlled under the Frequency Control Program.

Farley Units 1 and 2 B 3.4.12-10 Revision '52"'"

LTOP System B 3.4.12 REQUIREMENTS

''~V vent must only be to be OPERABLE.

Surveillance is required to be performed if the vent is being used to satisfy the pressure relief requirements of the LCO 3.4.12b.

RHR relief valves are verified OPERABLE by testing the relief setpoint. The setpoint verification ensures proper relief valve mechanical motion as well as verifying the setpoint. Testing is performed in accordance with the Inservice Testing which is based on the requirements of the ASME Code.Section XI (Ref. 7).

The RHR valve setpoints are verified in accordance with the Surveillance Frequency Control Program. Per the Inservice Testing Program. if the scheduled valve the relief setpoint 3% or the shall tested. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50. Appendix G.

2. Generic Letter 88-11.

3. ASME, and Code, III.

4. Chapter 5.2.2.4.

5. 10 CFR 50, Section 50.46.

6. 10 CFR 50. Appendix K.

7. ASME. Boiler and Pressure Vessel Code.Section XI.

Farley Units 1 and 2 B 3.4.12-11 Revision~

INSERT The is modified by two Note 1 allows for two charging pumps to be capable injecting into the ReS during pump swap operations, when one or more of the ReS cold is :s; 180°F, for a period of no more than 15 minutes provided that the ReS is in a non water solid condition and both RHR reHef valves are OPERABLE or the is vented via an opening of no than 5.7 square inches in area. A 5.7 square inch opening is equivalent to the throat size area of two RHR relief valves. This allows seal injection flow to be continually maintained. thus the for number one seal t1<>."<> ...,,,,, by on the and by preventing water from the seal. in the water cause wear on the seal and of seal injection pressure may cause the seal not to fully reseat when pressure is reapplied. Note 2 states that accumulator isolation is only required when the accumulator pressure is more than or at the maximum pressure for the existing temperature, as allowed by the PIT limit curves. This Note permits the accumulator discharge isolation valve Surveillance to performed only under these and temperature conditions.

Joseph M. Farley Nuclear Plant Request to Revise Technical Specifications Associated with the Low Temperature Overpressure Protection System and the Pressure and Temperature Limits Report Enclosure 3 Technical Specifications Clean Typed Pages

Definitions 1.1 1.1 Definitions AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel and temperature limits, including REPORT (PTLR) heatup and cooldown rates and Low Temperature Overpressure Protection System applicability temperature, for the current reactor vessel fluence period. These pressure and temperature limits shall determined for each period in accordance with Specification 5.6.6.

QUADRANT POWER TI QPTR shall be the ratio of the maximum upper excore RATIO (QPTR) detector calibrated output the average the upper excore detector calibrated outputs, or the ratio of maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER RTP shall be a total reactor core transfer to the (RTP) reactor coolant of MWt.

REACTOR TRI P The RTS RESPONSE TIME shall that time interval from SYSTEM (RTS) when the monitored parameter exceeds RTS trip setpoint TIME at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any of sequential, overlapping, or total steps so that entire response time is measured. In lieu of measurement, response time may verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn.

With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in determination of SDM; and Farley Units 1 and 2 1.1 Amendment No. (Unit 1)

Amendment No. (Unit

RCS Loops - MODE 4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Loops MODE 4 LCO 3.4.6 Two loops consisting of any combination of RCS loops and residual heat removal (RHR) loops shall be OPERABLE, and one loop shall in operation.

1. All reactor coolant pumps (RCPs) and RHR pumps may not be in operation for ::; 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a. No operations are permitted that would cause reduction of the RCS boron concentration; and

b. Core outlet temperature is maintained at least 10°F below saturation temperature.

2. No RCP shall be with any RCS cold leg temperature::; the Low Temperature Overpressure Protection (LTOP) System applicability temperature specified in the unless:

a. The secondary water temperature of steam generator (SG) is < above of the RCS cold leg temperatures; or

b. The pressurizer water volume is less than 770 cubic (24%

of wide range, cold, pressurizer level indication).

APPLICABILITY: MODE ACTIONS CONDITION REQUI RED ACTION A. required RCS loop A.1 Initiate action ..estore a Immediately inoperable. second loop to OPERABLE status.

Two RHR loops inoperable.

Farley Units 1 and 2 3.4.6-1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

RCS Loops MODE Loops Filled 3.4 REACTOR COOLANT SYSTEM (RCS)

RCS Loops - MODE Loops Filled LCO 3.4.7 One residual heat removal (RHR) loop shall be OPERABLE and in operation, and either:

a. additional RHR loop shall be or

b. The secondary side water level of at least two steam generators (SGs) shall be ~ 75% (wide range).

1. The RHR pump of the loop in operation may not be in operation for S 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a. No operations are permitted that would cause reduction of the RCS boron concentration; and

b. Core outlet temperature is maintained at saturation temperature.

2. One required RHR loop may inoperable for S 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided that the other RHR loop is OPERABLE and in operation.

No reactor coolant pump shall started with one or more RCS cold leg temperatures S the Low Temperature Overpressure Protection (LTOP) System applicability temperature specified in the PTLR unless:

a. The secondary side water temperature of SG is < 50°F above each of ReS cold leg temperatures; or

b. The water volume is than 770 cubic (24%

of wide range, cold, pressurizer level indication).

4. All RHR loops may be removed from operation during planned heatup to 4 when at least one RCS loop is in operation.

number of operating Reactor Coolant Pumps is limited to one at temperatures < 110°F with the exception that a second pump may started for the purpose of maintaining continuous flow while taking the operating pump out of service.

Farley Units 1 and 2 3.4.7-1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Safety 3.4.10 3.4 COOLANT 3.4.10 Pressurizer Safety Valves LCO 3.4.10 Three safety valves OPERABLE with lift settings

~2460 o psig.

APPLICABILITY:

cold leg temperatures> the Low Temperature Overpressure Protection System applicability temperature specified in the PTLR.

within the LCO limits during MODES 3 purpose of setting safety valves under ambient (hot) conditions. exception is allowed for 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months following entry into MODE 3 provided a preliminary cold setting was made prior to heatup.

ACTIONS CONDITION REQUIRED COMPLETION TIME A. One safety A.i Restore valve to 15 minutes valve inoperable. OPE B. Required Action and 6 as~;ociatE~d Completion not met.

Be in MODE 4 with any 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months RCS cold Two or more pressurizer temperatures:::; the safety inoperable. LTOP ..."eTern applicability temperature specified in Farley Units 1 2 3.4.10-1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Low Temperature Overpressure Protection (LTOP) System LCO 3.4.12 An LTOP System shall be OPERABLE with a maximum of one charging pump of injecting into the RCS when one or more of the RCS cold legs is ;5; 1 and a maximum of two charging pumps capable of injecting into the RCS when all of the cold are> 180°F and accumulators isolated and a or b below.

a. Two heat removal (RHR) suction relief valves with setpoints

5; 450 psig.

b. The RCS and an RCS vent of ~ 2.85 square inches.

1. With one or more of the RCS cold legs ::s; 180°F, two charging pumps may be capable of injecting into the during pump swap operations for a period of no more than 15 minutes provided that the RCS is in a non-water solid condition and both RHR relief valves are OPERABLE or the RCS is vented via an opening of no than 5.7 square inches in area.

2. Accumulator isolation is only required when accumulator is greater than or equal to maximum RCS pressure for the existing RCS cold leg temperature allowed by the PtT lirnit curves provided in PTLR.

APPLICABILITY: MODE 4 when the temperature of one or more RCS cold legs is s the

_UTOlrn applicability specified in PTLR, 6 when the reactor vessel head is on.

Farley Units 1 and 2 3.4.1 Amendment No. (Unit 1)

Amendment No. (Unit

LTOP System 3.4.12 ACTIONS LCO 3.0.4b is not applicable when entering MODE CONDITION REQLlI ACTION COMPLETION TIME A. More than the maximum A.i Initiate action to verify s Immediately charging the maximum required pump(s) capable of charging pump(s) injecting into the capable of injecting into RCS. the RCS.

B. An accumulator not B.1 Isolate affected 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months isolated when the accumulator.

accumulator pressure is greater than or equal to the maximum RCS pressure for existing cold leg temperature allowed in the PTLR.

C. Required Action and. C.1 Increase RCS cold leg 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion temperature to > the Time of Condition B not LTOP System met. applicability temperature in the PTLR.

C.2 Depressurize affected 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months accumulator than the maximum ReS pressure for existing cold leg temperature allowed in the PTLR.

Farley Units 1 and 2 3.4.1 Amendment No. (Unit 1)

Amendment No. (Unit 2)

LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 12.1 Verify a maximum of one charging pump is In accordance with capable of injecting into the RCS when one or more the Surveillance RCS cold legs is s 180"F. Frequency Control Program SR 3.4.12.2 Verify a maximum of two charging pumps are In accordance with capable of injecting into the RCS when all RCS cold the Surveillance legs are> 180°F. . Frequency Control Program SR 3.4.1 Verify each accumulator is isolated. In accordance with the Surveillance Frequency Control Program Verify RHR suction isolation valves are open for In accordance with required suction valve. the Surveillance Frequency Control

Program SR 3.4.1 Only required to be performed when complying with LCO 3.4.1 Verify RCS vent;::: 2.85 square inches open. In accordance with the Surveillance Frequency Control Program SR 3.4.1 Verify each required RHR suction relief valve In accordance with setpoint. the I nservice Testing Program In accordance with the Surveillance Frequency Control Program Units 1 and 2 12-4 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

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

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

5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. The reactor coolant system pressure and temperature limits, including heatup and cooldown rates and the LTOP System applicability temperature, shall be established and documented in the PTLR for the following:

LCO 3.4.3, "RCS Pressure and Temperature (PIT) Limits," and LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System."

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.

c. The PTLR shall be provided to the NRC upon issuance for each reactor fluence period and for any revision or supplement thereto.

5.6.7 EDG Failure Report If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures shall be reported within 30 days.

Reports on EDG failures shall include a description of the failures, underlying causes, and corrective actions taken per the Emergency Diesel Generator Reliability Monitoring Program.

(continued)

Farley Units 1 and 2 5.6-5 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Joseph M. Farley Nuclear Plant to Revise Technical Specifications Associated with the low Temperature Overpressure Protection System and the Pressure and Temperature limits Report Enclosure 4 Joseph M. Farley Nuclear Plant Unit 1 and Unit 2 Temperature limits Report

SOUTHERN'\'

COMPANY Energy to SeNle Your World*

Joseph Mil Farley Nuclear Plant Pressure Temperature Limits Report Unit 1 Revision 5 MONTH YEAR

Unit 1 Table of Contents List of ............................................................................................. 2 List of .................................................................................3 1.0 Pressure Temperature Limits Report (PTLR) ............................................... 5 Operating Limits ...................................................................................................5 RCS Pressureffemperature (Pff) Limits (LCO 5 RCP Operation Limits ....................................................................................... 5 LTOP System Applicability Temperature (LCO - 12)................................... 5 3.0 Reactor Vessel Material Surveillance Program ................................................... 12 4.0 Surveillance Credibility ...................................................... 13 5.0 Supplemental Data .......................................................... 13 6.0 References .........................................................................................................21

PTLR for FNP Unit 1 list of Tables 2*1 Farley Unit 1 EFPY Heatup Curve Data Points ..................................................8 Farley Unit 1 54 Cooldown Curve Data Points ............................................ 10 3-1 Surveillance Capsule Withdrawal Schedule .......................................................... 12 1 Comparison of Surveillance Material 30 ft-Ib Transition Temperature Shifts and Upper Shelf Energy with Regulatory Guide 1.99, Revision 2, Predictions ......................................................................................... 14 5-2 Calculation of Chemistry Factors Using Surveillance Capsule Data...................... 15 5-3 Reactor Vessel Toughness Table (Unirradiated) .................................................. 16 Reactor Vessel Fluence Projections at 54 EFPY .................................................. 17 Summary of ART Values for the Reactor Vessel Materials at the 1/4-T and 3/4-T Locations for EFPY....................................................... 18 5-6 Calculation of Adjusted Reference Temperature Values at 54 EFPY for the Limiting Reactor Vessel Material .......................................................................... 19 Pressurized Thermal Shock (RTPTs) Values for 54 EFPY ..................................... 20

PTLR for FNP Unit 1 List of Figures 2-1 Farley Unit 1 Reactor Coolant System Heatup 6 2-2 Farley Unit 1 Reactor Coolant System Cooldown Limitations .................................7

PTLR for FNP Unit 1 This intentionally blank.

PTLR for FNP Unit 1 1.0 ReS Pressure Temperature limits Report. (PTlR)

This PTLR for Farley Nuclear Plant* Unit 1 has been prepared in accordance with the requirement of Technical Specification (TS) 5.S.S. Revisions to the PTLR shall be provided to the NRC after issuance.

This report affects 3.4.3, PressurelTemperature Limits (PIT) Limits. All TS requirements associated with low temperature overpressure protection (LTOP) are contained in TS 3.4.12, RCS Overpressure Protection Systems.

2.0 Operating limits The limits for TS 3.4.3 are presented in the subsection which follows and were developed using the methodologies specified in The methodologies are contained in WCAP*14040-A, Revision 4[1]. The operability requirements associated with LTOP are specified in LCO 3.4.12 and were determined to adequately protect the RCS against brittle fracture in the event of an LTOP transient. The limitation on the number of operating reactor coolant pumps (RCPs) is necessary to assure operation consistent with pressure corrections incorporated in the PIT limits for flow losses associated with the RCPs.

2.1 RCS PressurelTemperature (PIT) Limits (LCO - 3.4.3) 2.1.1 The minimum boltup temperature is

.2 The RCS temperature rate-of-change limits are:

a. A maximum heatup of 100°F in anyone hour period.

b. A maximum cooldown of 100°F in anyone hour period.

c. A maximum temperature change of less than or equal to 1QaF in anyone hour period during inservice hydrostatic and testing operations above the heatup and cooldown limit curves.

2.1.3 The RCS PIT limits for heatup and cool down are specified by Figures 2-1 and respectively.

RCP Operation 2.2.1 The number of operating Reps is limited to one at temperatures less than 11 oaF with the exception that a second pump may started for the purpose of maintaining continuous flow while taking the operating out of service.

LTOP System Applicability Temperature (LCO - 3.4.12) 2.3.1 The Low Temperature Overpressure Protection (LTOP) System applicability temperature is

PTLR for FNP Unit 1 Revision 5 Page 6 of 22

Figure 2-1 Farley Unit 1 Reactor Coolant System Heatup Limitations[21 (Heatup Rates up to 1OO°F/hr) Applicable to 54 EFPY (adjusted to include 60 psi ~P at RCS temperatures ~ 110°F and 27 psi ~P at RCS temperatures < 110°F), Includes vessel flange requirements per 10 CFR 50, Appendix Gf3f.

Limiting Material: Lower Shell Plate 86919-1 with non-credible surveillance data Limiting ART Values at 54 EFPY: 1/4T = 191°F 3/4T = 166°F 2500 ...---- -:-

1 - -- ---j--I I ! I 2250 -\i

__I__

2000 . .--J~ -- !-

I 1--- - .

I 1750 , , *. I I

Q I .-,,

(J)

Q. 1500 ----- - - -.

( \)

~

I

(/)

(\) 1250 1

-"t -." - +.

! I 1-I

- - 1 Q. .[

I

~ I

(\)

(U I I i I "5 - .-:-. _ _ - L - _ :

(J 1000 I (ij I 0 i! , I 760 -- - -_.. - -, .--~-

i

- I -- --

I I

i I

I 500 - - - Criticality Limit based on inservice hydrostatic test

, temperatu re 1247"F) for the

1 service eriod u to 54 EFPY 250 -

j 1

f I l I I I

0 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Oeg. F)

PTLR for FNP Unit 1 Revision 5 Page 7 of 22 Figure 2*2 Farley Unit 1 Reactor Coolant System Cooldown Limitations[2]

(Cooldown Rates up to 100°F/hr) Applicable to 54 EFPY (adjusted to include 60 psi 6P at RCS temperatures ~ 110°F and 27 psi 6P at RCS temperatures < 110°F), Includes vessel flange requirements per 10 CFR 50, Appendix G[3f.

Limiting Material: Lower Shell Plate 86919-1 with non-credible surveillance data Limiting ART Values at 54 EFPY: 1/4T =191°F 3/4T = 166°F 2500 ~--~--~--~--~--~----~--~--~--~-------,

2250 '1 J! -'-! 1*! .i I I

...

i j

--+-

I 2000 --_.+ "---1

, I

- '--r-I

.' - j _ .

I l '1!

I I, ,I 1750 . -+ __ . ~__ . L___ . ~+- _.._

I i ;

i_ I

. ._...l-. - - "i . -* ~* t . - --'---1

: I I I '

G' I I I !

~ 1500 I

. .- i

"'-1 - .. ~ - , .. --- ! .- . ..

...:J

~ ;

I

(/J

...~ 1250 I i

..... _- I

. . ._ ;

I r .

a.

"0

.Q) I Acceptable

~ Operation

~ 1000 "-r ' i- - - j' -'

I Ii u i i

750 . ' -. ' .. ~- .. - . -r --

500 1- - . . ~

250 --- -.-- - .- -i-"

j o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Oeg. F)

PTLR for FNP Unit 1 Revision 5 Page 8 of 22 Table 2-1 Farley Unit 1 - 54 EFPY Heatup Curve Data Points[2j (adjusted to include 60 psi dP at RCS temperatures ~ 110°F and 27 psi dP at RCS temperatures < 110°F) 60°Flhr. 60°F/hr. 100°F/hr. 100°Flhr.

Leak Test Limit Heatup Criticality Heatup Criticality T P (psig) T (OF) P T P T P T P eF) (psi g) co F) (psig) eF) (psig) (OF) (psig) 229 2000 60 0 247 0 60 0 247 0 247 2485 60 594 247 561 60 574 247 541 65 594 247 561 65 574 247 541 70 594 247 561 70 574 247 542 75 594 247 561 75 574 247 542 80 594 247 561 80 574 247 545 85 594 247 561 85 574 247 545 90 594 247 561 90 574 247 549 95 594 247 561 95 574 247 550 100 594 247 561 100 574 247 554 105 594 247 561 105 574 247 556 110 594 247 561 110 574 247 561 110 561 247 561 110 541 247 561 115 561 247 561 115 541 247 561 120 561 247 561 120 542 247 561 125 561 247 561 125 545 247 561 130 561 247 561 130 549 247 561 135 561 247 561 135 554 247 561 140 561 247 561 140 561 247 561 145 561 247 561 145 561 247 561 150 561 247 561 150 561 247 561 155 561 247 561 155 561 247 561 160 561 247 561 160 561 247 561 165 561 247 561 165 561 247 561 170 561 247 561 170 561 247 561 175 561 247 822 175 561 247 670 180 561 247 855 180 561 247 693 180 561 247 892 180 561 247 718 180 822 247 932 180 670 247 746 185 855 247 977 185 693 247 777 190 892 247 1049 190 718 247 826

Table 2-1 (continued)

Farley Unit 1 - 54 Heatup Curve Data POints[2]

AP at RCS temperatures ~ 110°F RCS temperatures < 110°F) 60°Flhr.

6O°Flhr. Heatup Criticality p T T (psig) (OF) f'F}

195 932 200 977 1143 777 255 205 1027 1210 205 811 260 210 1082 210 849 265 215 1143 1365 215 891 270 220 1210 275 1455 220 938 275 225 1284 280 1552 225 990 280 230 1365 285 1638 230 1046 285 235 1455 290 1734 235 1109 290 240 1552 295 1839 240 1179 295 245 1638 300 1955 245 1255 300 250 1734 2084 250 1340 305 255 1839 1433 310 260 1955 1536 315 265 2084 265 1650 320 270 2225 270 1775 325 275 2382 275 1913 330 280 2066 285 2234 290 2419

PTLR for FNP Unit 1 Table 2-2 Farley Unit 1 - 54 EFPY Cooldown Curve Data Pointsl2]

(adjusted to include 60 psi AP at RCS temperatures ~ 110°F and 27 psi at RCS temperatures < 110°F) r-------------~------------~-- ----~-----

20" F/h r.

PTlR for FNP Unit 1 Table 2-2 (continued)

Unit 1 - 54 EFPY Cooldown Data Points[2]

(adjusted to include 60 psi AP at RCS temperatures ~ 110°F and psi AP at RCS temperatures < 110°F)

