View source for L-09-072, License Amendment Request to Incorporate the Use of Alternate Methodologies for Development of Reactor Pressure Vessel Pressure-Temperature Limit Curves and Request for Exemption from Certain Requirements Contained in 10 Cfr.

{{Adams
| number = ML091130228
| issue date = 04/15/2009
| title = License Amendment Request to Incorporate the Use of Alternate Methodologies for Development of Reactor Pressure Vessel Pressure-Temperature Limit Curves and Request for Exemption from Certain Requirements Contained in 10 Cfr.
| author name = Allen B
| author affiliation = FirstEnergy Nuclear Generation Corp
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000346
| license number = NPF-003
| contact person = 
| case reference number = L-09-072
| document type = Letter type:L, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 48
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:FENOC                  %                                                                      5501 North State Route 2 FirstEnergyNuclear Operating Company                                                           Oak Harbor,Ohio 43449 Barry S. Allen                                                                                          419-321-7676 Vice President - Nuclear                                                                           Fax. 419-321-7582 April 15, 2009 L-09-072                                                               10 CFR 50.12 10 CFR 50.60(b) 10 CFR 50.61 10 CFR 50.90 10 CFR 50, Appendix G ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

==SUBJECT:==

Davis-Besse Nuclear Power Station, Unit No. 1 Docket No. 50-346, License No. NPF-3 License Amendment Request to Incorporate the Use of Alternate Methodologies for the Development of Reactor Pressure Vessel Pressure-Temperature Limit Curves, and Request for Exemption From Certain Requirements Contained in 10 CFR 50.61 and 10 CFR 50, Appendix G FirstEnergy Nuclear Operating Company hereby requests amendment of the operating license for the Davis-Besse Nuclear Power Station, Unit No. 1. The proposed amendment would incorporate the use of alternate methodologies for the calculation of reactor pressure vessel beltline weld initial reference temperatures (RTNDT), the calculation of the adjusted reference temperatures (ART), and the development of the reactor pressure vessel (RPV) pressure-temperature (P-T) limit curves and the low temperature reactor coolant system (RCS) overpressure analysis into Technical Specification (TS) 5.6.4. The license amendment request also revises the reanalysis requirement for the low temperature RCS overpressure events from 21 to 32 Effective Full Power Years (EFPY) contained in Operating License (OL)
Condition 2.C(3)(d).
Enclosure A provides a description and evaluation of the proposed amendment.
Attachments to Enclosure A provide a copy of the existing OL/TS pages marked up to reflect the proposed amendment, re-typed OL/TS pages with the proposed amendment incorporated, and the TS Bases pages marked up to reflect the proposed amendment (for information only). Attachment 1 contains a listing of regulatory commitments associated with the implementation of the proposed amendment.
                                                                                                    .,Aooi'

Davis-Besse Nuclear Power Station, Unit No. 1 L-09-072 Page 2 of 2 Enclosure B contains a request for exemption from certain requirements of 10 CFR 50, Appendix G, "Fracture Toughness Requirements," and 10 CFR 50.61, "Fracture toughness requirements for protection against pressurized thermal shock events." The requested exemption would allow the use of an alternate method, described in Topical Report BAW-2308, Revisions 1-A and 2-A, "Initial RTNDT of Linde 80 Weld Materials," for determining the initial, unirradiated material reference temperatures of Linde 80 weld materials present in the beltline region of the Davis-Besse Nuclear Power Station, Unit No. 1 reactor pressure vessel.
Approval of the proposed license amendment and request for exemption are requested by April 30, 2010, to be implemented within 90 days of approval.
If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager - Fleet Licensing, at (330) 761-6071.
I declare under penalty of perjury that the foregoing is true and correct. Executed on April 5'-, 2009.
Sincerely, Barry S. Allen

==Attachment:==
: 1. Regulatory Commitment List

==Enclosures:==

A. Evaluation of Proposed License Amendment B. Request for Exemption cc:   NRC Region III Administrator NRC Project Manager NRC Resident Inspector Executive Director, Ohio Emergency Management Agency, State of Ohio (NRC Liaison)
Utility Radiological Safety Board

Attachment 1 L-09-072 Regulatory Commitment List Page 1 of 2 The following list identifies those actions committed to by FirstEnergy Nuclear Operating Company (FENOC) for the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS) in this document. Any other actions discussed in the submittal represent intended or planned actions by FENOC, are described only for information, and are not Regulatory Commitments. Please notify Mr. Thomas A. Lentz, Manager - Fleet Licensing, at (330) 761-6071 of any questions regarding this document or associated Regulatory Commitments.
Re-qulatory Commitment                                Due Date
: 1. Since the DBNPS is approaching 21                  1. 90 days after NRC approval of EFPY, revised P-T limit curves and LTOP              this License Amendment
[low temperature overpressure protection]            Request and Request for limits permitting operation beyond 21                 Exemption, but before EFPY have been developed. Since the                  exceeding 21 Effective Full revised P-T limit curves and LTOP limits             Power Years.
are based, in part, upon alternative methodologies which are not described in BAW-1 0046A, Revision 2, the revised curves and limits will not be implemented nor will operation beyond 21 EFPY be permitted at DBNPS until after this amendment request and the enclosed exemption request have been approved by the NRC and implemented at the site.
: 2. The LTOP reanalysis has been performed             2. 90 days after NRC approval of with a description in Section 3.2 of this             this License Amendment submittal. Since the reanalysis was                   Request and Request for performed using alternative methodologies             Exemption but before not yet approved by the NRC for use at                exceeding 21 Effective Full the DBNPS, the reanalysis is not effective            Power Years.
until after the proposed amendment request and the enclosed exemption request are NRC approved and implemented at the DBNPS.

Attachment 1 L-09-072 Regulatory Commitment List Page 2 of 2
. Reiqulatory Commitment                           Due Date
: 3. A Pressure-Temperature Limits Report will     3. Before exceeding 21 Effective be submitted to the NRC in accordance            Full Power Years.
with TS 5.6.4 after this proposed amendment request and the enclosed exemption request are approved by the NRC, and implemented by the DBNPS staff, but before the DBNPS operates beyond 21 EFPY.

Enclosure A L-09-072 EVALUATION OF PROPOSED LICENSE AMENDMENT Page 1 of 18

==Subject:==
License Amendment Request to incorporate the use of alternate methodologies for the calculation of the initial reference temperature (RTNDT), the calculation of the adjusted reference temperature (ART), and the development of the reactor pressure vessel (RPV) pressure-temperature (P-T) limit curves and the low temperature reactor coolant system (RCS) overpressure analysis into Technical Specification (TS) 5.6.4. The license amendment request also revises the reanalysis requirement for the low temperature RCS overpressure events from 21 to 32 Effective Full Power Years (EFPY) contained in Operating License (OL) Condition 2.C(3)(d).
1.0 

==SUMMARY==
DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Proposed Changes

===2.2 Background===

==3.0 TECHNICAL EVALUATION==

3.1   Evaluation of the Analytical Methods Used to Develop P-T Limit Curves 3.2   Evaluation of Low Temperature Overpressure Protection (LTOP) 3.3 Conclusion

==4.0 REGULATORY EVALUATION==

4.1  Significant Hazards Consideration 4.2 Applicable Regulatory Requirements/Criteria 4.3 Precedent 4.4 Conclusions

==5.0 ENVIRONMENTAL CONSIDERATION==

N N

Enclosure A L-09-072 EVALUATION OF PROPOSED LICENSE AMENDMENT Page 2 of 18

==6.0 REFERENCES==

Attachments:
: 1. Proposed Operating License and Technical Specification Changes (Mark Up)
: 2. Proposed Operating License and Technical Specification Changes (Re-typed - For Information Only)
: 3. Proposed Technical Specification Bases Changes (For Information Only)

Enclosure A L-09-072 Page,3 of 18 1.0 

==SUMMARY==
DESCRIPTION This evaluation supports a request to amend Operating License NPF-3 for the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS). The proposed amendment will incorporate the use of alternate methodologies for the calculation
--*of the initial reference temperatures (RTNDT) for welds Iih'1e:retctor pressure vessel (RPV) beltline region, the calculation of the adjusted reference temperatres (ART) for the RPV beltline region, and the development of the RPV pressure-temperature (P-T) limit curves and the low temperature reactor coolant system (RCS) overpressure analysis into Technical Specification (TS) 5.6.4, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)." The proposed amendment also revises the requirement to provide a reanalysis of the low temperature RCS overpressure events from 21 to 32 Effective Full Power Years (EFPY) contained in Operating License (OL)
Condition 2.C(3)(d).
The DBNPS is submitting this amendment request in order to use more current methods to develop the DBNPS P-T limit curves than is presently described within the DBNPSTS.
2.0 DETAILED DESCRIPTION 2.1    Proposed Changes TS 5.6.4 identifies the approved methodology for the analysis of RCS P-T limits.
The current methodology in use at the DBNPS is described in the NRC-approved Babcock and Wilcox Topical Report BAW-10046A, Revision 2, "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50 Appendix G." The proposed amendment will incorporate the use of several alternate methodologies for use in the development of the P-T limit curves and the low temperature RCS overpressure analysis, in addition to the use of BAW-1 0046A, Revision 2 into TS 5.6.4. One of the alternate methodologies is used for the calculation of the initial RTNDT values for welds in.
the RPV beltline region. An alternate methodology will be used in the calculation of the ART values for the RPV beltline region. An alternate methodology will be the use of axially oriented reference defects in axial welds and base metal and circumferentially oriented reference defects in circumferential welds. Another alternate methodology will be the use of K1c (material toughness property measured in terms of stress intensity factor, K1, which will lead to nonductile crack propagation). The alternate methodologies are described in the Babcock and Wilcox Owner's Group Topical Report BAW-2308, Revisions 1-A and 2-A, "Initial RTNDT of Linde 80 Weld Materials," March 2008; Oak Ridge National Laboratory ORNLITM 2006/530, "A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels;" and the American Society of Mechanical Engineers (ASME) Code Cases N-588 and N-640.

Enclosure A L-09-072 Page 4 of 18 OL Condition 2.C(3)(d) states that prior to operation beyond 21 EFPY the FirstEnergy Nuclear Operating Company (FENOC) shall provide to the NRC a re-analysis and proposed modifications, as necessary, to ensure continued means of protection against low temperature RCS overpressure events. The proposed amendment will revise the submittal requirement from 21 to 32 EFPY. contains a copy of TS 5.6.4 and OL Condition 2.C(3)(d) marked up with the: aforementioned changes. Attachment 2 contains a copy of TS 5.6.4 and OL Condition 2.C(3)(d) re-typed with the proposed changes incorporated.
Changes to the Bases of TS 3.4.12, "Low Temperature Overpressure Protection (LTOP)," are required to reflect the change in OL Condition 2.C(3)(d)-from 21 to 32 EFPY. Attachment 3 contains a copy of the TS Bases pages marked up with changes that support the proposed TS/OL changes. The Bases pages are provided for information only. These changes will be implemented using the DBNPS Technical Specifications Bases Control Program (TS 5.5.13) and evaluated under the DBNPS 10 CFR 50.59 program.
The alternate methodology used to calculate the initial RTNDT values for the welds in the RPV beltline region are described in the NRC-approved Topical. Report BAW-2308, Revision 1-A and Revision 2-A, "Initial RTNDT of Linde 80 Weld Materials" topical report. In accordance with the NRC Safety Evaluation for both revisions, an exemption request is required to be submitted to the NRC to use the alternate methodology. Enclosure B of this submittal contains the required exemption request for use of this methodology.

===2.2 Background===
10 CFR 50, Appendix G, "Fracture Toughness Requirements," specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, which the pressure boundary may be subjected to*.over its service lifetime. Appendix G also states that the ASME Boiler and Pressure:Vessel Code (Code) forms the basis for the requirements of this regulation.
10 CFR 50, Appendix G requires P-T limits and minimum temperature-requirements for the RPV. These limits are defined by the various operating conditions that the plant can be in and the reactor vessel pressure. The operating conditions include, but are not limited to, hydrostatic pressure and leak testing and normal operation (including heatup and cooldown). The P-T limits are developed in accordance with the methods of analysis and margins contained in the ASME Code, Section Xl, Appendix G.

Enclosure A L-09-072 Page 5 of 18 The P-T limits are used to prevent non-ductile failure of the reactor coolant pressure boundary during the aforementioned plant conditions. The P-T limits also prevent exceeding stress limits during cyclic operations. The RPV regions that have been analyzed tre the RPV head, the RPV nozzles, and the RPV beltline. The beltline region includes nozzle forgings and welds. The maximum allowable reactor vessel pressure is defined by the lowest of the three regions.
The P-T limit curves are a composite of the three regions, utilizing the lowest values such that at all temperatures, the maximum allowable preswure'is kept' below the controlling region's limit.
The low temperature overpressure protection (LTOP) system controls RCS pressure at low operating temperatures so that the integrity of the reactor coolant pressure boundary is not compromised by violating the P-T limits of 10 CFR 50, Appendix G. Each time the P-T limit curves are revised, the LTOP system must be reevaluated to ensure that its functional requirements can still be met.
The current analytical methods used to determine the DBNPS 21 EFPY P-T limit curves and the LTOP limits for the DBNPS are described in the NRC-approved BAW-10046A, Revision 2 topical report.
Since the DBNPS is approaching 21 EFPY, revised P-T limit curves and LTOP limits permitting operation beyond 21 EFPY have been developed. Since the revised P-T limit curves and LTOP limits are based, in part, upon alternative methodologies that are not described in BAW-i 0046A, Revision 2, the revised curves and limits will not be implemented nor will operation beyond 21 EFPY be permitted, at DBNPS until after this amendment request and the enclosed exemption request have been approved by the NRC and implemented at the site.
The revised P-T limit curves and LTOP limits were developed using alternative methodologies. An alternate methodology was used for the calculation of the initial RTNDT values for welds in the RPV beltline region. An alternate methodology was used for the calculation of the ART values for the RPV beltline region. Additionally, alternate methodologies incorporated the use of axially oriented reference defects in axial welds and base metal and circumferentially oriented reference defects in circumferential welds, and the use of Kic. These methodologies were used in addition to the methodologies described in BAW-10046A, Revision 2. The alternative methodologies are described in the-NRC-approved BAW-2308, Revisions 1-A and 2-A; Oak Ridge National Laboratory ORNL/TM 2006/530, "A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels;" and-the NRC-accepted ASME Code Cases N-588 and N-640.

Enclosure A L-09-072 Page 6 of 18 The alternative methodology for the calculation of the initial RTNDT values for the welds in the reactor vessel beltline region are described in the NRC-approved BAW-2308, Revisions 1-A and 2-A. As stated in the NRC Safety Evaluations for the topical report revisions, one of the conditions for licensees to reference-the topical report in licensing activities was to submit an exemptiori request to the NRC for review. This submittal contains the required exemption request as Enclosure B.
The alternative method to account for the adjustment in transition temperature caused by irradiation is based on equations documented in Oak Ridge National Laboratory ORNL/TM 2006/530, "A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels."
Since ASME Section Xl, Appendix G, was first developed, significant advances in fracture mechanics and the analysis of RPV integrity have been achieved. In general, advancements in knowledge are promulgated as ASME Code Cases.
ASME Code Cases N-588 and N-640, provide alternative methodologies to existing Code requirements for developing P-T operating limits and LTOP limits.
These alternatives to the Appendix G methodology have been accepted by the NRC as appropriate alternatives and are documented in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section Xl, Division 1."

==3.0 TECHNICAL EVALUATION==

3.1    Evaluation of the Analytical Methods Used to Develop P-T Limit Curves The analytical methods used to determine the revised P-T limit curves for the DBNPS are contained in the NRC-approved BAW-10046A, Revision 2, and in the NRC-approved BAW-2308, Revisions 1-A and 2-A; ORNL/TM 2006/530 equations; and the NRC-accepted ASME Code Cases N-588 and N-640.
BAW-10046A This topical report describes the Babcock and Wilcox practices, methods, and criteria for compliance with the requirements of 10 CFR 50, Appendix G. The topical report includes, but is not limited to, the description of the methodology for the development of the initial RTNDT values for ferritic materials of the beltline region and the calculational procedure used to determine the P-T limits of the RCS. The topical report was approved by the NRC as documented in a Safety Evaluation dated April 30,"1986.
Since the closure head and outlet nozzle regions of the reactor vessel do not receive significant irradiation during reactor operation, the material properties do not experience a significant change over plant life. Therefore, there are no

Enclosure A L-09-072 Page 7 of 18 requirements to monitor their reference transition temperatures. The methods which determine the maximum acceptable pressure for these regions are described within the topical report.
Since the reactor vessel beltline region does receive significant irradiation during reactor operation, the region experiences a significant change in RTNDT over the life of the plant. The methodologies described in Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," are used in the topical report to account for the effects of neutron embrittlement on the materials used in the reactor vessel beltline region. The RG includes, but is not limited to, the calculation of a material's adjusted reference temperature (ART).
The RG methodology for calculating the ART for the reactor vessel beltline materials is given in the following equation:
ART  = Initial RTNDT + ARTNDT  + Margin The ARTNDT term is an adjustment in the reference temperature caused by irradiation, and the Margin term is added to provide conservatism in the values of the ART. RG 1.99, Revision 2 contains the methodologies to calculate these two terms.
The proposed amendment requests to use the BAW-2308, Revisions 1-A and 2-A method for the determination of the initial RTNDT values for the RPV beltline welds instead of the methodology described within Topical Report BAW-10046A, Revision 2. However, the methodology in RG 1.99, Revision 2 was used to evaluate the RPV beltline forgings.
The P-T limit curves-were calculated using the procedure described within Topical Report BAW-10046A, Revision 2. Since, TS 5.6.4 currently lists Topical Report BAW-1 0046A, Revision 2 as describing the methods used for determining the DBNPS P-T limits curves, no further evaluation of the topical report is required.
BAW-2308 This topical report was developed to provide an alternate method for determining the initial RTNDT for Linde 80 beltline welds in Babcock and Wilcox fabricated reactor vessels. The alternative methodology was based on brittle-to-ductile transition range fracture toughness test data of the weld materials in accordance with ASTM Standard E1921 and using ASME Code Case N-629. This methodology would be used in lieu of the nil-ductility reference temperature parameter specified in ASME Section III Paragraph NB-2331. The use of ASME Section III Paragraph NB-2331 is specified in 10 CFR 50, Appendix G(II)(D)(i) and its use is also described in RG 1.99. Revision 1-A of the topical report was