Steady State T (oF) T (OF) P (psig) T (OF) P (psig) T eF} P (psig)

~~~~~~~~~

220 1293 1368 1879 1879 1879 1879 255 2016 255 2016 2016 255 2016 260 2167 260 2167 260 2167 260 2167 260 2167

--~---+--

265 2334 2334 265 2334 265 2334

Unit 1 3.0 Reactor Vessel Material Surveillance Program The reactor material surveillance program is in compliance with 10 CFR 50, Appendix H[4J, and is described in 5.4.3.6 of the Farley FSAR. Surveillance capsules are tested and the results reported in accordance with ASTM E185-82!51. The removal schedule is provided in Table 3-1.

neutron transport and evaluation used follow the guidance and meet requirements of Regulatory 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,,[6J. The results of the Z examination rNCAP*16964-NP, Revision 0 111 ) were to produce and 2-2.

Table 3-1 SUrveillance Capsule Withdrawal Schedule (al Capsule lead Removal Capsule location Factor EFPY (b) 343 107 287 110 340 a) Data from Table 7-1, WCAP-16964-NP. Revision 0 rn b) Effective Fu" Power Years from startup.

c) Plant-specific evaluation.

d) This fluance is not less than once or greater than twice the peak EOL lIuence for the initial40-year license term.

e) This fluence is not less than once or greater than twice the peak EOl fluence for a license renewal term 10 60 years.

f) This lIuence Is not less than once or greater than twice the EOl fluence for an additional license renewal term to 80 years.

PTLR for FNP Unit 1 4.0 Reactor Vessel Surveillance Data Credibility Regulatory Guide 1.99, Revision 2[81, describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1 Revision 2, describes the methodology for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Position C.2 can only be applied when two or more credible surveillance data become available from the reactor in question.

Per WCAP-881 019). the Unit 1 surveillance program was based on ASTM E185-73110I* All six surveillance capsules (Y [I1 J, U [121, X [131, W [141, V 115J, and Z [7]) have been removed from Unit 1 reactor vessel and analyzed. In accordance with the discussion of Regulatory Guide 1.99, ReviSion 2, there are five requirements that must be met for the surveillance data to be judged credible.

credibility conclusions for the Farley Unit 1 surveillance plate and weld are described below:

credibility evaluation of the Farley Unit 1 Lower Shell Plate 86919-1 surveillance material is documented in Appendix 0 of WCAP:"16964-NPI7l. The credibility evaluation concluded that the surveillance data for Lower Shell Plate 86919-1 is '-""'-:..:....;::==:::..

The credibility evaluation for weld Heat # 33A277 surveillance data is documented in Appendix 0 of WCAP-17365-NP[161* The evaluation into account data from Calvert Cliffs Unit 1 and Farley Unit 1. The credibility evaluation concluded that the surveillance data for weld Heat # 33A2n is credible.

5.0 Supplemental Data Tables 5-1 contains a comparison of surveillance material 30 ft-Ib transition temperature shifts and upper shelf energy with Regulatory Guide 1.99, Revision 2, predictions.

Table 5-2 shows the calculation of the surveillance material chemistry factors using surveillance capsule data.

Table 5-3 provides the unirradiated Farley Unit 1 reactor vessel toughness data.

Table 5-4 provides a summary of the reactor vessel fluence values at 54 Table 5-5 provides a summary of the ART values of Farley Unit 1 reactor vessel materials at 1/4-T and 314*T locations for 54 EFPY.

Table 5-6 shows the calculation of the ART values at 54 EFPY for the limiting Farley Unit 1 reactor vessel material (lower shell plate B6919-1).

5-7 provides AT PTS values for Farley Unit 1 for 54 EFPY.

PTLR for FNP Unit 1 Table Comparison Surveillance Material 30 fHb Transition Temperature Shift and Upper Shelf Decrease with Regulatory Guide 1 Revision 2, (a) 30 ft-Ib Transition Upper Shelf Energy Temperature Shift Decrease fluence Predicted Measured Predicted Measured Material Capsule (1019 n/cm2 *

(OF) (Of) (%) (%)

E> 1.0 MeV) y 0.612 U 1.73 Lower Shell X 3.06 126.7 Plate B6919*1 4.75 136.2 (Longltudinal) 143.4 21 29 1

Lower Shell X 3.06 126.7 110.8 30 12 Plate B6919-1 W 4.75 136.2 150.5 35 17 (Transverse)

V 7.14 143.4 161.7 39 21 Z 8.47 145.9 178.3 40 23 3

22 Surveillance 15 Program W 4.75 Weld Metal V 7.14 114.5 117.5 44 Z 8.47 116.5 113.5 46 Heat Affected Zone Material 7.14 Z 8.47 Data from Table 5-10 of WCAp*16964-NP, Revision 0[71.

PTLR for FNP Unit 1 Revision 5 Page 15 of 22 Table 5-2 Calculation of Chemistry Factors Using Surveillance Capsule Data Capsule f Material Capsule 19 (10 n/cm 2 , Fpa) .1RTNDT FF"'.1RTNDT FF2 (OF) (OF)

E> 1.0 MeV)

Y 0.612 0.862 64.6 55.7 0.744 U 1.73 1.151 110.0 126.6 1.324 Lower Shell X 3.06 1.295 129.2 167.1 1.678 Plate 86919-1 W 4.75 1.392 145.3 202.3 1.938 (Longitudinal)(b)

V 7.14 1.466 177.7 260.5 2.149 Z 8.47 1.492 202.2 301.6 2.225 Y 0.612 0.862 70.1 60.5 0.744 U 1.73 1.151 100.4 115.5 1.324 Lower Shell X 3.06 1.295 110.8 143.5 1.678 Plate 86919-1 W 4.75 1.392 150.5 209.5 1.938 (Transverse)(b)

V 7.14 1.466 161.7 237.1 2.149 Z 8.47 1.492 178.3 266.0 2.225 SUM: 2146.20 20.118 CFB69,9., =I(FF

ATNOT) + I( FF2) =(2146.20) + (20.118) = 106PF Y 0.612 0.862 108.38 (66 .9)(d) 93.47 0.744 Farley Unit 1 U 1.73 1.151 121.66 (75.1)(d) 140.00 1.324 Surveillance Weld X 3.06 1.295 141.59 (87.4)(d) 183.42 1.678 Material W 4.75 1.392 159.25 (98.3)(d) 221.69 1.938 (Heat # 33A277)(C)

V 7.14 1.466 190.35 (117.5)(d) 279.07 2.149 Z 8.47 1.492 183.87 (113.5)(d) 274.30 2.225 Calvert Cliffs Unit 1 263° 0.505 0.809 81.97 (50.4)(d) 66.34 0.655 Surveillance Weld 9]0 1.94 1.181 156.63 (1 04.5)(d) 185.00 1.395 Material (Heat # 33A277)(C) 284° 2.33 1.228 120.06 (78.0)(d) 147.49 1.509 SUM: 1590.78 13.618 CF Surv. Weld =I(FF

ATNOT) + L( FF2) :;: (1590.78) + (13.618) = 116.8°F NOTES:

(a) FF = fluence factor = f (026* 0.' log(I)).

(b) Information pertaining to Lower Shell Plate 86919-1 is taken from Table 0-1 of WCAP-16964-NP(7].

(c) Information pertaining to weld Heat # 33A277 is taken from Table 6.1-1 of WCAP-17506-NP[17J .

(d) To calculate the best fit chemistry factor (CF) as provided by Reg. Guide 1.99, Rev. 2, Position 2.1, the surveillance weld flRTNOT values have been adjusted to account for chemistry differences between the reactor vessel weld and the surveillance weld. For the Calvert Cliffs Unit 1 data, the surveillance weld flRTNOT values have also been adjusted to account for the temperature difference between the Farley Unit 1 and Calvert Cliffs Unit 1 reactor vessels. Pre-adjusted values are in parentheses. See Table 6.1-1 of WCAP-17506-NP[17/ for all details pertaining to the chemistry factor calculation for weld Heat # 33A277.

PTLR for FNP Unit 1 Revision 5 Page 16 of 22 Table 5-3 Reactor Vessel Toughness Table (Unirradiated) (a) 8eltline Material Cu Weight Ni Weight IRTNDT

% % (OF)

Closure Head Flange -- -- -50(d)

Vessel Flange -- -- 60 Inlet Nozzle 86917-1 0.16 0.83 60 Inlet Nozzle 86917-2 0.16 0.80 60 Inlet Nozzle 86917-3 0.16 0.87 60 Outlet Nozzle 86916-1 0.16 0.77 60 Outlet Nozzle 86916-2 0.16 0.78 60 Outlet Nozzle 86916-3 0.16 0.78 60 Upper Shell Forging 86914 0.16 0.684 30 Intermediate Shell Plate 86903-2 0.13 0.60 0 Intermediate Shell Plate 86903-3 0.12 0.56 10 Lower Shell Plate 86919-1 0.14 0.55 15 Lower Shell Plate 86919-2 0.14 0.56 5 Inlet/Outlet Nozzle to Upper Shell Girth Seams lO(e) 0.04 1.08 1-897 A--F Upper Shell to Intermediate Shell Circumferential -56(1) 0.197 0.06 Weld Seam 10-894 (Heat # 90099) (b)

Intermediate Shell Longitudinal Weld Seams -56(1)19-894 A & 8 (Heat # 33A277) (b) 0.258 0.165 Surveillance Weld (e) 0.14 0.19 -

Intennediate Shell to Lower Shell Circumferential -56(1) 0.205 0.105 Weld Seam 11-894 (Heat # 6329637) (b)

Lower Shell Longitudinal Weld Seams20-894 -56(1)

A & 8 (Heat # 90099) (b) 0.197 0.060 NOTES; (a) From Table 4.1-1 of WCAP-17506-Np ll 7J.

8 (b) Best-estimate copper and nickel from CE NPSD-1 039 11 ,.

(c) The surveillance weld is representative of intermediate shell longitudinal welds 19-894 A & B. Best-estimate copper and nickel values represent a single chemical analysis documented in WCAP-881 0, ReVision 0 191.

(d) Replacement closure head initial RT NOT value was taken from MHI-SNC-019S[19,.

(e) An estimation method using measured data was used to determine this initial RT NOT value. Therefore, a conservative value of 17°F is used for O'u and 0'1 in margin calculations.

(f) These initial RT NOT values are generic and taken from 10 CFR 50.61 paragraph (c)(l)(ii) of the 1-1-07 edition.

PTLR for FNP Unit 1 Table 5-4 Reactor VA!~~I Ruence Projections at 54 EFPY (a)

(10 19 n/cm 2, E> 1.0 MeV)

, I 54 Reactor Vessel Location Material Neutron Fluence

......---~--

Inlet Nozzle 86917-1 0.0349 Inlet Nozzle 86917-2 0.0190 Inlet Nozzle 86917-3 0.0139 Outlet Nozzle 86916-1 0.00922 Outlet Nozzle 86916-2 0.0126 I

Outlet Nozzle 86916-3 0.0231 Upper Shell Forging 86914 1.02 86903-2 Intermediate Shell Plates 5.93

& 86903-3 i 86919-1 l

Lower Shell Plates 5.81

& 86919-2 i

InleVOutlet Nozzle to Upper Shell 1-897 A-F 0.0349 Girth Seams Upper Shell to Intermediate Shell 10-894 1.02 Circumferential Weld Seam (Heat # 90099)

I I 1

Intermediate Shell Longitudinal Weld i 19-894 A & 8 1.83 Seams (Heat # 33A277)

Intermediate Shell to Lower Shell 11-894 5.81 Circumferential Weld Seam I (Heat # 6329637)

Lower Shell Longitudinal Weld I 20-894 A & 8 1.79 Seams (Heat # 90099)

NOTE:

(a) From Table 5.1-1 01 WCAP-1 . These values are also summarized in Table 2-1 of Attachment A of ALA-09 116[20J*

PTLR for FNP Unit 1 Revision 5 Page 18 of 22 Table 5-5 Summary of ART Values for the Reactor Vessel Materials at the 1/4T and 3/4T Locations for 54 EFPY (a) 1/4 T 3/4 T Material (oF) (OF)

Inlet Nozzle 86917-1 105 84 Inlet Nozzle 86917-2 90 76 Inlet Nozzle 86917-3 85 72 Outlet Nozzle 86916-1 78 69 Outlet Nozzle 86916-2 83 71 Outlet Nozzle 86916-3 94 78 Upper Shell Forging 86914 169 139 Intermediate Shell Plate 86903-2 156 134 Intermediate Shell Plate 86903-3 154 134 Lower Shell Plate 86919-1 179 156 LowerShell Plate 86919-1 Using non-credible SIC Data 191 (b) 166(b)

Lower Shell Plate 86919-2 170 147 Inlet/Outlet Nozzle to Upper Shell Girth Seams 55 50 1-897 A->F Upper Shell to Intermediate Shell Circumferential 89 66 Weld Seam 10-894 (Heat # 90099)

Intermediate Shell Longitudinal Weld Seams 140 107 19-894 A & 8 (Heat # 33A277)

Intermediate Shell Longitudinal Weld Seams19-894 A & 8 (Heat # 33A277) 111 80 Using credible SIC Data Intermediate Shell to Lower Shell Circumferential Weld Seam 11-894 (Heat # 6329637) 141 117 Lower Shell Longitudinal Weld Seams20-894 104 80 A & 8 (Heat # 90099)

NOTES:

(a) The ART values presented here are based on the reactor vessel surface lIuence values summarized in Table 5-4.

The values for the belt line materials are from Tables 4-10 and 4-11 of WCAP-17122-Np I2 1. The values for the extended beltline materials are summarized along with the values for the beilline materials in Tables 3-3 and 3-4 of Attachment A of ALA-09_116120I .

(b) Limiting 1/4T and 3/4T ART values. The PIT limit curves are based on these limiting ART values of 191 of and 166°F.

Unit 1 Table 5-6 Calculation of Adjusted Reference Temperature Values at EFPY for the Limiting Reactor Material - Lower Shell Plate 9-1

~ Parameter Value I

rating Period 54 EFPY location 1/4 T 3/4 T

Chemistry Factor, eF)(a) 106.7 106.7 Fluence, f (10 19 n/cm 2) (b) 3.622 1.408 Fluence Factor, FF :::; f (O.2S*0.1"log(l) 1.3343 1.0949

= l( FF 142.4 116.8 Initial RT NOT. I COF) (e) 15 15 Margin. M eF) (dJ 34 34 I

Adjusted Reference Temperature (ART), (OF) per Revision 2 (e) 191 166 Regulatory Guide 1 NOTES:

(a) Chemistry factor is taken from Table 5-2.

(b) Fluence is based on fsur! == 5.81 )( 10 19 (E,> 1.0 MeV), from Table 4*1 of WCAP*17122-NP, Revision 0[21. Farley Unil1 reactor vessel wall thickness is 7.875 inches in the beilline region.

(e) Initial RT NOT value is taken from Table 5-3.

(d) Margin = + <1/\2) o.S, for the lower shell plate 86919-1, <11 O"F and <1/\:::: 17°F.

(e) Per Regulatory Guide 1.99, Revision 2: ART (OF)::: LlRTNOT + I + M.

PTLR for FNP Unit 1 Revision 5 Page 20 of 22 Table 5-7 Pressurized Thermal Shock (RTpTs) Values for 54 EFPY (a)

Surface l\RTNDT Fluence I M .RTpTS Material CF FF (CF x FF)

(10 19 n/cm 2 , (OF) (OF) (OF)

(OF)

E > 1.0 MeV)

Inlet Nozzle 86917-1 123.3 0.0349 0.2397 29.6 60 29.6 119 Inlet Nozzle 86917-2 123 0.0190 0.1666 20.5 60 20.5 101 Inlet Nozzle 86917-3 123.7 0.0139 0.1365 16.9 60 16.9 94 Outlet Nozzle 86916-1 122.3 0.00922 0.1037 12.7 60 12.7 85 Outlet Nozzle 86916-2 122.5 0.0126 0.1280 15.7 60 15.7 91 Outlet Nozzle 86916-3 122.5 0.0231 0.1879 23.0 60 23.0 106 Upper Shell Forging 86914 120.1 1.02 1.0055 120.8 30 34.0 185 Intermediate Shell Plate 86903-2 91.0 5.93 1.4345 130.5 0 34.0 165 Intermediate Shell Plate 86903-3 82.2 5.93 1.4345 117.9 10 34.0 162 Lower Shell Plate 86919-1 97.8 5.81 1.4308 139.9 15 34.0 189 Lower Shell Plate 86919-1 202b )

106.7 5.81 1.4308 152.7 15 34.0 Using non-credible SIC Data Lower Shell Plate 86919-2 98.2 5.81 1.4308 140.5 5 34.0 180 InleVOutiet Nozzle to Upper Shell 54 0.0349 0.2397 12.9 10 36.4 59 Girth Seams 1-897 A-F Upper Shell to Intermediate Shell Circumferential Weld Seam 10-894 91.4 1.02 1.0055 91 .9 -56 65.5 101 (Heat # 90099)

Intermediate Shell Longitudinal Weld Seams19-894 A & B 126.3 1.83 1.1657 147.2 -56 65.5 157 (Heat # 33A277)

Intermediate Shell Longitudinal Weld Seams19-894 A & 8 116.8 1.83 1.1657 136.2 -56 44.0 124 (Heat # 33A277)

Using credible SIC Data Intermediate Shell to Lower Shell Circumferential Weld Seam 11-894 98.4 5.81 1.4308 140.8 -56 65.5 150 (Heat # 6329637)

Lower Shell Longitudinal Weld Seams20-894 A & 8 91 .4 1.79 1.1599 106.0 -56 65.5 116 (Heat # 90099)

NOTES:

(a) From Table 7.1-1 of WCAP-17506-NP(17].

(b) This limiting AT PTS value was calculated using the CF from the surveillance data and a full all margin of 17°F, since this surveillance data is not credible.

PTLR for FNP Unit 1 6.0 References

1. WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Umit " May 2004.

2. WCAP-17122-NP, Revision 0, "J. M. Farley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," A. Leicht and C. Heinecke, October 2009.

3. 10 CFR 50, Appendix "Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, December 1 1

4. 10 50 Appendix H, "Reactor Vessel Material Surveillance Program Requirements," Federal Register, Volume 60, No. 243, December 19,1995.

5. ASTM 85-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," in ASTM Standards, Section 3, American Society for Testing and Materials, 1982.

6. NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.

7. WCAP-16964-NP, Revision 0, of Capsule Z from the Southern Nuclear Operating Company Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program,"

J. M. Conermann and M. A. Hunter. October 2008.

8. Regulatory Guide 1.99, ReviSion 2, "Radiation Embrittlement of Reactor Materials,"

May 1

9. WCAP-8810, Revision 0, "Southern Alabama Power Company Joseph M. Nuclear Plant Unit NO.1 Reactor Vessel Radiation Surveillance Program," J. A. Davidson, et. aI.,

December 1976.

85-73, "Standard Recommended Practice for Surveillance for Nuclear Reactor

'essets," American Society for Testing and Materials, 1973.

11. WCAP-9717, Revision 0, "Analysis of Capsule Y from the Alabama Power Company Farley Unit No.1 Reactor Vessel Radiation Surveillance Program," Yanichko, et. aI., June 1980.

1 WCAP-10474, Revision 0, "Analysis of Capsule U from the Alabama Power Company Joseph M. Unit 1 Reactor Vessel Radiation Surveillance Program," R. Boggs, et aI.,

February 1984.

13. WCAP-11563, Revision 1, "Analysis of Capsule X from the Alabama Power Company Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program," R. P. Shogan, et. aI.,

September 1987.

14. WCAP-14196, ReviSion O,"Analysis of Capsule W from the Alabama Power Company Farley Unit 1 Reactor Vessel Radiation Surveillance Program," P. A. et. aI., February 1995.

15. WCAp*16221-NP, 0, "Analysis of Capsule V from the Southern Nuclear Operating Company, Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program, K. G. Knight, et. aL, March 2004.

16. WCAP-17365-NP, ReviSion 0, "Analysis of from the Calvert Cliffs Unit No.1 Reactor Vessel Radiation Surveillance Program," J. Long and J. I. Duo, March 2011.

17. WCAP-17506-NP, Revision 0, "Farley Units 1 and 2 Pressurized Thermal Shock Evaluations,"

B. A. December 2011.

18. CE NPSO-1039, Revision 2, "Best Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," Combustion Engineering Owners Group, June 1997.

19. Mitsublshi Heavy Industries, LTD, Kobe Shipyard & Machinery Works (MHI). MHI-SNC-01 Reactor Vessel Closure Head for Farley-1, "Certified Material Test Report," August 22,2003.

20. Westinghouse Letter ALA-09-116, "P-T Limit Curves with Margins for Instrumentation Errors and Extended Beltline Material Information," John M. Robinson, October 20, 2009.

SOUTHERN COMPANY Joseph Mil Farley Nuclear Plant Pressure Tem rature Limits Report Unit 2 Revision 5 MONTH YEAR

PTLR for FNP Unit 2 Table of Contents List of Tables ................................................................................................................... 2 List of ................................................................................................................. 3 1.0 Temperature Limits Report (PTLR) ............................................... 5 Operating Limits ................................................................................................... 5 2.1 Pressurerremperature (prr) Limits (LCO - 3.4.3) ................................... 5 RCP Operation ....................................................................................... 5 LTOP System Applicability Temperature (LCO - 3.4.12) .................................. 5 Reactor Vessel Material Surveillance Program ................................................... 12 4.0 Reactor Vessel Surveillance Data Credibility ...................................................... 13 5.0 Supplemental ................................................................................. 13 6.0 ...................................................................................... 21

PTLR for FNP Unit 2 of Tables Farley Unit 2 EFPY Heatup Curve Data Points ................................................ 8 Unit 2 Cooldown Curve Data ................................ 10 3~ 1 Surveillance Capsule Withdrawal Schedule .......... '" ........................................... 12 5-1 Comparison of Surveillance Material 30 Transition Temperature Shifts and Upper Shelf with Regulatory Guide 1.99, Revision 2, Predictions ....................................................................................... 14 5-2 Calculation of Chemistry Using Surveillance Capsule Data ................... 15 5-3 Reactor Vessel Toughness Table (Un irradiated) ................................................ 16 Fluence Projections at EFPY ................................................ 17 5-5 Summary of Values for Reactor Materials at the 1/4-T and 3/4-T Locations for 54 ................................................ 18 Calculation of Adjusted Reference Temperature Values at for the Limiting Material ........................................................................ 19 5-7 Thermal Shock (RT PTS) for 54 ................................... 20

list of Figures Farley Unit 2 Reactor Coolant System Heatup Limitations .................................... 6 Farley Unit 2 Reactor Coolant System Cooldown Limitations ............................... 7

ThiS page intentionally blank.

1.0 Res Pressure Temperature Limits Report (PTLR)

This PTLR for Farley Nuclear Plant - Unit 2 has prepared in accordance with the requirement of Technical Specification (TS) 5.6.6. Revisions to the shall provided to the NRC after This report affects TS RCS PressurefTemperature Limits. All TS requirements associated with low temperature protection (LTOP) are contained in TS 12, RCS Overpressure Protection Systems.

2.0 Operating Limits The limits for are presented in the subsection which follows and were developed using NRC-approved methodologies specified in TS 5.6.6. The methodologies are contained in WCAP-14040-A, Revision 411 ]. The operability

",'omonte:: associated with are in 3.4.12 and were determined to adequately protect RCS against brittle fracture in the event of an LTOP transient. The limitation on number operating reactor coolant pumps (RCPs) is to assure operation consistent with the pressure corrections incorporated in the PfT limits for flow associated with the RCPs.

RCS PressurefTemperature (PfT) Limits (LCO - 3.4.3) 2.1.1 The minimum boltup temperature is 60°F.

2.1.2 The RCS temperature rate-ot-change limits are:

a. A maximum heatup of 1 in anyone hour period.

b. A maximum cooldown of 100°F in anyone hour period.

c. A maximum temperature of than or to 1 in anyone hour period during inservice hydrostatic leak testing operations above heatup and cooldown limit curves.

2.1.3 The PfT limits for and cooldown are Figures 2-1 and respectively.

2.2 RCP Operation 2.2.1 number of operating Reps is limited to one at Res temperatures less than 11 with the exception a second pump be started for the purpose of maintaining continuous flow while taking operating pump out of service.

2.3 LTOP ....\1"',"',." Applicability Temperature (LeO - 3.4.12)

Low Temperature Overpressure Protection (LTOP) applicability temperature is

PTLR for FN P Unit 2 Revision 5 Page 6 of 22 Figure 2-1 Farley Unit 2 Reactor Coolant System Heatup limitations[2]

(Heatup Rates up to 1OO°F/hr) Applicable to 54 EFPY (adjusted to include 60 psi l1P at RCS temperatures ~ 110°F and 27 psi l1P at RCS temperatures < 110°F). Includes vessel flange requirements per 10 CFR 50, Appendix G.

limiting Material: Intermediate Shell Plate B7212-1 with credible surveillance data limiting ART Values at 54 EFPY: 1/4T =200°F 3/4T = 165°F 2500 ~---------------------~--------------~------~

Ileak Test limit I 2250 2000 ; ..

1750 8'

~

-Q1

J 1500 Unacceptable Operation II>

~

-. 1250 a..

"'0 Critical Limit ;

Q1

60 Deg. FIHr

~

"5 o

1000 Acceptable c;; Operation U

760 600 Criticality limit based on inservice hydrostatic test

+--- temperature (256 C F) for th e service period up to 54 EFPY 250 ,.

o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

PTLR for FNP Unit 2 Revision 5 Page 7 of 22 Figure 2-2 Farley Unit 2 Reactor Coolant System Cooldown Limitations[2]

(Cooldown Rates up to 100°F/hr) Applicable to 54 EFPY (adjusted to include 60 psi llP at RCS temperatures;:: 110°F and 27 psi llP at RCS temperatures < 110°F). Includes vessel flange requirements per 10 CFR 50, Appendix G.

Limiting Material: Intermediate Shell Plate B7212-1 with credible surveillance data Limiting ART Values at 54 EFPY: 1/4T = 200°F 3/4T = 165°F 2500 ~--------------------------------------------~

2250

__ I 2000 1750

-en

( !)

1500

---c.....

Q)

~

en II)

Q) c..

1250 1J

....1'0 Q)

1000

(.)

jij 0

750 o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

PTLR for FNP Unit 2 Revision 5 Page 8 of 22 Table 2-1 Farley Unit 2 - 54 EFPY Heatup Curve Data Points[2]

(adjusted to include 60 psi dP at RCS temperatures ~ 110°F and 27 psi dP at RCS temperatures < 11 O°F) 60°F/hr. 60°F/hr. 1OO°F/hr. 1OO°F/hr.

Leak Test Limit Heatup Criticality Heatup Criticality T P (psig) T (OF) P T P T P T P

, (OF) (psig) (OF) (psig) eF) (psjg) (OF) (psig) 238 2000 60 0 256 0 60 0 256 0 256 2485 60 594 256 561 60 575 256 542 65 594 256 561 65 575 256 542 70 594 256 561 70 575 256 544 75 594 256 561 75 575 256 544 80 594 256 561 80 575 256 546 85 594 256 561 85 575 256 547 90 594 256 561 90 575 256 551 95 594 256 561 95 575 256 551 100 594 256 561 100 575 256 556 105 594 256 561 105 575 256 557 110 594 256 561 110 575 256 561 110 561 256 561 110 542 256 561 115 561 256 561 115 542 256 561 120 561 256 561 120 544 256 561 i

I 125 561 256 561 125 547 256 561 I

I 130 561 256 561 130 551 256 561 135 561 256 561 135 556 256 561 I

140 561 256 561 140 561 256 561 I 145 561 256 561 145 561 256 561 150 561 256 561 150 561 256 561 155 561 256 561 155 561 256 561 160 561 256 561 160 561 256 561 165 561 256 561 165 561 256 561 170 561 256 561 170 561 256 561 175 561 256 561 175 561 256 561 180 561 256 828 180 561 256 675 180 561 256 862 180 561 256 698 180 828 256 900 180 675 256 724 185 862 256 941 185 698 256 752 190 900 256 987 190 724 256 784

PTlR for FNP Unit 2 Revision 5 Page 9 of 22 Table 2-1 (continued)

Farley Unit 2 - 54 EFPY Heatup Curve Data Points[2]

(adjusted to include 60 psi LlP at RCS temperatures ~ 110°F and 27 psi LlP at RCS temperatures < 110°F)

Leak Test 60°F/hr. 100°F/hr.

60°F/hr. Heatup 100°Flhr. Heatup Limit Criticality Criticality T p T T T T P P (psi g) P (psig) P (psig)

(OF) (psig) co F) (OF) CF) (OF) (psig) 195 941 256 1038 195 752 256 819 200 979 256 1094 200 784 256 858 205 1021 256 1130 205 819 256 910 210 1067 260 1175 210 858 260 949 215 1119 265 1237 215 901 265 1001 220 1175 270 1291 220 949 270 1059 225 1237 275 1350 225 1001 275 1123 230 1291 280 1416 230 1059 280 1194 235 1350 285 1488 235 1123 285 1272 240 1416 290 1568 240 1194 290 1359 245 1488 295 1656 245 1272 295 1454 250 1568 300 1753 250 1359 300 1559 255 1656 305 1861 255 1454 305 1675 260 1753 310 1979 260 1559 310 1803 265 1861 315 2110 265 1675 315 1944 270 1979 320 2254 270 1803 320 2099 275 2110 325 2414 275 1944 325 2250 280 2254 280 2099 330 2399 285 2414 285 2250 290 2399

PTLR for FNP Unit 2 Revision 5 Page 10 of 22 Table 2-2 Farley Unit 2 - 54 EFPY Cooldown Curve Data Points[2]

(adjusted to include 60 psi ~P at RCS temperatures ~ 110°F and 27 psi ~P at RCS temperatures < 110°F)

Steady State 20°F/hr. 40°F/hr. 60°F/hr. 100°F/hr.

T(OF) P (psig) T(OF) P (psig) T(OF) P (psig) T(OF) P (psig) T(OF) P (psig) ,