Enclosure A L-09-072 Page 8 of 18 approved by the NRC as documented in a Safety Evaluation dated August 4, 2005. Revision 2-A, which supplements Revision 1-A of the topical report, was approved by the NRC as documented in a Safety Evaluation dated March 24, 2008.
The Safety Evaluation for Revision 1-A of the topical report contained six conditions. Licensees had to satisfy four of these conditions in order to use the alternate methodology. The first three conditions describe performance requirements for use of the methodology. These conditions have been satisfied.
The fourth condition requires the submittal of an exemption request from the requirements of 10 CFR 50, Appendix G and 10 CFR 50.61 in order to use the methodology. Enclosure B contains the required exemption request to use the BAW-2308 topical report.
The topical report was revised to incorporate additional information from the pressurized water reactor owner's group to satisfy the remaining two conditions contained in the NRC Safety Evaluation for Revision 1-A. Revision 2-A of the topical report was approved by the NRC as documented in a Safety Evaluation dated March 24, 2008. This Safety Evaluation required licensees to follow the four conditions contained in the Safety Evaluation for Revision 1-A. The Safety Evaluation for BAW-2308, Revision 2-A also stated that the initial RTNDT values for certain materials contained in the Revision 1-A version are more conservative. These values apply to the DBNPS materials, hence, the revised DBNPS P-T limit curves are based upon the values contained in Revision 1-A.
ORNLITM 2006/530 Equations The revised DBNPS P-T limit curves were prepared utilizing, in part, the RG 1.99, Revision 2 ART methodology with two modifications. The first modification was the use of the initial RTNDT values documented in Topical Report BAW-2308 for the beltline weld materials. The second was the use of an alternative. method to account for the adjustment in transition temperature caused by irradiation in lieu of using the ARTNDT term. Equations, modified from those documented in Oak Ridge National Laboratory ORNLITM 2006/530, "A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels," were used to develop the alternative methodology for calculating the irradiation adjustment term. The equations provided in Section 7- of the ORNL/TM were included as part of a proposed rule, 10 CFR Part 50*61A, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," dated October 3, 2007 (72 FR 56275).
The alternative method was selected since it is an improvement in the RG 1.99, Revision 2 model due to the larger, better-balanced materials database, and the current understanding of the embrittlement mechanisms. The alternative method includes variables for copper, nickel, and fluence that are in RG 1.99, Revision 2,

Enclosure A L-09-072 Page 9 of 18 but also includes effects of irradiation temperature, neutron flux, phosphorous and manganese. As stated in the October 3, 2007 Federal Register Notice, the "existing requirements are based on unnecessarily conservative probabilistic fracture mechanic analyses."
These equations were adapted for use in the development of a temperature transition shift (TTS) term in lieu of the ARTNDT term of the RG 1.99, Revision -2 equation. The temperature transition shift (TTS) equation is:
TTS = MD + CRP MD is the Matrix Damage term, and is determined from the phosphorous and manganese content, fluence for the beltline material, and the irradiation temperature,. CRP is the Copper Rich Precipitate term, and is determined from the nickel, copper, and phosphorous content, fluence, and irradiation temperature of the materials.
When using the TTS term in lieu of the ARTNDT term in the RG 1.99, Revision 2 ART equation, the Margin term of the equation would also be affected. The RG guidance includes a condition that the oa term need not exceed 0.50 times the mean value of ARTNDT. When the TTS term is used, a similar YA,condition is used, that is, GA need not exceed 0.50 times the value of TTS.
The equations used to develop the DBNPS P-T limit curves are similar to those provided in the ORNL/TM with the following modifications.
The significant digits used in the DBNPS MD and CRP equations are greater than those contained in the ORNL/TM equations. However, this had no impact upon the results of the equations.
The DBNPS effective fluence term has a limit while the ORNLITM term does not.
Since the limit was not exceeded, there was no effect.
Since the CRP term for the reactor vessel beltline forgings is zero, the following CRP modifications would only impact the reactor vessel beltline weld values.
The ORNL/TM describes a temperature correction factor for the CRP term and why the inclusion of the temperature correction factor is not essential. However, the ORNL/TM indicates that for temperatures greater than 15 IF different than the base temperature of 543.1 OF the term starts to become significant. Since the DBNPS temperature is 561 IF, which is 17.9 IF above 543.1 0 F, the temperature factor was retained. The DBNPS CRP term is:
CRPDBNPS =     CRPX (Tc/543.1) 1 100

Enclosure A L-09-072 Page 10 of 18 Since Tc for the DBNPS is greater than 543.1 and the temperature correction term is being raised to the 1.100 power, it can be seen that this adjustment is greater than 1.0. Since the method multiplies the CRP by a value greater than 1, the TTSDBNPS will always be conservative (larger) when compared to the value that the ORNL/TM equations yields.
There also were differences in the f(Cue, P) term which is a component of the overall .CRP term. However, these differences had no effect on the results for the Linde 80 welds.
In the development of the DBNPS P-T limit curves, the ART for the limiting DBNPS reactor vessel beltline material (WF-182-1) was calculated using both the ARTNDT and the temperature modified TTS methodolgies. The result was that using TTS was conservative in both the 1/4t and the 3/4t locations. The beltline portion of the revised DBNPS composite P-T limit curve is based on this more conservative approach. Since this curve is the limiting component at temperatures above 200 OF and below 250 &deg;F, it is assured that composite P-T curve has been conservatively determined when compared to what would be determined using the RG 1.99, Revision 2 approach to calculating ARTNDT.
ASME Code Cases ASME Code Case N-588 provides an alternative procedure for assuming axially oriented reference defects in all axial welds and base metal and circumferentially Code Case N-640.
oriented reference defects in all circumferential welds. ASME allows the use of Kic (the material toughness property measured in terms of stress intensity factor, K1, which will lead to nonductile crack propagation) instead of KIA (the critical value of the stress intensity factor, K1, for crack arrest as a function of temperature) in the development of P-T limit curves.
NRC Regulatory Issue Summary 2004-04, "Use of Code Cases N-588, N-640, and N-641 in Developing Pressure-Temperature Operating Limits," states that these three ASME Code Cases have been accepted as appropriate alternatives to the ASME Code. This-acceptance has been documented in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section Xl,
-Division 1.."
Furthermore, the provisions of ASME Code Cases N-588, N-640, and. N-641 that are applicable to P-T limit curves and LTOP limits development have been incorporated into ASME Code Section Xl, Appendix G in the 1998 edition through 2000 addenda which is the edition and addenda codified in 10 CFR 50.55a, effective October 28, 2002.

Enclosure A L-09-072 Page 11 of 18 Though Regulatory Issue Summary 2004-04 accepts the use of ASME Code Cases N-588, N-640, and N-641, and they have been incorporated into the ASME Code; the use of Code Case N-641 is not part of this proposed amendment.
3.2 Evaluation of Low Temperature Overpressure Protection (LTOP)
As part of the revision to the DBNPS P-T limit curves, the LTOP limits were reanalyzed. The reanalysis was performed using BAW-10046A and ASME Code Case N-640. Since the reanalysis was performed using alternative methodologies not yet approved by the NRC for use at the DBNPS, the reanalysis will not be made effective and operation beyond 21 EFPY will not be
-permitted until the following occurs. First, the proposed amendment request and the enclosed exemption request are approved by the NRC. Second, the approved amendment request and exemption request are implemented at the DBNPS.
OL Condition 2.C(3)(d) requires FENOC to provide the NRC a reanalysis and proposed modifications, as necessary, to ensure continued means of protection against low temperature RCS overpressure events. The following information satisfies this requirement. No plant modifications are necessary to extend the protection against LTOP events to 32 EFPY.
The P-T limits for LTOP are obtained by taking 100% of the controlling normal operation heatup/cooldown limit. To support the development of an LTOP system effective temperature, the temperature difference is calculated between the 1/4t location of the reactor vessel wall and the RCS fluid for the condition when the metal temperature is at 160.8 IF + 50 IF, or 210.8 IF, as required by Article G-2215 of Section XI of the ASME Code, where 160.8 OF is the highest adjusted reference temperature and the 50 OF is a margin term. Temperature differences between the reactor coolant and the reactor vessel wall are evaluated for both the ramp and step heatup conditions when the metal temperature is lower and significantly lags the coolant temperature. During cooldown, the coolant temperature is always lower than the metal temperature, and therefore, it is not limiting in the development of the LTOP system effective temperature. The corresponding reactor coolant temperature during the step heatup transient is 235.0 OF, resulting in a temperature difference of 24.2 IF. The minimum LTOP enable temperature is therefore 235.0 OF. ASME Code Case N-640 allows the use of the Kic fracture toughness curve rather than the Kia curve. However, if this alternative is used, Case N-640 requires that LTOP systems limit the maximum pressure to 100% of the pressure allowed by the P-T limit curves. The minimum allowable pressure below 235 OF is controlled by the reactor pressure vessel head nil-ductility temperature (NDT) limit. Therefore, 625 psig is selected to be the uncorrected LTOP limit. This pressure value is corrected for sensor location for both the A Hot Leg and B Hot Leg pressure taps.