60 0 60 0 60 0 60 0 60 0 60 594 60 594 60 553 60 509 60 420 65 594 65 594 65 555 65 512 65 423 70 594 70 594 70 558 70 515 70 426 75 594 75 594 75 561 75 518 75 430 80 594 80 594 80 565 80 522 80 434 85 594 85 594 85 569 85 527 85 439 90 594 90 594 90 574 90 531 90 445 95 594 95 594 95 579 95 537 95 451 100 594 100 594 100 585 100 543 100 458 105 594 105 594 105 591 105 550 105 465 110 594 110 594 110 594 110 557 110 474 110 561 110 561 110 561 110 524 110 441 115 561 115 561 115 561 115 532 115 451 120 561 120 561 120 561 120 542 120 461 125 561 125 561 125 561 125 552 125 473 130 561 130 561 130 561 130 561 130 487 135 561 135 561 135 561 135 561 135 502 140 561 140 561 140 561 140 561 140 519 145 561 145 561 145 561 145 561 145 537 150 561 150 561 150 561 150 561 150 558 155 561 155 561 155 561 155 561 155 561 160 561 160 561 160 561 160 561 160 561 165 561 165 561 165 561 165 561 165 561 170 561 170 561 170 561 170 561 170 561 175 561 175 561 175 561 175 561 175 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 847 180 823 180 799 180 778 180 741 185 875 185 853 185 832 185 814 185 785 190 906 190 887 190 869 190 854 190 833 195 941 195 924 195 910 195 898 195 886 200 979 200 966 200 955 200 948 200 945 205 1021 205 1011 205 1005 205 1002 205 1002 210 1067 210 1062 210 1060 210 1060 210 1060 215 1119 215 1118 215 1118 215 1118 215 1118

PTLR tor FNP Unit 2 Table 2*2 (continued)

Farley Unit 2 - 54 Cooldown Curve Data Points[21 (adjusted to include 60 psi at RCS temperatures ~ 11 oaF and 27 psi at RCS temperatures < 110°F) 40°F/hr. 60°F/hr.

T (OF) P (psi g) T (OF) P (psig) 220 1175 225 1238 230 230 1307 1307 235 235 1384 1384 240 1468 240 1468 1468 1562 245 1562 1562 250 1665 1665 255 1779 255 1779 260 1906 1906 265 265 2045 2045 270 2199 275 2370

PTLR for FNP Unit 2 3.0 Reactor Vessel Material Surveillance Program The reactor vessel material surveillance program is in compliance with 10 CFR 50, Appendix H, and is described in Section 5.4.3.6 of the Farley FSAR. Surveillance are the reported in accordance with ASTM E185-82 131

The removal schedule is provided in Table 3~1. Consistent with specific requirements for Farley Unit 2 with the grantinQ of an exemption to Appendix H of 10 50 documented in NUREG-0117[4~ Figures 2-1 and 2-2 are based on the greater, or limiting value, of the following: (1) the actual shift in reference temperature for plate 8721 as determined by impact testing, or (2) predicted shift in reference temperature for weld seam 11-923 as determined by Regulatory Guide 1 Revision 2[51. The neutron transport and dosimetry evaluation methodologies used follow guidance meet the requirements of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence"161. Results from the reactor surveillance program will used to update Figures 2-1 and 2-2 if the results indicate the adjusted reference temperature (ART) for the limiting beltline material exceeds the ART used to generate the prr limits shown in Figures and 2~2 for the specified fluence period.

Table SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE (a)

Capsule Capsule Location Lead Removal Fluence (e) EFPY ib)

(Degree) Factor (n/cm2)

U 343 3.26 1.11 6.05 x 1018 I W 110 2.84 3.96 1.73 x 1 19 X 287 3.38 6.43 2.98 x 10 Z 340 2.98 13.85 4.92 x 10 19 (d)

Y 290 3.12 19.01 6.79 x 10 19 (e)

V 107 3.58 i 21.82 8.73 x 1019 if)

Notes:

a) Data from Table 7-1, WCAP-16918-NP, Revision 1111 b) Effective Full Power Years from plant startup.

c) Plant-specific evaluation.

d) This fluence is no! less than once or greater than twice the peak EOL fluence for the inilial40'year license term.

e) This fluence is not less than once or greater than twice the peak EOL fluence for a 20'year license renewal term to 60 years.

f) This fluence is not less than once or than twice the EOL fluence for an additional license renewal term to 80 years.

PTLR for FNP Unit 2 4.0 Reactor Vessel Surveillance Data Credibility Regulatory Guide 1 Revision 2, describes general procedures NRC staff for calculating the effects of neutron radiation embrittlement of alloy steels currently used for light-water-cooled reactor vessels. Position of Regulatory Guide 1.99, Revision the methodology for calculating adjusted temperature and Charpy upper*shelf energy of reactor vessel beltline materials surveillance capsule data. The methods Position C.2 can only be applied when two or more credible surveillance sets become from the reactor in question.

Per WCAP-8956[SJ, the Unit 2 surveillance was based on E185-73[9J.

At! six surveillance capsules (U [101, WillI, X ,Z i 131, Y [14 1, and V [71) have been removed from the Farley Unit 2 reactor vessel and analyzed. In accordance with the discussion of Regulatory Guide 1.99, Revision there are five requirements that must met the surveillance data to judged credible.

The credibility for the Unit 2 surveillance weld are described below:

The credibility evaluation of Farley Unit 2 materials is documented in WCAP-16918-NP, Revision 1[7l. credibility evaluation concluded that the surveillance data for Intermediate Shell B721 is credibility evaluation concluded that the surveillance for weld Heat # BOLA is 5.0 Supplemental Data Tables Table contains a comparison of measured surveillance material ft-Ib transition temperature shifts and upper shelf energy decreases with Regulatory Guide 1.99, 2, predictions.

Table shows the calculation of the material chemistry factors using surveillance capsule data.

Table 5-3 provides the unirradiated Farley Unit 2 reactor toughness data.

Table 5-4 provides a summary of reactor vessel fluence values at 54 EFPY.

Table provides a summary of the values of the Unit 2 reactor materials at the 1 and 3/4-T locations for 54 Table 5-6 shows the calculation of the ART values at 54 EFPY for the limiting Farley Unit 2 reactor vessel material.

Table 5*7 provides values for Farley Unit 2 for 54

PTLR for FNP Unit 2 Revision 5 Page 14 of 22 Table 5-1 Comparison of Surveillance Material 30 ft-Ib Transition Temperature Shift and Upper Shelf Energy Decrease with Regulatory Guide 1.99, Revision 2, Predictions (a) 30 ft-Ib Transition Upper Shelf Energy Temperature Shift Decrease Material Capsul Fluence(e) Predicted Measure Predicted Measure e (x 1019 (OF) (b) d (OF) (c) (%) (b) d (%)(d) n/cm 2 , E>

1.0 MeV)

Intermediate Shell U 0.605 128.0 105.5 26 27 Plate B7212-1 W 1.73 171.5 167.7 33 22 (Longitudinal) X 2.98 192.1 164.8 37 26 Z 4.92 208 .5 200.1 42 28 Y 6.79 217.2 214.2 45 36 V 8.73 222.9 218.3 48 34 Intermediate Shell U 0.605 128.0 124.0 26 27 Plate B7212-1 W 1.73 171.5 168.5 33 21 (Transverse) X 2.98 192.1 200.1 37 28 Z 4.92 208.5 195.8 42 29 Y 6.79 217.2 231.0 45 42 V 8.73 222.9 215.3 48 27 SUNeillance U 0.605 32.8 -28.4 17 8 Program W 1.73 44.0 7.0 22 0 Weld Metal X 2.98 49.2 -15.6 24 0 Z 4.92 53.4 10.2 27 8 Y 6.79 55.7 69.1 30 5 V 8.73 57.1 56.5 32 14 Heat Affected Zone U 0.605 --- 219.8 --- 30 Material W 1.73 --- 268.8 --- 20 X 2.98 -- - 230.5 --- 19 Z 4.92 --- 263.8 --- 20 Y 6.79 -- 269.6 --- 35 V 8.73 -- 322.4 --- 25 NOTES:

(a) Data from Table 5-10, WCAP-16918-NP, Revision 1 17]

(b) Based on Reg. Guide 1.99, Rev. 2 methodology using the mean weight percent values of copper and nickel of the surveillance material.

(c) Calculated using measured Charpy data plotted using CVGRAPH, Version 5.3.

(d) Values are based on the definition of upper shelf energy given in ASTM E185-82.

(e) The fluence values presented here are the calculated values, not the best estimate values.

PTLR for FNP Unit 2 Revision 5 Page 15 of 22 Table 5*2 Calculation of Chemistry Factors Using Surveillance Capsule Data Capsule f (10 19 n/em2 , FF (8) aRTNDT FF*aRTNDT Material Capsule FF2 E> 1.0 MeV) (OF) CF)

U 0.605 0.859 105.5 90.7 0.738 Intermediate Shell Plate 87212-1 W 1.73 1.151 167.7 193.0 1.324 (Longitudinal) X 2.98 1.289 164.8 212.4 1.662 Z 4.92 1.399 200.1 280.0 1.958 Y 6.79 1.458 214.2 312.3 2.125 V 8.73 1.496 218.3 326.6 2.238 Intermediate Shell U 0.605 0.859 124.0 106.5 0.738 Plate 87212-1 W 1.73 1.151 168.5 193.9 1.324 (Transverse) X 2.98 1.289 200 .1 258.0 1.662 Z 4.92 1.399 195.8 274.0 1.958 Y 6.79 1.458 231.0 336.8 2.125 V 8.73 1.496 215.3 322.1 2.238 SUM: 2906.13 20.091 CF =t(FF

6RT NDT) + t(FF2) =144.6 OF Weld Metal U 0.605 0.859 0.0 (c) 0.0 0.738 W 1.73 1.151 7.0 (b) 8.1 1.324 X 2.98 1.289 0.0 (c) 0.0 1.662 Z 4.92 1.399 10.2 (b) 14.3 1.958 Y 6.79 1.458 69.1 (b) 100.7 2.125 V 8.73 1.496 56.5 (b) 84.5 2.238 SUM: 207.59 10.046 CF =t{FF

6RTNOT) + t(FF2) =20.7 OF NOTES: (a) FF = Fluence Factor = F (0.28*0.1log f)

(b) 6RTNOT values from Table 4-1 were not multiplied by the ratio of 0.96 (from WCAP-14689, Rev. 6(15J Table 4, CF vessel + CFsurv weld = 36.8 + 38.2 = 0.96) to calculate the best fit chemistry factor (CF) as provided by Reg. Guide 1.99, Rev. 2, Position 2.1, since the ratio is less than one. This is a conservative approach.

(c) Actual measured 6RTNOT values are less than zero. Since physically a reduction should not occur, a value of zero is conservatively used.

PTLR for FNP Unit 2 Revision 5 Page 16 of 22 Table 5-3 Reactor Vessel Toughness Table (Unirradiated) (a)

Beltline Material Cu Weight Ni Weight % IRTNDT (OF)

Closure Head Flange - - -60(d}

Vessel Flange - - 60 Inlet Nozzle B7218-1 0.16 0.71 32 Inlet Nozzle B7218-2 0.16 0.68 50 Inlet Nozzle B7218-3 0.16 0.72 60 Outlet Nozzle B7217-1 0.16 0.73 60 Outlet Nozzle B7217-2 0.16 0.72 6 Outlet Nozzle B7217-3 0.16 0.72 48 Upper Shell Forging B7216-1 0.16 0.724 30 Intermediate Shell Plate B7203-1 0.14 0.60 15 Intermediate Shell Plate B7212-1 0.20 0.60 -10 Lower Shell Plate B721 0-1 0.13 0.56 18 Lower Shell Plate B7210-2 0.14 0.57 10 Inlet/Outlet Nozzle to Upper Shell Girth Seams 10(0) 0.07 1.04 1-926 A-F Upper Shell to Intermediate Shell Circumferential 0.153 0.077 -40 Weld Seam 10-923 (b) (Heat # 5P5622)

Upper Shell to Intermediate Shell Circumferential -56(1) 0.05 1.0 Weld Seam 10-923 (Heat # 51922)

Upper Shell to Intermediate Shell Circumferential -56(1) 0.09 0.06 Weld Seam 10-923 (Heat # 3P4767)

Intermediate Shell Longitudinal Weld Seam -56(1) 0.027 0.947 19-923 A (b) (Heat # HODA)

Intermediate Shell Longitudinal Weld Seam 0.027 0.913 -60 19-923 B (b) (Heat # BOLA)

Surveillance Weld (e) 0.028 0.89 -

Intermediate Shell to Lower Shell Circumferential -40 0.153 0.077 Weld Seam 11-923 (b) (Heat # 5P5622)

Lower Shell Longitudinal Weld Seams -70 0.051 0.096 20-923 A & B (b) (Heat # 83640)

NOTES:

(a) From Table 4.2-1 of WCAP-17506-NP(161.

(b) Best-estimate copper and nickel from CE NPSD-1039 1171.

(c) The best-estimate copper and nickel value represents the average of two chemistry measurements performed on the surveillance weld and documented in WCAP-8956 181 and WCAP-11438 [III. The surveillance weld is representative of intermediate shell longitudinal weld 19-923B.

(d) Replacement closure head initial RT NOT value was taken from MHI-SNC-0455F2[181.

(e) An estimation method using measured data was used to determine this initial AT NOT value. Therefore, a conservative value of 17°F is used for Ou and OJ in margin calculations.

(f) These initial ATNOT values are generic and taken from 10 CFA 50.61 paragraph (c)(1 Hii) of the 1-1-07 edition.

PTLR for FNP Unit 2 Revision 5 Page 17 of 22 Table 5-4 Reactor Vessel Fluence Projections at 54 EFPY (a)

(10 19 n/cm 2 , E> 1.0 MeV)

Reactor Vessel Location Material 54 EFPY Neutron Fluence Inlet Nozzle 87218-1 0.0449 Inlet Nozzle 87218-2 0.0254 i Inlet Nozzle 87218-3 0.0186 Outlet Nozzle 87217-1 0.0126 Outlet Nozzle 87217-2 0.0172 Outlet Nozzle 87217-3 0.0304 Upper Shell Forging 87216-1 1.09 87203-1 &

Intermediate Shell Plates 5.76 87212-1 87210-1 &

Lower Shell Plates 5.75 87210-2 Inlet/Outlet Nozzle to Upper Shell 1-926 A-+F 0.0449 Girth Seams Upper Shell to Intermediate Shell 10-923 1.09 Circumferential Weld Seam Intermediate Shell Longitudinal 19-923 A & 8 1.83 Weld Seams Intermediate Shell to Lower Shell 11-923 5.75 Circumferential Weld Seam Lower Shell Longitudinal Weld 20-923 A & 8 1.83 Seams i

NOTE:

(a) From Table 5.2-1 of WCAP-17506-NP(l61* These values are also summarized in Table 2-1 of Attachment B of ALA-09-116(191*

PTLR for FNP Unit 2 Revision 5 Page 18 of 22 Table 5-5 Summary of ART Values for the Reactor Vessel Materials at the 1/4-T and 3/4-T Locations for 54 EFPY (a)

Material 1/4-T (OF) 3/4-T (oF)

Inlet Nozzle 87218-1 83 60 Inlet Nozzle 87218-2 86 69 Inlet Nozzle 87218-3 89 75 Outlet Nozzle 87217-1 83 71 Outlet Nozzle 87217-2 34 20 Outlet Nozzle 87217-3 88 69 Upper Shell Forging 87216-1 172 141 Intermediate Shell Plate 87203-1 182 158 Intermediate Shell Plate 87212-1 223 187 Intermediate Shell Plate 87212-1 200(b) 16S(b)

Using credible SIC Data Lower Shell Plate 87210-1 172 150 Lower Shell Plate 87210-2 175 152 InleVOutlet Nozzle to Upper Shell Girth Seams 69 57 1-926 A .....F Upper Shell to Intermediate Shell Circumferential 82 55 Weld Seam 10-923 (Heat # 5P5622)

Upper Shell to Intermediate Shell Circumferential 70 42 Weld Seam 10-923 (Heat # 51922)

Upper Shell to Intermediate Shell Circumferential 39 19 Weld Seam 10-923 (Heat # 3P4767)

Intermediate Shell Longitudinal Weld Seam 33 17 19-923 A (Heat # HODA)

Intermediate Shell Longitudinal Weld Seam 16 -3 19-923 B (Heat # 80LA)

Intermediate Shell Longitudinal Weld Seam 19-923 B (Heat # 80LA) -17 -28 Using non-credible SIC Data Intermediate Shell to Lower Shell Circumferential 115 97 Weld Seam 11-923 (Heat # 5P5622)

Lower Shell Longitudinal Weld Seams -12 7

20-923 A & 8 (Heat # 83640)

NOTES:

(a) The ART values presented here are based on the reactor vessel surface fluence values summarized in Table 5-4.

The values for the beltline materials are from Tables 4-10 and 4-11 of WCAP-17123-NP, Revision 1(21. The values for the extended beltline materials are summarized along with the values for the belt/ine materials in Tables 3-3 and 3-4 of Attachment 8 of ALA-09-116(191.

(b) Limiting 1/4-T and 3/4-T ART values. The Pff limit curves are based on these limiting ART values of 200°F and 165°F.

PTLR for FNP Unit 2 Revision 5 Page 19 of 22 Table 5-6 Calculation of Adjusted Reference Temperature Values at 54 EFPY for the Limiting Reactor Vessel Material (a)

Parameter Intermediate Shell Plate 67212-1 Operating Period 54 EFPY Location %-T %-T Chemistry Factor, CF (OF) 144.6 144.6 Fluence, f (10 19 n/cm 2) (h) 3.591 1.396 Fluence Factor, FF 1.3324 1.0926 LlRT NOT = CF x FF (OF) 192.7 158.1 Initial RT NOT , I (OF) -10 -10 Margin, M (OF) (c) 17 17 Adjusted Reference Temperature 200(d) 165(d)

(ART), (oF) per Regulatory Guide 1.99, Revision 2 NOTES:

(a) From Tables 4-10 and 4-11 ~using credible surveillance capsule data) of WCAP-17123*NP, Revision 1 121.

2

=

(b) Fluence is based on fsurt (10 9 n/cm , E> 1.0 MeV) 5.76. The Farley Unit 2 reactor vessel wall thickness is 7.875 inches in the beltline region.

=

(c) Margin is calculated as M 2(0;2 + 0/) o.s. The standard deviation for the initial RT NOT margin term, 01, is OaF since the initial RTNOT is a measured value. The standard deviation for the 6RTNOT term, 06, is 1]oF for the plate, except that 0/1 need not exceed 0.5 times the mean value of 6RTNOT. In accordance with Regulatory Guide 1.99, Revision 2, Position 2.1, values of 0d may be cut in half when based on credible surveillance data.

(d) Limiting !/.I-T and %-T ART values.

PTLR for FNP Unit 2 Revision 5 Page 20 of 22 Table 5-7 Pressurized Thermal Shock (RTpts) Values for S4 EFPY (a)

Surface Fluence ARTNDT Material FF (CF x FF) I M RT pTS CF (10 19 n/cm2 ,

(OF) (OF) (OF) (OF)

E> 1.0 MeV)

Inlet Nozzle 87218-1 120.75 0.0449 0.2760 33.3 32 33.3 99 Inlet Nozzle 87218-2 120 0.0254 0.1990 23.9 50 23.9 98 Inlet Nozzle 87218-3 121 0.0186 0.1644 19.9 60 19.9 100 Outlet Nozzle 87217-1 121.25 0.0126 0.1280 15.5 60 15.5 91 Outlet Nozzle 87217-2 121 0.0172 0.1565 18.9 6 18.9 44 Outlet Nozzle B7217-3 121 0.0304 0.2213 26.8 48 26.8 102 Upper Shell Forging 87216-1 121.1 1.09 1.0241 124.0 30 34.0 188 Intermediate Shell Plate B7203-1 100.0 5.76 1.4292 142.9 15 34.0 192 Intermediate Shell Plate 87212-1 149.0 5.76 1.4292 213.0 -10 34.0 237 Intermediate Shell Plate B7212-1 214(b) 144.6 5.76 1.4292 206.7 -10 17.0 Using credible SIC Data Lower Shell Plate B7210-1 89.8 5.75 1.4289 128.3 18 34.0 180 Lower Shell Plate B721 0-2 98.7 5.75 1.4289 141.0 10 34.0 185 Inlet/Outlet Nozzle to Upper Shell 95 0.0449 0.2760 26.2 10 42.9 79 Girth Seams 1-926 A->F Upper Shell to Intermediate Shell Circumferential Weld Seam 10-923 74.1 1.09 1.0241 75.9 -40 56.0 92 (Heat # 5P5622)

Upper Shell to Intermediate Shell Circumferential Weld Seam 10-923 68 1.09 1.0241 69.6 -56 65.5 79 (Heat # 51922)

Upper Shell to Intermediate Shell Circumferential Weld Seam 10-923 46.3 1.09 1.0241 47.4 -56 58.3 50 (Heat # 3P4767)

Intermediate Shell Longitudinal Weld -56 54.7 42 36.8 1.83 1.1657 42.9 Seam 19-923 A (Heat # HODA)

Intermediate Shell Longitudinal Weld 26 36.8 1.83 1.1657 42.9 -60 42.9 Seam 19-923 B (Heat # BOLA)

Intermediate Shell Longitudinal Weld Seam 19-923 B (Heat # BOLA) 20.7 1.83 1.1657 24.1 -60 24.1 -12 Using non-credible SIC Data Intermediate Shell to Lower Shell Circumferential Weld Seam 11-923 74.1 5.75 1.4289 105.9 -40 56.0 122 (Heat # 5P5622)

Lower Shell Longitudinal Weld -70 43.5 17 37.3 1.83 1.1657 43.5 Seams20-923 A & 8 (Heat # 83640)

NOTES:

(a) From Table 7.2-1 of WCAP-17506-NP!161*

(b) This limiting RTpTs value was calculated using the CF from the surveillance data and a reduced 06 margin of 8.5°F, since this surveillance data is credible.

PTLR for Ft\IP Unit 2 6.0 References

1. WCAP-14040-A, Revision 4, Methodology Used to Develop Mitigating System RCS Heatup and Cooldown Umit Curves, May 2004 .

2. .:...:...:::::..:....::!--!..!....!..!::..::::..!..!.!-' Revision 1, J. M. Farley Unit 2 Heatup Cool down Umit for Normal Operation, July 2011.

3. ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," in ASTM Standards, 3, Society Testing Materials, 1982.

4. NUREG-0117, Supplement 5 to Safety Evaluation Report (NUREG-75/034),

Office of Nuclear Reactor Regulation, U. S. Regulatory Commission in matter of Power Company M. Farley Nuclear Plant Unit 2, Docket No. 50-364., March 19,1981.

5. ~~~;L.:::!.!.l~~~~..!.:..:::r.:':" Revision 2. "Radiation Embrittlement of Reactor "May 1988.

6. NRC Regulatory Guide 1.190, "Calculational Dosimetry Methods for Determining Pressure Neutron Fluence," March, 2001.

~~-...:...x.:::..~~' Revision 1. Analysis of V from the Southern Nuclear Operating Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program, April 2008.

8. WCAP-8956, Alabama Power Company Joseph M. Farley Nuclear Unit No.2 Reactor Vessel Radiation Surveillance Program, J. A. Davidson, et al., August 1977.

9. ASTM E185-73, Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels, American Society for Testing and Materials, 1973.

10. Analysis of Capsule U from Alabama Company M.

Farley 2 Reactor Radiation Surveillance Program, E. Yanichko, et October 1983.

11. Analysis of W from the Alabama Power Company Joseph Farley Unit 2 Reactor Vessel Radiation Surveillance S. E. Yanichko. et 1

Capsule X from the Power Company Joseph M.

Radiation Surveillance Program, E. et aI.,

13. Revision 1, of Capsule Z from Power Company Farley Unit 2 Reactor Vessel Radiation Surveillance Program, 2000.

14. Revision 1, Analysis of Capsule Y from the Southern Nuclear Operating Company Joseph M. Farley Unit 2 Reactor Radiation Surveillance Program, June 2008.

15. Revision 6, Farley Units 1 and 2 Heatup and Cooldown Limit Curves Normal Operation and PTlR Support Documentation, T. J. laubham, April 2001.

16. WCAP-17506-NP, Revision 0, Farley Units 1 and 2 Pressurized Thermal Shock Evaluations, A. Rosier, December 1.

17. CE NPSD-1039, Revision 2, Copper and Nickel Values In CE Fabricated Reactor Vessel Welds, Combustion Engineering Owners Group, June 1997.

18. Mitsubishi Heavy Industries, lTD, Kobe Shipyard & Machinery Works (MHI).

!.!!!..~~~!.i:Li::i:.!....!i:., Reactor Vessel Closure for Farley-2, Certified Material June 2004.

19. Westinghouse letter ALA-09-116, P-T limit Curves with Margins for Instrumentation Errors and Extended Beltline Material Information, John M.

Robinson, October 20,

Joseph M. Farley Nuclear Plant Request to Revise Technical Specifications Associated with the Low Temperature Overpressure Protection System and the Pressure and Temperature Limits Report Enclosure 5 WCAP-17122-NP, Revision 0, "J. M. Farley Unit 1 Heatup and Cool down limit Curves for Normal Operation," October 2009

Westinghouse Non-Proprietary Class 3 WCAP-17122-NP October 2009 Revision 0 J. M. Farley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation (8 Westinghouse

NON-PROPRIETARY CLASS 3 WCAP-17122-NP Revision 0 J. M. Farley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation A.E.

C. C. Heinecke*

Aging "PT'n"",t and License Renewal Services October 2009 Reviewer: B. A. Rosier*

Aging Management and License Renewal Services Approved:

Electronically approved records are authenticated in the electronic document management system.

Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355

ii RECORD OF REVISION Revision 0: Original Issue WCAP-17122-NP Revision 0

TABLE OF CONTENTS LIST OF TABLES ....................................................................................................................................... iv LIST OF FIGURES ..................................................................................................................................... vi EXECUTIVE Vll INTRODUCTION ........................................................................................................................ 1-1 2 FRACTURE TOUGHNESS PROPERTlES ................................................................................. 2-1 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS ................ 3-1 3.1 OVERALL APPROACH ................................................................................................. 3-1 3.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT ............................................................................................................ 3-1 3.3 CLOSURE HEADNESSEL FLANGE REQUIREMENTS ........................................... 3-S 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE ......................................... .4-1 S HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES....................... S-I 6 REFERENCES ............................................................................................................................. 6-1 APPENDIX A Thermal Stress un"",,,,,,] Factors WCAP-17I 22-NP Revision 0

LIST OF TABLES Table 2-1 of the Best Estimate Cu and Ni Weight Percent and Initial Values for the Unit I Reactor Vessel Beltline Materials ..................................................... 2-2 Table 2-2 Summary of the Initial RT NOT Values for the 1. M. Farley Unit I Closure Head and Vessel Flange ........................................ ,............. , ........................... ,., ......................................... 2-3 Table 2-3 Summary of the 1. M. Unit I Reactor Vessel Beltline Material Factors Per Regulatory Guide 1.99, Revision 2 ........................................................................... 2-3 Table 4-1 Fluence Values the J. M. Farley Unit I Reactor Vessel Beltline Materials ............... .4-2 Table 4-2 Fluence Values for the Vessel Surface, 1/4T and 3/4T Locations for the J. M. Farley Unit I Reactor Vessel Beltline Materials at 36 EFPY .............................................................. 4-3 Table 4-3 Fluence Values for the Vessel Surface. 1I4T and 3/4T Locations for the J. M. Unit 1 Reactor Vessel Beltline Materials at 54 EFPY .............................................................. 4-3 Table 4-4 Fluence Values for the Vessel 1I4T and 3/4T Locations for the J. M. Unit I Reactor Vessel Beltline Materials at 72 EFPY .............................................................. 4-3 Table 4-5 Fluence Factor Values at the il4T and 3/4T Locations for the 1. M. Unit I Reactor Vessel Bel tline Materials at 36 EFPY ............. ,............................................................... .4-4 Table 4-6 Fluence Factor Values at the 1/4T and 3/4T Locations for the j, M, Farley Unit I Reactor Vessel BeltEne Materials at 54 EFPY ........ ".".. " .... " ............................... " .. " ............. ".,,4-4 Table 4-7 Fluence Factor Values at the 1/4T and 3/4T Locations for the 1. M. Unit 1 Reactor Vessel Beltline Materials at 72 EFPY ..... "."" .." ........... " ......... , .............................. " ....... 4-4 Table 4-8 Adjusted Reference Evaluation for the 1. M. Farley Unit I Reactor Vessel Beltline Materials through 36 EFPY at the 1I4T Location ..... " .., ........................... " ...... .4-5 Table 4-9 Adjusted Reference Evaluation for the 1. M, Unit I Reactor Vessel Beltline Materials 36 EFPYat the 3/4T Location Table 4-10 Adjusted Reference Evaluation for the j, M. Farley Unit I Reactor Vessel Beltline Materials through 54 EFPY at the 1/4T Location ............................... "." ........ ..4-7 Table 4-11 Reference Evaluation for the 1. M, Unit 1 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location ............... " ...... " ...... "" ........".4-8 Table 4-12 Adjusted Reference Evaluation for the 1. M. Farley Unit 1 Reactor Vessel Beltline Materials 72 EFPYat the 1/4T Location ........ " ...... , ... " .."."",,............. 4-9 Table 4-13 Adjusted Reference Temperature Evaluation for the J, M. Farley Unit I Reactor Vessel Beltline Materials through 72 EFPY at the 3/4T Location " ...."""."..."".. "" ............ ",,4-10 Table 4-14 Summary of the Limiting ART Values Used in the Generation of the 1. M. Farley Unit I Heatup/Cooldown Curves ........ " .. " ... ,..... " ........ " ... " .......... ""., ...,""'" ..... "., ......... , ... ,.... 4-11 WCAP-17122-NP October Revision 0

.... r"'~rl""'ru Class 3 v Table 5-1 36 EFPY Heatup Curve Data Points Using the 1998 through the 2000 Addenda App. G Methodology (wI wI Flange Notch, wI Pressure Correction and wlo Uncertainties for Instrumentation ........................................ 5-9 Table 5-2 36 EFPY Cooldown Curve Data Points Using the 1998 through the 2000 Addenda App.