Enclosure A L-09-072 Page 12 of 18 Therefore, the LTOP pressure limit as measured from the A Hot Leg pressure tap is 625 - 85 = 540 psig. The LTOP pressure limit as measured from the B Hot Leg pressure tap is 625 - 60 = 565 psig. These LTOP limiting pressures are applicable from 70 OF up to 235 OF. Above 235 OF, the standard NDT limits for the limiting component (for example, the reactor pressure vessel head, beltline region, and outlet nozzles, for a given temperature) are applicable. The P-T values used to develop the curves in TS 3.4.12, "Low Temperature Overpressure Protection (LTOP)," are more conservative than the values calculated using Code Cases N-588 and N-640; therefore, no changes are proposed to TS 3.4.12.
OL Condition 2.C(3)(d) requires the submittal of a LTOP reanalysis prior to exceeding 21 EFPY. Since the LTOP reanalysis was performed using alternative methodologies not yet approved by the NRC for use at the DBNPS, the reanalysis will not be made effective until the following occurs. First, the proposed amendment request and the enclosed exemption request are approved by the NRC. Second, the approved amendment request and exemption request are implemented at the DBNPS. Since the proposed amendment request is expected to be approved and implemented prior to the DBNPS achieving 21 EFPY, compliance with the OL Condition would be satisfied.
The LTOP reanalysis is valid through 52 EFPY. Since the end of service for the DBNPS is 32 EFPY, the OL Condition will be revised to that point.
3.3 Conclusion The revised DBNPS P-T limit curves and LTOP reanalysis were developed using the NRC-approved Topical Report, BAW-10046A, Revision 2; the NRC-approved Topical Report BAW-2308, Revisions 1-A and 2-A; the modified ORNL/TM 2006/530 equations; and the NRC-accepted ASME Code Cases N-588 and N-640. Since the alternative methodologies are either approved or accepted for use by the NRC, or are conservative with respect to the existing approved methodologies, the alternative methodologies provide an acceptable means of satisfying 10 CFR 50, Appendix G, which governs the development of P-T limit curves and LTOP limits.
As a result, TS 5.6.4 needs to be revised to reflect the use of the aforementioned alternative methodologies. A Pressure-Temperature Limits Report will be submitted to the NRC in accordance with TS 5.6.4 after this proposed amendment request and the enclosed exemption request are approved by the NRC, and implemented by the DBNPS staff, but before the DBNPS operates beyond 21 EFPY.
The LTOP reanalysis has been performed with a description provided in Section-3.2'of this submittal. Since the reanalysis was performed using alternative methodologies not yet approved by the NRC for use at the DBNPS, the

Enclosure A L-09-072 Page 13 of 18 reanalysis is not effective until after the proposed amendment request and the enclosed exemption request are NRC approved and implemented at the DBNPS.
Since the reanalysis is required to be submitted to the NRC prior to 21 EFPY and given that the proposed amendment would be approved and implemented prior to the DBNPS achieving 21 EFPY, compliance with the OL Condition would be satisfied. The effective period for License Condition 2.C(3)(d) is being extended:
to 32 EFPY to ensure that the DBNPS re-examines the efficacy of its LTOP system prior to operating beyond that point in time.

==4.0 REGULATORY EVALUATION==

The FirstEnergy Nuclear Operating Company (FENOC) requests Nuclear Regulatory Commission (NRC) review and approval of an amendment for the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS) Technical Specifications (TS). The proposed amendment would incorporate the use of alternate methodologies for the calculation of the reactor vessel beltline initial reference transition temperatures (RTNDT), the calculation of the reactor vessel beltline adjusted reference temperatures (ART), the use of axially oriented reference defects in axial welds and base metal, the use of circumferentially oriented reference defects in circumferential welds, and the use of Kic in the development of the P-T limit curves and the low temperature RCS overpressure analysis, in addition to the use of the NRC-approved BAW-.10046A, Revision 2 topical report, into TS 5.6.4, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)." Additionally, the proposed amendment would revise the 21 Effective Full Power Years (EFPY) requirement in Operating License (OL) Condition 2.C(3)(d) to provide a reanalysis of the low temperature RCS overpressure events to 32 EFPY.
4.1    Siqnificant Hazards Consideration FENOC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below.
: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The amendment request proposes two changes to the TS/OL. The first change incorporates the use of alternative methodologies to develop the DBNPS P-T limit curves and LTOP limits into TS 5.6.4 to augment the existing listed methodology of BAW-10046A, Revision 2. The second change revises OL Condition 2.C(3)(d) to reflect the revised LTOP analysis is valid to 32 EFPY.

Enclosure A L-09-072 Page 14 of 18 The first change incorporates the use of Topical Report BAW-2308, Revisions 1-A and 2-A; modified ORNL/TM 2006/530 equations, and ASME Code Cases N-588 and N-640. The topical report and ASME code cases have been approved or accepted for use by the NRC (provided that any conditions/limitations are satisfied). The modified ORNLITM 2006/530 equations result in a more conservative ART value for the limiting reactor vessel component. The proposed additions to the methodologies for the reactor vessel P-T curve development provide an acceptable means of satisfying the requirements of 10 CFR 50, Appendix G. The proposed additions do not alter the design or function of any plant equipment. Therefore, the proposed additions do not affect the probability or consequences of any previously evaluated accidents, including reactor coolant pressure boundary failures.
The second change is considered administrative in nature and reflects the revised methodologies. It will not alter the design or operation of any plant equipment. Therefore, the proposed change does not affect the probability or consequences of any previously evaluated accidents.
: 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The amendment request proposes two changes to the TS/OL. The first change incorporates the use of alternative methodologies to develop the DBNPS P-T limit curves and LTOP limits into TS 5.6.4 to augment the existing listed methodology of BAW-1 0046A, Revision 2. The second change revises OL Condition 2.C(3)(d) to reflect that the revised analysis is valid to 32 EFPY.
The first change incorporates methodologies that either have been approved or accepted for use by the NRC (provided that any conditions/limitations are satisfied), or are conservative to current methodologies. The changes do not alter the design or function of any plant equipment. The P-T limit curves and LTOP limits will provide the same level of protection to the reactor coolant boundary as was previously evaluated. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
The second change is considered administrative in nature and reflects the revised methodologies. It will not alter the design or operation of any plant equipment. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Enclosure A L-09-072 Page 15 of 18
: 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The amendment request proposes two changes to the TS/OL. The first change incorporates the use of alternative methodologies to develop the DBNPS P-T limit curves and LTOP limits into TS 5.6.4 to augment the existing listed methodology of BAW-1 0046A, Revision 2. The second change revises OL Condition 2.C(3)(d) to reflect that the revised analysis, is valid to 32 EFPY.
The first change incorporates methodologies that either have been approved or accepted for use by the NRC (provided that any conditions/limitations are satisfied), or are conservative to current methodologies. The second change is considered administrative in nature and reflects the revised methodologies. The
.changes do not alter the design or function of any plant equipment. The P-T limit curves and LTOP limits will provide the same level of protection to the reactor coolant boundary as was previously evaluated. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
Based on the above, FENOC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
4.2 Applicable Regulatory Requirements/Criteria The following paragraphs describe relevant regulatory requirements associated with the proposed TS/OL change.
10 CFR 50, Appendix G, "FractureToughness Requirements" 10 CFR 50, Appendix G, requires that P-T limits be established for reactor pressure vessels during normal operating and hydrostatic or leak rate testing conditions. Further, Appendix G specifies that the requirements for these limits are based on the application of evaluation procedures provided in the ASME Boiler and Pressure Vessel Code, Section Xl, Appendix G.
The revised DBNPS P-T limit curves and LTOP reanalysis were developed using the NRC-approved Topical Report, BAW-10046A, Revision 2; the NRC-approved Topical Report BAW-2308, Revisions 1-A and 2-A; the modified ORNL/TM 2006/530 equations; and the NRC-accepted ASME Code Cases N-588 and N-640. Since the alternative methodologies are either approved or accepted for

Enclosure A L-09-072 Page 16 of 18 use by the NRC, or are conservative with respect to existing methodologies, the alternative methodologies provide an acceptable means of satisfying 10 CFR 50, Appendix G which governsthe development of P-T limit curves and LTOP limits.
Regulatory Guide 1.99, Revision 2, "RadiationEmbrittlement of Reactor Vessel Materials" Regulatory Guide (RG) 1.99, Revision 2 contains methodologies for determining the change in RTNDT and the.change in upper-shelf energy (USE) resulting from neutron radiation. The use of the RG provides an acceptable means of ensuring the requirements of 10 CFR 50, Appendix G are satisfied. The RG also indicates that there are other methodologies that could be used for complying with the specified portions of the regulation.
The RG states that the ART is calculated using the following equation:
ART = Initial RTNDT + ARTNDT +    Margin The factors are defined within the RG. This equation will be used to develop the ART for various materials. However, the proposed alternative methodologies provide different values for two of the factors.
One of the proposed methodology changes is the use of Topical Report BAW-2308, Revisions 1-A and 2-A to provide the Initial RTNDT values for the reactor vessel beltline welds. The topical report has been approved by the NRC with a condition that an exemption request needs to be submitted and approved by the NRC in order to use the topical report. The exemption request is contained in Enclosure B.
The other proposed methodology change uses a factor identified as TTS in place of the ARTNDT factor. This factor is also an adjustment in the reference temperature caused by irradiation. This factor is determined by the content of phosphorus, manganese, nickel, and copper in the material. as well as by the fluence and irradiation temperature. The use of TTS in place of ARTNDT in the determination of the ART results in values that are more conservative than would otherwise be generated using the ARTNDT factor for the limiting material. Since the RG indicates that other methodologies can be used to develop the ART, and the alternative methodologies are either approved by the NRC (provided that any conditions/limitations are satisfied) or are conservative to the RG, then the alternative methodologies would be considered an acceptable means for complying with the requirements of 10 CFR 50, Appendix G.

Enclosure A L-09-072 Page 17 of 18 American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section X1, Appendix G, "FractureToughness Criteriafor ProtectionAgainst Failure" As indicated in the description for 10 CFR 50, Appendix G, above, the proposed changes to the TS a'nTd-OL meet the requirements of the ASME Code,. Section XI, Appendix G as modified by the NRC-accepted ASME Code Cases N-588 and N-640.
4.3 Precedent An exemption request for use of BAW-2308, Revision 1-A at the Surry Power Station Unit Nos 1 and 2 was approved by NRC Safety Evaluation dated June 27, 2007 (ADAMS Accession No. ML071160287).
The use of ASME Code Cases N-588, N-640, and N-641 were incorporated into WCAP-14040, Revision 3, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," and approved by the NRC in a Safety Evaluation dated February 27, 2004. WCAP-14040, Revision 4 removed the requirement for exemption requests to use the code cases. The WCAP-14040, Revision 4 was approved for use at the Vogtle Electric Generating Plant, Units 1 and 2, and the Comanche Peak Steam Electric Station, Units 1 and 2 by NRC Safety Evaluations dated March 28, 2005 and February 22, 2007, respectively (ADAMS Accession No. ML050690216 and ML070320823, respectively).
4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Nuclear Regulatory Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

==5.0 ENVIRONMENTAL CONSIDERATION==

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative

Enclosure A L-09-072 Page 18 of 18 occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

==6.0   REFERENCES==
: 1. Babcock and Wilcox Owner's Group Topical Report BAW-2308, Revision 1-A, "Initial RTNDT of Linde 80 Weld Materials," August 2005
: 2. Babcock and Wilcox Owner's Group Topical Report BAW-2308, Revision 2-A, "Initial RTNDT of Linde 80 Weld Materials," March 2008
: 3. NRC Regulatory Issue Summary 2004-04, "Use of Code Cases N-588, N-640, and N-641 in Developing Pressure-Temperature Operating Limits,"
April 5, 2004
: 4. Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section Xl, Division 1," Revision 13
: 5. AREVA Document ANP-2718, "Appendix G Pressure-Temperature Limits For 52 EFPY, Using ASME Code Cases for Davis-Besse Nuclear Power Station,"
Revision 002, October 2008
: 6. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988
: 7. Oak Ridge National Laboratory ORNLITM 2006/530, "A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels," November 2007
: 8. Proposed Rule, 10 CFR Part 50, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events,"
dated October 3, 2007 (72 FR 56275)
: 9. 10 CFR 50, Appendix G, "Fracture Toughness Requirements,"
December 19, 1995
: 10. Babcock and Wilcox Topical Report BAW-10046A, Revision 2, "Methods of Compliance With Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G," June 1986

Attachment 1 Proposed Operating License and Technical Specification Changes (Mark Up)
(Eight Pages Enclosed)

o       cq 6 d                 P2o'10 .
2.C. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level FENOC is authorized to operate the facility at steady state reactor core power levels not in excess of 2817 megawatts (thermal).
Prior to attaining the power level, Toledo Edison Company shall comply with the conditions identified in Paragraph (3) (o) below and complete the preoperational tests, startup tests and other items identified in Attachment 2 to this license in the sequence specified. Attachment 2 is an integral part of this license.
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 279, are hereby incorporated in the license. FENOC shall operate the facility in accordance with the Technical Specifications.
(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the issuance of the license or within the operational restrictions indicated. The removal of these conditions shall be made by an amendment to the license supported by a favorable evaluation by the Commission:
(a) FENOC shall not operate the reactor in operational Modes I and 2 with less than three reactor coolant pumps in operation.
(b) Deleted per Amendment 6 (c) Deleted per Amendment 5 L-4                       Amendment No. 118', 279

2.C(3)(d)    Prior to operation be ondXEfective Full Power Years, FENOC shall pr    e t the NRC a reanalysis and proposed modifications, as necessary, to ensure continued means of protection against low temperature reactor coolant system overpressure events.
(e)  Deleted per Amendment 33 (f)  Deleted per Amendment 33 (g)  Deleted per Amendment 33 (h)  Deleted per Amendment 24 (i)  Deleted per Amendment 11 (j)  Revised per Amendment 3 Deleted per Amendment 28 (k) Within 60 days of.startup following the first (Ist) regularly scheduled refueling outage, Toledo Edison Company shall complete tests and obtain test results as required by the Commission to verify that faults on non-Class IE circuits would not propagate to the Class IE circuits in the Reactor Protection System and the Engineered Safety Features Actuation System.
(1)  Revised per Amendment 7 Deleted per Amendment 15 (m)  Deleted per Amendment 7 (n)  Deleted per Amendment 10 (o)  Deleted per Amendment 2 (p)  Deleted per Amendment 29 (q)  Deleted per Amendment-7 (r)  Deleted per Amendment 3D (s)  Toledo Edison Company shall be exempted from the requirements of Technical Specification 3/4.7.8.1 for the two (2) Americium-Beryllium-Copper startup sources to be installed or already installed for use during the first refueling cycle until such time as the sources are replaced.
(t)  Added per Amendment 83 Deleted per Amendment 122.
Amendment No. 2,3,7,10,11,14,.47,24,28,29,30,33,83,122,199,228 L-5

PO0       &#xfd;4,u~yd     //2cj0FA9
                            -6   -
2.C(4)   Fire Protection FENOC shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Updated Safety Analysis Report and as approved in the SERs dated July 26, 1979, and May 30, 1991, subject to the following provision:
FENOC may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
(5)  Deleted per Amendment No. 279.
(.6) Antitrust Conditions FENOC and FirstEnergy Nuclear Generation Corp. shall comply with the antitrust conditions delineated in Condition 2.E of this license as if named therein. FENOC shall not market or broker power or energy from the Davis-Besse Nuclear Power Station, Unit No. 1. FirstEnergy Nuclear Generation Corp. is responsible and accountable for the actions of FENOC to the extent that said actions affect the marketing or brokering of power or energy from the Davis-Besse Nuclear Power Station, Unit No. 1, and in any way, contravene the antitrust license conditions contained in the license.
L-6   Amendment No. 1Z,41,114,U,t    ,279

peO CHitJo 4         Pfoi oso0.
Reporting Requirements 5.6 5.0   ADMINISTRATIVE CONTROLS 5.6   Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.
5.6.1          Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.
The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.
5.6.2          Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit in the previous year shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.
5.6.3          CORE OPERATING LIMITS REPORT (COLR)
: a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
: 1. SL 2.1.1.1, "Reactor Core Safety Limits";
: 2. LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";
: 3. LCO 3.1.3, "Moderator Temperature Coefficient (MTC)";
: 4. LCO 3.1.7, "Position Indicator Channels," (SR 3.1.7.1 limits);
Davis-Besse                                     5.6-1                              Amendment 279

M0      C~40)6t        P12ofte~~a, Reporting Requirements 5.6 5.6   Reporting Requirements 5.6.3          CORE OPERATING LIMITS REPORT (COLR) (continued)
: 5. LCO 3.1.8, "PHYSICS TEST Exceptions - MODE 1";
: 6. LCO 3.1.9, "PHYSICS TEST Exceptions - MODE 2";
: 7. LCO 3.2.1, "Regulating Rod Insertion Limits";
: 8. LCO 3.2.2, "AXIAL POWER SHAPING ROD (APSR) Insertion Limits";
: 9. LCO 3.2.3, "AXIAL POWER IMBALANCE Operating Limits";
: 10. LCO 3.2.4, "QUADRANT POWER TILT (QPT)";
: 11. LCO 3.2.5, "Power Peaking Factors";
: 12. LCO 3.3.1,"Reactor Protection System (RPS) Instrumentation,"
Function 8 (Flux - AFlux - Flow) Allowable Value; and
: 13. LCO 3.9.1, "Boron Concentration."
: b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, as described in BAW-10179P-A, "Safety Criteria and Methodology for Acceptable Cycle Reload Analyses," or any other new NRC approved analytical methods used to determine core operating limits that are not yet referenced in the applicable approved revision of BAW-10179P-A. The applicable approved revision number for BAW-1 01 79P-A at the time of the reload analyses are performed shall be identified in the CORE OPERATING LIMITS REPORT (COLR). The COLR shall also list any new NRC approved analytical methods used to determine core operating limits that are not yet referenced in the applicable approved revision of BAW-10179P-A.
: c. As described in reference documents listed in accordance with the instructions given above, when an initialassumed power level of 102% of RTP is specified in a previously approved method, an actual value of 100.37% of RTP may be used when the input for reactor thermal power measurement of feedwater mass flow and temperature is from the Ultrasonic Flow Meter. The following NRC approved documents are applicable to the use of the Ultrasonic Flow Meter with a 0.37%
measurement uncertainty:
: 1. Caldon Inc. Engineering Report-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM&#xfd;TM System," Revision 0, dated March, 1997.
Davis-Besse                                    5.6-2                             Amendment 279