G Methodology {wI wI Notch, wI Pressure Correction and wlo Uncertainties for Instrumentation ............................................................................................ 5-11 Table 5-3 54 EFPY Heatup Curve Data Points the 1998 through the 2000 Addenda App. G Methodology (wI K1c

wi Flange Notch, wi Pressure Correction and w/o Uncertainties for Instrumentation 3 Table 5-4 54 EFPY Cooldown Curve Data Points the 1998 through the 2000 Addenda App.

G Methodology (wI wi Flange Notch, wi Pressure Correction and wlo Uncertainties for Instrumentation ............................................................................................ 5-15 Table 5-5 72 EFPY Heatup Curve Data Points Using the 1998 through the 2000 Addenda App. G Methodology (wI K1e

wi Flange Notch, wi Pressure Correction and wlo Uncertainties for Instrumentation Errors) .................................................................................................. 5-17 Table 5-6 72 EFPY Cooldown Curve Data Points Using the 1998 through the 2000 Addenda App.

G Methodology (wi K lc, wI Flange wI Pressure Correction and wlo Uncertainties for Instrumentation Errors) ............................................................................................ 5-19 Table A-I Values for 36, and 72 EFPY 100°FIhr Heatup Curves (w/o Margins for Instrument Errors) ............................................................................................................................. A-2 Table A-2 Kif Values for 36, 54, and 72 EFPY I OO°F/hr Cooldown Curves (w/o Margins for I nstrumen tErrors) ..

WCAP-17122-NP Revision 0

LIST OF FIGURES Figure 5-1 J. M. Farley Unit 1 Reactor Coolant System Limitations (Heatup Rates of 60 and lOO°F/hr) Applicable for 36 EFPY (without Margins for Instrumentation Errors and with Pressure Using the 1998 through the 2000 Addenda App. G Methodology (wIKlc) .............................................................................................................................. 5-3 5-2 J. M. Farley Unit I Reactor Coolant Cooldown Limitations (Cooldown Rates up to 1OO°Flhr) Applicable for 36 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) the J998 through the 2000 Addenda App. G Methodology (wIKle) ....................................................................................................... 5-4 J. M. Unit I Reactor Coolant System Limitations (Heatup Rates of 60 and 100°F/hr) Applicable 54 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda App. G Methodology (w/K 1c) ..............................................................................................................................5-5 Figure 5-4 J. M. Unit J Reactor Coolant Cooldown Limitations (Cool down Rates up to IOO°F/hr) Applicable for 54 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) the 1998 through the 2000 Addenda G Methodology (WIKle) ....................................................................................................... 5-6 5-5 J. M. Farley Unit I Reactor Coolant System Heatup Limitations (Heatup Rates of60 and 100°F/hr) Applicable for 72 EFPY (without Margins for Instrumentation Errors and with Pressure the I through the 2000 Addenda App. G Methodology (wIKle) .............................................................................................................................. 5-7 5-6 J. M. Farley Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°FIhr) Applicable for 72 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda G Methodology (w/Kl~)

WCAP-17122-NP Revision 0

EXECUTIVE

SUMMARY

This report provides the and results of the generation of heatup and cooldown pressure temperature (P-T) limit curves for normal operation of the 1. M. Farley Unit I reactor vessel. The heatup and cooldown p.T limit curves were using the highest adjusted reference temperature (ART) values pertaining to J. M. Unit I . The highest ART values pertaining to an axial weld or a plate/forging were those of lower shell plate B6919-1 (using surveillance data) at both 114 thickness (I/4T) and 3/4 thickness (3/4T) locations. The P-T curves made use of the K1c methodology detailed in the 1998 through the 2000 Addenda Edition of the ASME Code,Section XI, Appendix G, and ASME Code Case N-641.

The P-T limit curves were for 54 and 72 heatup rates of 60 and 100°F/hr, and cooldown rates of 0, 20, 40, 60 and 100 °Flhr. The curves were developed without margins for instrumentation errors. The curves include a pressure correction for the static and dynamic head loss between the reactor vessel beltline and the Residual Heat Removal (RHR) relief valves. These curves can be found in 5-1 through 5-6. Appendix A contains the thermal stress intensity factors for the maximum heatup and cooldown rates for each EFPY term.

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1 INTRODUCTION Heatup and cool down P- T limit curves are calculated the adjusted RTNDT (reference nil-ductility temperature) to the limiting beltline material of the reactor vesseL The RT NDT of the limiting material in the core region of the reactor vessel is detennined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ART NDT, and adding a The unirradiated is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-Ib of impact energy and 35-mil lateral expansion (normal to the working direction) minus 60°F.

increases as the material is exposed to fast-neutron radiation. to find the most limiting RT NDT at any time period in the reactor's life, ART NDT due to the radiation exposure associated with that time must be added to the unirradiated RT NDT (IRT NDT). The extent of the shi ft in is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, "Radiation Embriulement of Reactor Vessel Materials" I],

Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRT NOT + ART NDT + margins for uncertainties) at the 114T and 3/4T locations, where T is the thickness of the vessel at the beltline measured from the clad/base metal interface.

The heatup and cooldown poT limit curves documented in this report were generated using the most limiting ART values and the NRC approved methodology documented in WCAP-14040-A, Revision 4

[Reference 2J, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves." Specifically, the methodology of the 1998 through the 2000 Addenda Edition of ASME Section Appendix G 3] was used.

The calculated ART values for 36, and 72 EFPY are documented in Tables 4-8 4-13 of this report. The basis fluence projections are based on the values verified by Westinghouse in letter LTR-REA-09-112, Revision I (Reference 4].

The purpose of this report is to present the calculations and the development of the 1. M. Farley Unit heatup and cooldown P-T limit curves for 36, 54 and 72 EFPY. This report documents tlte calculated ART values and the development of the P- T limit curves for nonnal operation. The P-T curves herein were generated without instrumentation errors. The P-T curves contain a pressure correction for the static and dynamic head loss between the reactor vessel beltline and the RHR relief values. The P- T curves include limits for the vessel per the of IO CFR Part Appendix G [Reference 5J.

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3 2 FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic materials in the J. M. Farley Unit I reactor vessel are in Table 2-1. The unirradiated RTNOT values for the closure head and vessel are documented in Table 2-2.

The Regulatory Guide I Revision 2 methodology used to the heatup and cooldown P-T limit curves documented in this report is the same as that documented in WCAP-14040-A, Revision 4

[Reference The chemistry factors (CFs) were calculated using Regulatory Guide 1.99 Revision 2, Position 1.1 and 2.1. Position 1.1 uses the tables from the Regulatory Guide along with the best estimate copper and nickel weight which are presented in Table 2-1. Position 2.1 CFs are calculated based on the Charpy testing of irradiated surveillance capsule specimens. Table 2-3 summarizes the Position 1.1 and 2.1 CFs determined for the J. M Farley Unit I beltline materials.

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Westinghouse Non-Proprietary Class 3 2-2 Table 2-1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RT NDT Values for the J. M. Farley Unit 1 Reactor Vessel Beltllne Materials Fracture Material Description (1 ) Chemical Composition (1) Toughness Propertyc,)

Cu Ni Initial RT NDT Reactor Vessel Location Material ID #

wI. 0/0 wt. % CF)

Intermediate Shell (IS) Plate 86903-2 0.13 0.60 0 Intennediate Shell Plate 86903-3 0.12 0.56 10 Lower Shell (LS) Plate 86919-1 0.14 0.55 IS Lower Shell Plate 86919-2 0. 14 0.56 5 19-894 A & 8 _56<<)

IS Longitudinal Weld Seams 0.258 0.165 (Heat # 33 A277)

Surveillance Program Weld Metal(o)

(Heat # 33A277) 0.14 0.19 -- 11-894 _56<<)

IS to LS Circ. Weld Seam 0.205 0.105 (Heat # 6329637)

LS Longitudinal Weld 20-894 A& 8 _56<<)

0.197 0.06 Seams (Heat # 90099)

Notes for Table 2-1 :

(a) Information source for these. material properties is ALA-08-7S, Revision ° [Reference 61.

(b) Surveillance weld is representative of intermediate shelliongirudinal welds 19-894 A & B. Best estimate copper and nickel values represent a single chemical analysis documented in WCAP-88I 0, Revision ° [Reference 7] .

(c) Per ALA-08-75, Revision 0, all weld initial RT NOT values are generic, and are taken from 10 CFR 50.61 paragraph (c)(I)(ii) of the \-1-07 edition.

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Westinghouse Non-Proprietary Class 3 2-3 Table 2-2 Summary of the Initial RT NDT Values for the J. M. Farley Unit 1 Closure Head and Vessel Flange (a)

Material Identification Initial RT NDT Closure Head Flange -50 of Vessel Flange 60 OF Notes for Table 2-2:

(a) Infonnation source for the initial RT NDT values is ALA-08-75, Revision 0 [Reference 6).

Table 2-3 Summary of the J. M. Farley Unit 1 Reactor Vessel Beltline Material Chemistry Factors Per Regulatory Guide 1.99, Revision 2 CFper CFper Vessel Material Material Position 1.1 Position 2.1 CF) CF)

Intennediate Shell Plate B6903-2 91 ---

Intennediate Shell Plate B6903-3 82.2 --

Lower Shell Plate B6919-1 97.8 106.7 Lower Shell Plate B6919-2 98.2 --

IS Longitudinal Weld Seams19-894 A & B 126.3 118.5 SUlveillanee Program Weld Metal 19-894A& B 78.1 --

IS to LS Cire. Weld Seam 11-894 98.4 --

LS Longitudinal Weld Seams20-894 A& B 91.4 --

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3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3.1 OVERALL APPROACH The ASME for calculating the allowable limit curves for various heatup and cooldown rates that the total stress intensity factor, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be than the reference stress intensity KJc. for the metal temperature at that time. K,c is obtained from the reference fracture toughness curve, defined in the 1998 Edition through the 2000 Addenda of Section Appendix G of the ASME Code [Reference 3].

The curve is given by the following equation:

K[c =33.2+20.734 (1)

where, reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This curve is based on the lower bound of static critical KI values measured as a function of temperature on of SA-533 Grade B Class I, SA-508-1, SA-508-2, and SA-508-3 steeL 3.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT The equation for the heatup-cooldown is defined in Appendix G of the ASME Code as follows:

stress intensity factor caused by membrane (pressure) stress KII stress intensity factor caused by the thermal reference stress factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT C 2.0 for Level A and Level B service limits C 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical WCAP-17122-NP Revision 0

For membrane tension, the corresponding Kr for the postulated defect is:

K 1m == Mm X (pRi I t) (3) where, Mm for an inside surface flaw is Mm 1.85 for Ji 2, Mm for 2:::; Ji : :; 3.464 ,

Mm 3.21 for Ji > 3.464 Similarly, Mm for an outside surface flaw is given by:

1.77 for < 2, 0.893 Ji for 2 S; :::; 3.464, Mm 3.09 for Ji > 3.464 and p = internal pressure (ksi), Ri := vessel inner radius (in.), and t vessel wall thickness (in.).

For U""'UHI'l'. stress, the ",,,,,,,,p,,nont1, for the postulated defect is:

Klb = Mb

Maximum Stress, where Mb is two-thirds of Mm (4)

The maximum Kr produced radial thermal gradient for the postulated inside surface defect of 0-2120 is:

O.953xIO* 3 x CR x r's (5) where CR is the cooldown rate in or for a postulated outside surface defect (6) where HU is the heatup rate in °Flhr.

The through-wall difference associated with the maximum thermal can be determined from ASME Code,Section XI, G, 0-22]4-1. The temperature at any radial distance from the vessel surface can be determined from ASME Section Xl, Appendix 0, 0-2214-2 for the maximum thermal Kr.

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(a) The maximum thermal KI relationship and the relationship in G-2214-1 are applicable only for the conditions in G-2214.3(a)(l) and (2).

(b) Alternatively, the K for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time cooldown for a Y4-thickness inside surface defect using the relationship:

Kit (1.0359Co + 0.63220 + 0.4753C2 + 0.3855CJ) * (7) or similarly, Kit during heatup for a Y4-thickness outside surface defect using the relationship:

Kif:::: (1.043Co + O.630Cl + 0.48 (8) where the coefficients Co, C 1, and are determined from the thermal stress distribution at any time the heatup or cooldown the form:

(9) and x is a variable that the radial distance (in.) from the (Le., inside or outside) surface to any point on the crack front, and a is the maximum crack depth (in.).

Luual,IVI'" 3, 7, and 8 were implemented in the OPERLIM computer which is the program used to the pressure-temperature (P-T) limit curves. The P-T curve methodology is the same as that described in WCAP-14040-A, Revision 4 "Methodology Used to Develop Cold Overpressure and RCS Heatup and Cooldown Limit Curves" 2] Section 2.6 (equations 2.6.2-4 and 2.6.3-\).

At any time during the heatup or cooldown is determined by the metal temperature at the tip of a postulated flaw (the postulated flaw has a depth of 1I4 of the section thickness and a length of 1.5 times the section thickness per ASME Code,Section XI, paragraph G-2120), the appropriate value for and the reference fracture toughness curve (Equation I). The thermal stresses resulting from the through the vessel wall are calculated and then the (thermal) stress intensity Kit, for the reference flaw are computed, From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference 114T flaw G to Section XI of the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the vessel wall because the thermal gradients, which increase with cooldown rates, produce tensile stresses at the inside surface that would tend to open (propagate) the existing flaw. Allowable pressure-temperature curves are for steady-state (zero-rate) and each finite cooldown rate specified. From these curves, limit curves are constructed as the minimum of the steady-state or finite rate curve for each cooldown rate specified.

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The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is on the material at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows at any reactor coolant the L'1T across the vessel wall developed during cooldown results in a higher value of Kit at the 1/4T location for finite cooldown rates than for operation. if conditions exist so that the increase in K lc exceeds Klb the calculated aJIowable pressure will be than the value.

The above procedures are needed because there is no direct control on t"",,,,,,r,,h at the l/4T location and, allowable pressures could be lower if the rate of cooling is decreased at various intervals a cooldown ramp. The use of the composite curve eliminates this and ensures conservative operation of the system for the entire cooldown period.

Three calculations are required to detennine the limit curves for finite rates. As is done in the cooldown analysis, allowable relationships are developed for conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal at the crack lip the coolant temperature; therefore, the for the inside 1/4T flaw during heatup is lower than the Kic for the flaw during steady state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of thennal stresses and lower values do not offset each other, and the curve based on conditions no a lower bound of all similar curves for finite heatup rates when the 114T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that al any coolant the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The third portion of the heatup analysis concerns the calculation of the limitations for the case in which a 1J4T flaw located at the 1I4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be on an individual basis.

Following the generation of curves for the steady-state and finite heatup rate situations, the fmal limit curves are produced constructing a composite curve based on a of the steady-state and finite heatup rate data. At any given the allowable pressure is taken to be the least of the three values taken from the curves under consideration, The use of the composite curve is necessary to set conservative heat up limitations because it is possible for conditions to exist over the course of the ramp. the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

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II 3

3.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS 10 CFR Part 50, Appendix G [Reference 5J addresses the metal temperature of the closure head and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNDT by at least 120°F for normal operation when the pressure exceeds 20 percent of the hydrostatic test pressure (3107 for J. M. Farley Unit I), which is calculated to be 621 The limiting unirradiated RTNOT of 60°F occurs in the vessel flange of the J. M.

Unit I reactor so the minimum allowable temperature of this is 180°F at pressures greater than 621 (without instrument uncertainties). This limit is shown in Figures 5-1 through 5-6 wherever applicable.

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4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference renmeranLlre (ART) for each material in the beltline region is given by tbe following expression:

ART =: Initial RTNDT + LlRTNOT + (10)

Initial RTNOT is the reference temperature for the unirradiated material as defined in paragraph NB-233I of Section III of the AS ME Boiler and Pressure Vessel Code 8]. If measured values of the initial RTNDT for the material in question are not available, generic mean values for that class of material may be used, provided if there are sufficient test results to establish a mean and standard deviation for the class.

MTNOT is the mean value of the adjustment in reference temperature caused irradiation and should be calculated as follows:

LlRTNOT=CF* r028.0.10Iogf) (11)

To calculate LlRTNDT at any depth al 114T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.

e (*0.24.) (12) where x inches (vessel beltline thickness is 7.875 inches) is the depth into the vessel wall measured from the vessel cladfbase metal interface. The resultant fluence is then placed in Equation I I to calculate the LlRT NOT at the depth.

The Radiation Analysis Group evaluated the vessel fluence projections in LTR-REA-09 112, Revision I 4], and the results are in Table 4-1. The evaluation methods used in Reference 4 are consistent with the methods presented in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Selpoints and RCS and Cooldown Limit Curves" [Reference Tables 4-2 4-4 provide a summary of the vessel flucnce projections at the 1/4T and 3/4T locations for 36, 54 and 72 EFPY. Tables 4-5 through 4-7 contain the 1I4T and 3/4T calculated fluences and fluence factors, per Regulatory Guide 1.99, Revision 2, used to calculate the 36, 54 and 72 EFPY ART values for all beltline materials in the J. M. Farley Unit 1 reactor vessel.

Margin is calculated as M = 2 . The standard deviation for the initial RT NDT margin term (O'i) is O"F when the initial RTNDT is a measured value, and 17°F when a generic value is available. The standard deviation for the LlRTNOT margin term, 0'"" is 17°F for or and 8.5°F for plates or when credible surveillance data is used. For welds, O't. is equal to 28°F when surveillance capsule data is not used, and is 14°F (half the when credible surveillance capsule data is used. The value for O't.

need not exceed 0.5 times the mean value of LlRTNOT.

Contained in Tables 4-8 through 4-13 are the 36, 54 and 72 EFPY ART calculations at the 1I4T and 3/4T locations for of the J. M. Farley Unit I heatup and cooldown curves.

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Westinghouse Non-Proprietary Class 3 4-2 Table 4-1 Fluence Values for the J. M. Farley Unit 1 Reactor Vessel Beltline Materials Neutron Fluence (n/cm2, E > 1.0 MeV1 Reactor Vessel Location Material 36 EFPY 54 EFPY 72 EFPY B6903-2 Intermediate Shell Plates 4.02E+19 5.93E+19 7.84E+19

& B6903-3 B6919-1 Lower Shell Plates 3.98E+19 5.81E+19 7.65E+19

& B6919-2 19-894 IS Longitudinal Weld Seams 1.23E+l9 1.83E+19 2.42E+19 A&B IS to LS Cire. Weld Seam 11-894 3.98E+19 5.81E+19 7.65E+19 20-894 LS Longitudinal Weld Seams 1.21E+19 1.79E+19 2.36E+19 A&B WCAP-17122-NP October 2009 Revision 0

Westinghouse Non-Proprietary Class 3 4-3 Table 4-2 Fluence Values for the Vessel Surface, 1I4T and 3/4T Locations for the J. M. Farley Unit 1 Reactor Vessel 8eltline Materials at 36 EFPY Fluence, f 1I4T f 3/4 T f Region (xlO l9 nlcm 2, (xlO '9 nlcm l , (x 10 19 n/cm 2 ,

E> 1.0 MeV) E> 1.0 MeV) E> 1.0 MeV)

Intennediate Shell Plates 4.02 2.506 0.974 Lower Shell Plates 3.98 2.481 0.964 IS Longitudinal Weld Seams 1.23 0.767 0.298 IS to LS Cire. Weld Seam 3.98 20481 0.964 LS Longitudinal Weld Seams 1.21 0.754 0.293 Table 4-3 FJuence Values for the Vessel Surface, 1/4T and 3/4T Locations for the J. M. Farley Unit I Reactor Vessel 8eltJine Materials at 54 EFPY Fluence, f 114 T f 3/4 T f Region (xl0 19 nlcm 2, (xl0 19 n/cm 2, (x10 19 n/cm 2, E> 1.0 MeV) E> 1.0 MeV) E> 1.0 MeV)

Intennediate Shell Plates 5.93 3.697 10437 Lower Shell Plates 5.81 3.622 10408 IS Longitudinal Weld Seams 1.83 1.I4l 0.443 IS to LS Cire. Weld Seam 5 .81 3.622 10408 LS Longitudinal Weld Seams 1.79 1.116 00434 Table 4-4 Fluence Values for the Vessel Surface, 1/4T and 3/4T Locations for the J. M. Farley Unit I Reactor Vessel Beltline Materials at 72 EFPY Fluence, f 1/4 T f 3/4 T f Region (X10 19 n/cm 2, (xl0 19 n/cm2, (xl0 19 n/cm 2, E > 1.0 MeV) E> 1.0 MeV) E> 1.0 MeV)

Intennediate Shell Plates 7.84 4.888 1.900 Lower Shell Plates 7.65 4.769 1.854 IS Longitudinal Weld Seams 2042 1.509 0.586 1S to LS Cire. Weld Seam 7.65 4.769 1.854 I

, LS Longitudinal Weld Seams 2 .36 1.471 0.572 WCAP-17122-NP October 2009 Revision 0

Table 4-5 Fluence Factor Values at tbe 1/4T and 3/4T Locations for tbe J. M. Farley Unit 1 Reactor Vessel Beltline Materials at 36 EFPY 114 T f 3/4 Tf Region (Xl0 19 n/cml, 1I4T FF (XlO I9 n/cm", 3/4TFF E> 1.0 MeV) E> 1.0 MeV)

Intennediate Shell Plates 2.506 1.2468 . 0.974 0.99",

I Lower Shell Plates I 2.481 1.2443 0.964 0.9899

'" Id Seams I 0.767 0.9255 0.298 0.6686 IS to LS Circ. Weld Seam I 2.481 1.2443 0.964 I 0.9899 LS f .onDihl{iinlll Weld Seams I 0.754 0.9209 0.293 i 0.6644 Table 4-6 FlueDce Factor Values at the 1I4T and 3/4T Locations for the J. M. Farley Unit 1 Reactor Vessel Beltline Materials at 54 EFPY 114 T f 3/4T f Region (xl0 19 n/cm z, l/4TFF (d0 19 nlem1, 3/4TFF E > 1.0 MeV) E> 1.0 MeV)

Intennediate Shell Plates 3.697 1.3389 1.437 1.1005 Lower Shell Plates 3.622 1.0949 IS 0.7738 Table 4-7 FJuence Factor Values at the 1I4T and 3/4T Locations for the J. M. Unit 1 Reactor Vessel Beltline Materials at 72 EFPY 1I4Tf 3/4Tf Region (d0 19 D/cml, 1/4TFF (dO l9 n/cml, 3/4T F.F E> 1.0 MeV) E> 1.0 MeV)

Intennediate Shell Plates 4.888 1.3979 1.900 1.1756 Lower Shell Plates 4.769 1.3930 1.854 I 1691 1.1138 0.586 0.8506 1.3930 t .854 1.1691 1.1070 0.572 0.8436 WCAP*17122-NP Revision 0

Westinghouse Non-Proprietary Class 3 4-5 Table 4-8 Adjusted Reference Temperature Evaluation for the J. M. Farley Unit 1 Reactor Vessel Beltline Materials through 36 EFPY at the 1I4T Location 1I4T f Material CF (E+19 1I4T ARTNDT RTNDT(lJ) 0'1 O'A M ART Reactor Vessel Location (OF) n/cmI, E > FF (DF) (DF) (oF) (DF) eF) eF) 1.0 MeV)

Intermediate Shell Plate B6903-2 91 2.506 1.2468 113.5 0 0 17 34 147 Intermediate Shell Plate B6903-3 82.2 2.506 1.2468 102.5 10 0 17 34 146 Lower Shell Plate B6919-1 97.8 2.481 1.2443 121.7 15 0 17 34 171 without surveillance data Lower Shell Plate B6919-1 106.7 2.481 1.2443 132.8 15 0 17 34 182 with non-credible surveillance data(a) I Lower Shell Plate B6919-2 98.2 2.481 1.2443 122.2 5 0 17 34 161 Intennediate Shell Longitudinal Weld 19-894A& B 126.3 0.767 0.9255 116.9 -56 17 28 65.5 126 Seams without surveillance data Heat # 33A277 Intermediate Shell Longitudinal Weld 19-894A& B 14(')

118.5 0.767 0.9255 109.7 -56 17 44.0 98 Seams with credible surveillance data(') Heat # 33A277 Intermediate to Lower Shell 11-894 98.4 2.481 1.2443 122.4 -56 17 28 65.5 132 Circumferential Weld Heat # 6329637 20-894 A & B Lower Shell Longitudinal Weld Seams 91.4 0.754 0.9209 84.2 -56 17 28 65.5 94 Heat # 90099 Notes for Table 4-8:

(a) Per Appendix D of WCAP-16964-NP, Revision 0 [Reference 9], the surveillance data of the plate was deemed not credible and the surveillance data of the weld material was deemed credible. Since the surveillance data of the weld material was deemed credible, a reduced at. value is used.

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Westinghouse Non-Proprietary Class 3 4-6 Table 4-9 Adjusted Reference Temperature Evaluation for the J. M. Farley Unit 1 Reactor Vessel Beltline Materials througb 36 EFPY at the 3/4T Location 3/4T f Material CF (E+19 3/4T ARTNDT RTNDT(U) 0'1 (Jt.. M ART Reactor Vessel Location 2 (oF) n/cm , E > FF (oF) COF) (OF) COF) (OF) (oF) 1.0 MeV)

Intermediate Shell Plate B6903-2 91 0.974 0.9927 90.3 0 0 17 34 124 Intermediate Shell Plate B6903-3 82.2 0.974 0.9927 81.6 10 0 17 34 126 Lower Shell Plate B6919-1 97.8 0.964 0.9899 96.8 15 0 17 34 146 without surveillance dab!

Lower ShelJ Plate B6919-1 106.7 0.964 0.9899 105.6 15 0 17 34 155 I with non-credible surveillance data(a) I Lower Shell Plate B6919-2 98.2 0.964 0.9899 97.2 5 0 17 34 136 Intermediate Shell Longitudinal Weld 19-894 A & B 126.3 0.298 0.6686 84.4 -56 17 28 65.5 94 Seams without surveiUance data Heat # 33A277 Intermediate Shell Longitudinal Weld 19-894A & B 14(3) 118.5 0.298 0.6686 79.2 -56 17 44.0 67 Seams with credible surveillance dab!(a) Heat # 33A277 Intermediate to Lower Shell 11-894 98.4 0.964 0.9899 97.4 -56 17 28 65.5 107 Circumferential Weld Heat # 6329637 20-894A& B Lower Shell Longitudinal Weld Seams 91.4 0.293 0.6644 60.7 -56 17 28 65.5 70 Heat # 90099 Notes for Table 4-9:

(a) Per Appendix D ofWCAP-16964-NP, Revision 0 [Reference 9J, the surveillance data of the plate was deemed not credible and the surveillance data of the weld material was deemed credible. Since the surveillance data of the weld material was deemed credible, a reduced 06 value is used.

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Westinghouse Non-Proprietary Class 3 4-7 Table 4-10 Adjusted Reference Temperature Evaluation for the J. M. Farley Unit 1 Reactor Vessel Beltline Materials through S4 EFPY at the 1I4T Location 1I4T f Material CF (E+19 1I4T ARTNDT RTNDT(U) al a", M ART Reactor Vessel Location (OF) niemI, E > FF (DF) (OF) (OF) (DF) (OF) (DF) 1.0 MeV)

Intennediate Shell Plate B6903-2 91 3.697 1.3389 121.8 0 0 17 34 156 Intennediate Shell Plate B6903-3 82.2 3.697 1.3389 110.1 10 0 17 34 154 Lower Shell Plate B6919-1 97.8 3.622 1.3343 130.5 15 0 17 34 179 without surveiUance data Lower Shell Plate B6919-1 106.7 3.622 1.3343 142.4 15 0 17 34 191 with non-credible surveillance data(a)

Lower Shell Plate B6919-2 98.2 3.622 1.3343 131.0 5 0 17 34 170 Intennediate Shell Longitudinal Weld 19-894 A& B 126.3 1.141 1.0368 130.9 -56 17 28 65.5 140 Seams without surveillance data Heat # 33A277 Intennediate Shell Longitudinal Weld 19-894A& B 14(')

118.5 1.141 1.0368 122.9 -56 17 44.0 III Seams with credible surveillance data(a) Heat # 33A277 Intennediate to Lower Shell 11-894 98.4 3.622 1.3343 131.3 -56 17 28 65.5 141 Circumferential Weld Heat # 6329637 20-894A& B Lower Shell Longitudinal Weld Seams 91.4 1.116 1.0307 94 .2 -56 17 28 65.5 104 Heat # 90099 Notes for Table 4- \0:

(a) Per Appendix D of WCAP-16964-NP, Revision 0 [Reference 9], the surveillance data of the plate was deemed not credible and the surveillance data of the weld material was deemed credible. Since the surveillance data of the weld material was deemed credible, a reduced ali value is used.

WCAP-17122-NP October 2009 Revision 0

Westinghouse Non-Proprietary Class 3 4-8 Table 4-11 Adjusted Reference Temperature Evaluation for the J. M. Farley Unit 1 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location 3/4T f Material CF (E+l9 3/4T ARTNDT RTNDT(U) al a.1 M ART Reactor Vessel Location (OF) n/cm 2, E > FF (OF) eF) (OF) (OF) (oF) (oF) 1.0 MeV)

Intennediate Shell Plate B6903-2 91 1.437 1.1005 100.1 0 0 17 34 134 Intennediate Shell Plate B6903-3 82.2 1.437 1.1005 90.5 10 0 17 34 134 Lower Shell Plate B6919-1 97.8 10408 1.0949 107. 1 15 0 17 34 156 without surveillance data I Lower Shell Plate B6919-1 106.7 10408 1.0949 116.8 15 0 17 34 166 with non-credible surveillance data(a)

I Lower Shell Plate B6919-2 98.2 10408 1.0949 107.5 5 0 17 34 147 Intennediate Shell Longitudinal Weld 19-894 A& B 126.3 00443 0.7738 97.7 -56 17 28 65.5 107 Seams without surveillance data Heat # 33A277 Intermediate Shell Longitudinal Weld 19-894 A& B 14(')

118.5 0.443 0.7738 91.7 -56 17 44.0 80 Seams with credible surveillance data(>> Heat # 33A277 Intennediate to Lower Shell 11-894 9804 10408 1.0949 107.7 -56 17 28 65.5 117 Circumferential Weld Heat # 6329637 20-894A& B Lower Shell Longitudinal Weld Seams 91.4 00434 0.7678 70.2 -56 17 28 65.5 80 Heat # 90099 L..--- - - - -- - ~ - -

Notes for Table 4-11:

(a) Per Appendix D ofWCAP-16964-NP, Revision 0 [Reference 9], the surveillance data of the plate was deemed not credible and the surveillance data of the weld material was deemed credible. Since the surveillance data of the weld material was deemed credible, a reduced crt> value is used.

WCAP-17122-NP October 2009 Revision 0

Westinghouse Non-Proprietary Class 3 4-9 Table 4-12 Adjusted Reference Temperature Evaluation for the J. M. Farley Unit 1 Reactor Vessel BeItline Materials through 72 EFPY at the 1I4T Location 1I4T f Material CF (E+19 1I4T l1RT1mT RTNDT(L') <1. <1A M ART Reactor Vessel Location 1 (OF) n/cm , E> FF (OF) (OF) (OF) (oF) (OF) (OF) 1.0 MeV)

Intennediate Shell Plate B6903-2 91 4.888 1.3979 127.2 0 0 17 34 161 Intennediate Shell Plate B6903-3 82.2 4.888 1.3979 114.9 10 0 17 34 159 Lower Shell Plate B6919-1 97.8 4.769 1.3930 136.2 15 0 17 34 185 without surveillance data Lower Shell Plate B6919-1 106.7 4.769 1.3930 148.6 15 0 17 34 198 with non-<:redible surveillance data(a)

Lower Shell Plate B6919-2 98.2 4.769 1.3930 136.8 5 0 17 34 176 Intennediate Shell Longitudinal Weld 19-894 A & B 126.3 1.509 1.1 138 140.7 -56 17 28 65.5 150 Seams without surveillance data Heat # 33A277 I

Intennediate Shell Longitudinal Weld 19-894 A & B 14(a) 118.5 1.509 1.1138 132.0 -56 17 44.0 120 Seams with credible surveillance data(a) Heat # 33A277 Intennediate to Lower Shell 11-894 98.4 4.769 1.3930 137.1 -56 17 28 65.5 147 Circumferential Weld Heat # 6329637 20-894 A& B Lower Shell Longitudinal Weld Seams 91.4 1.471 1.1070 101.2 -56 17 28 65.5 111 Heat # 90099 Notes for Table 4-12:

(a) Per Appendix D ofWCAP-16964-NP, Revision 0 [Reference 9], the surveillance data of the plate was deemed not credible and the surveillance data of the weld material was deemed credible. Since the surveillance data of the weld material was deemed credible, a reduced crt. value is used.

WCAP-17122-NP October 2009 Revision 0

Westinghouse Non-Proprietary Class 3 4-10 Table 4-13 Adjusted Reference Temperature Evaluation for the J. M. Farley Unit 1 Reactor Vessel Beltline Materials through 72 EFPY at the 3/4T Location 3/4T f Material CF (E+19 3/4T ARTNDT RTNDT(U) 0'[ 0'.<1. M ART Reactor Vessel Location 1 (OF) nlcm , E> FF (oF) (OF) (oF) (oF) (oF) (oF) 1.0 MeV)

Intermediate Shell Plate 86903-2 91 1.900 1.1756 107.0 0 0 17 34 141 Intermediate Shell Plate B6903-3 82.2 1.900 l.l756 96.6 10 0 17 34 141 Lower Shell Plate B6919-1 97.8 1.854 1.1691 114.3 15 0 17 34 163 without surveillance data Lower Shell Plate B6919-1 106.7 1.854 I.l691 124.7 15 0 17 34 174 with non~redible surveillance data(a)

I Lower Shell Plate B6919-2 98.2 1.854 1.1691 114.8 5 0 17 34 154 Intermediate Shell Longitudinal Weld 19-894 A& B 126.3 0.586 0.8506 107.4 -56 17 28 65.5 117 Seams without surveillance data Heat # 33A277 Intermediate Shell Longitudinal Weld 19-894A& B 14(a) 118.5 0.586 0.8506 100.8 -56 17 44 .0 89 Seams with credible surveillance data(a) Heat # 33A277 Intermediate to Lower Shell 11-894 98.4 1.854 1.1691 115.0 -56 17 28 65.5 125 Circumferential Weld Heat # 6329637 20-894 A& B Lower Shell Longitudinal Weld Seams 91.4 0.572 0.8436 77.1 -56 17 28 65.5 87 Heat # 90099 Notes for Table 4-13:

(a) Per Appendix D ofWCAP-16964-NP, Revision 0 [Reference 9], the surveillance data of the plate was deemed not credible and the surveillance data of the weld material was deemed credible. Since the surveillance data of the weld material was deemed credible, a reduced at> value is used .

WCAP-17122-NP October 2009 Revision 0

Westinghouse Non-Proprietary Class 3 4-11 Contained in Table 4-14 is a summary of the limiting ART values used in the generation of the 1. M.

Farley Unit I reactor vessel poT limit curves. The limiting material for both the 1/4T location and the 3/4T location at 36, 54, and 72 EFPY is lower shell plate 86919-1 using non-credible surveillance data .

Table 4-14 Summary of the Limiting ART Values Used in the Generation of the J. M. Farley Unit 1 Heatup/Cooldown Curves Limiting ART (OF)

Lower Shell Plate B6919-1 with non-EFPY credible surveillance data 1I4T 3/4T 36 182 155 54 191 166 72 198 174 WCAP-17122-NP October 2009 Revision 0

5 HEATUP AND COOLDOWN PRESSURE~TEMPERATURE LIMIT CURVES limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline using the methods discussed in Sections 3 and 4 of this report. This approved methodology is also presented in WCAP-14040-A, Revision 4.

Figures 5-1, 5-3 and 5-5 the heatup curves without for possible instrumentation errors using heatup rates of 60 and 100"F/hr applicable for 36, 54 and 72 EFPY, respectively, with the "Flange-Notch" requirement and using the "Axial-flaw" methodology. 5-4 and 5-6 present the cooldown curves without for instrumentation errors using cool down rates of 0, 20, 40, 60 and 100°F/hr applicable for 36, 54 and 72 EFPY, respectively, with the "Flange-Notch" and using the "Axial-flaw" The heatup and cooldown curves were """pn,tpti the 1998 through the 2000 Addenda ASME Code Section XI, Appendix G Also, a pressure correction for the static and dynamic head loss between the reactor vessel belt line region and the RHR relief valves is included for both the and cooldown curves at each EFPY. These curves incorporate a pressure correction of 27 psi for less than llOoP, and 60 psi for temperatures than or equal to llO'F, associated with operation of one and three reactor coolant pumps, respectively

.. tp,rp""", 10].

Allowable combinations of temperature and pressure for specific rates are below and to the right of the limit lines shown in 5-1 througb 5-6. Tbis is in addition to other which must be met before the reactor is made critical, as discussed below in the following paragraphs.

The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 5-1, 5-3 and 5-5 (heatup curves only). The straight-line portion of the criticality limit is at the minimum permissible for the 2485 inservice hydrostatic test as by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in the 1998 through the 2000 Addenda ASME Code Section XI, Appendix G as follows:

where, Kim is the stress factor covered by membrane stress,

= 33.2 + 20.734 e[002 (T - RTNllTll, T is the minimum metal tt>",nprll and RT NDT is the metal reference nil-ductility temperature.

The criticality limit curve limits for core operation in order to provide additional actual power production. The pressure-temperature limits for core operation (except for low power physics tests) are that: I) the reactor vessel must be at a equal to or higher than the minimum temperature required for the inservice hydrostatic test, and 2) the reactor vessel WCAP-17122-NP Revision 0

must be at least 4Q"F than the minimum permissible in the pressure temperature curve for heatup and cooldown calculated as described in Section 4 of this report. For the and cooldown curves without for instrumentation errors, the minimum for the inservice hydrostatic leak tests for the 1. M. Farley Unit I reactor vessel at 36 EFPY is 238"F. The criticality limits for 54 and 72 EFPY are 247°F and 254°F, respectively. The vertical line drawn from these on the curve, a curve 4Q"F higher than the pressure temperature limit curve, constitutes the limit for core operation for the reactor vessel.

Figures 5-1 through 5-6 define all of the above limits for ensuring of non-ductile failure for the

1. M. Farley Unit I reactor vessel for 36, 54 and 72 EFPY with the "Flange-Notch" requirement, without instrumentation and with pressure correction. The data points used for the heatup and cool down pressure-temperature limit curves shown in Figures 5*1 through 5-6 are presented in Tables 5-1 through 5-6.

WCAP*17122*NP Revision 0

Westinghouse Non-Proprietary Class 3 5-3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate B6919-1 with non-credible surveillance data LIMITING ART VALUES AT 36 EFPY: 114T, 182°F 3/4T, 155°F Figure 5-1 J. M. Farley Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 36 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda App. G Methodology (wiKle) 2250 ..~- ._-j --- : .. I ILeak Test Limit Vi

*..* 1 * - ~ .

I I

2000 . . ' -- .+ " 1""

"I I

I 1750 '

I

-Cl

~ . . ->----.J.-- ~.

r 1500 I

-j-'

I Acceptable C1I

~

OperatIon I/)

i I I I/)

e 1250 .. ...L-_ ~ -+- ,.. I --

Il. I ; ,

"C C1I I Critical limit CI:I ~--":'.-1100 Oeg. F/Hr i

u 1000 - - -r

r

~

U 750 + - ',----t- ...

I I

500 - . _.- r - * * *** Criticality Limit based on inservice hydrostatic test temperature (238°F) for the service period up to 36 EFPY 250 o

o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

WCAP-17122-NP October 2009 Revision 0

Westinghouse Non-Proprietary Class 3 5-4 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate B69l9-l with non-credible surveillance data LIMITING ART VALUES AT 36 EFPY; 1/4T, 182°F 3/4T, 155°F Figure 5-2 J. M. Farley Unit 1 Reactor Coolant System Cool down Limitations (Cooldown Rates up to 100°F/hr) Applicable for 36 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda App. G Methodology (wIKle) 2500 Operlim Version:5,2 Run:31175 Operlim ,x1s Version: 5.2

~

I 2250 . - '- j--- - - ~

f I

2000 -- i ---

t, - f I i 1750 1- ,-- f - - .,-,- , -j ,

§'

' r

-~f! 1500 -I Unacceptable -',

Operation

", L ,

I

s 1/1 1/1 I ,

t f! 1250 ;- +- ----. --I- - - , "" " - r "

I

" 1 - ---,

a. i : ! ,I 1:1 CI.I

,g I

, **-- i-'-

i i

! ;

Acceptable Operation I

I

~ 1000 , ----4----. -

i '

iii o I j _ __ ...L... _ _ ,, _ _

750 --- ' j-' T-',-

I '

I I R::7.L-~+~-.l, I !

500 . \ . ... ~-- -. "~'-l-

I 250 ' --: - " .. , I .. -~ ... ,- ~ - ....

-100 o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Oe9. F)

WCAP-17122-NP October 2009 Revision 0

Westinghouse Non-Proprietary Class 3 5-5 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate B6919-1 with non-credible surveillance data LIMITING ART VALUES AT 54 EFPY: 1/4T, 191°F 3/4T, 166°F Figure 5-3 J. M. Farley Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and lOOOF/hr) Applicable for 54 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda App. G Methodology (WIKle) 2500 Openim Version:5.2 Run:32550 Operlim.xls Version: 5.2

,;

2250 -"'- . - ---_. -d--._ .._ . ~- . ,

ILeak Test Llmit~

I 2000 .-;- --- . - - +

. I

I 1750 . ., . -~: ----1" .. -'. .. .. ~ '-~-'"

Unacceptable

§' Operation en ..r

c. 1500 -

~

J

\

i

-. T- ' " . /_ . r=-":-:-:---:-:-:--::- ...... 'j I/)

I/) .

--j- .-

I: . . ., -_. , --, .

~ 1250

c. I I I "tJ I

~

B 1000 .

I

.. j ~ ...

~

o 750 500 -, .. Criticality Limit based on . . 1._.. ...

inservice hydrostatic test temperature (247°F) for the service period up to 54 EFPY 250 " 1- *!* .. *t o

o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

WCAP-17122-NP October 2009 Revision 0

Westinghouse Non-Proprietary Class 3 5-6 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate B6919-1 with non-credible surveillance data LIMITING ART VALUES AT 54 EFPY: 1I4T, 191°F 3/4T, 166°F Figure 5-4 J. M. Farley Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 54 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda App. G Methodology (wiKle) 2500 Operlim Version:5.2 Run:32550 Opertim .xls Version: 5.2 L. _.. I 2250 - I T.

--I *- -i 2000 . ..L...

I I 1 I ,

i I t 1

1750 .. ----+-- - i .' * ~t- * - -t-. --.!

§'

~ 1500 I u~accePt~ble Operation .j . --.J I

1 E ; 1

s II) I

-- ,. -r II) t

--.I. j -- ..1~~.~ ..- :

E 1250 a..

"0 t, I i

I I

.. ~ -I ' . - ,----

.! . I Acceptable J!! I 1 Operation

s 1000 . - -.. !- -~ -

~ !

CO o I 750 II r +I . --+. ... -!--- '

... '--~-. - .

;-'

1

! iI 500 _.+- - .!--- ..

250 - --- .--i- - ,

+ - - ..,- .~ -

o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Oeg. F)

WCAP-17122-NP October 2009 Revision 0

Westinghouse Non-Proprietary Class 3 5-7 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate B6919-1 with non-credible surveillance data LIMITING ART VALUES AT 72 EFPY: 1/4T, 198°F 3/4T, 174°F Figure 5-5 J. M. Farley Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 72 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda App. G Methodology (WIKle) 2500 2250 . ... 1* ..,,--_.. ~ " I .

..*.-- "- ' ....--- . . L.- .- ~-l~'--- ' 'r' - ~ . -!

I I I ILeak Test Li~it t-+ i i

! j : I i I I 2000 I--r I

.-t-._ .- .. .

I I

+ .. ---1---1'" J

'-1 I

i I I

. Critical Limit :

1750 .J. i 1-I

..*.** 1 ...-H. 60 Oe9. F/Hr ~_

+ t - .-

1

, I I I

S I Acceptable I i

en ...1 Operation I Q.

CI)

J 1500 - -l I/) I i ;

I/)

CI)

Q.

1250 -l _.f-_.

I j..

I I

~

'0 CI) u 1000

J -- . -

I I I

i ~-' -" - t.

I I

I .--1 CIS !

u 750 -! r*

500 . --. ... Criticality Limit based on in service hydrostatic test temperature (254°F) for the service period up to 72 EFPY I 250 * .. .. L.. ..... . ..

i o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Oeg. F)

WCAP*17122-NP October 2009 Revision 0

Westinghouse Non-Proprietary Class 3 5-8 MATERJAL PROPERTY BASIS LIMITING MATERJAL: Lower Shell Plate 86919-1 with non-credible surveillance data LIMITING ART VALUES AT 72 EFPY: 1/4T, 198°F 3/4T, 174°F Figure 5-6 J. M. Farley Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 72 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda App. G Methodology (wlK lc )

2500 Operlim Version :5.2 Run :30026 Operlim.xJs Version : 5.2 I

i i... , _1I 2250 -.,.\......

I

. ~~

-t --

j ,._... .. -., I "t ',- .' -- -- ,

i 2000 -+

I

-+.i i I

1750 ,-- _I ..~ - .. I,-

I

-en C)

I e:- 1500 .' , ~- ~ .. . ,. .j' .. \ - ~ .

Unacceptable

...:l II)

Operation 1/1 1/1 ! ,

II) 0-1250 ,-

, .. -t --I.

I

, _ _ 'H ""

'C :

....co II)

, __ J I

Acceptable I i Operation "3 1000 ... .... j .....,-- ' ._L ..__..

I i

. ... _ .. . -I I

(.)

co (J I

, i ,i i I

! , f \

- r-- ' ..- -~ . - . . ' _ _' --1.-._ ' ..._ , ~

750 ,

_ 1. _

I , ...-- ' .. --"j *.*

I I

500 .---'-1i '- 1 ..... -/ _ .

l __

--i- -- 4**- ----

250 . '" ' . , -,

I i I 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Oeg. F)

WCAP-17122-NP October 2009 Revision 0

Westinghouse Non-Proprietary Class 3 5-9 Table 5-1 36 EFPY Heatup Curve Data Points Using the 1998 through the 2000 Addenda App.

G Methodology (wI Kl<, wI Flange Notch, wI Pressure Correction and w/o Uncertainties for Instrumentation Errors)

Leak Test 60°F/hr 60°F/hr 10O°F/hr 100°F/hr Limit Heatup Criticality Heatup C riticaUty T P T P T P T P T P eF) (psig) (oF) (psig) (OF) (psig) (OF) (psig) eF) (psig) 220 2000 60 0 238 0 60 0 238 0 238 2485 60 594 238 561 60 591 238 558 65 594 238 561 65 591 238 559 70 594 238 561 70 591 238 560 75 594 238 561 75 591 238 560 80 594 238 561 80 591 238 561 85 594 238 561 85 591 238 561 90 594 238 561 90 591 238 561 95 594 238 561 95 591 238 561 100 594 238 561 100 591 238 561

\05 594 238 561 105 591 238 561 110 594 238 561 110 591 238 561 110 561 238 561 110 558 238 561 115 561 238 561 115 560 238 561 120 561 238 561 120 561 238 561 125 561 238 561 125 561 238 561 130 561 238 561 130 561 238 561 135 561 238 561 135 561 238 561 140 561 238 561 140 561 238 561 145 561 238 561 145 561 238 561 150 561 238 561 150 561 238 561 155 561 238 561 155 561 238 561 160 561 238 561 160 561 238 561 165 561 238 561 165 561 238 561 170 561 238 561 170 561 238 561 175 561 238 902 175 561 238 730 180 561 238 944 180 561 238 758 180 561 238 990 180 561 238 790 180 902 238 1075 180 730 238 848 185 944 240 1097 185 758 240 864 190 990 245 1159 190 790 245 907 195 1041 250 1228 195 825 250 955 200 \097 255 1304 200 864 255 1008 WCAP-17122-NP October 2009 Revision 0

NoY,.f"Y".....rll'l!lrv Class 3 5-10 Leak Test

, 60 0Flhr 60°Flhr 1000F/hr lOO°Flhr Limit Heatup Criticality Heatup Criticality T p T P T P T P T P (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig)

I 205 1159 260 1387 20 2 1066

"'v , ....u 265 \458 21 1130 215 1304 270 1535 215 270 1201

~

220 1387 275 1620 220 I 275 225 1458 280 1714 225 1130 280 367 230 1535 285 1818 230 1201 285 1462

! 235 1620 290 1932 235 1280 290 1568 I

240 1714 295 2058 240 1367 295 1685 ,

245 1818 3 2197 245 1462 300 =[ 1813 1932 305 2351 I 250 1568 305 1955 255 2058 255 J685 310 2111 260 2197 *1813 315 2283 265 2351 1955 320 2473 270 2111 275 2283 2473 WCAP-17122-NP Revision 0

Westinghouse Non-Proprietary Class 3 5-11 Table 5-2 36 EFPY Cooldown Curve Data Points Using the 1998 through the 2000 Addenda App. G Methodology (wi K. u wi Flange Notch, wi Pressure Correction and w/o Uncertainties for Instrumentation Errors)

Steady State 20° F/h r. 40°F/hr. 60 o F/hr. lOOoF/hr.

T(0F) P (psig) T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) 60 0 60 0 60 0 60 0 60 0 60 594 60 594 60 565 60 523 60 436 65 594 65 594 65 569 65 527 65 441 70 594 70 594 70 573 70 531 70 446 75 594 75 594 75 578 75 537 75 452 80 594 80 594 80 584 80 542 80 458 85 594 85 594 85 590 85 549 85 465 90 594 90 594 90 594 90 556 90 474 95 594 95 594 95 594 95 564 95 483

]00 594 ]00 594 100 594 100 573 100 493 105 594 105 594 105 594 105 583 105 505 110 594 110 594 110 594 110 594 110 517 110 561 110 561 110 56] 110 561 110 484 115 561 115 561 115 561 115 561 115 499 120 561 120 561 120 561 120 561 120 515 125 561 125 561 125 561 125 561 125 532 130 561 130 561 130 561 130 561 130 552 135 561 135 561 135 561 135 561 135 561 140 561 140 561 140 561 140 561 140 561 145 561 145 561 145 561 145 561 145 561 150 561 150 561 150 561 150 561 150 561 155 561 155 561 155 561 155 561 155 561 160 561 160 561 160 561 160 561 160 561 165 561 165 561 165 561 165 561 165 561 170 561 170 561 170 561 170 561 170 561 175 561 175 561 175 561 175 561 175 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 963 180 949 180 937 180 928 180 921 185 1004 185 993 185 985 185 980 185 980 190 1048 190 1041 190 1038 190 1038 190 1038 195 1097 195 1095 195 1095 195 1095 195 1095 WCAP-17122-NP October 2009 Revision 0

3 200 205 1212 205 1212 205 1212 205 210 1279 210 1279 210 1279 215 1352 215 1352 215 220 1434 220 1434 220 1523 225 1523 225 1523 225 230 1623 230 1623 235 1732 2 1732 1854 240 1854 240 1988 245 1988 245 1988 245 2136 250 2136 250 2136 250 255 2299 255 2299 255 2299 255 2299 255 2299 260 2480 260 2480 260 2480 260 2480 260 2480 WCAP-17122-NP Revision 0

Westinghouse Non-Proprietary Class 3 5-13 Table 5-3 54 EFPY Heatup Curve Data Points Using the 1998 through the 2000 Addenda App.

G Methodology (wi K le , wi Flange Notch, wi Pressure Correction and wlo Uncertainties for Instrumentation Errors)

Leak Test 60°Flhr 60°Flhr 100°F/hr 100°F/hr Limit Heatup Criticality I-ieatup Criticality T P T P T P T P T P (OF) (psig) (DF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) 229 2000 60 0 247 0 60 0 247 0 247 2485 60 594 247 561 60 574 247 541 65 594 247 561 65 574 247 541 70 594 247 56J 70 574 247 542 75 594 247 561 75 574 247 542 80 594 247 561 80 574 247 545 85 594 247 561 85 574 247 545 90 594 247 561 90 574 247 549 95 594 247 561 95 574 247 550 100 594 247 561 100 574 247 554 105 594 247 561 105 574 247 556 110 594 247 561 110 574 247 561 110 561 247 561 110 541 247 561 115 561 247 561 115 541 247 561 120 561 247 561 120 542 247 561 125 561 247 561 125 545 247 561 130 561 247 561 130 549 247 561 135 561 247 561 135 554 247 561 140 561 247 561 140 561 247 561 145 561 247 561 145 561 247 561 150 561 247 561 150 561 247 561 155 561 247 561 155 561 247 561 160 561 247 561 160 561 247 561 165 561 247 561 165 561 247 561 170 561 247 561 170 561 247 561 175 561 247 822 175 561 247 670 180 561 247 855 180 561 247 693 180 561 247 892 180 561 247 718 180 822 247 932 180 670 247 746 185 855 247 977 J85 693 247 777 190 892 247 1049 190 71 8 247 826 195 932 250 1082 195 746 250 849 200 977 255 1143 200 777 255 891 WCAP-17122-NP October 2009 Revision 0

Westinghouse Non-Proprietary Class 3 5-14 Leak Test 60°F/hr 60°F/hr lOO°F/hr lOO°F/hr Limit Heatup Criticality Heatup Criticality T p T P T P T P T P (OF) (psig) (DF) (psig) (DF) (psig) eF) (psig) (DF) (psig) 205 1027 260 1210 205 811 260 938 210 1082 265 1284 210 849 265 990 215 1143 270 1365 215 891 270 1046 220 1210 275 1455 220 938 275 1109 225 1284 280 1552 225 990 280 1179 230 1365 285 1638 230 ]046 285 1255 235 1455 290 1734 235 ]109 290 1340 240 1552 295 1839 240 1179 295 1433 245 1638 300 1955 245 1255 300 1536 250 1734 305 2084 250 1340 305 1650 255 1839 310 2225 255 1433 3\0 1775 260 1955 315 2382 260 1536 315 ]913 265 2084 265 1650 320 2066 270 2225 270 ]775 325 2234 275 2382 275 1913 330 2419 280 2066 285 2234 290 2419 WCAP-17122-NP October 2009 Revision 0

Westinghouse Non-Proprietary Class 3 5-15 Table 5-4 54 EFPY Coo/down Curve Data Points Using the 1998 through the 2000 Addenda App. G Methodology (wI K le , wI Flange Notch, wI Pressure Correction and wlo Uncertainties for Instrumentation Errors)

Steady State 20oF/hr. 40°F/hr. 60°F/hr. 100oFlhr.

T(OF) P (psig) . T(OF) P (psig) T(°F) P (psig) T(OF) P (psig) T(OF) P (psig) 60 0 60 0 60 0 60 0 60 0 60 594 60 594 60 558 60 515 60 427 65 594 65 594 65 561 65 519 65 431 70 594 70 594 70 565 70 522 70 435 75 594 75 594 75 569 75 527 75 440 80 594 80 594 80 574 80 531 80 445 85 594 85 594 85 578 85 537 85 451

, 90 594 90 594 90 584 90 543 90 458 I

95 594 95 594 95 590 95 549 95 465 100 594 100 594 100 594 100 556 100 474 105 594 105 594 105 594 105 565 105 483 110 594 110 594 110 594 110 574 110 494 110 561 110 561 110 561 110 541 110 461 115 561 115 561 115 561 115 551 115 473 I

120 561 120 561 120 561 120 561 120 486 125 561 125 561 125 561 125 561 125 500 130 561 130 561 130 561 130 561 130 517 135 561 135 561 135 561 135 561 135 535 140 561 140 561 140 561 140 561 140 555 145 561 145 561 145 561 145 561 145 561 150 561 150 561 150 561 150 561 150 561 155 561 155 561 155 561 155 561 155 561 160 561 160 561 160 561 160 561 160 561 165 561 165 561 165 561 165 561 165 561 170 561 170 561 170 561 170 561 170 561 175 561 175 561 175 561 175 561 175 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 900 180 880 180 862 180 846 180 823 185 934 185 917 185 902 185 890 185 875 190 971 190 957 190 946 190 938 190 933 195 1012 195 1002 195 995 195 991 195 991 WCAP-17122-NP October 2009 Revision 0