Reporting Requirements 5.6 5.6   Reporting Requirements 5.6.3          CORE OPERATING LIMITS REPORT (COLR) (continued)
: 2. Caldon Inc. Engineering Report-1 57P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM /TM or LEFM CheckPlusTM System," Revision 5, dated October, 2001.
: d. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
: e. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.4          Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
: a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
: 1. LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits."
: 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 the following documents:
: 1. BAW-10046A, Rev. 2, "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50 Appendix G,"
1N5FJtT                      June 1986.
: c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
5.6.5          Post Accident Monitoring Report When a report is required by Condition B or F of LCO 3.3.17, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6-3                               Amendment 279 Davis-Besse                                    5.6-3                               Amendment 279

INSERT
: 2. ASME Code Section XI, Appendix G, 1995 Edition with Addenda through 1996, as modified.by the alternative procedures provided in ASME Code Case N-640 and ASME Code Case N-588;
: 3. BAW-2308, Revision 1-A and Revision 2-A, "Initial RTNDT of Linde 80 Weld Materials," August 2005 and March 2008, respectively; and
: 4. The use of the Transition Temperature Shift (TTS) model in lieu of ARTNDT to account for adjustments in material properties due to radiation embrittlement as approved by NRC Safety Evaluation for License Amendment Number [this amendment].

AJo0         Cqwi&#xfd;6            fI20 fcr Reporting Requirements 5.6 5.6   Reporting Requirements 5.6.6          Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.8, "Steam Generator (SG) Program." The report shall include:
: a. The scope of inspections performed on each SG;
: b. Active degradation mechanisms found;
: c. Nondestructive examination techniques utilized for each degradation mechanism;
: d. Location, orientation (if linear), and measured sizes (if available) of service induced indications;
: e. Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism;
: f. Total number and percentage of tubes plugged or repaired to date;
: g. The results of condition monitoring, including the results of tube pulls and in-situ testing;
: h. The effective plugging percentage for all plugging and tube repairs in each SG; and
: i. Repair method utilized and the number of tubes repaired by each repair method.
5.6.7          Remote Shutdown System Report When a report is required by Condition C of LCO 3.3.18, "Remote Shutdown System," a report shall be submitted within the following 30 days. The report shall outline the action taken, the cause of the inoperability, and the plans and schedule for restoring the control circuit or transfer switch of the Function to OPERABLE status.
Davis-Besse                                      5.6-4                              Amendment 279

Attachment 2 Proposed Operating License and Technical Specification Changes (Re-typed - For Information Only)
(Three Pages Enclosed)

                                      -5    -
2oC(3)  d)  Prior to operation beyond 32 Effective Full Power Years, FENOC shall provide to the NRC a reanalysis and proposed modifications, as necessary, to ensure continued means of protection against low temperature reactor coolant system overpressure events.
(e)  Deleted per Amendment      33 (f)  Deleted per Amendment      33 (g)  Deleted per Amendment      33 (h)  Deleted per Amendment      24 (i)  Deleted per Amendment 11 (j)  Revised per Amendment 3 Deleted per Amendment 28 (k)  Within 60 days of startup following the first          (ist) regularly scheduled refueling outage, Toledo         Edison Company shall complete tests       and obtain test   results as required by the Commission to verify that         faults on non-Class IE circuits      would not propagate to    the Class IE circuits     in the Reactor Protection System     and the Engineered Safety Features Actuation System.
(1)  Revised per Amendment 7 Deleted per Amendment 15 (m)  Deleted per Amendment      7 (n)  Deleted per Amendment 10 (o)  Deleted per Amendment      2 (p)  Deleted per Amendment      29 (q)  Deieted per Amendment 7 (r)  Deleted per Amendment      30 (s)  Toledo Edison Company shall be exempted from the requirements of Technical Specification 3/4.7.8.1 for the two (2) Americium-Beryllium-Copper startup sources to be installed     or already installed     for use during the first    refueling cycle until      such time as the sources are replaced.
(t)  Added per Amendment 83 Deleted per Amendment 122 Amendment No. 2,3,7,10, ll5,17,24,28,29,30,33,83,122,ol99,-2 L-5

Reporting Requirements 5.6 5.6   Reporting Requirements 5.6.3          CORE OPERATING LIMITS REPORT (COLR) (continued)
: 2. Caldon Inc. Engineering Report-1 57P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM /TM or LEFM CheckPlus TM System," Revision 5, dated October, 2001.
: d. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
: e. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.4          Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
: a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
: 1. LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits."
: 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 the following documents:
: 1. BAW-10046A, Rev. 2, "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50 Appendix G,"
June 1986;
: 2. ASME Code Section XI, Appendix G, 1995 Edition with Addenda through 1996, as modified by the alternative procedures provided in ASME Code Case N-640 and ASME Code Case N-588;
: 3. BAW-2308, Revision 1-A and Revision 2-A, "Initial RTNDT of Linde 80 Weld Materials," August 2005 and March 2008, respectively; and
: 4. The use of the Transition Temperature Shift (TTS) model in lieu of ARTNDT to account for adjustments in material properties due to radiation embrittlement as approved by NRC Safety Evaluation for License Amendment Number [this amendment].
: c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
Davis-Besse                                   5.6-3                              Amendment 279,

Reporting Requirements 5.6 5.6   Reporting Requirements 5.6.5          Post Accident Monitorinq Report When a report is required by Condition B or F of LCO 3.3.17, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6-3a                             Amendment 279, Davis-Besse Davis-Besse                                   5.6-3a                             Amendment 279,

Attachment 3 Proposed Technical Specification Bases Changes (For Information Only)
(Five Pages Enclosed)

INFO RATIO ONLY                                                             LTOP B 3.4.12 B 3.4 REACTOR COOLANT SYSTEM (RCS)                               ,     1 o'J B 3.4.12 Low Temperature Overpressure Protection (LTOP)
BASES BACKGROUND         LTOP controls RCS pressure at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is not compromised by violating the pressure and temperature (P/T) requirements of 10 CFR 50, Appendix G (Ref. 1). The reactor vessel is the limiting RCPB component for providing such protection. LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," provides the allowable combinations for operational pressure and temperature during cooldown, shutdown, and heatup to keep from violating the Reference 1 limits.
The reactor vessel material is less tough at reduced temperatures than at normal operating temperature. Also, as vessel neutron irradiation accumulates, the material becomes less resistant to pressure stress at low temperatures (Ref. 2). RCS pressure must be maintained low when temperature is low and must be increased only as temperature is increased.
Operational maneuvering during cooldown, heatup, or any anticipated operational occurrence must be controlled to not violate LCO 3.4.3.
Exceeding these limits could lead to brittle fracture of the reactor vessel.
LCO 3.4.3 presents requirements for administrative control of RCS pressure and temperature to prevent exceeding the P/T limits.
This LCO provides RCS overpressure protection in the applicable MODES by ensuring an adequate pressure relief capacity through the Decay Heat Removal (DHR) System relief valve.
The DHR System relief valve provides overpressure protection for the RCS during low temperature operations. RCS and DHR Systems are monitored for temperature and pressure. Maintaining the relief setpoint within the limits of the LCO ensures the Reference 1 limits will be met in any event in the LTOP analysis.
If system pressure exceeds the lift setpoint of the DHR System relief valve, it will open. As the relief valve opens, coolant is released and pressure decreases. When the relief valve reset is reached, below the LTOP pressure limit, the relief valve closes.
APPLICABLE         Safety analyses (Ref. 3) demonstrate that the reactor vessel can be SAFETY             adequately protected against overpressurization transients during ANALYSES           shutdown. In MODES 1 and 2, and portions of MODE 3, the pressurizer safety valves will prevent RCS pressure from exceeding the Reference 1 limits. For the remaining portions of MODE 3, overpressure protection is Davis-Besse                           B 3.4.12-1                                      Revision 0