Westinghouse Non-Proprietary Class 3 5-16 Steady State 20°Flhr. 40°F/hr. 60°F/hr. lOO°Flhr.

i

! TeF) P (psig) TeF) P (psig) T(°F) P (psig) T(°F) P (psig) TeF) P (psig) 200 1058 200 1051 200 1049 200 1049 200 1049 205 1108 205 1106 205 1106 205 1106 205 1106 210 1164 210 1164 210 1164 210 1164 210 1164 I 215 1225 215 1225 215 1225 215 1225 215 1225 220 1293 220 1293 220 1293 220 1293 220 1293 225 1368 225 1368 225 1368 225 1368 225 1368 230 1451 230 1451 230 1451 230 1451 230 1451 235 1542 235 1542 235 1542 235 1542 235 1542 240 1644 240 1644 240 1644 240 1644 240 1644 245 1756 245 1756 245 1756 245 1756 245 1756 250 1879 250 1879 250 1879 250 1879 250 1879 255 2016 255 2016 255 2016 255 2016 255 2016 260 2167 260 2167 260 2167 260 2167 260 2167 265 2334 265 2334 265 2334 265 2334 265 2334 WCAP-17122-NP October 2009 Revision 0

Westinghouse Non-Proprielary Class 3 5-17 Table 5-5 72 EFPY Heatup Curve Data Points Using the 1998 through the 2000 Addenda App.

G Methodology (wi Kit, wi Flange Notch, wi Pressure Correction and w/o Uncertainties for Instrumentation Errors)

Leak Test 60°F/hr 60°F/hr 100°Flhr 100°F/hr Limit Heatup Criticality Heatup Criticality T P T P T P T P T P eF) (psig) eF) (psig) eF) (psig) (OF) (psig) (OF) (pslg) 236 2000 60 0 254 0 60 0 254 0 254 2485 60 594 254 561 60 563 254 530 65 594 254 561 65 563 254 530 70 594 254 561 70 563 254 530 75 594 254 561 75 563 254 532 80 594 254 561 80 563 254 533 85 594 254 561 85 563 254 534 90 594 254 561 90 563 254 536 95 594 254 561 95 563 254 538 100 594 254 561 100 563 254 541 ,

105 594 254 561 105 563 254 544 110 594 254 561 110 563 254 548 110 561 254 561 110 530 254 550 115 561 254 561 115 530 254 556 120 561 254 561 120 530 254 558 125 561 254 561 125 532 254 561

[30 561 254 561 130 534 254 561 135 561 254 561 135 538 254 561 140 561 254 561 140 544 254 561 145 561 254 561 145 550 254 561 150 561 254 561 150 558 254 561 155 561 254 561 155 561 254 561 160 561 254 561 160 561 254 561 165 561 254 561 165 561 254 561 170 561 254 561 170 561 254 561 175 561 254 773 175 561 254 634 180 561 254 801 180 561 254 653 180 561 254 833 180 561 254 675 180 773 254 867 180 634 254 698 185 801 254 905 185 653 254 725 190 833 254 948 190 675 254 754 195 867 254 1036 195 698 254 815 200 905 255 1046 200 725 255 822 WCAP-17122-NP October 2009 Revision 0

Westinghouse Non-Proprietary Class 3 5-18 Leak Test 60°F/hr 60°F/hr lOO°F/hr lOO°F/hr Limit Heatup Criticality Heatup Criticality T p T P T P T P T P (OF) (psig) (OF) (pslg) (OF) (psig) (OF) (pslg) (OF) (psig) 205 948 260 1103 205 754 260 861 210 995 265 1166 210 786 265 905 215 1046 270 1236 215 822 270 953 220 1103 275 1312 220 861 275 1007 225 1166 280 1397 225 905 280 1066 230 1236 285 1490 230 953 285 1131 235 1312 290 1594 235 1007 290 1203 240 1397 295 1694 240 1066 295 1282 245 1490 300 1795 245 1131 300 1370 250 1594 305 1907 250 1203 305 1467 255 1694 310 2030 255 1282 310 1573 260 1795 315 2166 260 1370 315 1691 265 1907 320 2316 265 1467 320 1820 270 2030 325 2482 270 1573 325 1963 275 2166 275 1691 330 2121 280 2316 280 1820 335 2295 285 2482 285 1963 340 290 2121 295 2295 WCAP-17122-NP October 2009 Revision 0

Westinghouse Non-Proprietary Class 3 5-19 Table 5-6 72 EFPY Cooldown Curve Data Points Using the 1998 through the 2000 Addenda App. G Methodology (wI Kle , wI Flange Notch, wI Pressure Correction and wlo Uncertainties for Instrumentation Errors)

Steady State 20°F/hr. 40°F/hr. 60°F/hr. lOO°F/hr.

T(OF) P (psig) T(OF) P (psig) T(°F) P (psig) T(°F) P(psig) T(°F) P (psig) 60 0 60 0 60 0 60 0 60 0 60 594 60 594 60 554 60 511 60 422 65 594 65 594 65 557 65 513 65 425 70 594 70 594 70 560 70 517 70 428 75 594 75 594 75 563 75 520 75 432 80 594 80 594 80 567 80 524 80 437 85 594 85 594 85 571 85 529 85 442 90 594 90 594 90 576 90 534 90 447 95 594 95 594 95 581 95 539 95 454 100 594 100 594 100 587 100 546 100 461 105 594 105 594 105 594 105 553 105 469 110 594 110 594 110 594 110 560 110 478 110 561 110 561 110 561 110 527 1\0 445 115 561 115 561 115 561 115 536 115 455 120 561 120 561 120 561 120 546 120 466 125 561 125 561 125 561 125 557 125 479 i 130 561 130 561 130 561 130 561 130 493 135 561 135 561 135 561 135 561 135 509 140 561 140 561 140 561 140 561 140 526 145 561 145 561 145 561 145 561 145 546 150 561 150 561 150 561 150 561 150 561 155 561 155 561 155 561 155 561 155 561 160 561 160 561 160 561 160 561 160 561 165 561 165 561 165 561 165 561 165 561 170 561 170 561 170 561 170 561 170 561 175 561 175 561 175 561 175 561 175 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 858 180 834 180 812 180 792 180 758 185 887 185 866 185 847 185 830 185 803 190 920 190 901 190 885 190 871 190 853 195 956 195 940 195 927 195 918 195 909 WCAP-17122-NP October 2009 Revision 0

"r('\I~rlpt"r\l Class 3 lOO°F/br.

P(psig) T(°F) P (psig) 969 200 969 205 1026 205 1026 205 1026 210 1084 210 1084 210 1084 1141 215 1141 215 1141 215 220 1200 220 1200 220 1200 225 1265 225 1265 225 1265 230 1337 230 1337 230 1337 235 1417 235 1417 235 1417 235 240 1505 240 1505 240 1505 240 1602 245 1602 245 1602 245 1602 1710 250 1710 250 1710 250 1710 1828 255 1828 255 1828 255 1828 1960 260 1960 260 1960 260 265 2105 265 2443 WCAP- I7122-NP October Revision 0

6 REFERENCES

1. Regulatory Guide I Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U. S.

Nuclear Regulatory Commission, May 1988.

2. WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," 1. D. Andrachek, et aI., May 2004.

3. Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division I, "Fracture Toughness Criteria for Protection Against Failure."

4. Westinghouse Letter L1'R-REA-09-112, Revision 1, "]. M. Farley Units I and 2 Updated BeltHne Fluence," B. W. 21,2009.

5. Code of Federal Regulations, 10 CFR Part Appendix "Fracture Toughness Requirements,"

U.S. Nuclear Regulatory Washington, D. Federal Volume 60, No.

dated December \9, 1995.

6. Westinghouse Letter ru.... n.-'vu- Revision 0, "Southern Nuclear Operating Joseph M.

Farley Nuclear Plant Unit Transmittal of Pressure Temperature Limits Report," E. C. Arnold, October 2, 2008.

7. WCAP-88 10, Revision 0, "Southern Alabama Power Company Joseph M. Nuclear Plant Unit No. I Reactor Vessel Radiation Surveillance Program," ]. A. Davidson, et. ai., December 1976.

8. ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division I, Subsection NB, Section NB-2300, "Fracture Requirements for Material."

9. WCAP-16964-NP, Revision 0, "Analysis of Capsule Z from the Southern Nuclear Company Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program," J. M.

Conermann and M. A. Hunter, October 2008,

10. WCAP*14689, Revision 6, "Farley Units I and 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation," T. 1. Laubham, April 2001.

WCAP.. 17122-NP Revision 0

_l-'r",nnpt~Ml Class 3 APPENDIX A THERMAL STRESS INTENSITY FACTORS (KIT)

The following pages contain the thennal stress intensity (KI!) for the maximum and cooldown rates. The vessel radii to the 1/4T and 3/4T locations are as follows:

.. 114T Radius 80.625"

.. 3/4T Radius"" 84.562" WCAP-17122-NP Revision 0

3 Table A-I KII Values for 36, 54, and 72 EFPY 100°FIhr Hestup Curves (w/o Margins for Instrument Errors)

Vessel Temperature Vessel Temperature Water 1/4T Thermal Stress J/4T Thermal Stress

@ 1I4T Location for @ J/4T Location for Temp. Intensity Factor Intensity Factor (oF) lOOoF/hr Heatup lOO"F/hr Heatup (KSI SQ. RT. IN.) (KSI SQ. RT. IN.)

("F) ("F) 60 56.130 -0.987 55.065 0.493 65 58.927 -2.377 55.425 1.455 70 62.129 -3.521 56.315 2.377 75 65.562 -4.586 57.748 3.208 on 69.262 -5.475 59.641 3.929 85 73.079 -6.273 61.944 4.558 90 71.089 -6.948 64.601 5.101 95 81.193 -7.553 67.562 5.578 100 85.435 -8.069 70.788 5.991 89.755 -8.531 74.238 6.353 94.171 -8.928 77.881 6.671 98.650 -9.285 81.690 6.951 103.196 -9.594 85.642 7.198 107.790 -9.875 89.717 7.418 110 I PAll -10.118 93.898 7.612 11" 117.114 -10.341 98.171 7.785

~

140 -10.535 102.523 7.940

-10.715 106.944 8.080 150 -10.873 111.424 8.206 155 -11.020 115.955 8.320 160 140.945 -I Ll51 120.529 8.423 165 145.773 -11.275 125.142 8.519 170 150.613 -11.385 129.788 8.606 155.467 -11.491 134.462 8.687 180 160.330 -11.586 139.161 8.762 i 185 165.204 -11.678 143.881 8.833 190 170.083 -11.763 148.620 8.899 195 174.972 -11.845 153.374 8.961 200 179.864 -11.920 158.143 9.020 205 184.764 11.995 162.923 9.077 210 189.666 -12.064 167.713 9.131 WCAP-17122-NP Revision 0

Westinghouse Non-Proprietary Class 3 A-3 Table A-2 KII Values for 36,54, and 72 EFPY 100°F/hr Cooldown Curves (w/o Margins for Instrument Errors)

Vessel Temperature lOOoF/hr Cooldown Water @ 1/4T Location for 1/4T Thermal Stress Temp. lOO°F/hr Cooldown Intensity Factor eF) eF) (KSI SQ. RT. IN.)

210 232.426 13.510 205 227.352 13.454 200 222.278 13.398 195 217.204 13.342 190 212. 131 13.286 185 207.057 13.230 I

I 180 201.983 13.175 175 196.909 13.119 170 191.836 13.063 165 186.762 13 .008 160 181.688 12.952 155 176.615 12.897 150 171.541 12.842 145 166.468 12.786 J40 161.395 12.731 135 156.322 12.676 130 151.249 12.622 [

125 146.176 12.567 120 141.103 12.5J2 115 \36 .031 12.457 110 130.958 12.403 i 105 125 .886 12.349 100 120.813 12.295 95 115.741 12.240 90 110.669 12. 187 I 85 105 .597 12.133 80 100.526 12.079 75 95.454 12.025 70 90.382 J 1.972 65 85 .311 J 1.919 60 80.241 11.865 WCAP-17122-NP October 2009 Revision 0

Joseph M. Farley Nuclear Plant Request to Revise Technical Specifications Associated with the Low Temperature Protection System and the Pressure and Temperature Limits Report EnclosureS WCAP-17123-NP, Revision 1, "J. M. Farley Unit 2 Heatup and Cool down Limit Curves for Normal Operation," July 2011

Westinghouse Non-Proprietary Class 3 WCAP-17123-NP July 2011 Revision 1 J. M. Farley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation (e Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17123-NP Revision 1 J. M. Farley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation E.J.Long*

Aging Management and License Renewal Services July 2011 Reviewer: B. A. Rosier*

Aging Management and License Renewal Services Approved: A. E. L1oyd*. Acting Manager and License Renewal Services "Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Drive

r"nh"nrv Township, PA 16066

Class 3 II RECORD OF REVISION Revision 0: Issue Revision I: Revision 0 of this incorrectly documents the heat number the Lower Shell Longitudinal Weld Seams20-923 A & B as 83649. The correct heat number for this material is 83640. This typographical error was documented in the Westinghouse Corrective Actions Process (CAPs) as Issue (IR) # 11-165-C017. This Revision makes the correction to this material's heat number to this CAPs YR.

WCAP-17123-NP Revision I

P'MP'' Class 3 iii TABLE OF CONTENTS LIST OF TABLES ....................................................................................................................................... iv LIST OF FIGURES ..................................................................................................................................... vi EXECUTIVE

SUMMARY

......................................................................................................................... vii INTRODUCTION ........................................................................................................................ J-l 2 FRACTURE TOUGHNESS PROPERTIES ................................................................................. 2-1 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS ................ 3-1 3.l OVERALL APPROACH ................................................................................................. 3-1 3.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT ............................................................................................................ 3-1 3.3 CLOSURE HEADNESSEL FLANGE REQUIREMENTS ........................................... 3-5 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE ......................................... .4-1 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES ....................... 5-1 6 REFERENCES ............................................................................................................................. 6-1 APPENDIX A Thermal Stress Intensity Factors (Kit) ....................................................... A-l WCAP-17123-NP Revision I

LIST OF TABLES Table 2-1 Summary of the Best Estimate Cu and Ni Percent and Initial RT NOT Values for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials ..................................................... 2-2 Table 2-2 Summary of the Initial RT NDT Values for the J. M. Unit 2 Closure Head and Vessel Flange .............................................................................................................................. 2-3 Table 2-3 Summary of the J. M. Unit 2 Reactor Vessel Beltline Material Chemistry Factors per Regulatory Guide 1.99, Revision 2 [Reference I] ..................................................... 2-3 Table 4-1 Fluence Values for the J. M. Unit 2 Reactor Vessel Beltline Materials .............. ..4-2 Table 4-2 Fluence Values for the Vessel Surface, 114T and 3/4T Locations for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials at 36 EFPY............................................................ ..4-3 Table 4-3 Fluence Values at the Vessel 1/4T and 3/4T Locations for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials at 54 EFPY................................................................ .4-3 Table 4-4 Fluence Values at the Vessel Surface, 1/4T and 3/4T Locations for the 1. M. Farley Unit 2 Reactor Vessel Beltline Materials at 72 EFPY................................................................ .4-3 Table 4-5 Fluence Factor Values at the 1I4T and 3J4T Locations for the 1. M. Farley Unit 2 Reactor Vessel Beltline Materials at 36 EFPY ............................................................................. .4-4 Table 4-6 Fluence Factor Values at the 1/4T and 3/4T Locations for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials at 54 EFPY ............................................................................. .4-4 Table 4-7 Fluence Factor Values at the 1/4T and 3/4T Locations for the 1. M. Unit 2 Reactor Vessel Beltline Materials at. 72 EFPY Table 4-8 Reference Temperature Evaluation for the 1. M. Unit 2 Reactor Vessel Beltline Materials through 36 EFPY at the 1/4T Location ............................................ ..4-5 Table 4-9 Adjusted Reference Temperature Evaluation for the 1. M. Farley Unit 2 Reactor Vessel Beltline Materials through 36 EFPYat the 3/4T Location ............................................. .4-6 Table 4-10 Adjusted Reference Temperature Evaluation for the 1. M. Unit 2 Reactor Vessel BeltHne Materials 54 EFPYat the 1/4T Location ............................................. .4-7 Table 4-11 Adjusted Reference Evaluation for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location ............................................. .4-8 Table 4-12 Adjusted Reference Evaluation for the 1. M. Unit 2 Reactor Vessel Beltline Materials through 72 EFPY at the 1/4T Location .............................................. 4-9 Table 4-13 Adjusted Reference Temperature Evaluation for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials 72 EFPY at the 3/4T Location ........................................... .4- JO Table 4-14 Summary of the ART Values Used in the Generation of the J. M. Farley Unit 2 Heatup/Cooldown Curves ......................................................................... ,.................. ,.4-11 WCAP-17123-NP July 2011 Revision I

Class 3 v Table 5-1 36 EFPY Curve Data Points the 1998 through the 2000 Addenda G Methodology (wI Klc. wi Notch, wi Pressure Correction and wlo Uncertainties for Instrumentation Errors) .................................................................................................... 5-9 Table 5-2 36 EFPY Cooldown Curve Data Points Using the 1998 through the 2000 Addenda App.

G Methodology (wi Kle , wi Notch, wi Pressure Correction and wlo Uncertainties for Instrumentation Errors) ............................................................................................ 5-11 Table 5-3 54 EFPY Curve Data Points Using the 1998 through the 2000 Addenda App. G Methodology (wI K le , wI Flange Notch, wI Pressure Correction and wlo Uncertainties for Instrumentation .......................................................................... 5-13 Table 5-4 54 EFPY Cooldown Curve Data Points the 1998 the 2000 Addenda App.

G Methodology (wI KIc. wI Notch, wI Pressure Correction and wlo Uncertainties for Instrumentation ............................................................................................ 5-\5 Table 5-5 72 EFPY Heatup Curve Data Points the 1998 through the 2000 Addenda App. G Methodology (wi K Ic , wI Notch, wi Pressure Correction and wlo Uncertainties for Instrumentation ....... 5-17 Table 5-6 72 EFPY Cooldown Curve Data Points the 1998 the 2000 Addenda G Methodology (wi wI Flange Notch, wi Pressure Correction and wlo Uncertainties for Instrumentation ............................................................................................ 5-19 Table A-I Kit Values for 54 and 72 EFPY 100°F/hr Heatup Curves (w/o Margins for Instrument Table A-2 KI! Values for 36,54 and 72 EFPY IOO°F/hr Cooldown Curves (w/o for Instrument ........................................ A-3 WCAP-17123-NP Revision I

3 LIST OF FIGURES 5-1 J. M. Farley Unit 2 Reactor Coolant System Limitations (Hearup Rates of 60 and 1OO°F/hr) Applicable for 36 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) the 1998 through the 2000 Addenda App. G Methodology (wIK,e) .............................................................................................................................. 5-3 Figure 5-2 J. M. Farley Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to WO°F/hr) Applicable for 36 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) the 1998 through the 2000 Addenda App. G Methodology (WIKle) ....................................................................................................... 5-4 Figure 5-3 J. M. Unit 2 Reactor Coolant System Hearup Limitations (Hearup Rates of 60 and IOO°F/hr) for 54 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda App. G Methodology (wlKre) .............................................................................................................................. 5-5 5-4 J. M. Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to WO°F/hr) Applicable for 54 EFPY (without for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda App. G Methodology (wIKle) ....................................................................................................... 5-6 Figure 5-5 J. M. Farley Unit 2 Reactor Coolant System Hearup Limitations Rates of 60 and 100°F/hr) Applicable for 72 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda App. G Methodology Figure 5-6 I. M. Farley Unit 2 Reactor Coolant System Cool down Limitations (Cooldown Rates up to JOO°F/hr) Applicable for 72 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) the 1998 through the 2000 Addenda G Methodology (WIKle) ....................................................................................................... 5-8 WCAP-17 J23-NP July 2011 Revision I

N"".l'rnnn(,j'lIl'V Class 3 vii EXECUTIVE

SUMMARY

This report provides the methodology and results of the generation of heatup and cooldown pressure temperature (P-T) limit curves for normal operation of the 1. M. Farley Unit 2 reactor vessel. The heatup and cooldown PoT limit curves were using the adjusted reference (ART) values pertaining to 1. M. Farley Unit 2. The highest ART values pertaining to an axial weld or a were those of intermediate shell plate B7212-1 (using surveillance at both 1/4 thickness (1I4T) and 3/4 thickness (3/4T) locations. The poT curves use of the methodology detailed in the 1998 through the 2000 Addenda Edition of the ASME Code,Section XI, Appendix and ASME Code Case N -641.

The P-T limit curves were generated for 36, 54 and 72 EFPY using heatup rates of 60 and 100°F/hr and cooldown rates of 0, 40, 60 and lOO°F/hr. The curves were developed without margins for instrumentation errors. The curves include a pressure correction for the static and dynamic head loss between the reactor vessel beltline region and the Residual Heat Removal (RJIR) relief valves. These curves can be found in Figures 5-1 through 5-6. Appendix A contains the thermal stress factors for the maximum heatup and cooldown rates for each EFPY term.

WCAP-17123-NP

3 1 INTRODUCTION Heatup and cooldown PoT limit curves are calculated using the adjusted RTNDT (reference nil-ductility temperature) corresponding to the limiting beltline material the reactor vessel. The adjusted RTNDT the limiting material in the core of the reactor vessel is detennined by using the unirradiated reactor vessel material fracture toughness properties, the radiation-induced t.RTNDT, and adding a margin. The un irradiated RTNDT is designated as the higher of either the drop nil-ductility transition (NDTT) or the temperature at which the material exhibits at least 50 ft-Ib of impact energy and 35-mB lateral (nonnal to the major direction) minus 60°F.

RTNDT increases as the material is exposed to fast-neutron radiation. to find the most limiting at any time period in the reactor's t.RTNDT due to the radiation exposure associated with that time period must be added to the un irradiated (IRTNDT). The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials" [Reference I].

Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature values (IRT NDT + t.RTNDT + margins for uncertainties) at the surface, 1/4T and 3/4T locations, where T is the thickness of the vessel at the belt line measured from the clad/base metal interface.

The heatup and cooldown poT limit curves documented in this report were generated using the most limiting ART values and the NRC approved methodology documented in WCAP-14040-A, Revision 4 "Methodology Used to Cold Mitigating Setpolnts and ReS Heatup and Cooldown Limit Curves." the K10 methodology of the 1998 through the 2000 Addenda Edition of ASME Section Xl, Appendix G [Reference 3] was used.

The calculated ART values for 54 and 72 EFPY are documented in Tables 4-8 through 4-13 of this report. The design basis fluence projections are based on the values verified by Westinghouse in letter LTR-REA-09-112, Revision I [Reference 4].

The purpose of this is to present the calculations and the development of the J. M. Unit 2 heatup and cooldown poT limit curves for 36, 54 and 72 EFPY. This report documents the calculated ART values and the developmel'it of the P-T limit curves for nonnal operation. The P-T curves herein were generated without instrumentation errors. The P-T curves contain a pressure correction for the static and dynamic head loss between the reactor vessel beltline region and the RRR relief valves. The poT curves include limits for the vessel flange per the requirements of 10 CFR Part 50, Appendix G WCAP-17123-NP July2011 Revision I

Prnnr;,.I".v Class 3 2-1 2 FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic materials in the J. M. Farley Unit 2 reactor vessel are in Tab Ie 2*1. The unirradiated RTNDT values for the closure head and vessel flange are documented in Table 2-2.

The Regulatory Guide 1.99, Revision 2 methodology used to develop the and cooldown P-T limit curves documented in this is the same as that documented in WCAP*14040-A, Revision 4

[Reference 2]. The chemistry factors (CFs) were calculated using Regulatory Guide 1.99 Revision 2, Position 1.1 and 2.1. Position 1.1 uses the tables from the Regulatory Guide along with the best estimate copper and nickel weight which are in Table 2-1. Position 2.1 CFs are calculated based on the Charpy testing of irradiated surveillance capsule Table 2*3 summarizes the Position 1.1 and 2.1 CFs determined for the J. M. Unit 2 beltline materials.

WCAP-17123-NP Revision I

Westinghouse Non-Proprietary Class 3 2-2 Table 2-1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RTI'iDT Values for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials Fracture Chemical Material Description(a) Toughness Composition(a)

Property(a)

Cu Ni Initial RT NDT Reactor Vessel Location Material ID #

wt. % wt. % CF)

Intermediate Shell (IS) Plate B7203-1 0. 14 0.60 15 Intermediate Shell Plate B7212-l 0.20 0.60 -10 Lower Shell (LS) Plate B7210-1 0.13 0.56 18 Lower Shell Plate B7210-2 0.14 0.57 10 19-923 A _56(b)

IS Longitudinal Weld Seam 0.027 0 .947 (Heat # HODA)19-923 B IS Longitudinal Weld Seam 0.027 0.913 -60 (Heat # BOLA)19-923 B

-?Surveillance Data (Heat # BOLA) 0.028 0.89 -60 I

11-923 IS to LS Circ. Weld Seam 0.153 0.077 -40 (Heat # 5P5622)20-923 A&B LS Longitudinal Weld Seams 0.051 0.096 -70 (Heat # 83640)

Notes for Table 2-1 :

(a) Information source for these material properties is WCAP-14689, Revision 6 [Reference 6). unless otherwise noted.

(b) Estimated per 10CFR 50.61 [Reference 8].

WCAP-17123-NP July 2011 Revision 1

Westinghouse Non-Proprietary Class 3 2-3 Table 2-2 Summary of the Initial RT I\OT Values for the J. M. Farley Unit 2 Closure Head and Vessel Flange Material Identification Initial RTNDT Closure Head Flange(a) -60°F Vessel Flange(b) 60°F Notes for Table 2-2 :

(a) Original J. M. Farley Closure Head was replaced. New material properties are contained in MHI SNC-0455F2 [Reference 7].

(b) Initial RT NOT for the Vessel Flange is taken from WCAP-14689. Revision 6 [Reference 6].

Table 2-3 Summary or the J. M. Farley Unit 2 Reactor Vessel Beltline Material Chemistry Factors per Regulatory Guide 1.99, Revision 2 [Rererence I)

CFper CFper Material Vessel Material Position Position lD#

1.1 ('F) 2.1 CF)

Intermediate Shell Plate B7203-1 100.0 -

Intermediate Shell Plate B7212-1 149.0 144.6 I Lower Shell Plate B72IO-1 89.8 -

I Lower Shell Plate B7210-2 98.7 - 19-923 A IS Longitudinal Weld Seam (Heat # HODA) 36.8 -

J 9-923 IJ IS Longitudinal Weld Seam 36.8 20.7 (Heat # BOLA)19-923 B

~Surveillanee Data (Heat # BOLA) 38.2 -

1 L-923 IS to LS eire. Weld Seam (Heat # 5P5622) 74.1 - 20-923 A&B LS Longitudinal Weld Seam (Heat # 83640) 37.3 -

WCAP-17123-NP July 20]1 Revision I

3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3.1 OVERALL APPROACH The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates that the total stress intensity factor, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity K1e

for the metal temperature at that time. is obtained from the reference toughness curve, defined in the 1998 through the 2000 Addenda of Section XI, Appendix G of the ASME Code 3].

The Kic curve is given by the following equation:

K IC 33.2 + 20.734 *e[002(T -RT,mr)) (I)

where, (ksivin.) reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This K 1c curve is based on the lower bound of static critical Kr values measured as a function of temperature on specimens of SA-533 Grade B Class 1, SA-508-1, and SA-508-3 steel.

3.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT The ITlH,prnilna equation for the heatup-cooldown ,m""v",,, is defined in Appendix G of the ASME Code as follows:

(2)

KIm stress factor caused by membrane (pressure) stress stress intensity factor caused by the thermal gradients reference stress intensity factor as a function ofthe metal tenl0erature T and the metal reference nil-ductility temperature RTNDT C 2.0 for Level A and Level B service limits C 1.5 for hydrostatic and leak test conditions which the reactor core is not critical WCAP-17123-NP

For membrane tension, the corresponding for the postulated defect is:

Kim M", X (pR; II) (3) where, Mm for an inside surface flaw is by:

Mm 1.85 for < 2, Mm 0.926.fi for 2:::;.fi :::; 3.464, 3.21 for > 3.464 Similarly, Mm for an outside surface flaw is given by:

Mm 1.77 for .fi < 2, Mm 0.893.fi for 2'5...Ji '5. 3.464, Mm 3.09 for .fi > 3.464 and p internal pressure (ksi), Ri vessel inner radius (in.), and t vessel wall thickness (in.).

For h",r,ti .."" stress, the ('(Iff:,,Clnl1. K j for the postulated defect is:

where Mb is two-thirds ofM m (4)

The maximum K, produced radial thermal gradient for the postulated inside surface defect of G-2120 is:

= 0.953x 10-3 x CR x (5) where CR is the cooldown rate in °Flhr., or for a postulated outside surface defect Kit O.753xlO* 3 X HU x (6) where HU is the heat up rate in °Flhr.

The through-wall temperature difference associated with the maximum thermal KI can be detennined from ASME Code,Section XI, Appendix G, G-2214-1. The temperature at any radial distance from the vessel surface can be determined from ASME Code, Section Xl, Appendix G, Fig. G-2214-2 for the maximum thenna! K,.

WCAP-17123-NP July 2011 Revision L

(a) maximum thermal relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions in G-2214.3(a)( I) and (b) Alternatively, the Kl for radial thermal gradient can be calculated for any thermal stress distribution and at any "~'VU,,,,y time cooldown for a V4-thickness inside surface defect the relationship:

== (L0359Co + v.v..., ...... ....., + +0.38550)* (7) or similarly, Kit during for a Y4-thickness outside surface defect the relationship:

+O.630C! +OA8IC2 + OAOIC,) >10 (8) where the coefficients Ch and C) are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

Co + CI(X / a) + (9) and x is a variable that rl'irl'~'l'nlt<: the radial distance (in.) from the appropriate (Le., inside or outside) surface to any on the crack front and a is the maximum crack Note that Equations 3, 7, and 8 were implemented in the OPERLIM code, which is the program used to the pressure-temperature (P-T) limit curves. The P-T curve methodology is the same as that described in WCAP-14040-A, Revision 4 "Methodology Used to Cold Overpressure Mitigating Setpoints and RCS Heatup and Cooldown Limit Curves" 2] Section 2.6 (equations 2.6.2-4 and 2.6.3-1).

At any time during the heatup or cooldown is determined the metal at the tip of a postulated flaw postulated flaw has a depth of 1/4 of the section thickness and a length of 1.5 times the section thickness per ASME Code, Section Xl, paragraph G-2(20), the appropriate value for and the reference fracture toughness curve (Equation I). The thermal stresses resulting from the through the vessel wall are calculated and then the corresponding (thermal) stress for the reference flaw are computed. From Equation 2, the pressure stress intensity from these, the allowable pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference 1/4T flaw of Appendix G to Section XI of the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the vessel wall because the thermal gradients, which increase with increasing cooldown rates, produce tensile stresses at the inside surface that would tend to open (propagate) the flaw. Allowable curves are for steady-state and each finite cooldown rate curves, composite limit curves are constructed as the minimum of the steady-state or finite rate curve for each cooldown rate specified.

WCAP-17123-NP July 2011 Revision 1

Class 3 The use of the composite curve in the cooldown is necessary because control of the cooldown procedure is on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 114T vessel location is at a temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any reactor coolant temperature, the AT (temperature) across the vessel wall developed during cooldown results in a higher value of at the l/4T location for finite cooldown rates than for operation. Furthermore, if conditions exist so that the increase in KIc exceeds Kit. the calculated allowable pressure during cooldown will be than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures could be lower if the rate of cooling is decreased at various intervals a cooldown ramp. The use of the curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as weH as finite heatup rate conditions assuming the presence of a 114T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the for the inside 1/4T flaw during is lower than the K 1c for the flaw during steady-state conditions at the same coolant During heatup, especially at the end of the conditions may exist so that the effects of compressive thermal stresses and lower K1c values do not offset each other, and the curve based on conditions no lower bound of all similar curves for finite heatup rates when the 114T flaw is considered.

cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for and finite rates is obtained.

The third portion of the heatup analysis concerns the calculation of the pressure~temperature limitations for the case in which a 114T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside the thermal established at the outside surface during heatup stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heat up and the time coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with rates, each heatup rate must be on an individual basis.

Following the generation of curves for the steady-state and finite heatup rate U""'V"~. the final limit curves are produced by constructing a curve based on a DOIm~nV*DOI comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the least of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

WCAP-17123-NP Revision I

Class 3 3-5 3.3 CLOSURE HEADNESSEL FLANGE REQUIREMENTS 10 CFR Part 50, Appendix G [Reference 5] addresses the metal temperature of the closure head and vessel flange regions. This rule states that the metal temperature of the closure flange must exceed the material unirradiated by at least 120°F for normal operation when the pressure exceeds 20 of the hydrostatic test pressure (3107 for J. M. Unit 2), which is calculated to be 621 psig. The limiting unirradiated of60"F occurs in the vessel of the J. M.

Farley Unit 2 reactor vessel, so the minimum allowable temperature of this region is 180°F at pressures than 621 (without instrument This limit is shown in 5-1 through 5-6 wherever applicable.

WCAP-17123-NP Revision 1

4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide I Revision 2, the adjusted reference temperature (ART) for each material in the beltline is by the following ART Initial RT NDT + L\RTNDT + Margin (10)

Initial RT NDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section IU of the ASME Boiler and Pressure Vessel Code [Reference 9]. If measured values of the initial for the material in question are not mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.

is the mean value of the adjustment in reference temperature caused irradiation and should be calculated as follows:

= CF '" r028-0101ogl) (II)

To calculate L\RT NDT at any depth (e.g., at 1I4T or 3/4T), the following formula must first be used to attenuate the fluence at the depth.

"t(dcplh xl -- ""CC t,u I~ '" e (.IU4x) ( 12) where x inches belt line thickness is 7.875 inches) is the depth into the vessel wall measured from the vessel c1adlbase metal interface. The resultant tluence is then in Equation 11 to calculate the L\RTNDT at the specific depth.

The Westinghouse Radiation and Analysis Group evaluated the vessel fluence projections in LTR-REA J 12, Revision I [Reference 4], and the results are presented in Table 4-1. The evaluation methods used in Reference 4 are consistent with the methods presented in WCAP-14040-A. Revision 4, "Methodology Used to Cold Setpoints and ReS Beatup and Cooldown Limit Curves" [Reference 2]. Tables 4-2 4-4 provide a summary of the vessel fluence projections at the 1I4T and 3/4T locations for 36, 54 and 72 EFPY. Tables 4-5 through 4-7 contain the 1/4T and 3/4T calculated fluences and tluence factors, per Regulatory Guide 1 Revision 2. used to calculate the 36, 54 and 72 EFPY ART values for all beltline materials in the J. M. Farley Unit 2 reactor vessel.

Margin is calculated as M 2 . The standard deviation for the initial RTNOT margin term (Oi) is OaF when the initial RTNOT is a measured value and 17°F when a generic value is available. The standard deviation for the L\RT NOT term, Ot., is 17°F for plates or and 8.5°F for plates or when credible surveillance data is used. For welds, 06 is equal to 28°F when surveillance capsule data is not used, and is 14°F (hal f the value) when credible surveillance data is used. The value for Ot.

need not exceed 0.5 times the mean value of L\RT NDT.

Contained in Tables 4-8 through 4-13 are the 54 and 72 EFPY ART calculations at the 114T and 3/4T locations for of the 1. M. Unit 2 and cooldown curves.

WCAP-17123-NP July 2011 Revision 1

Westinghouse Non-Proprietary Class 3 4-2 Table 4-1 FJuence Values for the J. M. Farley Unit 2 Reactor Vessel BeItline Materials Neutron Fluenee [nIemI, E > 1.0 MeV)

Reactor Vessel Location Material 36EFPY 54 EFPY 72 EFPY Intennediate Shell Plate 87203-1 3.90E+19 5.76E+19 7.63E+19 Intennediate Shell Plate B7212-1 3.90E+19 5.76E+ 19 7.63E+19 19-923 A IS Longitudinal Weld Seam 1.24 E+ I 9 1.83E+19 2.42E+19 (Heat # HODA) 19-9238 IS Longitudinal Weld Seam 1.24E+19 1.83E+19 2.42E+19 (Heat # BOLA)11-923 IS to LS Cire. Weld Seam 3.89E+19 5.75E+19 7.61E+19 (Heat # 5P5622)

Lower Shell Plate B7210-1 3.89E+19 5.75E+19 7.6IE+19 Lower Shell Plate B7210-2 3.89E+ 19 5.75E+19 7.6IE+19 LS Longitudinal 20-923A

1. 24E+19 1.83E+19 2.4IE+19 Weld Seam (Heat # 83640)

LS Longitudinal 20-9238 I. 24E+19 1.83E+19 2.41E+19 Weld Seam (Heat # 83640)

WCAP-17123-NP July 20 II Revision I

Westinghouse Non-Proprietary Class 3 4-3 Table 4-2 Fluence Values for the Vessel Surface, 1I4T and 3/4T Locations for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials at 36 EFPY Fluenee, f (Xl0 19 nlem l , 114 T f(xlO l9 nlem l , 3/4 T f (dO* 9 nleml, Region E> 1.0 MeV) E> 1.0 MeV) E> 1.0 MeV)

Intermediate Shell Plates 3.90 2.431 0.945

~Surveillance Data 3.90 2.431 0.945 Lower Shell Plates 3.89 2.425 0.943 IS Longitudinal Weld Seams 1.24 0.773 0.300

~ Surveillance Data 1.24 0.773 0.300 IS to LS Circ. Weld Seam 3.89 2.425 0.943 LS Longitudinal Weld Seams 1.24 0.773 0.300 Table 4-3 Fluence Values at the Vessel Surface, \/4T and 3/4T Locations for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials at 54 EFPY Fluenee, f(x10 19 nlem l , 114 T f (xlO* 9 nlem 2, 3/4 T f (xlo* 9 niemI, Region E> 1.0 MeV) E> 1.0 MeV) E> 1.0 MeV)

Intermediate Shell Plates 5.76 3.591 1.396

~Surveillance Data 5.76 3.591 1.396 I

Lower Shell Plates 5.75 3.585 1.393 IS Longitudinal Weld Seams 1.83 1.141 0.443

~Surveillance Data 1.83 1.141 0.443 IS to LS Circ. Weld Seam 5.75 3.585 1.393 LS Longitudinal Weld Seams 1.83 1.141 0.443 Table 4-4 Fluencc Values at the Vessel Surface, 1/4T and 3/4T Locations for theJ. M. Farley Unit 2 Reactor Vessel Beltline Materials at 72 EFPY Fluenee, f (xlo '9 niemI, 114 T f (x 10 19 nlemZ, 3/4 T f (xlO* 9 nlem 2, Region E> 1.0 MeV) E> 1.0 MeV) E> 1.0 MeV)

Intermediate Shell Plates 7.63 4.757 1.849

~Surveillance Data 7.63 4.757 1.849 Lower Shell Plates 7.61 4.744 1.844 IS Longitudinal Weld Seams 2.42 1.509 0.586

~Surveillance Data 2.42 1.509 0.586 IS to LS Circ. Weld Seam 7.61 4.744 1.844 LS Longitudinal Weld Seams 2.41 1.502 0.584 WCAP-17123-NP July 2011 Revision I

Westinghouse Non-Proprietary Class 3 4-4 Table 4-5 Fluence Factor Values at the l/4T and 3/4T Locations for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials at 36 EFPY 1I4T f (xl01~ niemI, 3/4T f (x101~ niemI, Region 1I4T FF 3/4T FF E> 1.0 MeV) E> 1.0 MeV)

Intennediate Shell Plates 2.431 1.2392 0.945 0.9842

~Surveillanee Data 2.431 1.2392 0.945 0.9842 I Lower Shell Plates 2.425 l.2386 0.943 0.9834 IS Longitudinal Weld Seams 0.773 0.9278 0.300 0.6707

~Surveillanee Data 0.773 0.9278 0.300 0.6707 IS to LS eire. Weld Seam 2.425 1.2386 0.943 0.9834 LS Longitudinal Weld Seams 0.773 0.9278 0.300 0.6707 Table 4-6 Fluence Factor Values at the 1/4T and 3/4T Locations for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials at 54 EFPY 1I4T f(xl01~nleml, 3/4T f (xlO '9 n/em 1, Region 1I4T FF 3/4T FF E> 1.0 MeV) E> 1.0 MeV)

Intennediate Shell Plates 3.591 1.3324 1.396 1.0926

~Surveillanee Data 3.591 1.3324 1.396 1.0926 Lower Shell Plates 3.585 1.3320 1.393 1.0921 IS Longitudinal Weld Seams 1.141 1.0368 0.443 0.7738

~Surveil1anee Data 1.141 1.0368 0.443 0.7738 IS to LS eire. Weld Seam 3.585 1.3320 1.393 1.0921 LS Longitudinal Weld Seams \.141 1.0368 0.443 0.7738 Table 4-7 Fluence Factor Values at the 1/4T and 3/4T Locations for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials at 72 EFPY 1/4T f (x10 19 nlem\ 3/4T f (xIOI~ nlcml, Region 1I4T FF 3/4T FF E> 1.0 MeV) E> 1.0 MeV)

Intennediate Shell Plates 4.757 1.3924 1.849 1.1685

~Surveillanee Data 4.757 1.3924 1.849 1.1685 Lower Shell Plates 4.744 1.3919 1.844 1.1678 IS Longitudinal Weld Seams 1.509 1.1138 0.586 0.8506

~Surveillanee Data 1.509 1.1138 0.586 0.8506 IS to LS eire. Weld Seam 4.744 1.3919 1.844 1.1678 LS Longitudinal Weld Seams 1.502 1.1127 0.584 0.8494 WCAP-17l23-NP July 20 II Revision I

Westinghouse Non-Proprietary Class 3 4-5 Table 4-8 Adjusted Reference Temperature Evaluation for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials through 36 EFPY at the 1I4T Location 1I4T f Material CF 1I4T RTNDT(U) ~RTNDT 0, 04 M ART Reactor Vessel Location (XlO I9 n/cm 2, (OF) FF (oF) (oF) (oF) (OF) (OF) (OF)

E> 1.0 MeV)

Intermediate Shell Plate B7203-1 100.0 2.431 1.2392 [5 [23.9 0 17 34.0 173 Intermediate Shell Plate B7212-1 149.0 2.431 1.2392 -[0 [84.6 0 17 34.0 209

-7 Surveillance Data B7212-1 144.6 2.431 1.2392 -10 179.2 0 8.5(' ) 17.0 186 Lower Shell Plate B7210-1 89.8 2.425 1.2386 18 111.3 0 17 34.0 163 Lower Shell Plate B7210-2 98.7 2.425 1.2386 10 122.3 0 17 34.0 166 Inter. Shell Longitudina[ [9-923 A 36.8 0.773 0.9278 -56 34 .1 17 17.1 48.2 26 Weld Seam (Heat # HODA)

Inter. Shell Longitudinal 19-923 B 36.8 0.773 0.9278 -60 34. 1 0 17. 1 34.1 8 Weld Seam (Heat # BOLA)19-923 B

-7 Surveillance Data 20.7 0.773 0.9278 -60 19.2 0 9.6 19.2 -22 (Heat # BOLA)

IS to LS Circumferential 11-923 74.1 2.425 1.2386 -40 91.8 0 28 56.0 108 Weld (Heat # 5P5622)

Lower Shell Longitudinal 20-923 A&B 37.3 0.773 0.9278 -70 34.6 0 17.3 34.6 -I Weld Seams (Heat # 83640)

Note:

(a) Per WCAP-16918-NP, Revision 1 [Reference 10], the intermediate shell plate B7212-1 surveillance data was deemed credible. Therefore, a reduced all value is used.

WCAP-17123- NP Ju[y 2011 Revision I

Westinghouse Non-Proprietary Class 3 4-6 Table 4-9 Adjusted Reference Temperature Evaluation for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials through 36 EFPY at the 3/4T Location 3/4T f Material CF 3/4T RTNDT(U) ARTNDT (JI (JA M ART Reactor Vessel Location (xlO"n/cm 2 ,

CO F) FF COF) (oF) (oF) COF) (oF) (oF)

E> 1.0 MeV)

Intermediate Shell Plate B7203-1 100.0 0.945 0.9842 15 98.4 0 17 34.0 147 Intermediate Shell Plate B7212-1 149.0 0.945 0.9842 -10 146.6 0 17 34.0 171

-7Surveillance Data B72l2-1 144.6 0.945 0.9842 -10 142.3 0 8.5(a) 17.0 149 Lower Shell Plate B72I0-1 89.8 0.943 0.9834 18 88.3 0 17 34.0 140 Lower Shell Plate B72IO-2 98.7 0.943 0.9834 10 97.1 0 17 34.0 141 Inter. Shell Longitudinal Weld Seam 19-923 A (Heat # HODA) 36.8 0.300 0.6707 -56 24.7 17 12.3 42.0 II l Inter. Shell Longitudinal 19-923 B 36.8 0.300 0.6707 -60 24.7 0 12.3 24.7 -11 Weld Seam (Heat # BOLA)19-923 B

-7 Surveillance Data 20.7 0.300 0.6707 -60 13.9 0 6.9 13.9 -32 (Heat # BOLA)

IS to LS Circumferential 11-923

74. 1 0.943 0.9834 -40 72.9 0 28 56.0 89 Weld (Heat # 5P5622)

Lower Shell Longitudinal 20-923 A&B 37.3 0.300 0.6707 -70 25.0 0 12.5 25.0 -20 Weld Seams (Heat # 83640)

Note:

(a) PerWCAP-16918-NP, Revision I [Reference 10], the intermediate shell plate B7212-1 surveillance data was deemed credible. Therefore, a reduced crt. value is used.

WCAP-17123-NP July 2011 Revision 1

Westinghouse Non-Proprietary Class 3 4-7 Table 4-10 Adjusted Reference Temperature Evaluation for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 1I4T Location 1I4T f Material CF l!4T RTNDT(U) ARTNDT 0"1 O"A M ART Reactor Vessel Location (xtO I ' n/cm 2, (oF) FF (OF) eF) eF) (oF) eF) eF)

E> 1.0 MeV)

Intennediate Shell Plate B7203-1 100.0 3.591 1.3324 15 133.2 0 17 34.0 182 Intennediate Shell Plate B7212-1 149.0 3.591 1.3324 -10 198.5 0 17 34.0 223

-7Surveillance Data 87212-1 144.6 3.591 1.3324 -10 192.7 0 8.5(8) 17.0 200 I Lower Shell Plate B721O-1 89.8 3.585 1.3320 18 119.6 0 17 34.0 172 I

Lower Shell Plate B7210-2 98.7 3.585 1.3320 10 131.4 0 17 34.0 175 I Inter. Shell Longitudinal 19-923 A 36.8 l.141 1.0368 -56 38.1 17 19.1 51.1 33 Weld Seam (Heat # HODA)

Inter. Shell Longitudinal 19-923 B 36.8 1.141 1.0368 -60 38 .1 0 19.1 38. 1 16 Weld Seam (Heat # BOLA)19-923 B

-7 Surveillance Data 20.7 1.141 1.0368 -60 21.5 0 10.7 21.5 -17 (Heat # BOLA)

IS to LS Circumferential 11-923 74 .1 3.585 1.3320 -40 98 .7 0 28 56.0 115 Weld (Heat # 5P5622)

Lower Shell Longitudinal 20-923 A&B 37.3 1.141 1.0368 -70 38.7 0 19.3 38.7 7 Weld Seams (Heat # 83640)

Note:

(a) Per WCAP-16918-NP, Revision I [Reference 10], the intennediate shell plate B7212-1 surveillance data was deemed credible. Therefore, a reduced CJt; value is used.

WCAP-17123-NP July 2011 Revision I

Westinghouse Non-Proprietary Class 3 4-8 Table 4-11 Adjusted Reference Temperature Evaluation for tbe J. M. Farley Unit 2 Reactor Vessel Beltline Materials tbrougb 54 EFPY at the 3/4T Location 3/4T f Material CF 3/4T RTNUT{U) ARTNUT 0'& M ART Reactor Vessel Location (xlO l9 D/cm z, 0'1 (OF)

(OF) FF eF) (oF) (OF) (OF) (OF)

E> 1.0 MeV)

Intermediate Shell Plate B7203- 1 100.0 1.396 1.0926 15 109.3 0 17 34.0 158 !

I Intermediate Shell Plate B7212- 1 149.0 1.396 1.0926 -10 162.9 0 17 34.0 187 I

-?SurveiUance Data B7212-1 144.6 1.396 1.0926 -10 158.1 0 8.5(a) 17.0 165 Lower Shell Plate B72IO-1 89.8 1.393 \.0921 18 98.0 0 17 34.0 150 Lower Shell Plate B72IO-2 98.7 1.393 1.0921 10 107.7 0 17 34.0 152 Inter. SheIl Longitudinal 19-923 A 36.8 0.443 0.7738 -56 28.5 17 14.2 44.3 17 Weld Seam (Heat # HODA)

Inter. Shell Longitudinal 19-923 B 36.8 0.443 0.7738 -60 28.5 0 14.2 28.5 -3 Weld Seam (Heat # BOLA)19-923 B

-? Surveillance Data 20.7 0.443 0.7738 -60 16.0 0 8.0 16.0 -28 (Heat # BOLA)

IS to LS Circumferential 11-923 74.1 1.393 1.0921 -40 80.9 0 28 56.0 97 Weld (Heat # 5P5622)

Lower Shell Longitudinal 20-923A&B 37.3 0.443 0.7738 -70 28.9 0 14.4 28.9 -12 Weld Seams (Heat # 83640)

Note:

(a) Per WCAP-16918-NP, Revision I [Reference 10], the intermediate sheIl plate B7212-1 surveillance data was deemed credible. Therefore, a reduced all value is used.

WCAP-17123-NP July 2011 Revision I

Westinghouse Non-Proprietary Class 3 4-9 Table 4-12 Adjusted Reference Temperature Evaluation for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials through 72 EFPY at the 1I4T Location 1I4T f Material CF l!4T RTNOT(U) dRTNOT (1, (16 M ART Reactor Vessel Location (Xl0 19 n/cm 2, (OF) FF (OF) (oF) (oF) (OF) (OF) (OF)

E> 1.0 MeV)

Intennediate Shell Plate B7203-1 100.0 4.757 1.3924 15 139.2 0 17 34.0 188 Intennediate Shell Plate B7212-1 149.0 4.757 1.3924 -10 207.5 0 17 34.0 231

-7SurveiUance Data B7212- 1 144.6 4.757 1.3924 -10 201.3 0 8.5(0) 17.0 208 Lower Shell Plate B72IO-1 89.8 4.744 1.3919 18 125.0 0 17 34 .0 177 Lower Shell Plate B72IO-2 98.7 4.744 1.3919 10 137.4 0 17 34.0 181 I Inter. Shell Longitudinal 19-923 A 36.8 1.509 l.\138 -56 41.0 17 20.5 53.3 38 Weld Seam (Heat # HODA) I Inter. Shell Longitudinal 19-923 B 36.8 1.509 1.\ 138 -60 41.0 0 20.5 41.0 22 I Weld Seam (Heat # BOLA)19-923 B

-7 Surveillance Data 20.7 1.509 1.1138 -60 23. 1 0 11.5 23.1 -14 (Heat # BOLA)

IS to LS Circumferential 11-923 74.1 4.744 1.3919 -40 103.1 0 28 56.0 119 Weld (Heat # 5P5622)

Lower Shell Longitudinal 20-923A&B 37.3 1.502 1.1127 -70 41.5 0 20.8 41.5 13 Weld Seams (Heat # 83640)

Note:

(a) Per WCAP-16918-NP, Revision I [Reference 10], the intennediate shell plate B7212-1 surveillance data was deemed credible. Therefore ,

a reduced (JA value is used.

WCAP-17123-NP July 2011 Revision I

Westinghouse Non-Proprietary Class 3 4-10 Table 4-13 Adjusted Reference Temperature Evaluation for the J. M. Farley Unit 2 Reactor Vessel Beltline Materials through 72 EFPY at the 3/4T Location 3/4T f Material CF 3/4T RTNDT(U) ARTNDT 0") O"d M ART Reactor Vessel Location (xl0 19 n/cm\ E >

(oF) FF (oF) (oF) (oF) (oF) (oF) (oF) 1.0 MeV)

Intennediate Shell Plate B7203-1 100.0 1.849 1.1685 15 116.8 0 17 34.0 166 Intennediate Shell Plate B7212-1 149.0 1.849 1.1685 -10 174.1 0 17 34.0 198

~SurveiUance Data B7212-1 144.6 1.849 1.1685 -10 169.0 0 8.5(0) 17.0 176 Lower Shell Plate B7210-1 89 .8 1.844 1.1678 18 104.9 0 17 34.0 157 Lower Shell Plate B72IO-2 98.7 1.844 1.1678 10 115.3 0 17 34.0 159 Inter. Shell Longitudinal Weld 19-923 A 36.8 0.586 0.8506 -56 31.3 17 15.7 46.2 22 Seam (Heat # HODA) I Inter. Shell Longitudinal Weld 19-923 B 36.8 0.586 0.8506 -60 31.3 0 15.7 31.3 3 Seam (Heat # BOLA)19-923 B

~ Surveillance Data 20.7 0.586 0.8506 -60 17.6 0 8.8 17.6 -25 (Heat # BOLA)11-923 IS to LS Circumferential Weld 74.1 1.844 1.1678 -40 86.5 0 28 56.0 103 (Heat # 5P5622)

Lower Shell Longitudinal 20-923A&B 37.3 0.584 0.8494 -70 31.7 0 15.8 31.7 -7 Weld Seams (Heat # 83640)

Note:

(a) Per WCAP-16918-NP, Revision 1 [Reference 10J, the intennediate shell plate B7212-1 surveillance data was deemed credible. Therefore, a reduced (JA value is used.

WCAP-17123-NP July 2011 Revision 1

_l'r"nrllP!l1rv Class 3 4-11 Contained in Table 4-14 is a summary of the limiting ART values used in the of the 1. M.

Farley Unit 2 reactor vessel poT limit curves. The limiting material for both the 1I4T location and the 3/4T location at 36, 54, and 72 EFPY is Intermediate Shell Plate B7212-1. Note that Regulatory Guide 1.99 Revision 2 I] allows the use of a lower factor, when credible surveillance vaL'i>Uj,,, data is to calculate ART. The values listed in Table 4-14 are based on credible surveillance capsule data.

Table 4-14 Summary of the Limiting ART Values Used in the Generation of the J. M. Unit 2 Heatup/Cooldown Curves Limiting ART (oF)

Intermediate Shell Plate B7212-1 EFPY with credible surveiUance data Ii 1/4T 3/4T I 36 186 149 II 54 200 165 1 72 208 176 WCAP-17123-NP Revision I

5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal and cooldown of the primary reactor coolant have been calcu tated the pressure and temperature in the reactor vessel beltline region using the methods discussed in Sections 3 and 4 of this report. This approved methodology is also presented in WCAP-14040-A, Revision 4.

5-1,5-3 and 5-5 the limiting instrumentation errors using heatup rates of 60 and lOO°F/hr for 36, 54 and 72 respectively, with the "Flange-Notch" and using the "Axial-flaw" methodology. 5-2, 5-4 and 5-6 the limiting cooldown curves without for instrumentation errors cooldown rates of 0, 20, 40, 60 and lOO°F/hr applicable for 54 and 72 with the "Flange-Notch" and using the "Axial-flaw" methodology. The heatup and cool down curves were generated the 1998 the 2000 Addenda ASME Code Section XI, G. Also, a pressure correction for the static and dynamic head loss between the reactor vessel beltline and the RHR relief valves is included for both the heatup and cooldown curves at each EFPY. These curves incorporate a pressure correction of 27 for temperatures less than IIO*F and 60 for greater than or equal to IIO*F, associated with operation of one and three reactor coolant pumps, respectively 6].

Allowable combinations of temperature and pressure for specific rates are below and to the right of the limit lines shown in 5-1 through 5-6. This is in addition to other criteria, which must be met before the reactor is made as discussed in the following paragraphs.

The reactor must not be made critical until pressure-temperature combinations are to the of the limit tine shown in 5-1, 5-3 and 5-5 (heatup curves only). The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test, as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic testis defined in the 1998 through the 2000 Addenda ASME Code Section Appendix G as follows:

where, is the stress factor covered by membrane (pressure) stress, IT. RTNDT)]