WORMTIONONLYLTOP                               B 3.4.12 BASES APPLICABLE SAFETY ANALYSES (continued) provided by operating procedures. At nominally 280&deg;F to 140 0 F, overpressure prevention falls to the OPERABLE DHR System relief valve.
Below 140&deg;F, credible overpressurization sources are secured.
The actual temperature at which the pressure in the P/T limit curve falls below the pressurizer safety valve setpoint increases as vessel material toughness decreases due to neutron embrittlement. Each time the P/T limit curves are revised, the LTOP will be re-evaluated to ensure that its functional requirements can still be met with the DHR System relief valve and operating procedures.
Transients that are capable of overpressurizing the RCS have been identified and evaluated. These transients relate to either mass input or heat input: actuating the High Pressure Injection (HPI) System, discharging the Core Flooding Tanks (CFTs), energizing the pressurizer heaters, failing the makeup control valve open, losing decay heat removal, and starting a reactor coolant pump (RCP).
The DHR System relief valve (DH-4849), which is in the suction line to the decay heat pumps, has been sized to pass 1800 gpm at the nominal set pressure of 320 psig. The flow rate is based on the maximum -developed runout flow (900 gpm per pump) with both HPI pumps running simultaneously. This flow rate is considered to cause the worst credible pressure transient. The opening of a CFT isolation valve was not considered because power is removed from the valve once it is closed upon plant cooldown and depressurization. Other postulated occurrences, makeup control valve failing to open, loss of DHR System cooling, all pressurizer heaters energizing, do not produce a pressure excursion as severe as that produced by the two HPI pumps. Although the pressurizer, by procedure, cannot be solid, for the purpose of analysis it was considered to go solid during the transient. The DHR System relief valve is a Seismic Class I Nuclear Class 2 bellows type of safety-relief valve. It should be noted that the postulation of both HPI pumps starting during DHR System operation is made only for the purpose of sizing the DHR System relief valve. The possibility of this event occurring due to either a single operator error or a single spurious signal is precluded by the design of the Safety Features Actuation System.
The Reference 3 analyses demonstrate the DHR System relief valve can maintain RCS pressure below limits.
The DHR System relief valve is placed in service before RCS temperature is reduced below 2800 F. Above this temperature, the pressurizer safety valves and operating procedures provide the reactor Davis-Besse                         B 3.4.12-2                                      :Revision 0

INFORATMNONLYLTOP B 3.4.12 BASES APPLICABLE SAFETY ANALYSES (continued) vessel pressure protection. The         sel materials were ass           have a neutron irradiation accumulati neq/ue+         effective full power    ars (EFPYs) of operation.                   1zc A-T,e "N"-4Q       3 Z.
Reference 3 contains the acceptance limits that satisfy the LTOP requirements. Any change to the RCS must be evaluated against these analyses to determine the impact of the change on the LTOP acceptance limits.
As required by License Condition 2.C(3)(d), prior to operation beyo*.24    .2r 3 Effective Full Power Years, a reanalysis and proposed modifications as necessary, to ensure continued means of protection for LTOP events wi be provided to the NRC. The P/T limits are periodically modified as the reactor vessel material toughness decreases due to embrittlement induced by neutron irradiation. Revised P/T limits are determined using neutron fluence projections and the results of examinations of the reactor vessel material irradiation surveillance specimens.
LCO              For low temperature overpressure protection, Davis-Besse relies on the four-inch DHR System relief valve (DH-4849) with a lift setpoint
                < 330 psig.. This relief valve is located on the DHR System suction line from the RCS. The RCS to DHR System isolation valves (DH-1 1 and DH-12) must be open and control power removed from the valve operators for the DHR System relief valve to be OPERABLE. Control power can be removed either in the control room or at the motor control center (by removing fuses, opening breakers, or racking breakers out).
APPLICABILITY   This LCO is applicable in MODES 4 and 5, and in MODE 6 when the reactor vessel head is on. The Applicability is established by fracture mechanics analyses. The pressurizer safety valves provide overpressure protection to meet LCO 3.4.3 P/T limits in MODES 1, 2, and 3. With the vessel head off, overpressurization is not possible.
LCO 3.4.3 provides the operational P/T limits for all MODES.
LCO 3.4.10, "Pressurizer Safety Valves," requires the pressurizer safety valves OPERABLE to provide overpressure protection during MODES 1, 2, and 3.
ACTIONS         A.1 and A.2 With the DHR System relief valve inoperable due to one or both RCS to DHR System isolation valves closed, the overpressure protection flow path is isolated. The flow path must be restored by opening the RCS to DHR System isolation bypass valves (DH-21 and DH-23), within 1 hour.
Davis-Besse                        B 3.4.12-3                                        Revision 0

LTOP FO M TO ON~LY                                     B 3.4.12 A)L)0    a'4I~4VCO-9&#xfd; BASES ACTIONS A.1 and A.2 (continued)
After opening, the RCS to DHR System isolation bypass valves must be verified open every 24 hours.
The 1 hour Completion Time reflects the importance of the action and provides time for a timely opening of the RCS to DHR System isolation bypass valves. To ensure they remain in the open position, the positions of the RCS to DHR System isolation bypass valves are required to be verified every 24 hours. RCS to DHR System isolation bypass valves are manual valves and do not have remote position indication.
B.1 With control power available to one or both of the RCS to DHR System isolation valves, the overpressure protection flow path could be inadvertently isolated. The control power must be removed from the valves within 1 hour to ensure the valves will remain open during system operation.
The 1 hour Completion Time reflects the importance of the action and provides time for a timely removal of control power.
C. 1 Ifthe DHR System relief valve is inoperable for reasons other than the relief flow path (Condition A or B), the DHR System relief valve must be restored to OPERABLE status within 8 hours. The 8 hour Completion Time is acceptable due to the low probability of an overpressure event.
D.1, D.2, D.3, and D.4 If any Required Action and Completion Time of Condition A, B, or C is not met, other compensatory actions must be taken to minimize the probability and consequences of an LTOP event. Without an OPERABLE relief path for overpressure protection, the RCS water addition capabilities must be limited. Within 1 hour both HPI pumps must be disabled (e.g., by opening motor supply breakers), and within 8 hours the makeup pump suction automatic transfer to the borated water storage tank on low makeup tank level must be disabled. Makeup tank level must be verified to be < 73 inches within 8 hours to minimize volume. Furthermore, without an overpressure relief path, RCS pressure and pressurizer level Davis-Besse                    B 3.4.12-4                                    Revision 0

WFORATION ONLY                                              LTOP B 3.4.12
                     /VD       Cvl1$AJ&A~t BASES ACTIONS D.1, D.2, D.3, and D.4 (continued) must be verified to be in the Acceptable Region of Figure 3.4.12-1 or 3.4.12-2 (depending on the MODE) within 8 hours to ensure an overpressure condition cannot occur. These Figures do not include instrument error uncertainties.
SURVEILLANCE SR 3.4.12.1 REQUIREMENTS Verification of the flow path from the RCS to the DHR System relief valve is required every 24 hours. This verification is performed by checking RCS to DHR System isolation valves in the open position with control power removed from the valve operator. This Surveillance ensures the overpressure relief flow path is aligned and remains aligned. Removal of control power ensures the flow path is not inadvertently closed.
The Frequency is adequate based on operating experience.
Manual operation is required to close the isolation valves or energize control power. Valve operations are administratively controlled by procedure. In this configuration the isolation valves will not inadvertently close.
SR 3.4.12.2 Verification of the DHR System relief valve lift setpoint must be performed to ensure LTOP requirements can be met.
Overpressure protection of the RCS is ensured by the DHR System relief valve, which relieves pressure and prevents the RCS from exceeding the Pressure/Temperature Limits.
The DHR System relief valve setpoint is verified in accordance with the Inservice Testing (IST) Program for proper operation and correct lift setting of < 330 psig. This lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. The IST Program specifies the testing and frequency, as directed by ASME Code.
REFERENCES   1. 10 CFR 50, Appendix G.
: 2. Generic Letter 88-11.
: 3. UFSAR, Section 9.3.5.
Davis-Besse                       B 3.4.12-5                                      Revision 0

Enclosure B L-09-072 REQUEST FOR EXEMPTION Page 1 of 7

==Subject:==
Request for exemption from certain requirements of 10 CFR 50, Appendix G, "Fracture Toughness Requirements," and 10 CFR 50.61, "Fracture toughness requirements for protection against pressurized thermal shock events." The requested exemption would allow the use of an alternate method, described in Topical Report BAW-2308, Revisions 1-A and 2-A, "Initial RTNDT of Linde 80 Weld Materials," for determining the initial, unirradiated material reference temperatures of Linde 80 weld materials present in the beltline region
-of the Davis-Besse Nuclear Power Station, Unit No. 1 reactor pressure vessel.

==1.0 INTRODUCTION==

==2.0 BACKGROUND==

3.0 PROPOSED EXEMPTION 4.0 LINDE 80 WELD VALUES

Enclosure B L-09-072 Page 2 of 7 1.0 Introduction In accordance with the provisions of 10 CFR 50.60(b) and 10 CFR 50.12, the FirstEnergy Nuclear Operating Company (FENOC) is submitting a request for exemption from certain requirements of 10 CFR 50, Appendix G, "Fracture Toughness Requirements," and 10 CFR 50.61, "Fracture toughness requirements for protection against pressurized thermal shock events." The requested exemption would allow use of an alternate method, as described in Topical Report BAW-2308, Revisions 1-A and 2-A, "Initial RTNDT of Linde 80 Weld Materials" for determining the initial, unirradiated material reference temperatures of the Linde 80 weld materials present in the beltline region of the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS) reactor pressure vessel (RPV).

===2.0 Background===
10 CFR 50, Appendix G(II)(D)(i) and 10 CFR 50.61(a)(5) require that the pre-service or unirradiated condition RTNDT be evaluated according to the procedures in the American Society for Mechanical Engineers (ASME) Code, Section III, Paragraph NB-2331, which requires Charpy V-notch impact tests and drop weight tests.
Topical Report BAW-2308, Revisions 1-A and 2-A provide an NRC-approved alternate method for determining the initial, unirradiated material reference temperatures of the Linde 80 weld materials present in the beltline region of the DBNPS RPV. BAW-2308, Revision 2-A is a supplement to Revision 1-A, and incorporated additional test data and a re-evaluation of the reference temperature, To, determination, as requested by the NRC in the Safety Evaluation for Revision 1-A of the topical report. As stated in the NRC Safety Evaluation for Revison 2-A of the topical report dated March 24, 2008, the Conditions and Limitations, Items (1) through (4), contained in the Safety Evaluation for Revision 1-A of the topical report must be satisified in order for licensees to reference the topical report in specific licensing applications.
Condition and Limitation (4) as stated in the NRC Safety Evaluation for BAW-2308, Revision 1-A:
       "Any licensee who wants to utilize the methodology of TR BA W-2308, Revision 1 as outlined in items (1) through (3) above, must request an exemption, per 10 CFR 50.12, from the requirements of Appendix G to 10 CFR Part 50 and 10 CFR 50.61 to do so. As part of a licensee's exemption request, the NRC staff expects that

Enclosure B L-09-072 Page 3 of 7 the licensee will also submit information which demonstrates what values the licensee proposes to use for ARTNDT and the margin term for each Linde 80 weld in its RPV through the end of its facility's current operatinglicense."
In the above quotation, Condition and Limitation (1) provides criteria associated with the use of the NRC-accepted values of initial (unirradiated) reference temperature, IRTTo, and the corresponding uncertainty term, cOh, to define the initial heat-specific or generic properties of its facility's Linde 80 welds.
Condition and Limitation (2) requires that a minimum chemistry factor of 167.0&deg;F be applied when the methodology of Regulatory Guide 1.99, Revision 2, is used to assess the shift in nil-ductility transition temperature due to irradiation.
Condition and Limitation (3) requires that a value of a, = 28.0&deg;F be used, to determine the margin term, as defined in BAW-2308, Revision 1-A, and Regulatory Guide 1.99, Revision 2.
DBNPS is planning to revise its Pressure-Temperature limit curves for operation beyond 21 Effective Full Power Years and desires to use the BAW-2308 initial RTNDT values in the revision. Hence, exemption from certain requirements of 10 CFR 50, Appendix G is required. However, the DBNPS is not currently planning to revisethe pressurized thermal shock (PTS) reference temperature (RTPTS) values. Where the RTPTS is the RTNDT for reactor vessel materials calculated using the end of life fluence and the methodology described in 10 CFR 50.61. Though not needed at this point, exemption from certain requirements, as described below, of 10 CFR 50.61 would be beneficial should the DBNPS decide to revise the RTPTS in the future.
3.0 Proposed Exemption The exemption requested by FENOC addresses portions of the following regulations:
(1) 10 CFR Part 50, Appendix G which sets forth 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 system may be subjected over its service lifetime; and

Enclosure B L-09-072 Page 4 of 7 (2) 10 CFR 50.61, which sets forth fracture toughness requirements for protection against pressurized thermal shock (PTS).
The exemption from Appendix G to 10 CFR 50 is to replace the required use of the existing Charpy V-notch and drop-weight-based methodology with the use of an alternate methodology that incorporates the use of fracture toughness test data for evaluating the integrity of the Linde 80 weld materials present in the DBNPS RPV beltline regions. The alternate methodology employs direct fracture toughness testing per the Master Curve methodology based on use of ASTM Standard Method E 1921 (1997 and 2002 editions), and ASME Code Case N-629. The exemption is required since 10 CFR 50, Appendix G requires that for the pre-service or unirradiated condition, RTNDT be evaluated by Charpy V-notch impact tests and drop weight tests according to the procedures in the ASME Code, Paragraph NB-2331.
The exemption from 10 CFR 50.61 is to use an alternate methodology to allow the use of fracture toughness test data for evaluating the integrity of the Linde 80 weld materials present in the DBNPS RPV beltline regions, based on the use of the ASTM E 1921 (1997 and 2002 editions) and ASME Code Case N-629. The exemption is required because the methodology for evaluating RPV material fracture toughness in 10 CFR 50.61 requires that the pre-service or unirradiated condition be evaluated using Charpy V-notch impact tests and drop weight tests according to the procedures in the ASME Code, Paragraph NB-2331.
Additionally, the NRC Safety Evaluation for BAW-2308, Revision 1-A, concludes that an exemption is required to address issues related to 10 CFR 50.61 inasmuch as the BAW-2308, Revision 1-A, methodology, as modified and approved by the NRC staff, represents a significant change to the 10 CFR 50.61 methodology for determining the RTPTS value for Linde 80 weld material. The changes in the BAW-2308, Revision 1-A, methodology with respect to the 10 CFR 50.61 methodology, include the requirements for use of a minimum chemistry factor of 167&deg;F and a value of GA = 28.0&deg;F for Linde 80 weld materials.
10 CFR 50.12 states that the Commission may grant an exemption from requirements contained in 10 CFR 50 provided that: 1) the exemption is authorized by law, 2) the exemption will not result in an undue risk to public health and safety, 3) the exemption is consistent with the common defense and security, and 4) special circumstances, as defined in 10 CFR 50.12(a)(2) are
.present. The requested exemption to allow the use of BAW-2308, Revisions 1-A and 2-A, as the basis for the Linde 80 weld material initial properties at the DBNPS satisfy these requirements as described below.

Enclosure B L-09-072 Page 5 of 7
: 1. The requested exemption is authorized by law.
No law exists which precludes the activities covered by this exemption request. 10 CFR 50.60(b) allows the use of alternatives to 10 CFR 50, Appendix G when an exemption is granted by the Commission under 10 CFR 50.12.
In addition, 10 CFR 50.61 permits other methods for use in determining the initial material properties provided such methods are approved by the Director, Office of Nuclear Reactor Regulation.
: 2. The requested exemption does not present an undue risk to the public health and safety.
The proposed material initial properties basis described in BAW-2308, Revisions 1-A and 2-A represent an NRC-approved methodology for establishing IRTTO values for Linde 80 welds. BAW-2308, Revisions 1-A and 2-A, includes conservatisms to ensure that use of the proposed initial material properties basis does not increase the probability of occurrence or the consequences of an accident at the DBNPS and will not create the possibility for a new or different type of accident that could pose a risk to public health and safety.
The use of this proposed approach ensures that the intent of the requirements specified in 10 CFR 50, Appendix G and 10 CFR 50.61 are satisfied.
The requested exemption is consistent with the NRC staff requirements specified in the Safety Evaluation for the approved BAW-2308, Revision 1-A and Revision 2-A; consequently, the exemption does not present an undue risk to the public health and safety.
: 3. The requested exemption will not endanger the common defense and security.
The requested exemption is specifically concerned with RPV material properties and is consistent with NRC staff requirements specified in the Safety Evaluation Report for approved BAW-2308, Revision 1-A and Revision 2-A. Consequently, the requested exemption will not endanger the common defense and security.

Enclosure B L-09-072 Page 6 of 7
: 4. Special circumstances are present which necessitate the request for an exemption to the regulations of 10 CFR 50, Appendix G and 10 CFR 50.61.
Pursuant to 10 CFR 50.12(a)(2), the NRC will not consider granting an exemption to the regulations unless special circumstances are present.
The requested exemption meets the special circumstances of paragraph 10 CFR 50.12(a)(2)(ii) since application of these regulations in this particular circumstance is not necessary to achieve the underlying purpose of the regulations.
The underlying purpose of 10 CFR 50, Appendix G and 10 CFR 50.61 is to protect the integrity of the reactor coolant pressure boundary by ensuring that each reactor vessel material has adequate fracture toughness. Application of ASME Section III, paragraph NB-2331 in the determination of initial material properties was developed based on the level of knowledge existing in the-early 1970s concerning RPV materials.
Since the early 1970s, the level of knowledge concerning these topics has greatly expanded. This increased knowledge level permits relaxation of the ASME III NB-2331 requirements via application of BAW-2308, Revision 1-A and Revision 2-A, while maintaining the underlying purpose of the ASME Code and NRC regulations to ensure an acceptable margin of safety is maintained.
The proposed change in reactor vessel material initial properties will continue to satisfy the intent of 10 CFR 50, Appendix G and 10 CFR 50.61, thus justifying the exemption request. Issuance of an exemption from the criteria of these regulations to permit the use of BAW-2308, Revision 1-A and Revision 2-A for the DBNPS will not compromise the safe operation of the reactor, and will ensure that RPV integrity is maintained.
4.0 Linde 80 Weld Values As described in the NRC Safety Evaluation for BAW-2308, Revision 1-A, Condition and Limitation (4), the licensee is required to provide the values proposed for ARTNDT and the margin term for each Linde 80 weld in its RPV through the end of its facility's current operating license. Table 1 provides these values.

Enclosure B L-09-072 Page 7 of 7 Table 1 A RTNDT, OF        Margin, OF      Adjusted Reference Initial Material  Reference                                           Temperature, OF Beltline Materials    Ident. Temperature oF1/4A3/4                   T        3/4T       1/4AT        %T Location Location  Location   Location Location  Location Nozzle Belt to Upper Shell Circ. WF-233      -47.6       92.9       57.4    65.7       65.7    111.0      75.5 Weld (OD 91%)
Upper Shell to SherLower WF-182-1    -80.2       175.9      126.2   59.0       59.0    154.7      105.0 Shell Circ.
Weld}}