Kic == 33.2 + 20.734 T is the minimum permissible metal temperature, and is the reference nil-ductility 'I-"nnne'r",

The criticality limit curve pressure-temperature limits for core operation in order to provide additional margin during actual power production. The limits for core operation for low power physics are that: I) the reactor vessel must be at a temperature to or higher than the minimum temperature required for the inservice hydrostatic test, and 2) the reactor vessel WCAP-17123-NP July 2011 Revision I

.PTj~nr;i"'I"rv Class 3 5-2 must be at least 40°F than the minimum permissible temperature in the corresponding pressure curve for and cooldown calculated as described in Section 4 of this For the and cooldown curves without margins for instrumentation errors, the minimum for the inservice hydrostatic leak test for the J. M. Unit 2 reactor vessel at 36 EFPY is 242°F. The limits for 54 and 72 EFPY are 256°F and 264°F, The vertical line drawn from these on the curve, intersecting a curve 40°F than the pressure-limit curve, constitutes the limit for core operation for the reactor vessel.

5-1 5-6 define all of the above limits for prevention of non-ductile failure for the J. M. Farley Unit 2 reactor vessel for 36, 54 and 72 EFPY WiUl the "Flange-Notch" requirement, without instrumentation and with pressure correction. The data points used for the heatup and cool down limit curves shown in 5-1 through 5-6 are in Tables 5-1 5-6.

WCAP-17123-NP Revision 1

Westinghouse Non-Proprietary Class 3 5-3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intennediate Shell Plate B7212-1 with credible surveillance data LIMITING ART VALUES AT 36 EFPY: 1/4T, 186°F 3/4T, 149°F Figure 5-1 J. M. Farley Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 36 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda App. G Methodology (wiKle) 2500 2250 ~,.~.

ILeak Test limit I -.

2000 ,I 1750 Unacceptable

. --i

-!:2 Operation I

(I)

D. 1500 ~ .. t .__

I 1-- -t- ... - ;

Q)

I en en

~-"

Q)

D.

1250 I'" . ..

I "C

Q)

I (J

1000 iij U

750 I

-r---~- -,-i

"' r ---- r 500 " ..... - Criticality Limit based on I Inservlce hydrostatic test

+- temperature (242'F) for the service period up to 36 EFPY 250 O-~~++~~~~~~+r~~~~~~~~~~~~~~~

o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Oe9. F)

WCAP-17123-NP July 2011 Revision I

Westinghouse Non-Proprietary Class 3 5-4 MATERIAL PROPERTY BASIS LIMITING MATERIAL : Intennediate Shell Plate B7212-1 with credible surveillance data LIMITING ART VALUES AT 36 EFPY: 1I4T, 186°F 3/4T, 149°F Figure 5-2 J. M. Farley Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 36 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda App. G Methodology (w/KI<)

2500Tr===========C=~======~------:------:-----~-------~

Operllm Version:5.2 Run:30256 Operlim. x1s Version: 5.2 i

2250 I .~ . ...-'----. .I ,,.

.... _I . , . ~-. +

2000 . , -, ~ , '

i

.... - .. ~ {, . ,-1 I

1750 ..- .

I

. -~ . ' -- " :

(!)

(i)

I :

!:. 1500 1-I Q)

II

J 1/1 1/1

.... 1250 Q) a..

't:J

..- . ~ -, +--

... .., I . I Q)

I

~ I --1., _L_

J 1000 ,

U "iij U

, i 750 *r ~* * .- ;.. . . L.

, i,* * - \

Cooldown 500 Rates

'F/Hr 250 I

_ ,. J . _. . . .

o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

WCAP.. 17123..NP July 201 1 Revision 1

Westinghouse Non-Proprietary Class 3 5-5 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intennediate Shell Plate 872) 2-1 with credible surveillance data UMITING ART VALUES AT 54 EFPY: 1I4T, 200°F 3/4T, 165°F Figure 5-3 J. M. Farley Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 54 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda App. G Methodology (wiKle) 2500rr=============~========,----'------------~--,

Ope~ im Version:5.2 Run:1180 Ope~im . x1s Version: 5,2 ILeak Test"r Limit..."I ,

I 2250 " . 'V **

T "T-- r-I, 2000 r-

} - .. "

1750 --'r, ' .. ~

1** **

I

§'

~ 1500 1---

...__J

' r-- "

CD

~

J In In 1250 ,t CD

~

a.

"'0 CD

~

J 1000 nI U

750 500 ,- Criticality limit based on inservlce hydrostatic test

+ - temperature (256*F) for the i service period up to 54 EFPY ,

250

- : '~l' ~~~u: I

.~-- ~-,

I I o

o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

WCAP-17123-N P July 2011 Revision 1

Westinghouse Non-Proprietary Class 3 5-6 MATERIAL PROPERTY BASIS LIMITING MATERlAL: Intermediate Shell Plate B7212-1 with credible surveillance data LIMITING ART VALUES AT 54 EFPY: 1/4T, 200°F 3/4T, 165°F Figure 5-4 J. M. Farley Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 54 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda App. G Methodology (w/K 1<)

2500~~~~~~~~~~~~r-'---~------~--~---,

Oper1im Version :5.2 Run:llBO Oper1im .xls Version: 5 .2 I

2250 -

-'r --f I

I 2000 ,,;"

1750 .~ ,,- "

6' ena.. I Q)

J 1500 - "i -- . i III III Q) a..

1250 J i

-~ !.

- I i

-. -j.. 4 -1._

I "C

n; Q)

.!l!

J

(.) 1000 . - , ---~

U 750 .I* .

I "i i ' r' 1

500 *' - 1' " Cooldown . ~

steady-state 1._ -20 250

-40

-60

-100 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Oe9. F)

WCAP-17123-NP July 2011 Revision I

Westinghouse Non-Proprietary Class 3 5-7 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intennediate Shell Plate B7212-1 with credible surveillance data LIMITING ART VALUES AT 72 EFPY: 1/4T, 208°F 3/4T, 176°F Figure 5-5 J. M. Farley Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 1000F/hr) Applicable for 72 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda App. G Methodology (wIKle) 2500 Operlim Version:5.2 Run:1720 Operlim.xls Version: 5.2 2250 -I 2000 1750 C)

~ 1500 .,1, __ R , \ ,

l!!

l en en l!! 1250 . _L

(

a..

"C CII 1'0

'S u 1000 -- -! ~

-;

U 750 '.---- . .. :-

Criticality limit based on I I===+-~.....J ~ inservice hydros~atic test 500* _ .... J, , temperature (264 F) for the service period up to 72 EFPY I

250 - - _ L..

I

~ .. ,

I i

, ~ Boltup I I I Tem9' o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

WCAP-17123-NP July 2011 Revision I

Westinghouse Non-Proprietary Class 3 5-8 MATERIAL PROPERTY BASIS LIMITING MATERlAL: Intennediate Shell Plate B7212-1 with credible surveillance data LIMITING ART VALUES AT 72 EFPY: 1I4T, 208°F 3/4T, 176°F Figure 5-6 J. M. Farley Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°Flbr) Applicable for 72 EFPY (without Margins for Instrumentation Errors and with Pressure Correction) Using the 1998 through the 2000 Addenda App. G Methodology (WIKle) 2500 Opertim Version:5.2 Run:1720 Opertim.>ds Version: 5.2 I

I i 2250

-t-I

~

I 2000

-___1 I

, ~- ..,,...

~r I

1750 - ... --+.- .--. ___1 . --" - ---t.. . " . ~-

I I I i

6'

~ 1500 ,-,,--,.. - ,..

.r:-:--..L-,--;-:--,'

e

~

III III i l -L _I ~

e 1250 11.

.---.i _ __

t--

-B "C

Q) iii 1000 I

.... ~ . _. I

+

u I

750 f*- - f' ---r-I 500 .._.j . Cooldown .t ..

I 250 o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Oeg. F)

WCAP-17123-NP July 2011 Revision I

Westinghouse Non-Proprietary Class 3 5-9 Table 5-1 36 EFPY Heatup Curve Data Points Using the 1998 through the 2000 Addenda App. G Methodology (wI KIt> wI Flange Notch, wI Pressure Correction and wlo Uncertainties for Instrumentation Errors) 60°Flhr lOO°Flhr lOO°Flhr Leak Test Limit 60°F/hr Heatup Criticality Heatup Criticality T T T P T P T P P (psig) P (pslg)

(OF) eF) (oF) (psig) (oF) (psig) (oF) (psig) 224 2000 60 0 242 0 60 0 242 0 242 2485 60 594 242 561 60 594 242 561 65 594 242 561 65 594 242 561 70 594 242 561 70 594 242 561 75 594 242 561 75 594 242 561 80 594 242 561 80 594 242 561 85 594 242 561 85 594 242 561 90 594 242 561 90 594 242 561 95 594 242 561 95 594 242 561 100 594 242 561 100 594 242 561 105 594 242 561 105 594 242 561 110 594 242 561 110 594 242 561 110 561 242 561 110 561 242 561 115 561 242 561 115 561 242 561 120 561 242 561 120 561 242 561 125 561 242 561 125 561 242 561 130 561 242 561 DO 561 242 561

\35 561 242 561 135 561 242 561 140 561 242 561 140 561 242 561 145 561 242 561 145 561 242 561 150 561 242 561 150 561 242 561 155 561 242 561 155 561 242 561 160 56! 242 561 160 56! 242 561 165 561 242 561 165 561 242 561 170 561 242 561 170 561 242 561 175 561 242 561 175 561 242 561 180 561 242 955 180 561 242 768 180 561 242 1002 180 561 242 801 180 934 242 1054 180 768 242 836 185 971 242 1095 185 801 242 876 190 1012 242 1130 190 836 242 940 195 1058 245 1164 195 876 245 969 200 1108 250 1225 200 920 250 1023 WCAP-17123-NP July 2011 Revision 1

Westinghouse Non-Proprietary Class 3 5-10 60°Flhr IOO°F/hr IOO°F/hr Leak Test Limit 60°F/hr Heatup Criticality Heatup Criticality T T T P T P T P (oF)

P (pslg) P (psig)

(OF) (oF) (psig) (OF) (psig) (oF) (psig) 205 1164 255 1279 205 969 255 1083 210 1225 260 1338 210 1023 260 1148 215 1279 265 1402 215 1083 265 1221 220 1338 270 1473 220 1148 270 1302 225 1402 275 1552 225 1221 275 1390 230 1473 280 1638 230 1302 280 1488 235 1552 285 1734 235 1390 285 1596 240 1638 290 1839 240 1488 290 1715 245 1734 295 1956 245 1596 295 1847 250 1839 300 2084 250 1715 300 1973 255 1956 305 2226 255 1847 305 2093 II 260 2084 310 2383 260 1973 310 2225 265 2226 265 2093 315 2371 270 2383 270 2225

, 275 2371 WCAP-17123-NP July 2011 Revision 1

Westinghouse Non-Proprietary Class 3 5-11 Table 5-2 36 EFPY Cooldown Curve Data Points Using the 1998 through the 2000 Addenda App. G Methodology (wi K. c, wI Flange Notch, wI Pressure Correction and wlo Uncertainties for Instrumentation Errors)

Steady State 20°F/hr. 40°F/hr. 60°F/hr. lOO°F/hr.

T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) T(OF) P (psig) T(°F) P (psig) 60 0 60 0 60 0 60 0 60 0 60 594 60 594 60 562 60 519 60 432 i 65 594 65 594 65 565 65 523 65 436 70 594 70 594 70 569 70 527 70 441 i 75 594 75 594 75 574 75 532 75 446 I

i 80 594 80 594 80 579 80 537 80 452 85 594 85 594 85 584 85 543 85 459 90 594 90 594 90 591 90 550 90 466 95 594 95 594 95 594 95 557 95 475 100 594 100 594 100 594 100 565 100 484 105 594 105 594 105 594 105 574 105 495 110 594 110 594 110 594 110 584 110 506 110 561 110 561 110 561 110 551 110 473 115 561 115 561 115 561 115 561 115 487 120 561 120 561 120 561 120 561 120 501 125 561 125 561 125 561 125 561 125 517 130 561 130 561 130 561 130 561 130 536 135 561 135 561 135 561 135 561 135 556 140 561 140 561 140 561 140 561 140 561 145 561 145 561 145 561 145 561 145 561 150 561 150 561 150 561 150 561 150 561 155 561 155 561 155 561 155 561 155 561 160 561 160 561 160 561 160 561 160 561 165 561 165 561 165 561 165 561 165 561 170 561 170 561 170 561 170 561 170 561 175 561 175 561 175 561 175 561 175 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 934 180 917 180 902 180 890 180 876 185 971 185 957 185 946 185 938 185 933 190 1012 190 1002 190 995 190 991 190 991 195 1058 195 1051 195 1049 195 1049 195 1049 WCAP-17123-NP July2011 Revision I

Westinghouse Non-Proprietary Class 3 5-12 Steady State 20°F/hr. 40°Flhr. 60°Flhr. 100°F/hr.

T(OF) P (psig) TeF) P (psig) TeF) P (psi g) T(OF) P (psig) T(OF) P (psig) 200 1108 200 1106 200 1106 200 1106 200 1106 205 1164 205 1164 205 1164 205 1164 205 1164 210 1225 210 1225 210 1225 210 1225 210 1225 215 1293 215 1293 215 1293 215 1293 215 1293 220 1368 220 1368 220 1368 220 1368 220 1368 225 1451 225 1451 225 1451 225 1451 225 1451 230 1542 230 1542 230 1542 230 1542 230 1542 235 1644 235 1644 235 1644 235 1644 235 1644 240 1756 240 1756 240 1756 240 1756 240 1756 245 1879 245 1879 245 1879 245 1879 245 1879 I 250 2016 250 2016 250 2016 250 2016 250 2016 255 2167 255 2167 255 2167 255 2167 255 2167 260 2334 260 2334 260 2334 260 2334 260 2334 WCAP-17123-NP July 2011 Revision 1

Westinghouse Non-Proprietary Class 3 5-13 Table 5-3 54 EFPY Heatup Curve Data Points Using the 1998 through the 2000 Addenda App. G Methodology (wI Krc> wI Flange Notch, wI Pressure Correction and wlo Uncertainties for Instrumentation Errors) 60°F/hr IOO°F/hr IOO°F/hr Leak Test Limit 60°F/hr Heatup Criticality Heatup Criticality T T T P T P T P P (psig) P (psig)

(oF) (OF) (oF) (psi g) (oF) (psig) (oF) (psi g) 238 2000 60 0 256 0 60 0 256 0 256 2485 60 594 256 561 60 575 256 542 65 594 256 561 65 575 256 542 70 594 256 561 70 575 256 544 75 594 256 561 75 575 256 544 80 594 256 561 80 575 256 546 85 594 256 561 85 575 256 547 90 594 256 561 90 575 256 551 95 594 256 561 95 575 256 551 100 594 256 561 100 575 256 556 105 594 256 561 105 575 256 557 110 594 256 561 110 575 256 561 110 561 256 561 110 542 256 561 115 561 256 561 115 542 256 561 120 561 256 561 120 544 256 561 125 561 256 561 125 547 256 561 130 561 256 561 130 551 256 561 135 561 256 561 135 556 256 561 140 561 256 561 140 561 256 561 145 561 256 561 145 561 256 561 150 561 256 561 150 561 256 561 155 561 256 561 155 561 256 561 160 561 256 561 160 561 256 561 165 561 256 561 165 561 256 561 170 561 256 561 170 561 256 561 175 561 256 561 175 561 256 561 180 561 256 828 180 561 256 675 180 561 256 862 180 561 256 698 180 828 256 900 180 675 256 724 185 862 256 941 185 698 256 752 190 900 256 987 190 724 256 784 195 941 256 1038 195 752 256 819 200 979 256 1094 200 784 256 858 WCAP-17 I 23-NP july 201 I Revision 1

Westinghouse Non*Proprietary Class 3 5*14 60°Flhr lOO°Flhr lOO°F/hr Leak Test Limit 60°F/hr Heatup Criticality Heatup Criticality T T T P T P T P P (psig) P (psig)

(oF) eF) e F) (psig) (oF) (psig) eF) (pslg)

I 205 1021 256 1130 205 819 256 910 210 1067 260 1175 210 858 260 949 215 1I 19 265 1237 215 901 265 1001 220 1175 270 1291 220 949 270 1059 225 1237 275 1350 225 1001 275 1123 230 1291 280 1416 230 1059 280 1194 235 1350 285 1488 235 1123 285 1272 I

240 1416 290 1568 240 1194 290 1359 245 1488 295 1656 245 1272 295 1454 250 1568 300 1753 250 1359 300 1559 255 1656 305 1861 255 1454 305 1675 260 1753 310 1979 260 1559 310 1803 265 1861 315 2110 265 1675 315 1944 270 1979 320 2254 270 1803 320 2099 275 2110 325 2414 275 1944 325 2250 280 2254 280 2099 330 2399 285 2414 285 2250 290 2399 WCAP*17123*NP July 2011 Revision I

Westinghouse Non-Proprietary Class 3 5-15 Table 5-4 54 EFPY Cooldown Curve Data Points Using the 1998 through the 2000 Addenda App. G Methodology (wI Kin wI Flange Notch, wI Pressure Correction and wlo Uncertainties for Instrumentation Errors)

Steady State 20°Ffhr. 40°F/hr. 6O°F/hr. lOO°Ffhr.

,I T(OF) P (psig) T(°F) P (psig) TeF) P (psig) T(°F) P (pslg) T(°F) P (psig) 60 0 60 0 60 0 60 0 60 0 60 594 60 594 60 553 60 509 60 420 65 594 65 594 65 555 65 512 65 423 70 594 70 594 70 558 70 515 70 426 75 594 75 594 75 561 75 518 75 430 80 594 80 594 80 565 80 522 80 434 85 594 85 594 85 569 85 527 85 439 90 594 90 594 90 574 90 531 90 445 95 594 95 594 95 579 95 537 95 451 100 594 100 594 100 585 100 543 100 45 8 105 594 105 594 105 591 105 550 105 465 110 594 . 110 594 110 594 110 557 110 474 110 561 110 561 110 561 110 524 110 441 :

115 561 115 561 115 561 115 532 115 451 120 561 120 561 120 561 120 542 120 461 125 561 125 561 125 561 125 552 125 473 130 561 130 561 130 561 130 561 130 487 135 561 135 561 135 561 135 561 135 502 140 561 140 561 140 561 140 561 140 519 145 561 145 561 145 561 145 561 145 537 150 561 150 561 150 561 150 561 150 558 155 561 155 561 155 561 155 561 155 561 160 561 160 561 160 561 160 561 160 561 165 561 165 561 165 561 165 561 165 561 170 561 170 561 170 561 170 561 170 561 175 561 175 561 175 561 175 561 175 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 847 180 823 180 799 180 778 180 741 185 875 185 853 185 832 185 814 185 785 190 906 190 887 190 869 190 854 190 833 195 941 195 924 195 910 195 898 195 886 WCAP-17123-NP July 2011 Revision I

Westinghouse Non-Proprietary Class 3 5-16 Steady State 20°F/hr. 40°F/hr. 60°F/hr. IOO°F/hr.

T(OF) P (psig) T(°F) P (psig) T(°F) P (psig) T(0F) P (psig) T(°F) P (psig) 200 979 200 966 200 955 200 948 200 945 205 1021 205 1011 205 1005 205 1002 205 1002 210 1067 210 1062 210 1060 210 1060 210 1060 215 1119 215 1118 215 1118 215 1118 215 1118 220 1175 220 1175 220 1175 220 1175 220 1175 225 1238 225 1238 225 1238 225 1238 225 1238 230 1307 230 1307 230 1307 230 1307 230 1307 235 1384 235 1384 235 1384 235 1384 235 1384 240 1468 240 1468 240 1468 240 1468 240 1468 245 1562 245 1562 245 1562 245 1562 245 1562 250 1665 250 1665 250 1665 250 1665 250 1665 255 1779 255 1779 255 1779 255 1779 255 1779 260 1906 260 1906 260 1906 260 1906 260 1906 265 2045 265 2045 265 2045 265 2045 265 2045

, 270 2199 270 2199 270 2199 270 2199 270 2199 275 2370 275 2370 275 2370 275 2370 275 2370 WCAP-17123-NP July 2011 Revision I

Westinghouse Non-Proprietary Class 3 5-17 Table 5-5 72 EFPY Heatup Curve Data Points Using the 1998 through the 2000 Addenda App. G Methodology (wi K le , wi Flange Notch, wi Pressure Correction and wlo Uncertainties for Instrumentation Errors) 60°F/hr lOO°F/hr lOO°F/hr Leak Test Limit 60°F/hr Heatup Criticality Heatup Criticality T T T P T P T P P (psig) P (psig)

(oF) (oF) (OF) (pslg) (oF) (psig) (OF) (psig) 246 2000 60 0 264 0 60 0 264 0 264 2485 60 594 264 561 60 560 264 527 65 594 264 561 65 560 264 527 70 594 264 561 70 560 264 528 I, 75 594 264 561 75 560 264 529

~ 80 85 594 594 264 264 561 561 80 85 560 560 264 264 530 531 90 594- 264 561 90 560 264 534 95 594 264 561 95 560 264 535 100 594 264 561 100 560 264 539 105 594 264 561 105 560 264 540 110 594 264 561 110 560 264 546 110 561 264 561 110 527 264 546 115 561 264 561 115 527 264 546 120 561 264 561 120 527 264 553 125 561 264 561 125 529 264 554 130 561 264 561 130 531 264 561 135 561 264 561 135 535 264 561 140 561 264 561 140 540 264 561 145 561 264 561 145 546 264 561 150 561 264 561 150 553 264 561 155 561 264 561 155 561 264 561 160 561 264 561 160 561 264 561 165 561 264 561 165 561 264 561 170 561 264 561 170 561 264 561 175 561 264 561 175 561 264 561 180 561 264 762 180 561 264 626 180 561 264 789 180 561 264 645 180 762 264 819 180 626 264 665 185 789 264 852 185 645 264 688 190 819 264 889 190 665 264 713 195 852 264 930 195 688 264 74l

,i 200 889 264 975 200 713 264 772 WCAP-17123-NP July 20ll Revision I

Westinghouse Non-Proprietary Class 3 5-18 60°F/hr lOO°Flhr lOO°F/hr Leak Test Limit 60°F/hr Heatup Criticality Heatup Criticality T T T P T P T P P (psig) P (psig)

(oF) eF) (oF) (psig) eF) (psig) (oF) (psig) 205 930 264 1024 205 741 264 806 210 975 264 1127 210 772 264 878 I

I 215 1024 265 1139 215 806 265 886 220 1079 270 1200 220 844 270 932 i 225 1139 275 1258 225 886 275 984 i 230 1200 280 1314 230 932 280 1040 235 1258 285 1376 235 984 285 1103 240 1314 290 1444 240 1040 290 1172 245 1376 295 1519 245 1103 295 1248 I 250 1444 300 1602 250 1172 300 1332 I

255 1519 305 1694 255 1248 305 1425 260 1602 310 1795 260 1332 310 1528 !

265 1694 315 1906 265 1425 315 1641 270 1795 320 2029 270 1528 320 1765 275 1906 325 2165 275 1641 325 1902 280 2029 330 2315 280 1765 330 2054 285 2165 335 2481 285 1902 335 2221 290 2315 290 2054 340 2405 295 2481 295 2221 300 2405 WCAP-17123-NP July 2011 Revision 1

Westinghouse Non-Proprietary Class 3 5-19 Table 5-6 72 EFPY Cooldown Curve Data Points Using the 1998 through the 2000 Addenda App_ G Methodology (wI KIt> wI Flange Notch, wI Pressure Correction and wlo Uncertainties for InstrumentatIon Errors)

Steady State 20°Flhr. 40°F/hr. 60°F/hr. lOO°Flhr.

T(OF) P (psig) T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) 60 0 60 0 60 0 60 0 60 0 60 594 60 591 60 548 60 505 60 415 65 594 65 593 65 551 65 507 65 417 70 594 70 594 70 553 70 509 70 420 75 594 75 594 75 556 75 512 75 423 80 594 80 594 80 559 80 515 80 426 85 594 85 594 85 562 85 519 85 430 90 594 90 594 90 566 90 523 90 435 95 594 95 594 95 570 95 528 95 440 100 594 100 594 100 575 100 533 100 445 105 594 105 594 105 581 105 538 105 452 110 594 110 594 110 586 110 545 110 459 110 561 110 561 110 553 110 512 110 426 115 561 115 561 115 560 115 519 115 434 120 561 120 561 120 561 120 526 120 443 125 561 [25 561 125 561 125 535 125 453 130 561 130 561 130 561 130 545 130 465 135 561 135 561 135 561 135 556 135 477 I 140 561 140 561 140 561 140 561 140 491 145 561 145 561 145 561 145 561 145 507 150 561 150 561 150 561 150 561 150 525 155 561 155 561 155 561 155 561 155 544 160 561 160 561 160 561 160 561 160 561 165 561 165 561 165 561 165 561 165 561 170 561 170 561 170 561 170 561 170 561 175 561 175 561 175 561 175 561 175 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 808 180 779 180 752 180 726 180 680 185 832 185 805 185 781 185 757 185 717 190 858 190 834 190 812 190 791 190 757 195 887 195 866 195 846 195 829 195 803 WCAP-17123-NP July 2011 Revision 1

Westinghouse Non-Proprietary Class 3 5-20 Steady State 20°F/hr. 40°F/hr. 60°F/hr. lOO°F/hr.

TeF) P (psig) T(OF) P (psi g) T(OF) P (psig) T(OF) P (psig) T(°F) P (psig) 200 920 200 901 200 885 200 871 200 853 205 956 205 940 205 927 205 917 205 908 210 995 210 983 210 974 210 969 210 969 215 1039 215 1031 215 1026 215 1025 215 1025 220 1087 220 1083 220 1083 220 1083 220 1083 225 1141 225 1141 225 1141 225 1141 225 1141 230 1200 230 1200 230 1200 230 1200 230 1200 235 1265 235 1265 235 1265 235 1265 235 1265 240 1337 240 1337 240 1337 240 1337 240 1337 245 1417 245 1417 245 1417 245 1417 245 1417 250 1505 250 1505 250 1505 250 1505 250 1505 255 1602 255 1602 255 1602 255 1602 255 1602 260 1710 260 1710 260 1710 260 1710 260 1710 265 1828 265 1828 265 1828 265 1828 265 1828 270 1960 270 1960 270 1960 270 1960 270 1960 275 2105 275 2105 275 2105 275 2105 275 2105 280 2265 280 2265 280 2265 280 2265 280 2265 285 2443 285 2443 285 2443 285 2443 285 2443 WCAP-17123-NP July 2011 Revision)

Westinghouse Non-Proprietary Class 3 6-1 6 REFERENCES I. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U. S.

Nuclear Regulatory Commission, May 1988.

2. WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Set points and RCS Heatup and Cooldown Limit Curves," J. D. Andrachek, et aI., May 2004.

3. Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division I, "Fracture Toughness Criteria for Protection Against Failure."

4. Westinghouse Letter LTR-REA-09-112, Revision I, "J. M . Farley Units I and 2 Updated Beltline Fluence," B. W. Amiri, dated September 21,2009.

5. Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements,"

U.S. Nuclear Regulatory Commission, Washington, D. C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

6. WCAP-14689, Revision 6, "Farley Units I and 2 Heatup and Cooldown Limit Curves for Nonnal Operation and PTLR Support Documentation," T. 1. Laubham, April 200 I.

7. MHI-SNC-0455F2, "Certified Material Test Report - Reactor Vessel Closure Head for Joseph M.

Farley Nuclear Plant-2," June 2004.

8. 10 CFR Part 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events", Federal Register, Volume 60, No . 243, dated December 19, 1995, effective January 18,1996.

9. ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division I, Subsection NB, Section NB-2300, "Fracture Toughness Requirements for Material."

10. WCAP-16918-NP, Revision I, "Analysis of Capsule V from the Southern Nuclear Operating Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program," N. R. Jurcevich and G. A. Fischer, April 2008 .

WCAP-17123-NP July 2011 Revision 1

I\lnrl.iJrAnrIPI,,'r'\J Class 3 A-I APPENDIX A THERMAL STRESS INTENSITY FACTORS (KIT)

The following pages contain the thermal stress intensity factors (KII ) for the maximum heatup and cooldown rates. The vessel radii to the 1I4T and 3/4T locations are as follows:

1/4T Radius = 80.625"

3/4T Radius 84.562" WCAP-17123-NP

Westinghouse Non-Proprietary Class 3 A-2 Table A-I Kit Values for 36, 54 and 72 EFPY lOO°F/hr "eatup Curves (w/o Margins for Instrument Errors)

Vessel Temperature Vessel Temperature Water 1I4T Thermal Stress 3/4T Thermal Stress

@ 1I4T Location for @ 3/4T Location for Temp. Intensity Factor Intensity Factor lOO°F/hr "eatup lOO°F/hr Heatup (OF) (KSI SQ. RT. IN.) (KSI SQ. RT. IN.)

(OF) (OF) 60 56.130 -0.987 55.065 0.493 65 58.927 -2.377 55.425 1.455 70 62. 129 -3.521 56.315 2.377 75 65.562 -4.586 57.748 3.208 80 69.262 -5.475 59.641 3.929 85 73.079 -6.273 61.944 4.558 90 77.089 -6.948 64.601 5.101 95 8l.l93 -7.553 67.562 5.578 100 85.435 -8.069 70.788 5.991 105 89.755 -8.531 74.238 6.353 110 94.171 -8.928 77.881 6.671 115 98.650 -9.285 81.690 6.951 120 103.196 -9.594 85.642 7.198 125 107.790 -9.875 89.717 7.418 130 112.433 -10.118 93.898 7.612 135 117.114 -10.341 98.171 7.785 140 121.829 -10.535 102.523 7.940 145 126.574 -10.715 106.944 8.080 150 131.343 -10.873 111.424 8.206 155 136. 136 -11.020 115.955 8.320 160 140.945 -1l.l51 120.529 8.423 165 145.773 -11.275 125.142 8.519 170 150.613 -11.385 129.788 8.606 175 155.467 -11.491 134.462 8.687 180 160.330 -11.586 139.161 8.762 185 165.204 -11.678 143.881 8.833 190 170.083 -11.763 148.620 8.899 195 174.972 -11.845 153 .374 8.961 200 179.864 -11.920 158.143 9.020 205 184.764 -11.995 162.923 9.077 210 189.666 -12.064 167.713 9.131 WCAP-I7123-NP July 2011 Revision I

Westinghouse Non-Proprietary Class 3 A-3 Table A-2 KIf Values for 36, 54 and 72 EFPY lOO°F/hr Cooldown Curves (w/o Margins for Instrument Errors)

Vessel Temperature IOO°F/hr Cooldown Water @ 1I4T Location for 1I4T Thermal Stress I Temp. IOO°F/hr Cooldown Intensity Factor eF) eF) (KSI SQ. RT. IN.)

210 232.426 13.510 205 227.352 13.454 200 222.278 13.398 195 217.204 13.342 190 212.131 13 .286 185 207.057 13.230 180 201.983 13.175 175 196.909 13.119 170 191.836 13.063 165 186.762 13.008 160 181.688 12.952 155 176.615 12.897 150 171.541 12.842 145 166.468 12.786 140 161.395 12.731 135 156.322 12.676 130 151.249 12.622 125 146.176 12.567 120 141.103 12.512 115 136.031 12.457 110 130.958 12.403 105 125.886 12.349 100 120.813 12.295 95 115.741 12.240 90 110.669 12.187 85 105.597 12.133 80 100.526 12.079 75 95.454 12.025 70 90.382 11.972 65 85.311 11.919 60 80.241 11.865 WCAP-17123-NP July 2011 Revision 1

NL-12-0868, Request to Revise Technical Specifications Associated with the Low Temperature Overpressure Protection System and the Pressure and Temperature Limits Report

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NL-12-0868, Request to Revise Technical Specifications Associated with the Low Temperature Overpressure Protection System and the Pressure and Temperature Limits Report

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